ML20265A198

R. E. Ginna Nuclear Power Plant, Application to Revise Technical Specifications for Steam Generator Tube Inspection Frequency

ML20265A198
Person / Time
Site:	Ginna
Issue date:	09/21/2020
From:	David Gudger Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML20265A198 (67)

Text

200 Exelon Way Kennett Square, PA 19348 www.exeloncorp.com 10 CFR 50.90 September 21, 2020 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 R. E. Ginna Nuclear Power Plant Renewed Facility Operating License No. DPR-18 NRC Docket No. 50-244

Subject:

Application to Revise R. E. Ginna Technical Specifications for Steam Generator Tube Inspection Frequency In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests an amendment to the Technical Specifications (TS) for R. E. Ginna Nuclear Power Plant (Ginna).

The proposed License Amendment Request (LAR) revises Ginna Technical Specifications (TS) 5.5.8, Steam Generator (SG) Program, to reflect a proposed change to the required SG tube inspection frequency. This request is for a one-time change to modify the SG inspection frequency from the current wording No steam generator shall operate more than 72 effective full power months or three refueling outages (whichever is less) without being inspected to add the phrase with the exception that each steam generator is to be inspected during the fourth refueling outage, in G1R44, following inspections that were completed in refueling outage G1R40.

EGC is scheduled to perform the next Ginna SG tube inspection during the Ginna Cycle 43 Refueling Outage (G1R43), which is scheduled to commence on October 4, 2021. The cycle 42 Refueling Outage (G1R42) had originally scoped in the SG tube inspections. Due to Covid-19 precautions, these inspections were deferred to reduce the number of personnel and activities onsite during the pandemic. Because the fall 2021 refueling outage will not include a full core offload, performing SG tube inspections would result in a significant manpower burden, an increased dose, and the addition of higher risk activities. These issues are not considered commensurate with the derived safety benefits. The technical analysis provided in Section 3.0 of this application demonstrates the acceptability of approving this one-time exception to the SG inspection frequency.

The operational experience of the of the Ginna Replacement SGs (RSGs), as described in the enclosure to this submittal, demonstrates that the proposed change to the schedule for the SG inspections is appropriate and will result in a reduction of dose to personnel and risk to the plant. Furthermore, the Ginna RSG operational assessment and experience supports the proposed TS change.

U.S. Nuclear Regulatory Commission Application to Revise R. E. Ginna Technical Specifications for Steam Generator Tube Inspection Frequency September 21, 2020 Page 2 The enclosure to this submittal provides a description and technical evaluation of the proposed change, a regulatory evaluation, and a discussion of environmental considerations. Attachment 1 to the enclosure provides the existing Ginna TS pages marked-up to show the proposed change. Attachment 2 to the enclosure provides the existing Ginna TS pages retyped to show the proposed change. There are no TS Bases changes. Attachment 3 to the enclosure provides the Framatomes Ginna Steam Generator Operational Assessment Deferral of the Fall 2021 Tube Inspection.

EGC has determined that there are no significant hazard considerations associated with the proposed change and that the change qualifies for a categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9).

The proposed change has been reviewed by the Ginna Plant Operations Review Committee in accordance with the requirements of the EGC Quality Assurance Program.

EGC requests approval of the proposed amendment by April 2, 2021. Once approved, the amendment shall be implemented, prior to the refueling outage. The basis for this expedited schedule is to avoid significant contingency costs (approximately $1 million) associated with preparation for a potential need to inspect during the fall 2021 refueling outage.

As a compensatory measure, during the operating cycle between G1R43 and G1R44, Ginna will administratively limit allowable SG leakage to 30 gpd/SG, from the current allowable operational leakage of 100 gpd/SG.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation,"

paragraph (b), EGC is notifying the State of New York of this application for license amendment by transmitting a copy of this letter and its attachments to the designated State Official.

Should you have any questions concerning this letter, please contact Jessie Hodge at (610) 765-5532.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 21st day of September 2020.

Respectfully, David T. Gudger Sr Manager - Licensing and Regulatory Affairs Exelon Generation Company, LLC

U.S. Nuclear Regulatory Commission Application to Revise R. E. Ginna Technical Specifications for Steam Generator Tube Inspection Frequency September 21, 2020 Page 3

Enclosure:

Evaluation of Proposed Changes Attachments: 1. Markup of Technical Specifications Pages

2. Retyped Version of Technical Specifications Pages

3. Framatome Document No. 51-9317069-000 cc: USNRC Region I, Regional Administrator w/ attachments USNRC Senior Resident Inspector, Ginna "

USNRC Project Manager, Ginna "

A. L. Peterson, NYSERDA "

Enclosure Evaluation of Proposed Changes R. E. Ginna Nuclear Power Plant Renewed Facility Operating License No. DPR-18 Docket No. 50-244

Subject:

Application to Revise R. E. Ginna Technical Specifications for Steam Generator Tube Inspection Frequency 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY ANALYSIS

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration 4.4 Conclusions

6.0 REFERENCES

ATTACHMENTS

1. Proposed TS Changes (Mark-Ups)

2. Proposed TS Changes (Final Typed)

3. Framatome Document No. 51-9317069-000

1.0

SUMMARY

DESCRIPTION In accordance with the provisions of Title 10 of the Code of Federal Regulations (10 CFR) 50.90, Application for amendment of license, construction permit, or early site permit, Exelon Generation Company, LLC (EGC) is requesting a license amendment to Operating License No. DPR-18 for the R. E. Ginna Nuclear Power Plant (Ginna). The proposed License Amendment Request (LAR) revises Ginna Technical Specifications (TS) 5.5.8, Steam Generator (SG) Program, to reflect a proposed change to the required SG tube inspection frequency. This request is for a one-time change to modify the SG inspection frequency from the current wording No steam generator shall operate more than 72 effective full power months or three refueling outages (whichever is less) without being inspected to add the phrase with the exception that each steam generator is to be inspected during the fourth refueling outage, in G1R44, following inspections that were completed in refueling outage G1R40.

The operation experience of the Ginna Replacement SGs (RSGs), as described in this enclosure, demonstrates that the proposed change to the schedule for the SG inspection is appropriate and will result in a reduction of dose to personnel, and risk to the plant. The Ginna RSG Operational Assessment (OA) and experience supports the proposed TS change.

2.0 DETAILED DESCRIPTION The following is a detailed description of the proposed Ginna TS changes (red texts).

• TS 5.5.8 d.2 is revised as follows:

Inspect 100% of the tubes at sequential periods of 144, 108, 72, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected, with the exception that each SG is to be inspected during the fourth refueling outage, in G1R44, following inspections completed in refueling outage G1R40. to this enclosure provides the existing Ginna TS page mark-up to show the proposed change. Attachment 2 to this enclosure provides the existing TS page retyped to show the proposed change. There are no TS Bases changes associated with this LAR. to the enclosure provides the Framatomes Ginna Steam Generator Operational Assessment Deferral of the Fall 2021 Tube Inspection.

The operational experience of the Ginna RSGs, as described in Section 3.0 to this enclosure, demonstrates that the proposed change to the schedule for the SG inspections is appropriate and will result in a reduction of dose to personnel and risk to the plant. Table 1 shows the current outage schedule with SG inspections. With this Amendment Request, we are proposing to perform the cycle 42 inspection at the end of cycle 43.

Table 1: R. E. Ginna SG Inspection Scope Timeline End of Refueling Cumulative Cycle Date Outage SG EFPY Inspection Scope, SG A and SG B 37 Spring G1R38 16.552

• Visual examination of hot and cold leg 2014 channel heads

• Full length bobbin probe examination of 100%

of the tubes

• +PointTM probe examination:

- Periphery (incl. no-tube lane) 3 tubes deep at TTS+/-3, hot and cold legs

- All dents and over expansions at TTS

- Miscellaneous areas of special interest, including tubes with proximity indications

• Sludge lancing / TTS SSI / FOSAR

• In SG B only, SSI upper internals, feedring, upper tube bundle 38 Fall 2015 G1R39 17.949 N/A 39 Spring G1R40 19.404

• Visual examination of hot and cold leg 2017 channel heads

• Full length bobbin probe examination of 75%

of the tubes

• Array probe examination of periphery (incl. no-tube lane) 4 tubes deep from tube end to first support, hot and cold legs; and previous proximity indications

• +PointTM probe examination:

- All dents and over expansions at TTS

- Miscellaneous areas of special interest

• Sludge lancing / TTS SSI / FOSAR

• SSI upper internals 40 Fall 2018 G1R41 20.804 N/A 41 Spring G1R42 22.195 N/A 2020

42 Fall 2021 G1R43 23.655*

• Visual examination of hot and cold leg channel heads

• Full length bobbin probe examination of 100%

of the tubes

• Array probe examination of periphery (incl. no-Original Plan tube lane) 4 tubes deep from tube end to first support, hot and cold legs; and previous proximity indications

• +PointTM probe examination:

- All dents and over expansions at TTS

- Miscellaneous areas of special interest

• Sludge lancing / TTS SSI / FOSAR

• SSI upper internals 43 Spring G1R44 25.115* N/A Original Plan 2023

• Estimates. EOC 42 and 43 EFPY are based on a cycle length of 1.46 EFPY which bounds prior actual cycle lengths.

A reduction in dose will be achieved by performing the SG inspections during a full core offload, compared to a fuel shuffle refueling outage. The dose reduction will be the result of fewer SG activities and additional shielding of personnel during refueling outages with full core offloads because secondary side water covering the SG tube bundle provides shielding and reduces exposure for all activities in containment. Typical dose values for SG inspections vary depending on outage scope and activities. However, during the G1R42 SG inspection outage the associated activities would have accounted for estimated 10.03 person-rem. Because the G1R43 refuel outage is a fuel shuffle outage, additional dose would occur due to the requirement to install nozzle dams prior to fuel shuffle. The avoided dose for Ginna personnel under the proposed amendment and planned SG inspection schedule would be approximately 2.0 person-rem, and would increase the duration of the outage by 3 days.

Many of the evolutions performed for SG inspections pose an increased risk to the plant and personnel due to high dose (e.g., heavy lifts, confined space activities). An example of plant configuration improvement and risk reduction, as a result of not performing an SG inspection during a fuel shuffle outage, is the elimination of the need for a plant mid-loop hold for installation of SG nozzle dams. This evolution is performed by personnel in a confined space internal to the SG channel head under a high radiation environment with the plant at a reduced primary water inventory. For any SG inspections, there are heavy lifts associated with moving equipment onto the refuel floor, into containment, and to the SG platforms. The primary and secondary manways must also be rigged off and installed back on the RSGs.

Personnel performing these activities could potentially be working in a locked high radiation area.

While risk is minimized as much as possible through plant processes and procedures, performing 100% SG inspection scope in the proposed outage (G1R44) will reduce the total

number of such associated activities. This will yield a corresponding reduction of personnel dose exposure while improving plant and personnel safety.

For the reasons cited above, since the first inspection after SG replacement, Ginna has not performed any SG inspections during a fuel shuffle outage. Specifically, although the Ginnas Technical Specification allows three (3) operating cycles between SG inspections and the Operational Assessments have supported it, Ginna has elected to only operate two (2) cycles, solely to avoid having to perform SG inspections during fuel shuffle outages.

This pattern has been in place from 1997 to the most recent full core offload refueling outage in Spring 2020 where a SG inspection was planned but cancelled just prior to the outage due to COVID-19 personnel safety concerns. This recent unplanned SG inspection deferral (permitted by Ginnas Technical Specification) has created the need for this one-time License Amendment Request to return to performing SG inspections only during full core offload refueling outages.

3.0 TECHNICAL ANALYSIS

3.1 System Description The R. E. Ginna Nuclear Power Plant (Ginna) design has two (2) recirculating design steam generators (SG) designed and fabricated by Babcock and Wilcox (B&W) of Cambridge, Ontario, Canada. These replacement steam generators were installed in May 1996 (G1R26). The nomenclature used for fabrication and subsequent in-service inspections is SG A and SG B. Each BWI steam generator contains 4765 tubes. The tubing material is thermally treated Inconel 690 having a nominal outer diameter (OD) of 0.750 inch and a nominal wall thickness of 0.043 inch. The nominal thickness of the tubesheet is 25.25 inches, with a full depth hydraulic expansion of all the tubes into the tubesheet material. The steam generators were designed and fabricated to the ASME Boiler and Pressure Vessel Code,Section III Division 1, 1986 Edition with no Addenda.

The tubes are supported in the straight section by eight 410 stainless steel lattice grid supports which are comprised of high, medium, and low bars. The tubes are supported in the U-bend by ten 410 stainless steel fan bar/ collector bars and the number of supports varies by row. The lower row U-bend region, Row 1 through Row 18 received additional thermal stress relief following the tube bending process. The Row 1 and Row 2 U-bends are crossover designed to maximize the bend radius and to minimize ovality.

The Ginna divider plates are fabricated from Alloy 690 material and the weld material used to weld the divider plate to the seat bar is Inconel 152, which is an Alloy 690 compatible weld material. The seat bar, which is manufactured by weld buildup, is made of Inconel 82 (I-82) material applied to an I-82 weld overlay on the tubesheet. The weld overlay and the closing seam between the primary head and the tubesheet have been subjected to post weld heat treatment process to a minimum temperature to improve resistance to stress corrosion cracking.

The Ginna feedring design includes materials that are not susceptible to flow accelerated corrosion. The feedring is fabricated from seamless Ferritic Chrome Moly steel SA-335 GR P22 and the J-tubes are fabricated from Inconel 690 SB-167 material.

The Ginna steam drum internals are fabricated from carbon steel. The steam drum contains two levels of steam separation equipment. Each steam generator uses 85 CAP (curved arm

primary) cyclones and 85 secondary cyclones. The function of the cyclones is to separate the steam from the steam/water mixture generated in the boiling section of the tube bundle.

3.2 Technical Analysis The current TS requirement is to inspect 100% of the tubes at sequential periods of 144, 108, 72, and, thereafter, 60 effective full power months. At the time of the Fall 2021 (G1R43) refueling outage the Ginna steam generators will be approximately 16 EFPM into the 72 EFPM inspection interval. At the time of the Spring 2023 (G1R44) refueling outage the Ginna Steam Generators will be approximately 34 EFPM into the 72 EFPM Inspection Interval. The proposed one-time revision allows the inspection deferral of each SG inspection until the spring of 2023 (G1R44) which is 4 cycles after the last inspection was performed in the Spring of 2017 (G1R40).

Significant operating experience has been gained over the past 24 years since the Ginna replacement steam generators were placed in service, this combined with industry operating experience of steam generators containing alloy 690 tubing and provides justification for deferring the inspection by one operating cycle.

To date the Ginna SGs have operated for over 265 EFPM with minimal degradation and a minimal number of tubes requiring plugging. Active degradation mechanisms at Ginna include lattice grid wear and foreign object wear, however, it is important to note that no foreign object wear has been detected at the top of tubesheet location.

To justify the deferral of any inspections the SG Program requires assessments to ensure safe SG inspection intervals that are based on measurable parameters that monitor SG performance, such as results of prior SG tube inspections and operational leakage.

Objective criteria to assess performance are established based on deterministic analyses and performance history. In addition, the TS requirements on operational leakage require a plant shutdown if the limits are exceeded. This ensures that the failure to meet a performance criterion, while undesirable, will not result in an immediate safety concern.

Therefore, the proposed one-time extension of the existing SG inspection frequency is acceptable.

3.2.1 Recent operational experience summary 3.2.1.a Trends of primary to secondary leakage No primary to secondary side leakage has been noted for operating Cycles 40 and 41. All trends are below 3 gallons per day (gpd) (Figure 1). The primary to secondary leak rate determination utilizes the condenser offgas method, which uses steam jet air ejector (SJAE) flow. During unit start-ups, these flows are normally higher due to systems being returned to service. The high SJAE flow causes a higher primary to secondary leak rate indication during start-up and post outage which are not related to actual primary to secondary leakage. The other larger step changes are caused by baseline flow changes in air ejector flow.

Primary to secondary leak rates are quantified/confirmed through periodic sampling of the primary coolant system and the condenser off gas/air ejectors, with leak rate calculated based on mass-balance of noble gas isotopes. Leak rates are continuously monitored using on-line radiation monitors and supporting software available in the control room.

Figure 1: Primary to Secondary Leak Rate for Operating Cycles 40 and 41 3.2.1.b Main Steam Pressure Trending Main steam pressure at Ginna has remained steady over the past two cycles. Ginna uses poly acrylic acid (PAA) injection into the main feedwater to enhance iron removal from the steam generators via blowdown. PAA injection initially began in July of 2012 and resulted in a steam pressure increase of several pounds. Over the past several cycles since 2012 steam pressure has decreased but remained consistent over the past two cycles.

Corresponding turbine governor valve positions remain at approximately 45% which is a considerable margin to valves wide open.

3.2.1.c Summary of the most recent primary and secondary (e.g., FOSAR) inspections, detected degradation and its location The "Steam Generator Tube Inspection Report, End of Cycle 39 Refueling Outage, May 2017" (WPLNRC-1003166) (Reference 5) and R.E. Ginna Station EOC 39 Steam Generator Condition Monitoring and Operational Assessment (CMOA) 0192-AST-101215 Revision 0 (Reference 2) contain the high-level summary of the most recent primary and secondary inspections, including degradation detected and location of the degradation.

3.2.1.d Number of tubes plugged and reason for plugging The number of tubes plugged for Ginna will not adversely impact tube integrity, as noted in Table 2 below. Table 3 represents the reasons for tube plugging during each SG inspection outage.

Table 2: Ginna Plugging Percentage Plugging Summary SG A SG B TOTAL (Post G1R40)

Total Tubes Plugged 1 5 6 Total PCT Plugged 0.02% 0.11% 0.06%

Plugging Limit 10% 10% 10%

Table 3: Ginna Plugging History by Degradation Mechanism Lattice Foreign Plant SG Fan Bar IGA Date Outage Grid Object Other Total EFPY EFPY Wear SCC Wear Wear 04/1996 G1R26 19.735 0 0 0 0 0 2* 2 10/1997 G1R27 20.975 1.24 Inspection SKIP 3/1999 G1R28 22.227 2.492 0 0 0 0 0 0 9/2000 G1R29 23.588 3.853 Inspection SKIP 3/2002 G1R30 24.967 5.232 0 0 0 0 0 0 9/2003 G1R31 26.348 6.613 Inspection SKIP 3/2005 G1R32 27.755 8.019 0 0 4 0 0 4 10/2006 G1R33 29.235 9.449 Inspection SKIP 4/2008 G1R34 30.657 10.939 0 0 0 0 0 0 9/2009 G1R35 31.992 12.257 Inspection SKIP 4/2011 G1R36 33.515 13.780 0 0 0 0 0 0 10/2012 G1R37 34.865 15.130 Inspection SKIP 4/2014 G1R38 36.276 16.540 0 0 0 0 0 0 10/2015 G1R39 37.672 17.936 Inspection SKIP 4/2017 G1R40 39.125 19.404 0 0 0 0 0 0 10/2018 G1R41 40.539 20.804 Inspection SKIP 4/2020 G1R42 41.930 22.195 Inspection SKIP TOTAL 0 0 4 0 2 6

• Tubes plugged at manufacture due to manufacturing defect.

3.2.1.e Relevant operating experience that could impact tube integrity Deposit Loading:

Deposit (sludge) loading has been tracked since SG replacement and is critical to monitoring steam generator thermal performance. Deposits are transported to the steam generators in the main feedwater where they settle out in the sludge on the top of the tubesheet, plate out in the tubes or are removed from the system via SG blowdown. By tracking the deposits removed via blowdown and sludge lancing the overall deposit loading can be determined. Table 4 below shows the cumulative deposit loading since the steam generators were replaced at the end of cycle 25. Figure 2 is a graph of the cumulative deposit loading and shows how only after the end of cycle 36 has the cumulative deposit loading began to significantly decrease. This can be contributed to several factors including

improved chemistry, high pressure inner bundle sludge lancing, traditional sludge lancing and use of PAA both on line and during wet layup. Figure 3 shows the positive effect PAA has had on the SG iron mass balance since PAA injection began in 2012.

Based on the relatively low total deposit loading amounts for both Ginna SGs, as shown below, the deposits do not impose an adverse impact to tube integrity. Sludge lancing and FOSAR have been performed during each SG inspection outage and were most recently performed during the 2017 RFO in both steam generators.

Table 4: Ginna Cumulative Sludge / Deposit Loading End of Cycle 26 27 28 29 30 31 32 33 34 35 36 37 38 39 40 41 Deposit /

581 1006 1483 1888 2050 2086 2156 2153 2242 2225 2300 2173 2172 2173 2123 2108 Sludge (lb.)

Figure 2: Cumulative Sludge / Deposit Loading (Cycle 26 - Present) 2400 2200 2000 1800 Deposit / Sludge (lb) 1600 1400 1200 1000 800 600 400 26 27 28 29 30 31 32 33 34 35 36 37 38 39 40 41 42 End of Cycle

Figure 3: Steam Generator Iron Mass Balance 2011 - Present Chemistry (Operating Cycles 40 and 41):

There were no adverse changes to SG Chemistry in Cycle 40 or Cycle 41.

Cycle 40 SG Chemistry:

• No CEI-R points were accrued this cycle, indicating excellent chemistry control and practices.

• A total of 54.6 pounds of iron was transported to the 2 steam generators this cycle and a total of 91.7 pounds of iron was removed from the steam generators via blowdown.

• Polyacrylic acid (PAA) was injected to the feedwater during all of Cycle 40. The injection rate was adjusted throughout the cycle to balance the blowdown iron removal rate with the feedwater input rate.

Cycle 41 SG Chemistry:

• No CEI-R points were accrued this cycle, indicating excellent chemistry control and practices.

• A total of 44.7 pounds of iron was transported to the 2 steam generators this cycle and a total of 59.3 pounds of iron was removed from the steam generators via blowdown.

• Polyacrylic acid (PAA) was injected to the feedwater during all of Cycle 41. The injection rate was adjusted throughout the cycle to balance the blowdown iron removal rate with the feedwater input rate.

Steam Drum:

Degradation of the secondary moisture separator baseplates has been identified in both steam generators. These were inspected, measured and analyzed during G1R40 (Spring 2017), it has been determined that the degradation will not result in any perforations (holes) of the baseplate in the four cycles following G1R40 that have the potential to affect moisture carryover. The secondary moisture separators were originally scheduled to be inspected using laser scanning methodology during G1R42 (2020) to obtain a second quantitative measurement. This measurement was to be a second in a series of measurements to better trend the degradation rate of the secondary moisture separators. Due to the current COVID-19 Pandemic, the decision was made to defer the secondary separator inspections.

During the scoping phase for the 2020 outage, conservative estimates were made on the amount of secondary separator degradation expected by the 2020 outage and it was deemed that no repair strategy was required. This inspection was intended as data gathering to better trend the degradation, rather than to repair perforations, therefore, deferring this inspection does not challenge any operational parameters until the next scheduled outage (Spring 2023), Attachment 3.

Contingency plans will be available for the Spring 2023 outage to ensure readiness to repair the secondary separators in the unlikely event the degradation detected is greater than projected. Contingency tooling and processes are available to weld patch plates in select severely degraded separators, this repair process has been successfully implemented at other utilities within the United States and Canada.

Therefore, based on the operating experience summary described above and industry operating experience, it was concluded that one additional operating cycle is justified with no adverse consequences to extend the planned SG inspection until the next refueling outage (G1R44).

Foreign Object Review:

Existing foreign objects remaining in the steam generators have been identified at the top of tubesheet (TTS). The visual inspections conducted during G1R40 identified four (4) objects that remain within the SGs. These are identified below in Table 5. All of the objects that remain are considered a Priority 3, are native to the SGs and pose no threat to tube integrity due to tube wear. Therefore, these objects satisfy Operational Assessment criteria for foreign objects until the next scheduled SG inspection during G1R44.

Table 5: Objects Identified and Remaining in the Steam Generators SG Row Col Leg Elevation Description Inspection State Priority Part Visible -

A 92 76 Hot TS Pile of Flakes 3 Unable to Remove Legacy Part Visible - Not A 76 98 Hot TS 3 Sludge Rock Removed Legacy Part Visible - Not A 61 109 Hot TS 3 Sludge Rock Removed Part Visible - Not B 1 71 Cold TS Legacy Sludge Rock 3 Removed Foreign Material:

A search of issue reports related to Foreign material, was performed for Ginna operating Cycles 40 and 41 and no indications of ingress of foreign material into the Steam Generators occurred from other systems during that time frame.

3.2.2. Condition Monitoring (CM) During Refueling Outage G1R40 (May 2017)

A summary of the Condition Monitoring (CM) results for G1R40 was submitted to the NRC on September 26, 2017, (Reference 5). The detailed inspection results and inputs used to perform the CM assessment are provided in the CM and Operational Assessment (OA) developed for G1R40 (Reference 2).

3.2.2.a For Each Degradation Mechanism Detected, the Most Limiting as Found Condition Compared to the Tube Performance Criteria For each degradation mechanism detected at the last Ginna steam generator inspection in G1R40 (April 2017), the most limiting as found condition was compared to the tube performance criteria and is provided in Table 6 below.

The limiting case degraded condition detected at G1R40 for each degradation mechanism was less than the condition monitoring limit; therefore, Structural Integrity Performance Criteria (SIPC) were satisfied. The severity of the limiting degraded condition for each degradation mechanism is expressed as a percent through wall (%TW) depth and compared to the CM limit calculated for a given bounding axial extent. The SIPC is met if the as found worst case depth ("Maximum Depth Recorded" in Table 6) is less than the allowable depth

("CM Limit Depth" in Table 6). As can be seen very few degradation mechanisms were identified at Ginna during G1R40, all of which have a large margin to the condition monitoring limit.

Table 6: Summary of Condition Monitoring performance for Existing Degradation Mechanisms during G1R40 (April 2017) (Reference 2)

Maximum Projected Degradation Max Depth Margin %TW Depth CM Limit Mechanism at G1R40 to CM Growth Recorded Depth Detected at from Limit per ETSS at G1R40 (%TW)

G1R40 G1R38 OA (%TW) EFPY

(%TW)

(%TW)

None Fan Bar Wear N/A N/A N/A N/A N/A Identified Lattice Grid Wear 10 31.9 51.3 41.3 0.3 96910.1 Foreign Object 39 37 58.5 21.5 0 27901.1 Wear (Note 1)

Note:

(1) Existing Foreign Object (FO) Wear is not predicted to grow due to FOs no longer present at location indicated.

3.2.2.b Discuss any tubes that required flaw profiling to demonstrate condition monitoring was met.

No tubes required flaw profiling to demonstrate that the CM limit was not exceeded.

3.2.3 Operational Assessment (OA) Supporting an Additional Operating Cycle 3.2.3.a The existing degradation mechanisms observed at Ginna which require an OA are as follows:

• Mechanical wear from lattice grid support structures

• Mechanical wear from foreign objects In addition to the existing degradation mechanisms above, potential degradation mechanisms were considered in light of Ginnas request to operate for 4-cycles between SG inspections. An OA for potential degradation mechanisms predicts the behavior of postulated flaws that could have been present at or prior to the last SG inspection in G1R40 and those that could have initiated during the 4-cycle operating period. The potential degradation mechanisms for which an OA was performed are:

• Mechanical wear from fan bar supports

• Tube to tube wear 3.2.3.b Inspection Strategy Details During G1R40 Inspection for the Degradation Mechanisms Described Above are as Follows:

• All mechanical wear mechanisms - Full-length bobbin inspections of 75% of in-service tubes using qualified techniques were used to detect mechanical wear. In addition, at the TTS where the bobbin probe's detection capabilities are challenged, supplemental array probe testing was used. These exams included full coverage of the tube periphery where the flow velocities and operating experience have shown a greater susceptibility to foreign object wear. The top of tubesheet array (X-Probe'1) scope was approximately 28% of all tubes on the hot and cold legs. Bobbin or +Point'1 probe was used to depth size all detected wear at structures. The +Point' probe was used to depth and length size all foreign object wear.

3.2.3.c Operational assessment summary for all degradation mechanisms; including predicted margin to the tube integrity performance criteria at G1R44 (April 2023):

The technical justification for deferring the G1R43 SG tube examination by one operating cycle due to mechanical wear at structures and mechanical wear from foreign objects is based on a new operational assessment (OA), Attachment 3, performed in accordance with EPRI Steam Generator Integrity Assessment Guidelines (IAGL) (Reference 1). This OA supplements the current OA (Reference 2) from the end of operating Cycle 42 condition to the end of operating Cycle 43, thus justifying operation of the SGs for four operating cycles between SG eddy current inspections. The OA provided in Attachment 3 will fully support the deferral of the G1R43 SG inspections until the next refueling outage (G1R44) such that, for the existing and potential degradation mechanisms:

1. Structural integrity performance criterion (SIPC) margin requirement of three times normal operating pressure (3xNOPD) on tube burst will be satisfied at G1R44 for the existing and potential degradation, and

2. Accident-induced leakage performance criteria (AILPC) for the limiting accident condition will be met for the end of Cycle 43 condition.

SG tubing is subject to two types of degradation; existing degradation, or degradation modes previously observed within the Ginna SGs, and potential degradation, or degradation modes not yet observed within the Ginna SGs but judged to have a meaningful likelihood of occurrence based on operation of similar units or laboratory testing.

To the required technical justification for deferral of the planned SG examinations during G1R43, Attachment 3, provides the OA predictions for all existing degradation mechanisms and all the potential degradation mechanisms necessary to justify extended operation of the Ginna SGs until G1R44 (April 2023). A summary of the methodology and results of the OA analysis are presented below.

Methodology for Mechanical Wear Mechanisms This evaluation will examine degradation that was detected, characterized and accepted for continued operation at the most recent inspection outage. It must also consider degradation 1

+Point and X-Probe are trademarks or registered trademarks of Zetec, Inc., its subsidiaries and/or affiliates in the United States of America and may be registered in other countries through the world. All rights reserved.

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that may have existed but was not detected, either because of imperfect inspection probability of detection (POD) or because the tube was not examined.

The G1R40 inspection included full length bobbin inspection of 75% of the tubes; therefore, 25% of the tubes were last examined during G1R38. Consequently, the OA considers two scenarios to project the limiting G1R44 flaw depth: a) degradation detected during G1R40 and returned to service, and b) degradation present, but undetected during G1R38.

Mechanical wear from lattice grid support structures Progression of G1R40 Wear (Scenario (a))

Only four indications of lattice grid wear were detected during G1R40. All four had been reported during the previous inspection (G1R38) and they had experienced negligible change in depth during the operating period between G1R38 and G1R40 (specifically, the depth of one flaw increased by 1%TW). The maximum reported depth was 10%TW.

Using the maximum measured depth (10%TW) from G1R40 and adjusting to conservatively account for NDE sizing uncertainty in order to determine the upper bound depth (UBD) from G1R40. Using the calculation in section 4.1.1.1 of Framatome document 51-9317069-000 (Attachment 3) a upper bound depth (UBD) from G1R40 of 25%TW was determined.

For the OA, a reasonable growth rate must be assumed for lattice grid wear. Since only four indications of lattice grid wear were detected, a statistically valid upper 95th percentile growth rate cannot be calculated. As shown in Table 7, there was essentially no change in depth for these indications. Based on the small population and the lack of measured growth, a bounding growth rate of 2.0%TW per EFPY was used in the OA.

Using an operating period of 5.711 EFPY between G1R40 and G1R44 and a growth rate of 2.0%TW per EFPY yields an end of cycle 43 upper bound depth of 36.4%TW.

Conservatively assuming the structural limit of a 3.2 inch long uniformly deep wear flaw is 51.3%TW (Reference 4). Since the projected value of EOC43 UBD is less than 51.3%TW, it is concluded that lattice grid wear reported during G1R40 will not challenge the structural integrity performance criteria prior to the next inspection outage (G1R44).

Table 7: Lattice Grid Wear Indications Max %TW Prior Delta Depth Delta Growth SG Row Col Location ETSS Outage %TW Current EFPY per Depth Growth Outage EFPY A 1 25 05H - 1.76 96910.1 10% TW 10% TW 0 2.871 0 B 2 20 06C - 1.62 96910.1 6% TW 6% TW 0 2.871 0 B 78 24 01H + 0.97 96910.1 8% TW 8% TW 0 2.871 0 B 78 24 01H + 1.26 96910.1 7% TW 6% TW 1 2.871 0.3 Wear Undetected during G1R38 (Scenario (b))

The EPRI SG Integrity Assessment Guidelines (Reference 1) specify that the beginning of cycle (BOC) UBD for degradation present but not reported must be assumed equal to the depth corresponding to a probability of detection of 0.95. During G1R38, bobbin probe ETSS 96004.1 was used to detect lattice grid wear. Utilizing the industrys MAPOD methodology (Reference 3) in conjunction with Ginna-specific ECT noise measurements, the depth at POD=0.95 was determined to be 7%TW. However, as an added measure of conservatism, BOC38 UBD was assumed to be 20%TW. Note that because POD curves reflect actual flaw depths rather than measured flaw depths, no adjustment for sizing uncertainty is required.

Adjusting the BOC38 UBD upward to reflect depth growth during the 8.562 EFPY period between G1R38 and G1R44, yields the EOC43 UBD of 37.1%TW. Based on the small population and the lack of measured growth, a bounding growth rate of 2.0%TW per EFPY was used in the OA.

Since the projected value of EOC43 UBD is less than 51.3%TW, it is concluded that lattice grid wear postulated as present but undetected during G1R38 will not challenge the structural integrity performance criteria prior to the next inspection outage (G1R44).

Mechanical Wear from Fan Bar Support Structures Fan bar wear has not been identified in the Ginna SGs. Consequently, the OA must only address fan bar wear postulated to have been present but not detected during the G1R38 examination (i.e., Scenario (b) in above). During G1R38, bobbin probe ETSS 96004.1 was used to detect fan bar wear. Utilizing the industrys MAPOD methodology (Reference 3) in conjunction with Ginna-specific ECT noise measurements, the depth at POD=0.95 was determined to be 7%TW. However, as an added measure of conservatism, BOC38 UBD was assumed to be 20%TW. Note that because POD curves reflect actual flaw depths rather than measured flaw depths, no adjustment for sizing uncertainty is required.

Adjusting the BOC38 UBD upward to reflect depth growth during the 8.562 EFPY period between G1R38 and G1R44, yields the EOC43 UBD of 37.1%TW, Attachment 3, assuming a hypothetical fan bar wear growth rate of 2.0%TW per EFPY exists.

Since the projected value of EOC43 UBD is less than 51.3%TW, Attachment 3, it is concluded that fan bar wear postulated as present but undetected during G1R38 will not challenge the structural integrity performance criteria prior to the next inspection outage (G1R44).

Tube-to-Tube Mechanical Wear No tube-to-tube wear has been identified in the Ginna SGs. Therefore, only wear that is postulated to have been present but not detected during the G1R38 examination is evaluated in this section (i.e., Scenario (b) from above).

During the G1R38 examination, bobbin probe technique ETSS 13091.1 was used for the detection and sizing of tube-to-tube wear. As with the lattice grid and fan bar wear mechanisms, the bobbin probe provides excellent detection capability for tube-to-tube wear.

A review of ETSS 13091.1 reveals that all 40 of the test samples, some as shallow as 7%TW and 16 less than 20%TW, were detected during the development of this technique.

In the absence of Ginna noise measurement data for these techniques, the depth at POD=0.95 and thus, BOC38 UBD, was assumed to be 20%TW. It should be noted that during the G1R38 inspection, all proximity indications were also examined with +PointTM probes (ETSS 13901.1), providing additional detection capability in those regions considered most susceptible to tube-to-tube wear.

Without any history of tube-to-tube wear in the Ginna SGs, or in any other recirculating SG manufactured by BWXT Canada, the postulation of tube-to-tube wear and the assumption of future growth rate are highly speculative. For the purpose of this OA the future growth rate of hypothetical tube-to-tube wear is assumed to be 2%TW/EFPY. Adjusting the BOC38 UBD upward to reflect depth growth during the 8.562 EFPY period between G1R38 and G1R44, yields the EOC43 UBD:

EOC43 UBD = BOC38 UBD + (8.562 EFPY)(2.0%TW/EFPY)

EOC43 UBD = 37.1%TW Due to the tendency of tube-to-tube wear to be significantly longer than the support wear evaluated in earlier sections, the structural limit used earlier is not applicable. For this degradation mechanism a flaw length of 30 inches was assumed and the corresponding structural limit was determined to be 50.0%TW (Reference 4). Since the projected value of EOC43 UBD is less than 50.0%TW, it is concluded that tube-to-tube wear will not challenge the structural integrity performance criteria prior to the next inspection outage (G1R44).

Mechanical Wear Due to Foreign Objects Only three (3) foreign object wear flaws have been identified to-date in the Ginna SGs after over 19 EFPY of operation. All three are located at lattice grid support intersections, were identified during the G1R38 inspection (or earlier). These indications exhibited no growth during the period from G1R38 to G1R40 and no foreign objects remained at the affected locations. The flaws did not exceed the plugging limit (maximum depth was 37%TW, see Table 8 below) and did not challenge the SG performance criteria at G1R40. Since the objects that caused the wear in the tubes are no longer present at the locations of the wear indications, there is no mechanism to cause future growth.

Table 8: Foreign Object Wear Indications Max %TW Prior Delta Depth Delta Growth SG Row Col Location ETSS Outage %TW Current EFPY per Depth Growth Outage EFPY A 91 50 05H + 0.35 27901.1 37% TW 37% TW 0 2.871 0 A 53 85 03H - 2.07 27901.1 21% TW 21% TW 0 2.871 0 B 2 78 02H - 1.84 27901.1 25% TW 25% TW 0 2.871 0 During the spring 2017 refueling outage at Ginna (G1R40), Foreign Object Search and Retrieval (FOSAR) was performed at the top of the secondary tubesheet. Although no new foreign object wear indications were reported during the EOC39 inspections, foreign objects were removed that consisted of 18 pieces of flexitallic gasket, three pieces of weld slag, and one small machine curl.

The EOC39 inspection scope for foreign objects and associated wear was extensive and included both visual and eddy current inspections. Visual inspections included both the annulus and no-tube lane at the top of the tubesheet in both steam generators. The eddy current examinations included tubesheet array probe inspections of all tubes within about five tubes of the periphery and no-tube lane on both the hot and cold sides of the tubesheets in both steam generators. With these extensive inspections and subsequent part removal, there is reasonable assurance that no parts capable of causing significant tube degradation remain in the tube bundle.

Despite the extensive inspections and removal of multiple parts, the OA considers the potential for tube degradation from parts entering the bundle during the next inspection interval. The objects that were found during the EOC39 outage were consistent, in terms of both quantity and type of material, with findings of previous inspections at Ginna. This indicates that the parts are not likely to have come from a single, recent intrusion event.

Rather, its indicative of minor intrusion events occurring over multiple cycles. Therefore, it is expected that similar material may also be detected during the next inspection during the EOC43 outage. Since no new foreign object wear indications were detected during EOC39 after two cycles of operation, there is reasonable assurance that any foreign object wear that has the potential to be generated during two additional cycles of operation will meet the structural integrity performance criterion. The applicable structural limit is 58.3%TW for volumetric degradation with limited circumferential extent and bounding assumed length of 0.5 inches (Reference 4).

Since no wear exceeding the structural criteria is expected, there is reasonable assurance that the operational leakage and accident leakage performance criteria will not be exceeded by foreign object wear prior to the next tube examination in each steam generator.

Operational Assessment Conclusion

There is reasonable assurance that with a deferral of the next Ginna SG examination from G1R43 to G1R44, the SG structural and leakage integrity performance criteria will remain satisfied throughout the operating period preceding G1R44. Table 9 summarizes the projected margin to SIPC and AILPC at G1R44 for each tube degradation mechanism evaluated.

Table 9: Ginna SG Tube Integrity Margin Summary SIPC AILPC Degradation Mechanism Projected Upper Margin to Limit Limit (at Bound Depth at bounding length) G1R44 Limit G1R44 Projection Lattice Grid Support Wear 51.3 %TW 37.1 %TW 14.2 %TW 250 GPD Zero Leakage Fan Bar Support Wear (Potential) 51.3 %TW 37.1 %TW 14.2 %TW 250 GPD Zero Leakage Tube-to-tube Wear (Potential) 50.0 %TW 37.1 %TW 12.9 %TW 250 GPD Zero Leakage Foreign Object Wear 58.3 %TW < 58.3 %TW N/A 250 GPD Zero Leakage 3.2.4 Mitigating Strategies As a compensatory measure, during the operating cycle between G1R43 and G1R44, Ginna will administratively limit allowable SG leakage to 30 gpd/SG, from the current allowable operational leakage of 100 gpd/SG.

The normal Mode 1 (Power Operation) requirement under the EPRI Primary-to-Secondary Leak Guidelines state that Action Level 1 is reached at a leakage rate of 30 gallons per day (gpd) However, the guidelines require a site to enter an "Increased Monitoring" condition when total primary-to-secondary leakage is detected to be equal to or greater than 5 gpd.

After entering the increased monitoring condition, radiation monitors alert/alarm set points are reset, as necessary, to above their existing baseline reading (but not over 30 gpd) to permit detection of rapidly increasing leakage.

The EGC Primary-to-Secondary Leak Program procedure (CY-AP-120-340) (Reference 6) currently in effect has lower administrative limits on Primary-to-Secondary leakage in order to ensure Ginna is prepared to quickly respond should the leakage rate increase. Steam Generator Management Program Monitoring condition is entered when normal radiochemical grab sampling and process radiation monitors indicate leakage of greater than or equal to 3 gpd. This describes the condition in which leakage has been detected and quantified and is greater than or equal to 3 gpd but is not in a range that can be accurately monitored by most radiation monitors. When this occurs, Engineering is notified, and the appropriate Corrective Action Processes are initiated to document and track the excursion. The activities when this condition is entered are described in EGC procedure ER-AP-420-0051, "Conduct of Steam Generator Management Program Activities."

(Reference 7)

When operational leakage is equal to or greater than 3 gpd is confirmed during the operating period between inspections, at the next outage, in situ pressure testing, tube pull, or analysis should be performed to quantify the expected accident leak rate to assess compliance with accident leakage performance criteria. In addition, prior to entering an outage, an action plan is developed to address means of identifying the defective tube(s),

flowchart sampling methods to bound the defect and provide reasonable assurance that unit restart is prudent.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria The following NRC requirements and guidance document are applicable to the review of the proposed change.

The proposed change has been evaluated to determine whether applicable regulations and requirements continue to be met. EGC has determined that the proposed change does not require any exemptions or relief from the applicable regulatory requirements. The following current applicable regulations and regulatory requirements were reviewed in making this determination:

Criterion 14, Reactor Coolant Pressure Boundary The reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, or rapidly propagating failure, and of gross rupture.

Compliance with GDC 14 is described in Section 3.1.2.2.5 of the Ginna UFSAR.

Criterion 15, Reactor Coolant System Design The reactor coolant system and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences.

Compliance with GDC 15 is described in Section 3.1.2.2.6 of the Ginna UFSAR.

Criterion 16, Containment Design Reactor containment and associated system shall be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.

Compliance with GDC 16 is described in Section 3.1.2.2.7 of the Ginna UFSAR.

Criterion 30, Quality of reactor coolant pressure boundary Components, which are part of the reactor coolant pressure boundary, shall be designed, fabricated, erected, and tested to the highest quality standards practical.

Means shall be provided for detecting and, to the extent practical, identifying the location of the source of reactor coolant leakage.

Compliance with GDC 30 is described in Section 3.1.2.4.1 of the Ginna UFSAR.

Criterion 31, Fracture prevention of reactor coolant pressure boundary The reactor coolant pressure boundary shall be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions (1) the boundary behaves in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the boundary material under operating, maintenance, testing, and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of irradiation on material properties, (3) residual, steady state and transient stresses, and (4) size of flaws.

Compliance with GDC 31 is described in Section 3.1.2.4.2 of the Ginna UFSAR.

Criterion 32, Inspection of reactor coolant pressure boundary Components, which are part of the reactor coolant pressure boundary, shall be designed to permits (1) periodic inspection and testing of important areas and features to assess their structural and leaktight integrity, and (2) an appropriate material surveillance program for the reactor pressure vessel.

Compliance with GDC 32 is described in Section 3.1.2.4.3 of the Ginna UFSAR.

4.2 Precedent While there is no exact precedent for this LAR, precedence does exit for one-time changes to SG inspection frequencies. For example:

• In Reference 8, NRC issued a license amendment for Arkansas Nuclear One, Unit 2, which granted a one-time change to revise the steam generator inservice inspection frequency TS requirements to allow a 40-month inspection interval after one inspection, rather than after two consecutive inspections.

• In Reference 9, NRC issued a license amendment for South Texas Project (STP), Unit 1, which granted a one-time change to extend the steam generator inservice inspection frequency TS requirements from 40 months to 44 months.

• In Reference 10, NRC issued a license amendment for Virgil C. Summer Nuclear Station, Unit No. 1, which granted a one-time change to extend the steam generator inservice inspection frequency TS requirements from 40 months to 58 months.

• In Reference 11, NRC issued a license amendment for Braidwood Station, Unit 2, which granted a one-time change to extend the steam generator inservice inspection frequency TS requirements from a maximum of 2 cycles to 3 cycles.

• In Reference 12, Byron Station, Unit 2, submitted a license amendment which asked for a one-time change to extend the steam generator inservice inspection frequency TS requirements from a maximum of 2 cycles to 3 cycles.

4.3 No Significant Hazards Consideration Exelon Generation Company, LLC (EGC), proposes changes to the Technical Specifications (TS), Appendix A of Renewed Facility Operating License No. DPR-18 for R. E. Ginna Nuclear Power Plant (Ginna). The proposed change revises Ginna TS 5.5.8, Steam Generator (SG) Program, to require SG tube inspection frequency from 72 Effective Full Power Months (EFPM) or three refueling outages, whichever is shorter, to add one four-refueling outage inspection interval (G1R40 to G1R44).

Ginna operating experience supports this TS revision. No adverse impact to safety and reliability is expected as a result of such a change.

An evaluation of the proposed change has been performed in accordance with 10 CFR 50.91(a)(1) regarding no significant hazards considerations using the standards in 10 CFR 50.92(c). A discussion of these standards as they relate to this amendment request follows:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The implementation of the proposed amendment has no significant effect on either the configuration of the plant or the manner in which it is operated based on the improved RSG design and reliability, the Inservice Inspection (ISI) data, and Operational Assessments (OAs). The consequences of a hypothetical failure of a tube remain bounded by the current SG Tube Rupture (SGTR) analysis described in the Ginna Updated Final Safety Analysis Report (UFSAR). A main steam line break, feedwater line break, or an ATWS event will not cause a SGTR because the SG tubes will still meet their structural and leakage performance criteria. Therefore, EGC has concluded that the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated in the Ginna UFSAR.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change will not alter any plant design basis or postulated accidents resulting from potential SG tube degradation. The proposed change does not affect the design of the Replacement SGs (RSGs), the method of operation, nor the reactor coolant chemistry controls. No new equipment is being introduced, and installed equipment is not being operated in a new or different manner. The proposed change will not give rise to new failure modes. In addition, the proposed change does not impact any other plant systems or components. The proposed amendment has no effect on either the configuration of the plant, nor the manner in which it is operated.

Therefore, EGC concludes that this proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The steam generator tubes are an integral part of the reactor coolant pressure boundary and, as such are relied upon to maintain the primary system pressure and inventory. Revising the SG tube inservice inspection frequency will not alter their function or design. The improved design of the RSGs [Alloy 690 thermally treated (Alloy 690TT) tubes], the ISI data and OAs also provide reasonable assurance that significant tube degradation is not likely to occur. Therefore, EGC concludes that this proposed change does not involve a significant reduction in a margin of safety.

Based on the above, EGC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c) and, accordingly, a finding of no significant hazards consideration is justified.

4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement.

However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

1. Electric Power Research Institute (EPRI) Report 3002007571, "Steam Generator Management Program: Steam Generator Integrity Assessment Guidelines," Revision 4, June 2016

2. Technical Evaluation 0192-AST-101215 Rev. 000, R. E. Ginna Station EOC 39 Steam Generator Condition Monitoring and Operational Assessment (CMOA) dated May 11th 2017

3. EPRI Report 3002010334, Model Assisted Probability of Detection Using R (MAPOD-R), Version 2.1, September 2017

4. BWXT Document 0192-AST-101038, R. E. Ginna Station EOC 39 SG Degradation Assessment, Revision 1, May 2017

5. "Steam Generator Tube Inspection Report, End of Cycle 39 Refueling Outage, May 2017" (WPLNRC-1003166) (ML17276A348)

6. CY-AP-120-340, Primary-to-Secondary Leak Program procedure, Revision 10

7. ER-AP-420-0051, "Conduct of Steam Generator Management Program Activities, Revision 24

8. NRC letter to Entergy Operations, Inc., Arkansas Nuclear One, Unit No. 2 -

Issuance of Amendment Re: One-Time Change of Steam Generator Tube Inspection Frequency (TAC No. MB6808), dated May 28, 2003 (ML031490475).

9. NRC letter to STP Nuclear Operating Company, South Texas Project, Unit 1 -

Issuance of Amendment Re: Onetime Extension to Steam Generator Inservice Inspection Frequency (TAC No. MC1046), dated June 8, 2004 (ML041610073).

10. NRC letter to South Carolina Electric & Gas Company, Virgil C. Summer Nuclear Station, Unit No. 1 - Issuance of Amendment Re: One-Time Extension of the Steam Generator Inspection Frequency (TAC No. MB7312), dated October 29, 2003 (ML033020450).

11. NRC letter to Exelon Generation Company, LLC, Braidwood Station, Unit 2 -

Issuance of Amendment No. 209, Revision Technical Specification 5.5.9, Steam Generator (SG) Program, for One-Time Revision to Frequency of SG Tube Inspections (Exigent Circumstances) (EPID L-2020-LLA-0069), dated May 1, 2020 (ML20111A000).

12. Exelon Generation Company, LLC letter to NRC, Byron Station, Units 1 and 2Application for Revision to TS 5.5.9, Steam Generator (SG) Program, for a One-Time Deferral of Steam Generator Tube Inspections, dated July 10, 2020 (ML20195B158).

ATTACHMENT 1 Markup of Technical Specifications Pages (red texts)

R. E. Ginna Nuclear Power Plant Renewed Facility Operating License No. DPR-18 Docket No. 50-244 Revised Technical Specifications Pages TS Pages 5.5.8

Programs and Manuals 5.5

3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE."

c. Provisions for SG tube repair criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40%

of the nominal tube wall thickness shall be plugged.

d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial, and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

1. Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.

2. Inspect 100% of the tubes at sequential periods of 144, 108, 72, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected, with the exception that each SG is to be inspected during the fourth refueling outage in G1R44 following inspections completed in refueling outage G1R40.

3. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

e. Provisions for monitoring operational primary to secondary LEAKAGE R.E. Ginna Nuclear Power Plant 5.5-6 Amendment 110,

ATTACHMENT 2 Markup of Technical Specifications Pages R. E. Ginna Nuclear Power Plant Renewed Facility Operating License No. DPR-18 Docket No. 50-244 Revised Technical Specifications Pages TS Pages 5.5.8

Programs and Manuals 5.5

3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE."

c. Provisions for SG tube repair criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40%

of the nominal tube wall thickness shall be plugged.

d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial, and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

1. Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.

2. Inspect 100% of the tubes at sequential periods of 144, 108, 72, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected, with the exception that each SG is to be inspected during the fourth refueling outage in G1R44 following inspections completed in refueling outage G1R40.

3. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

e. Provisions for monitoring operational primary to secondary LEAKAGE R.E. Ginna Nuclear Power Plant 5.5-6 Amendment 110,

ATTACHMENT 3 R.E. Ginna Nuclear Power Plant NRC Docket No. 50-244 Framatome Document No. 51-9317069-000 R. E. Ginna Steam Generator Operational Assessment Deferral of the Fall 2021 Tube Inspection

Technical Evaluation of 51-9317069-000 Rev. 001 R. E. Ginna Steam Generator Operational Assessment Deferral of the Fall 2021 Steam Generator Inspections Passport ATI: 04369718-02 9/18/2020 Prepared By: Date:

Mike Bodak - Fleet Steam Generator Engineer Reviewed By: Date: 9/18/2020 Patrick Creegan - Fleet Steam Generator Engineer ITPR: Date: 9/18/2020 C. Lee Friant - Independent Third-Party Review (ITPR)

Approved By: Date: 9/18/2020 Heather Malikowski - Asset Protection Manager Page 1 of 35

REVISION

SUMMARY

Revision Revision Description Date 000 9/17/2020 Original Issue Attachment 2, Owner Acceptance Review Checklist, was revised to 001 9/18/2020 reflect the evaluation of impacts due to an additional accident event (i.e., ATWS) by Framatome.

Page 2 of 35

TABLE OF CONTENTS 1.0 REASON FOR EVALUATION ............................................................. 4 2.0 DETAILED EVALUATION: ................................................................. 4

3.0 CONCLUSION

......................................................................................... 4

4.0 REFERENCES

........................................................................................ 5 5.0 DAR RESPONSES: ................................................................................. 5 ATTACHMENT 1 ................................................................................................ 6 ATTACHMENT 2 ................................................................................................ 7 Page 3 of 35

1.0 REASON FOR EVALUATION The purpose is to demonstrate that the primary and secondary side examinations currently planned for refueling outage G1R43 (end of Cycle 42, fall 2021) may be safely deferred by one additional operating cycle to G1R44 (spring 2023). A License Amendment Request for a One-Time Extension of the Steam Generator Tube Inspections was communicated to the NRC on 8/31/2020. The final LAR submittal will occur prior to start of the G1R43 outage.

The purpose of this EC Evaluation is to review and accept the primary side Operational Assessment (OA), including the secondary side foreign object evaluation OA completed by Framatome. These documents evaluate the Ginna Unit 1 Steam Generators (SG) anticipated condition during the next operating period (operational assessment) up through G1R44.

Based on LS-AA-117 Revision 18, Section 4.7.5. Engineering shall approve all technical information (e.g., design analyses, configuration changes, and technical evaluations) provided in support of licensing correspondence under the appropriate Engineering processes. All technical information provided by a vendor shall also be reviewed and approved under the appropriate Engineering processes.

2.0 DETAILED EVALUATION:

The Framatome SG Operational Assessments (51-9317069-000) as documented in Attachment 1 have been reviewed by Exelon Tube Integrity to confirm that the assumptions made are consistent with plant design parameters and with the way that the steam generators are operated.

No unverified assumptions exist in the operational assessment that would have the potential to affect safe and reliable plant operation. All critical input parameters are clearly identified, understood and are consistent with appropriate plant design parameters. All engineering judgments and decisions have been reviewed and are clearly documented, justified, and support safe and reliable plant operation. The operational assessments (OA) conclude that there is reasonable assurance that operation of the Ginna Unit 1 SGs for a total of 4 cycles of operation starting in cycle 40, will not cause any of the three performance criteria (structural integrity, accident-induced leakage integrity, and operational leakage integrity) to be exceeded. One additional Cycle of operation up through G1R44 requires a Tech Spec change through the LAR process, which will be submitted to the NRC for approval in September 2020.

Note: See Attachment 2 for the Owners Acceptance Review of the Framatome Document 51-9317069-000 Revision 0.

3.0 CONCLUSION

Consistent with the requirements of the Exelon Steam Generator (SG) Program ER-AP-420-0051, Tube Integrity GLs, and G1R40 SG inspection results, this document evaluates the anticipated condition at the end of the next operating period (operational assessment) up to G1R44. The operational assessments (OA) conclude that there is reasonable assurance that operation of the Ginna Unit 1 steam generators for up to four operating cycles starting from cycle Page 4 of 35

G1R40 will not cause any of the three performance criteria to be exceeded. Operation for one additional operating cycle to G1R44 will require a change to the Ginna Unit 1 Technical Specification regarding maximum inspection intervals (one-time, four cycles). This will be submitted to the NRC for approval in September 2020.

4.0 REFERENCES

: Framatome Document 51-9317069-000, Rev. 0, R. E. Ginna Steam Generator Operational Assessment Deferral of the Fall 2021 Steam Generator Inspections Revision 0 5.0 DAR RESPONSES:

The applicable DAR items have been reviewed against the G1R43 revised Operational Assessments, no additional reviews were required.

Review Method and Results of Review:

A detailed independent technical review was performed by Pat Creegan, a qualified Exelon Steam Generator Program Engineer. Inputs and reasonableness of the output have been verified.

The Framatome evaluation results are consistent with the purpose of the Operational Assessment deferral of SG inspection to G1R44.

Page 5 of 35

ATTACHMENT 1 Framatome R. E. Ginna Steam Generator Operational Assessment Deferral of the Fall 2021 Steam Generator Inspections Document No.: 51-9317069-000.

Page 6 of 35

20004-025 (02/27/2018)

Framatome Inc.

Engineering Information Record Document No.: 51 - 9317069 - 000 R. E. Ginna Steam Generator Operational Assessment Deferral of the Fall 2021 Steam Generator Inspections Page 7 of 35 Page 1 of 21

20004-025 (02/27/2018)

Document No.: 51-9317069-000 R. E. Ginna Steam Generator Operational Assessment Deferral of the Fall 2021 Steam Generator Inspections Safety Related? YES NO Does this document establish design or technical requirements? YES NO Does this document contain assumptions requiring verification? YES NO Does this document contain Customer Required Format? YES NO Signature Block Pages/Sections Name and P/LP, R/LR, M, Prepared/Reviewed/

Title/Discipline Signature A-CRF, A Date Approved or Comments Kent Colgan KA COLGAN LP All Advisory Engineer 9/3/2020 Craig Kelley CJ KELLEY LR All Supervisory Engineer 9/3/2020 Wayne Belden WD BELDEN A All Manager, SG Perf.

Engineering 9/4/2020 Note: P/LP designates Preparer (P), Lead Preparer (LP)

M designates Mentor (M)

R/LR designates Reviewer (R), Lead Reviewer (LR)

A-CRF designates Project Manager Approver of Customer Required Format (A-CRF)

A designates Approver/RTM - Verification of Reviewer Independence Project Manager Approval of Customer References (N/A if not applicable)

Name Title (printed or typed) (printed or typed) Signature Date Mike Epling Project Manager MW EPLING 9/3/2020 Page 8 of 35 Page 2

20004-025 (02/27/2018)

Document No.: 51-9317069-000 R. E. Ginna Steam Generator Operational Assessment Deferral of the Fall 2021 Steam Generator Inspections Record of Revision Revision Pages/Sections/

No. Paragraphs Changed Brief Description / Change Authorization 000 All Original release Page 9 of 35 Page 3

Document No.: 51-9317069-000 R. E. Ginna Steam Generator Operational Assessment Deferral of the Fall 2021 Steam Generator Inspections Table of Contents Page SIGNATURE BLOCK ............................................................................................................................. 2 RECORD OF REVISION ....................................................................................................................... 3 LIST OF TABLES .................................................................................................................................. 5 LIST OF FIGURES ................................................................................................................................ 5 1.0 PURPOSE ................................................................................................................................. 6

2.0 BACKGROUND

......................................................................................................................... 6 2.1 Regulatory Requirements ............................................................................................... 6 2.2 SG Design and OA Inputs............................................................................................... 6 2.3 Inspection Timeline ......................................................................................................... 8 2.4 Recent Operating Experience ......................................................................................... 8 2.5 Degradation Mechanisms ............................................................................................. 10 3.0 ASSUMPTIONS ....................................................................................................................... 11 4.0 OPERATIONAL ASSESSMENT............................................................................................... 11 4.1 Tube Degradation - Structural Integrity ........................................................................ 11 4.1.1 Lattice Grid Support Wear .............................................................................. 11 4.1.2 Fan Bar Support Wear ................................................................................... 13 4.1.3 Tube-to-Tube Wear ........................................................................................ 14 4.1.4 Foreign Object Wear ...................................................................................... 15 4.2 Tube Degradation - Leakage Integrity ........................................................................... 16 4.3 Secondary Side Components ....................................................................................... 16 4.4 Operational Assessment Conclusion ............................................................................ 18 5.0 COMPUTER FILES .................................................................................................................. 18

6.0 REFERENCES

......................................................................................................................... 20 APPENDIX A : ABBREVIATIONS AND ACRONYMS ...................................................................... 21 Page 10 of 35 Page 4

Document No.: 51-9317069-000 R. E. Ginna Steam Generator Operational Assessment Deferral of the Fall 2021 Steam Generator Inspections List of Tables Page Table 2-1: Basic Inputs ......................................................................................................................... 7 Table 2-2: R. E. Ginna SG Inspection Scope Timeline .......................................................................... 9 Table 2-3: Evaluated Degradation Mechanisms .................................................................................. 10 Table 4-1: Lattice Grid Wear Indications ............................................................................................. 12 Table 4-2: Foreign Object Wear Indications ........................................................................................ 15 Table 4-3: Most Advanced Secondary Separator Base Plate FAC ...................................................... 17 Table 4-4: Ginna SG Tube Integrity Margin Summary......................................................................... 18 Table 5-1: Computer Files .................................................................................................................. 19 List of Figures Page Figure 2-1: NOPD since G1R40............................................................................................................ 8 Page 11 of 35 Page 5

Document No.: 51-9317069-000 R. E. Ginna Steam Generator Operational Assessment Deferral of the Fall 2021 Steam Generator Inspections 1.0 PURPOSE This report provides an operational assessment (OA) of R. E. Ginna steam generators (SGs). The purpose is to demonstrate that the primary and secondary side examinations currently planned for refueling outage G1R43 (end of Cycle 42, fall 2021) may be safely deferred by one additional operating cycle to G1R44 (spring 2023).

2.0 BACKGROUND

2.1 Regulatory Requirements Technical specification 5.5.8, Exelon SG Program procedures [7, 8], and NEI 97-06 [1] require that a forward looking operational assessment be performed to determine if the steam generator tubing will continue to meet specific structural and leakage integrity performance criteria throughout the operating period preceding the next inspection. The performance criteria are:

• Structural Integrity Performance Criteria (SIPC) - Margin of 3.0 against burst under normal steady state power operation and a margin of 1.4 against burst under the most limiting design basis accident. Additional requirements are specified for non-pressure accident loads. Based upon the guidance of Reference [6, Section 3.7.2], it is concluded that these non-pressure accident loads are not relevant for the degradation types and locations that could credibly occur in the Ginna SGs prior to G1R44.

• Operational Leakage Performance Criteria - The operational leakage limit is specified in LCO 3.4.13, RCS Operational Leakage as 150 gallons per day through any one SG [8].

• Accident Induced Leakage Performance Criteria (AILPC) - Leakage shall not exceed the value assumed in the limiting accident analysis (locked rotor and rod ejection): 250 GPD per SG [8].

2.2 SG Design and OA Inputs The Ginna power plant design incorporates two recirculating steam generators (SG) designed and fabricated by Babcock and Wilcox of Cambridge, Ontario, Canada; installed as replacements in May 1996. The thermally treated Inconel 690 tubes were installed with a full depth hydraulic expansion into the tubesheet. The tubes are supported in the straight section by eight 410 stainless steel lattice grid supports which are comprised of high, medium, and low bars. The tubes are supported in the U-bend by ten 410 stainless steel fan bar / collector bars and the number of supports varies by row. The lower row U-bend region (row 1 through row 18) received additional thermal stress relief following the tube bending process. The row 1 and row 2 U-bends are crossover designed to maximize the bend radius and to minimize ovality.

The feedring and J-tubes were fabricated with materials that are not susceptible to flow accelerated corrosion: seamless ferritic chrome moly steel SA-335 GR P22 and Inconel 690 SB-167, respectively. The steam drum internals are fabricated from carbon steel. The design incorporates two levels of steam separation equipment: 85 curved arm primary (CAP) cyclones and 85 secondary cyclones.

Page 6 of 21 Page 12 of 35

Document No.: 51-9317069-000 R. E. Ginna Steam Generator Operational Assessment Deferral of the Fall 2021 Steam Generator Inspections Table 2-1 identifies basic input information used in this evaluation. The nominal full power steady state primary to secondary pressure differential (NOPD) value shown in the table was confirmed to be conservative based upon a review of Ginna operating data through 8-10-2020; illustrated in Figure 2-1.

Table 2-1: Basic Inputs Tube Material Thermally Treated Alloy 690 (A690TT) [8]

Tube OD 0.750 inch [2]

Tube Wall Thickness 0.043 inch [2]

Number of Installed Tubes (per SG) 4765 [8]

Mean Tube Material Yield plus 121.32 ksi [2]

Ultimate Strength (SY+SU) at Operating Temperature Standard Deviation of (SY+SU) 1.464 ksi [2]

Primary Hot Leg Temperature 608 °F [8]

Nominal Full Power Steady State 1461 psi [2]

Primary to Secondary Pressure Differential (NOPD)

MSLB Pressure Differential 2560 psi [8]

(MSLB PD)

Assumed Operating Duration 102.74 EFPM from G1R38 to G1R44 8.562 EFPY 6 cycles Assumed Operating Duration 68.52 EFPM from G1R40 to G1R44 5.710 EFPY 4 cycles Page 7 of 21 Page 13 of 35

Document No.: 51-9317069-000 R. E. Ginna Steam Generator Operational Assessment Deferral of the Fall 2021 Steam Generator Inspections Figure 2-1: NOPD since G1R40 2.3 Inspection Timeline Table 2-2 provides a brief description of the Ginna SG inspections performed or to be performed from the spring 2014 (G1R38) outage through the spring 2023 (G1R44) outage. A detailed description of the inspection history and the most recent SG inspection scope (G1R40) is provided in various References [2, 3, 5]. As discussed above, the purpose of this OA is to determine if the planned G1R43 examinations may be safely deferred until G1R44. The table shows the scope as-planned prior to the deferral.

2.4 Recent Operating Experience Based on the recommendations of Reference [6, Section 11.2.4], Ginna and industry operating experiences that might impact SG tube integrity or the assessment of tube integrity must be considered when evaluating extended inspection intervals. Important factors include chemistry excursions and transients, sustained power level changes, unusual operating transients, foreign material intrusion events, and industry identification of new degradation mechanisms. An extensive evaluation of these factors was performed in the spring of 2020 to assess the acceptability of the G1R42 inspection skip [9]. The evaluation identified no Ginna or industry operating experiences that would undermine the existing integrity assessments and identified no new degradation mechanisms as credible threats to Ginna SG integrity. A more recent review of industry operating Page 8 of 21 Page 14 of 35

Document No.: 51-9317069-000 R. E. Ginna Steam Generator Operational Assessment Deferral of the Fall 2021 Steam Generator Inspections experience by the author in support of this OA similarly identified no concerns beyond those addressed in this report. Going forward, Exelon has identified no plans to implement power uprates, load following operations, or other operational changes that could meaningfully impact Ginna SG degradation initiation and growth through G1R44.

Table 2-2: R. E. Ginna SG Inspection Scope Timeline End of Refueling Cumulative Cycle Date Outage SG EFPY [9] Inspection Scope, SG A and SG B 37 Spring 2014 G1R38 16.552

• Visual examination of hot and cold leg channel heads

• Full length bobbin probe examination of 100% of the tubes TM

• +Point probe examination:

- Periphery (incl. no-tube lane) 3 tubes deep at TTS+/-3, hot and cold legs

- All dents and over expansions at TTS

- Miscellaneous areas of special interest, including tubes with proximity indications

• Sludge lancing / TTS SSI / FOSAR

• In SG B only, SSI upper internals, feedring, upper tube bundle 38 Fall 2015 G1R39 17.949 N/A 39 Spring 2017 G1R40 19.404

• Visual examination of hot and cold leg channel heads

• Full length bobbin probe examination of 75% of the tubes

• Array probe examination of periphery (incl. no-tube lane) 4 tubes deep from tube end to first support, hot and cold legs; and previous proximity indications TM

• +Point probe examination:

- All dents and over expansions at TTS

- Miscellaneous areas of special interest

• Sludge lancing / TTS SSI / FOSAR

• SSI upper internals 40 Fall 2018 G1R41 20.804 N/A 41 Spring 2020 G1R42 22.194 N/A 42 Fall 2021 G1R43 23.654*

• Visual examination of hot and cold leg channel heads

• Full length bobbin probe examination of 100% of the tubes

• Array probe examination of periphery (incl. no-tube lane) 4 tubes deep from tube end to first support, hot and cold legs; Original Plan and previous proximity indications TM

• +Point probe examination:

- All dents and over expansions at TTS

- Miscellaneous areas of special interest

• Sludge lancing / TTS SSI / FOSAR

• SSI upper internals Page 9 of 21 Page 15 of 35

Document No.: 51-9317069-000 R. E. Ginna Steam Generator Operational Assessment Deferral of the Fall 2021 Steam Generator Inspections 43 Spring 2023 G1R44 25.114* N/A Original Plan

• Estimated EOC 42 and 43 EFPY values are based on a cycle length of 1.46 EFPY which bounds prior actual cycle lengths.

2.5 Degradation Mechanisms The most recent degradation assessment [2] provides detailed discussion regarding the susceptibility of Ginna SG tubes to various degradation mechanisms. The only two tube degradation mechanisms identified to date (i.e., existing mechanisms) in the Ginna SGs are summarized in Table 2-3. The table also includes two degradation mechanisms that have not been identified but which could reasonably be expected to occur (i.e. potential mechanisms). Based upon a review of Ginna operations and industry operating experience (Section 2.4), it is concluded that no other degradation mechanisms pose a credible threat to Ginna SG tube integrity prior to G1R44. Each degradation mechanism identified in this table is evaluated within this OA.

Table 2-3: Evaluated Degradation Mechanisms Evaluated in Type Mechanism Detection Strategy Section Existing Lattice grid support wear

• Full length bobbin probe 4.1.1 examinations Potential Fan bar support wear

• Full length bobbin probe 4.1.2 examinations Potential Tube-to-tube wear

• Full length bobbin probe 4.1.3 examinations TM

• +Point or array probe examination of proximity indications 1

Existing Foreign object wear

• Hot and cold leg array probe 4.1.4 TM or +Point probe examinations of TTS periphery tubes

• Full length bobbin probe examinations

• TTS SSI / FOSAR 1

To-date, foreign object wear has only been identified at lattice supports; therefore, it is currently a potential degradation mechanism at the TTS.

Page 10 of 21 Page 16 of 35

Document No.: 51-9317069-000 R. E. Ginna Steam Generator Operational Assessment Deferral of the Fall 2021 Steam Generator Inspections 3.0 ASSUMPTIONS Various assumptions used during the course of this evaluation are identified within the applicable section discussions. None of the assumptions require verification beyond this reports independent review and approval process.

4.0 OPERATIONAL ASSESSMENT The condition monitoring and operational assessment (CMOA) performed during G1R40 [3]

determined that the Ginna SGs satisfied the technical specification performance criteria and concluded that the SGs could be safely operated until G1R43 (fall 2021). The OA herein evaluates the acceptability of one additional cycle of operation prior to performing the next SG inspection during G1R44.

References [3] and [5] provide detailed tabulations of G1R40 inspection results as well as the CM assessment of the G1R40 findings. No additional tube examinations have been performed since G1R40; therefore, that information will not be repeated in this report.

4.1 Tube Degradation - Structural Integrity The OA must evaluate degradation that was detected, characterized and accepted for continued operation at the most recent inspection outage. It must also consider degradation that may have existed but was not detected, either because of imperfect inspection probability of detection (POD) or because the tube was not examined.

The bobbin probe is the principal detection method used for existing and potential degradation in the Ginna SGs. The G1R40 bobbin probe inspection included 75% of the tubes; hence, 25% of the tubes were most recently examined with bobbin probes during G1R38. Consequently, this OA considers two scenarios to project the limiting G1R44 flaw depth: a) degradation detected during G1R40 and returned to service, and b) degradation present but undetected during G1R38.

Note that although degradation may also be postulated to have been missed during the G1R40 examination, the growth interval from G1R40 to G1R44 is less limiting than the interval from G1R38 to G1R44. Therefore, this evaluation is based on the more conservative assumption of undetected degradation during G1R38.

The following sections summarize the evaluations performed for the degradation mechanisms discussed in Section 2.4.

4.1.1 Lattice Grid Support Wear 4.1.1.1 Progression of G1R40 Wear (Scenario (a))

Only four indications of lattice grid wear were detected during G1R40. All four had been reported during the previous inspection (G1R38) and they had experienced negligible change in depth during the operating period between G1R38 and G1R40 (specifically, the depth of one flaw increased by 1%TW).

The maximum reported depth was 10%TW [3] based on +PointTM ETSS 96910.1. See Table 4-1 below.

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Document No.: 51-9317069-000 R. E. Ginna Steam Generator Operational Assessment Deferral of the Fall 2021 Steam Generator Inspections Table 4-1: Lattice Grid Wear Indications The maximum measured depth (10%TW) must be adjusted to conservatively account for NDE sizing uncertainty in order to determine the upper bound depth (UBD) at the beginning of Cycle 40 (BOC40). This was accomplished by applying the sizing uncertainty parameters for ETSS 96910.1 to generate the upper 95% probability / 50% confidence (95%/50%) depth estimate. The ETSS 96910.1 (Revision 11) sizing uncertainty parameters are: slope=0.95, intercept=6.70, and standard error of regression=5.36. The adjustment was performed as follows:

BOC40 UBD = upper 95%/50% maximum depth of in-service lattice grid wear at the beginning of Cycle 40 BOC40 UBD = (0.95)(10)+(6.70)+(1.645)(5.36)2 BOC40 UBD = 25.0%TW For the OA, a conservative growth rate must be assumed for lattice grid wear going forward. Since only four indications of lattice grid wear were detected, a statistically valid upper 95%/50% growth rate cannot be calculated. As discussed above the existing lattice grid wear demonstrated negligible growth from G1R38 to G1R40. Based on the small population and the absence of meaningful growth, a bounding growth rate of 2.0%TW per EFPY was assumed.

Adjusting the BOC40 UBD upward to reflect depth growth during the 5.711 EFPY period between G1R40 and G1R44, yields the end of Cycle 43 (EOC43) UBD:

EOC43 UBD = BOC40 UBD + (5.710 EFPY)(2.0%TW/EFPY)

EOC43 UBD = 36.4%TW 2

In accordance with [6], analyst sizing uncertainty is adequately represented by the standard error of regression, often referred to as the technique uncertainty.

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Document No.: 51-9317069-000 R. E. Ginna Steam Generator Operational Assessment Deferral of the Fall 2021 Steam Generator Inspections The projected EOC43 UBD must be compared with the appropriate structural limit. In this case the appropriate limit conservatively accounts for uncertainties in material strength and in the burst pressure relationship. The structural limit for a conservatively assumed 3.2 inch long uniformly deep wear flaw is 51.3%TW [2]. Since the projected value of EOC43 UBD is less than 51.3%TW, it is concluded that lattice grid wear reported during G1R40 will not challenge the structural integrity performance criteria prior to the next inspection outage (G1R44).

4.1.1.2 Wear Undetected during G1R38 (Scenario (b))

Reference [6] specifies that the BOC UBD for degradation present but not reported must be assumed equal to the depth corresponding to a probability of detection of 0.95. During G1R38, bobbin probe ETSS 96004.1 (Revision 13) was used to detect lattice grid wear. The ability of this technique to detect wear is excellent. A review of ETSS 96004.1 reveals that all indications in the dataset were detected, including wear as shallow as 4%TW. Utilizing the industrys MAPOD methodology [12] in conjunction with Ginna-specific ECT noise measurements, the depth at POD=0.95 was determined to be 7%TW. However, as an added measure of conservatism, BOC38 UBD was assumed to be 20%TW. Note that because POD curves reflect actual flaw depths rather than measured flaw depths, no adjustment for sizing uncertainty is required.

Adjusting the BOC38 UBD upward to reflect depth growth during the 8.562 EFPY period between G1R38 and G1R44, yields the EOC43 UBD:

EOC43 UBD = BOC38 UBD + (8.562 EFPY)(2.0%TW/EFPY)

EOC43 UBD = 37.1%TW Since the projected value of EOC43 UBD is less than 51.3%TW, it is concluded that lattice grid wear postulated as present but undetected during G1R38 will not challenge the structural integrity performance criteria prior to the next inspection outage (G1R44).

4.1.1.3 Conclusion Neither lattice grid wear flaws returned to service following the G1R40 inspection, nor those postulated to be present but undetected during the G1R38 inspection, are projected to violate the structural limit at G1R44. Therefore, there is reasonable assurance that lattice grid wear will not violate the structural integrity performance criteria during the operating period preceding G1R44.

4.1.2 Fan Bar Support Wear As discussed earlier, fan bar wear has not been identified in the Ginna SGs. Consequently, the OA must only address fan bar wear postulated to have been present but not detected during the G1R38 examination (i.e., Scenario (b) in Section 4.1).

During the G1R38 examination, bobbin probe technique ETSS 96041.1 (Revision 4) was used for the detection of fan bar wear. The bobbin probe provides excellent detection capability for this degradation mechanism. Utilizing the industrys MAPOD methodology [12] in conjunction with Ginna-specific ECT noise measurements, the depth at POD=0.95 was determined to be 15%TW.

However, as an added measure of conservatism, BOC38 UBD was assumed to be 20%TW.

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Document No.: 51-9317069-000 R. E. Ginna Steam Generator Operational Assessment Deferral of the Fall 2021 Steam Generator Inspections In the absence of any fan bar wear, it was assumed for the purpose of this OA that the future growth rate of hypothetical fan bar wear will be 2%TW/EFPY. Adjusting the BOC38 UBD upward to reflect depth growth during the 8.562 EFPY period between G1R38 and G1R44, yields the EOC43 UBD:

EOC43 UBD = BOC38 UBD + (8.562 EFPY)(2.0%TW/EFPY)

EOC43 UBD = 37.1%TW Since the projected value of EOC43 UBD is less than the structural limit of 51.3%TW at a bounding length of 3.2 inches [2], it is concluded that fan bar wear will not challenge the structural integrity performance criteria prior to the next inspection outage (G1R44).

4.1.3 Tube-to-Tube Wear No tube-to-tube wear has been identified in the Ginna SGs. Therefore, only wear that is postulated to have been present but not detected during the G1R38 examination is evaluated in this section (i.e., Scenario (b) from Section 4.1).

During the G1R38 examination, bobbin probe technique ETSS 13091.1 (Revision 1) was used for the detection of tube-to-tube wear. As with the lattice grid and fan bar wear mechanisms, the bobbin probe provides excellent detection capability for tube-to-tube wear. A review of ETSS 13091.1 reveals that all 40 of the test samples, some as shallow as 7%TW and 16 less than 20%TW, were detected during the development of this technique. In the absence of Ginna noise measurement data for these techniques, the depth at POD=0.95 and thus, BOC38 UBD, was assumed to be 20%TW. It should be noted that during the G1R38 inspection, all proximity indications were also examined with +PointTM probes (ETSS 13901.1), providing additional detection capability in those regions considered most susceptible to tube-to-tube wear.

Without any history of tube-to-tube wear in the Ginna SGs, or in any other recirculating SG manufactured by BWXT Canada, the postulation of tube-to-tube wear and the assumption of future growth rate are highly speculative. For the purpose of this OA the future growth rate of hypothetical tube-to-tube wear is assumed to be 2%TW/EFPY. Adjusting the BOC38 UBD upward to reflect depth growth during the 8.562 EFPY period between G1R38 and G1R44, yields the EOC43 UBD:

EOC43 UBD = BOC38 UBD + (8.562 EFPY)(2.0%TW/EFPY)

EOC43 UBD = 37.1%TW Due to the tendency of tube-to-tube wear to be significantly longer than the support wear evaluated in earlier sections, the structural limit used earlier is not applicable. For this degradation mechanism a flaw length of 30 inches was assumed and the corresponding structural limit was determined to be 50.0%TW [10]. Consistent with the structural limits discussed earlier, this limit conservatively accounts for uncertainties in material strength and in the burst pressure relationship.

Since the projected value of EOC43 UBD is less than 50.0%TW, it is concluded that tube-to-tube wear will not challenge the structural integrity performance criteria prior to the next inspection outage (G1R44).

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Document No.: 51-9317069-000 R. E. Ginna Steam Generator Operational Assessment Deferral of the Fall 2021 Steam Generator Inspections 4.1.4 Foreign Object Wear Only three (3) foreign object wear flaws have been identified to-date in the Ginna SGs after over 19 EFPY of operation (Table 4-2). All three are located at lattice grid support intersections, were identified during the G1R38 inspection (or earlier), and exhibited no growth during the period from G1R38 to G1R40. Based upon +PointTM and bobbin probe examinations, no foreign objects remained at the affected locations. These flaws did not exceed the plugging limit (maximum depth was 37%TW) and did not challenge the SG performance criteria at G1R40. The affected tubes were returned to service following G1R40. Since the objects that caused the wear in the tubes that were returned to service are no longer present at the locations of the wear indications, there is no mechanism to cause future growth. Consequently, these flaws will not grow to exceed the applicable structural limit (58.3%TW) for volumetric degradation with limited circumferential extent and bounding assumed length of 0.5 inches [2]. Therefore, there is reasonable assurance that these flaws will not violate the performance criteria going forward to G1R44.

Table 4-2: Foreign Object Wear Indications The G1R40 inspection scope for foreign objects and associated wear was extensive and included both visual and eddy current inspections. Visual examinations performed following sludge lancing included both the annulus and no-tube lane at the top of the tubesheet in both steam generators.

These examinations viewed all rows and columns into the tube bundle interior at all peripheral and no-tube lane locations. The eddy current examinations included tubesheet array probe inspections of all periphery tubes (4 tubes deep), including the no-tube lane, on both the hot and cold legs of the tubesheets in both steam generators. This examination provided excellent foreign object and foreign object wear detection capability. In addition, although not specifically qualified to detect foreign object wear at the top of tubesheet, the bobbin probe does provide a significant detection capability at these locations and is the primary means of detection elsewhere in the tube bundle. A large sample of the tubes in both SGs was examined full length with bobbin probes. With these Page 15 of 21 Page 21 of 35

Document No.: 51-9317069-000 R. E. Ginna Steam Generator Operational Assessment Deferral of the Fall 2021 Steam Generator Inspections extensive inspections and subsequent object retrievals, there is reasonable assurance that no parts capable of causing significant tube degradation remained in the SGs following G1R40.

The potential development of new foreign object wear during the operating period through G1R44 must be considered. It is difficult to predict if and when foreign object wear will occur. However, by examining the aggregate operating history of the Ginna SGs with respect to foreign object wear, a judgment of the risk can be developed. During G1R40, 18 pieces of Flexitallic gasket, three pieces of weld slag, and one small machine curl were identified and removed from the SGs. These objects were consistent, both in terms of quantity and type of material, with findings of previous inspections at Ginna. This suggests that the parts did not originate from a single, recent intrusion event, but rather from minor intrusion events occurring over multiple cycles. Despite the presence of these objects for various operating times, no foreign object wear has been identified at the top of tubesheet to-date. Since no new foreign object wear indications of any depth were detected during G1R40 after two cycles of operation, either at the top of tubesheet or elsewhere in the tube bundles; there is reasonable assurance that any new foreign object wear generated during the operating period prior to G1R44 will not violate the structural integrity performance criterion.

The results of FOSAR, array probe examinations, bobbin coil examinations, and Ginnas history with foreign objects, provide reasonable assurance that deferral of the SG examinations to G1R44 will not generate foreign object wear that violates the SG structural integrity performance criteria. In the unlikely event that significant degradation does occur, primary to secondary leakage monitoring procedures in place at Ginna provide a high degree of confidence of safe unit shutdown without challenging the performance criteria.

4.2 Tube Degradation - Leakage Integrity Consistent with Reference [6, Section 9.6], for volumetric degradation of the type and length detected or postulated to occur in the Ginna SGs (i.e., support wear and/or tube-to-tube wear), the onset of pop-through leakage and burst are coincident. This means that if the degradation satisfies the structural integrity performance criteria (i.e., does not burst at 3xNOPD) then the degradation will not pop-through and leak at the much lower accident pressure differential (i.e., MSLB PD), or the still lower normal operating pressure differential (i.e., NOPD).

As discussed in Sections 4.1.1, 4.1.2, 4.1.3, and 4.1.4, there is reasonable assurance that no degradation will violate the structural integrity performance criteria prior to a G1R44 SG examination. Therefore, there is reasonable assurance that the accident-induced leakage performance criteria and the normal operating leakage performance criteria will remain satisfied throughout the operating period preceding G1R44.

4.3 Secondary Side Components Degradation of SG secondary side components that could impact the ability of the SGs to perform their intended safety functions must be considered. The principal concern is component degradation that could produce foreign objects that could, in-turn, impact tube integrity. During prior outages the condition of secondary side components has been routinely monitored. Results of potential significance to tube integrity are discussed below.

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Document No.: 51-9317069-000 R. E. Ginna Steam Generator Operational Assessment Deferral of the Fall 2021 Steam Generator Inspections Previous inspections have identified flow accelerated corrosion (FAC) of the secondary separator base plates. If the FAC progresses local perforations could occur. Conceivably, with further progression, local perforations could eventually join and allow the center portion of the base plate to separate and become a foreign object. Two plants of similar design have identified perforations but neither visual nor eddy current inspections have identified associated foreign objects or foreign object wear.

During G1R40, the base plates of all 85 secondary separators in both steam generators were visually inspected and were also inspected using laser profilometry. Table 4-3 provides the through-thickness degradation depth of the five most significant FAC locations in each SG. No perforations were observed during the visual inspections and the deepest base plate degradation was measured as 51% through the plate [11].

To evaluate the acceptability of deferring the G1R43 secondary side inspections to G1R44, the future growth of this FAC must be considered. The most recent OA review [9] which evaluated operation through G1R43, assumed that the FAC began in 2011 when it was first observed. This is a very conservative assumption since the FAC is known to have initiated prior to the outage of detection. In fact, it is most likely that the FAC began when the SGs initially operated following SG replacement (G1R26, 1996), but could have accelerated as a result of the 17% power uprate implemented in 2006 (G1R33). Assuming the FAC did not begin until 2008 (G1R34), one cycle after the uprate, the FAC growth rate is conservatively estimated as 6.0%/EFPY.

Prior analyses have shown that perforation due to operating pressure differentials across the base plate will not occur if the through-thickness degradation is less than 90% [13]. Table 4-33 provides the projected G1R44 depth for the deepest five FAC locations in each SG. Since the degradation depth is projected to remain below 90%, there is reasonable assurance that secondary separator FAC will not generate foreign objects within the SGs prior to G1R44. These findings support the deferral of secondary side inspections until G1R44.

Table 4-3: Most Advanced Secondary Separator Base Plate FAC Projected G1R44 G1R40 Through- Through-Thickness Steam Generator Separator Number Thickness Depth (%) Depth (%)

A 73 32 66 A 15 33 67 A 54 33 67 A 26 39 73 A 46 51 85 B 80 29 63 B 85 33 67 B 69 34 68 B 79 36 70 B 71 38 72 Page 17 of 21 Page 23 of 35

Document No.: 51-9317069-000 R. E. Ginna Steam Generator Operational Assessment Deferral of the Fall 2021 Steam Generator Inspections 4.4 Operational Assessment Conclusion There is reasonable assurance that with a deferral of the next Ginna SG examination from G1R43 to G1R44, the SG structural and leakage integrity performance criteria will remain satisfied throughout the operating period preceding G1R44. Table 4-4 summarizes the projected margin to SIPC and AILPC at G1R44 for each tube degradation mechanism evaluated.

Table 4-4: Ginna SG Tube Integrity Margin Summary SIPC AILPC Degradation Mechanism Projected Upper Limit (at bounding Bound Depth at length) G1R44 Limit G1R44 Projection Lattice Grid Support Wear 51.3 %TW 37.1 %TW 250 GPD Zero Leakage Fan Bar Support Wear 51.3 %TW 37.1 %TW 250 GPD Zero Leakage Tube-to-tube Wear 50.0 %TW 37.1 %TW 250 GPD Zero Leakage Foreign Object Wear 58.3 %TW < 58.3 %TW 250 GPD Zero Leakage 5.0 COMPUTER FILES Table 5-1 identifies computer files used in the course of this operational assessment. All files were transferred to the following Framatome ColdStor directory: \cold\General-Access\51\51-9317069-000\official.

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Document No.: 51-9317069-000 R. E. Ginna Steam Generator Operational Assessment Deferral of the Fall 2021 Steam Generator Inspections Table 5-1: Computer Files File Size Modified Date Filename File Type (kB) Description and Time CMOA EOC37 G1R38.pdf PDF 1528454 Ginna SG CMOA based on G1R38 inspection 7/29/2020 8:49 AM results CMOA EOC39 G1R40 0192-AST-101215 - rev 000.pdf PDF 1312528 Ginna SG CMOA based on G1R40 inspection 7/29/2020 8:49 AM results DA EOC39 G1R40 0192-AST-101038 - rev 001.pdf PDF 2551209 Ginna SG DA prior to G1R40 7/29/2020 8:49 AM Skip DA-OA EOC41 G1R42.pdf PDF 289507 Assessment to support the planned skip of SG 7/29/2020 8:49 AM inspections during G1R42 Ginna SG Delta P.xls MS Excel 261120 Primary to secondary pressure differential data 8/14/2020 8:36 AM from spring 2017 to 8-10-2020 Ahat_96004_1_more_conservative.csv MS Excel 562 Actual depth vs signal amplitude for ETSS 96004.1 4/25/2019 12:10 PM wear samples Ahat_I-96041_1.csv MS Excel 409 Actual depth vs signal amplitude for ETSS 96041.1 8/18/2020 10:06 AM wear samples GinnaFanBarBobP1Noise.csv MS Excel 4637 Limiting eddy current noise distribution for Ginna 8/18/2020 10:41 AM fan bar intersections (SG 11, EOC39)

GinnaLatBarBobP1Noise.csv MS Excel 2475 Limiting eddy current noise distribution for Ginna 8/18/2020 10:18 AM lattice bar intersections (SG 21, EOC39)

CALCULATORV1 XL2010 check bwxt.xlsm MS Excel 746161 Flaw Handbook Calculator file used to verify 8/19/2020 4:40 PM structural limits from EOC39 DA CALCULATORV1 XL2010 Tube-to-tube wear.xlsm MS Excel 745893 Flaw Handbook Calculator file used to determine 8/19/2020 4:49 PM structural limit for tube-to-tube wear Ginna EOC39 Noise.xlsx MS Excel 39678358 EOC39 eddy current noise measurements for both 8/18/2020 8:58 AM Ginna SGs mapod_fan_bar.png PNG 266293 MAPOD results for Ginna fan bar wear 8/18/2020 10:44 AM mapod_Lattice_Bar.png PNG 259618 MAPOD results for Ginna lattice bar wear 8/18/2020 10:28 AM Page 19 of 21 Page 25 of 35

Document No.: 51-9317069-000 R. E. Ginna Steam Generator Operational Assessment Deferral of the Fall 2021 Steam Generator Inspections

6.0 REFERENCES

References identified with an (*) are maintained within R. E. Ginna Records System and are not retrievable from Framatome Records Management. These are acceptable references per Framatome Administrative Procedure 0402-01, Attachment 7. See page 2 for Project Manager Approval of customer references.

1. NEI 97-06, Steam Generator Program Guidelines, Rev. 3, March 2011

2. *BWXT Document 0192-AST-101038, R. E. Ginna Station EOC 39 SG Degradation Assessment, Revision 1, May 2017

3. *BWXT Document 0192-AST-101215, R. E. Ginna Station EOC 39 SG CMOA, Revision 0, May 2017

4. *BWXT Document S000112-TECR-000003, R. E. Ginna Station EOC 37 SG CMOA, Revision 0, May 2014

5. *Exelon Letter, NRC Docket No. 50-244, 2017 SG Tube Inspection Report, September 26, 2017 (ML17276A348)

6. EPRI Report 3002007571, SG Integrity Assessment Guidelines, Revision 4, June 2016

7. *Exelon Procedure ER-AP-420, SG Management Program, Revision 18

8. *Exelon Procedure ER-AP-420-0051, "Conduct of SG Management Program Activities,"

Revision 24

9. *Exelon Technical Evaluation ATI#04330850-01, G1R40 Degradation and Operational Assessment Review for G1R42, April 2020

10. EPRI Report 3002003048, Steam Generator Management Program: Flaw Handbook Calculator (SGFHC) for Excel 2010 v1.0, June 2014

11. *BWXT Document 322Q-LR-01, Ginna SG A and B (G1R40 inspection) Assessment of Secondary Separator Baseplate Degradation, Revision 0, May 2017

12. EPRI Report 3002010334, Model Assisted Probability of Detection Using R (MAPOD-R),

Version 2.1, September 2017

13. *BWXT Document 321U-B001 R.E. Ginna Replacement Steam Generators Maximum Allowable Degradation of Secondary Separator Support Plate February 22, 2017.

Page 20 of 21 Page 26 of 35

Document No.: 51-9317069-000 R. E. Ginna Steam Generator Operational Assessment Deferral of the Fall 2021 Steam Generator Inspections APPENDIX A: ABBREVIATIONS AND ACRONYMS AILPC Accident Induced Leakage Performance Criteria MSLB Main Steam Line Break BOC Beginning Of Cycle NOPD Normal Operating Pressure Differential CM Condition Monitoring Assessment OA Operational Assessment CMOA Condition Monitoring and Operational Assessment OD Outer Diameter DA Degradation Assessment POD Probability Of Detection ECT Eddy Current Test SG Steam Generator EFPY Effective Full Power Years SIPC Structural Integrity Performance Criteria EOC End Of Cycle SSI Secondary Side Inspection ETSS Examination Technique Specification Sheet TSH Tube Sheet Hot FAC Flow Assisted Corrosion TSP Tube Support Plate FOSAR Foreign Object Search And Retrieval TTS Top of Tubesheet GPD Gallons Per Day UBD Upper Bound Depth ID Inner Diameter Page 21 of 21 Page 27 of 35

ATTACHMENT 2 Framatome Document 51-9317069-000, Revision 0 Owners Acceptance Review Page 28 of 35

ATTACHMENT 3 Owners Acceptance Review Checklist for External Technical Evaluations Page 1 of 2 Technical Evaluation No.: 51 - 9317069 - 000 __________ EC:

Subject:

R. E. Ginna Steam Generator Operational Assessment Deferral of the Fall 2021 Steam Generator Inspections (September 2020) _______ FCMS:

ECDE: Framatome, Inc _____________________________ ATI: 04369718-02 Contract #: _612756____________________ Release #: _00008_

Yes No N/A Is the Technical Evaluation prepared in accordance with Procedure CC-AA-1.

309-101?

If the subject of the Technical Evaluation is Safety-related, ASME code-

2. related, or Augmented Quality, did the external organization perform an Independent Review and an Engineering Manager approval?

Does the Technical Evaluation clearly state the reason for the evaluation

3. (plant condition outside of expected range, non-conforming condition, response to a technical question or technical interpretation)?

Does the Technical Evaluation state what acceptance criteria or controlled 4.

range has been exceeded that requires this evaluation?

Does the Technical Evaluation state why the condition is being evaluated 5.

rather than fixed?

Does the Technical Evaluation verify conformance with applicable design and 6.

configuration control requirements?

Verify that the Technical Evaluation does not include any of the following:

• Process computer change

• Item equivalency required to be evaluated by SM-AA-300

7.

• A configuration change that should be processed per CC-AA-103-100, CC-AA-104, or CC-AA-112

• Hardware performance outside of established design envelop

• Operability Determination that should be evaluated by OP-AA-108-115 Does the Technical Evaluation disposition all of the required items that were 8.

established in the HU-AA-1212 Pre Job Brief?

Does the Technical Evaluation confirm the correctness of inputs, mathematics

9. and reasonableness of the output?

Does the Technical Evaluation verify the appropriateness of method, assumptions, accuracy, completeness, and compliance with design bases,

10. design criteria, codes and standards as well as licensing commitments and quality requirements?

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ATTACHMENT 3 Owners Acceptance Review Checklist for External Technical Evaluations Page 2 of 2 Technical Evaluation No.: 51 - 9317069 - 000 Yes No N/A

11. Are Engineering Judgments clearly documented and justified?

Does the Technical Evaluation state whether there are other components or 12.

systems to which the condition applies?

Does the Technical Evaluation verify applicable design considerations and

13. impacts per CC-AA-102, Design Input and Configuration Change Impact Screening have been identified and addressed?

Are the existing margins adequate to bound the combination of uncertainties 14.

associated with both the evaluation methodology and the actual evaluation?

Does the Technical Evaluation verify that the results accomplish the stated 15.

purpose?

Is it apparent that the Technical Evaluation does not contain any incorrect, 16.

false or misleading statements?

If there are recommendations or suggestions for follow-up actions, have they

17. been resolved or have Action Tracking assignments been issued to track such resolution?

Are the recommendations, results, and conclusions reasonable compared to 18.

the inputs and assumptions?

Do the conclusions and findings restate the key limitations, conditions and 19.

prerequisites?

Do the conclusions and findings describe the significant risks or highly

20. sensitive inputs that may have significant effects on the conclusion with minor changes in inputs?

List the critical characteristics of the product, and validate those critical characteristics.

Critical Characteristics: Requirement that the probability of a SG tube burst at 3X delta P is less than the limit of 0.05 and the probability of leakage at postulated worst case accident conditions exceeding the applied AILPC limit of 250 gpd/SG is less than the limit of 0.05.

Validation: For all degradation mechanisms evaluated in the Operational Assessment, the probability of burst at 3X delta P is less than the limit of 0.05 and the probability of leakage at worst case accident conditions exceeding the applied AILPC limit of 250 gpd is less than the limit of 0.05. During SQR, it was

21. identified that an Anticipated Transient Without Scram (ATWS) event may create a primary-to-secondary pressure differential on the order of or slightly exceeding that assumed due to a Main Steam Line Break (MSLB) value used in the OA. Subsequently, Framatome has assessed the impact of a limiting accident differential pressure of 2575 psid, which bounds both ATWS and MSLB events and determined that it has no impact on the results or conclusions to the Ginna 4-cycle Operational Assessment (see attached Framatome email). Therefore, extending the inspection interval to four-cycles will satisfy the tube integrity requirements of NEI 97-06 and justifies deferring the G1R43 SG inspections to G1R44 in April 2023.

Create an SFMS entry as required by CC-AA-4008. SFMS Number: N/A Page 30 of 35

September 18, 2020 FRAMATOME-20-01981 Peter Logar Exelon Corporation Ginna Nuclear Power Plant 1503 Lake Road Ontario, NY 14519

Subject:

Transmittal of Framatome Document 51-9317069-000, R. E. Ginna Steam Generator Operational Assessment Deferral of the Fall 2021 Steam Generator Inspections

References:

1. Exelon PO # 00612756 Release 08 dated August 12, 2020

2. E-mail from Carl L. Friant (Exelon) to Kent Colgan (Framatome)

Subject:

RE:

Ginna Deferral OA, dated Thursday, September 17, 2020 5:19 PM (enclosed)

Mr. Logar, The subject document is being submitted to you today as the final deliverable for the Reference 1 contract.

In the enclosed Reference 2 correspondence a change was identified in the limiting accident condition from a postulated Main Steam Line Break (MSLB) to an Anticipated Transient Without Scram (ATWS) resulting in an increase in the primary-to-secondary pressure differential of 2560 psid to 2575 psid and the question was asked if there was an impact to the Operational Assessment.

As discussed in Section 4.2 of the OA, for the volumetric degradation that is expected to occur in the Ginna SGs, and modeled in the OA, the onset of pop-through leakage and burst are coincident. As the degradation evaluated satisfied the structural integrity performance criteria at a pressure differential much greater than the limiting accident pressure differential, there is no impact to the OA from this increase to 2575 psid.

Additionally, all flaws considered in the OA are expected to be axial with limited circumferential extent, and therefore loads from any accident condition would have no impact on the OA results.

Should you have any questions please feel free to contact me at 434-832-3654.

Sincerely, Michael Epling, PMP Principal Project Manager Installed Base - North America 22709_MR-VA-3 (01/12/2018)

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September 18, 2020 FRAMATOME-19-01981 c: Carl (Lee) Friant (Exelon)

Patrick Creegan (Exelon)

George Wrobel (Exelon)

Heather Malikowski (Exelon)

Michael Bodack (Exelon)

Damon Peters (Exelon)

Chris Donavan (Exelon)

Wayne Belden (Framatome)

Kent Colgan (Framatome)

Craig Kelly (Framatome)

T1.2 - F.505410

Attachment:

Reference 2 email 22709_MR-VA-3 (01/12/2018)

Page 32 of 35

EPLING Mike (FRA-IB)

From: Friant, Carl L:(Exelon Nuclear) <carl.friant@exeloncorp.com>

Sent: Thursday, September 17, 2020 5:19 PM To: COLGAN Kent (FRA-IB)

Cc: Creegan, Patrick M:(Exelon Nuclear); Wrobel, George:(Exelon Nuclear); Malikowski, Heather M:(Exelon Nuclear); Logar Jr, Peter D:(Exelon Nuclear); BELDEN Wayne (FRA-IB); MIX Michael (FRA-IB); EPLING Mike (FRA-IB); Bodak, Michael J:(Exelon Nuclear);

Creegan, Patrick M:(Exelon Nuclear); Peters, Damon J:(Exelon Nuclear)

Subject:

RE: Ginna Deferral OA Security Notice: Please be aware that this email was sent by an external sender.

Kent, During todays PORC review of our Ginna SG Deferral LAR, it was discovered that Ginnas Limiting Accident primaryto secondary differential pressure is slightly higher than that assumed in the OA you prepared for the skip. The postulated accident event is called an ATWS (Anticipated Transient Without Scram) versus the previously assumed limiting MSLB (Main Steam Line Break) event. The ATWS event max differential pressure is 2575 psid versus the assumed value used in your OA and Flaw Handbook Calculations of 2560 psid (see Table 21 of the Framatome OA), or 15 psid higher. Based on your understanding of the Integrity Assessment guidelines, does this higher psid mean this OA needs to be revised? Or can we (with your help) technically justify that the slight pressure increase will not have an effect on passing the leakage criteria? Also, I believe the loads on the tube during an MSLB event are factored into the leakage model whereas the tube loads for this ATWS event are unknown at this time.

We can set up a call to discuss further. We are interested in resolving this quickly so our LAR submittal is not delayed and before Framatome staff becomes unavailable due to fall outage commitments.

Thanks, Lee Friant 6673131394 From: COLGAN Kent (Framatome) <Kent.Colgan@framatome.com>

Sent: Friday, September 4, 2020 10:46 AM To: Bodak, Michael J:(Exelon Nuclear) <Michael.Bodak@exeloncorp.com>

Cc: Friant, Carl L:(Exelon Nuclear) <carl.friant@exeloncorp.com>; Creegan, Patrick M:(Exelon Nuclear)

<patrick.creegan@exeloncorp.com>; Wrobel, George:(Exelon Nuclear) <george.wrobel@exeloncorp.com>; Malikowski, Heather M:(Exelon Nuclear) <Heather.Malikowski@exeloncorp.com>; Logar Jr, Peter D:(Exelon Nuclear)

<Peter.LogarJr@exeloncorp.com>; BELDEN Wayne (Framatome) <Wayne.Belden@framatome.com>; MIX Michael (Framatome) <michael.mix@framatome.com>; EPLING Mike (Framatome) <Michael.Epling@framatome.com>

Subject:

RE: Ginna Deferral OA

Mike, See attached, signed OA.

Thanks for your business!

Kent Kent Colgan Advisory Engineer 1

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Steam Generator Services Office: (434) 832-2925 Cell: (434) 534-5160 Email: kent.colgan@framatome.com From: Bodak, Michael J:(Exelon Nuclear) [1]

Sent: Thursday, September 03, 2020 3:15 PM To: COLGAN Kent (FRA-IB)

Cc: Friant, Carl L:(Exelon Nuclear); Creegan, Patrick M:(Exelon Nuclear); Wrobel, George:(Exelon Nuclear); Malikowski, Heather M:(Exelon Nuclear); Logar Jr, Peter D:(Exelon Nuclear)

Subject:

RE: Ginna Deferral OA Security Notice: Please be aware that this email was sent by an external sender.

Kent, Everything looks good. Go ahead and send the final copy!

Thanks for all your work on this.

Mike From: COLGAN Kent (Framatome) <Kent.Colgan@framatome.com>

Sent: Wednesday, September 2, 2020 3:46 PM To: Bodak, Michael J:(Exelon Nuclear) <Michael.Bodak@exeloncorp.com>

Subject:

[EXTERNAL] Ginna Deferral OA EXTERNAL MAIL. Do not click links or open attachments from unknown senders or unexpected Email.

Mike, Were a little ahead of schedule so I thought you might like another look at this before we release it.

Ill plan to release it on Friday unless you tell me otherwise.

Ive attached two versions: one with all changes tracked and comments intact, and one with all changes accepted and comments deleted.

Kent This Email message and any attachment may contain information that is proprietary, legally privileged, confidential and/or subject to copyright belonging to Exelon Corporation or its affiliates ("Exelon"). This Email is intended solely for the use of the person(s) to which it is addressed. If you are not an intended recipient, or the 2

Page 34 of 35

employee or agent responsible for delivery of this Email to the intended recipient(s), you are hereby notified that any dissemination, distribution or copying of this Email is strictly prohibited. If you have received this message in error, please immediately notify the sender and permanently delete this Email and any copies.

Exelon policies expressly prohibit employees from making defamatory or offensive statements and infringing any copyright or any other legal right by Email communication. Exelon will not accept any liability in respect of such communications. -EXCIP 3

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