Text

Exigent License Amendment Request (LAR)20-003 Revision to Technical Specification (TS) 5.5.9, Unit 1 Model D76 and Unit 2 Model D5 Steam Generator Program
	ML20101M879
Person / Time
Site:	Comanche Peak
Issue date:	04/10/2020
From:	Thomas McCool; Vistra Energy
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	CP-20200260, TXX-20025
	Download: ML20101M879 (89)
	v • d • e

Thomas P. McCool Site Vice President Luminant P.O. Box 1002 6322 North FM 56 Glen Rose, TX 76043 o 254.897.6042 CP-20200260 TXX-20025 U. S. Nuclear Regulatory Commission Ref 10 CFR 50.90 ATTN: Document Control Desk 10 CFR 50.91(b)(1)

Washington, DC 20555-0001 04/10/2020

SUBJECT:

COMANCHE PEAK NUCLEAR POWER PLANT DOCKET NOS. 50-445 AND 50-446 EXIGENT LICENSE AMENDMENT REQUEST (LAR)20-003 REVISION TO TECHNICAL SPECIFICATION (TS) 5.5.9, UNIT 1 MODEL D76 AND UNIT 2 MODEL D5 STEAM GENERATOR (SG) PROGRAM

Dear Sir or Madam:

Pursuant to 10CFR50.90 and 10 CFR 50.91, Vistra Operations Company LLC (Vistra OpCo) hereby requests an exigent amendment to the Comanche Peak Nuclear Power Plant (CPNPP) Unit 1 Operating License (NPF-87) and CPNPP Unit 2 Operating License (NPF-89) by incorporating the attached change into the CPNPP Units 1 and 2 Technical Specifications. The one-time exigent license amendment is required due to unforeseen issues as a result of the current COVID-19 virus pandemic.

The proposed change will revise Technical Specification 5.5.9 entitled UNIT 1 MODEL D76 AND UNIT 2 MODEL D5 STEAM GENERATOR (SG) PROGRAM. The proposed change will implement a one-time change to Unit 2 steam generator inspection frequency. This one-time exigent license amendment is required to minimize CPNPP Unit 2 spring 2020 refueling outage scope to limit personnel exposure to the COVID-19 virus pandemic.

The Enclosure provides a description and assessment of the proposed changes. Attachment 1 provides the existing TS pages and marked TS pages to show the proposed changes. No change is proposed to the current TS Bases as a result of this license amendment request. Attachment 2 provides an Operational Assessment describing the impact of deferring the Unit 2 steam generator inspections for spring 2020 refueling outage to the fall 2021 refueling outage.

Vistra OpCo has determined that the proposed change does not involve a significant hazards consideration pursuant to 10 CFR 50.92(c), and there are no significant environmental impacts associated with the change. The CPNPP Station Operations Review Committee has reviewed the proposed license amendment. In accordance with 10 CFR 50.91(b)(1), a copy of the proposed license amendment is being forwarded to the State of Texas.

6555 SIERRA DRIVE IRVING, TEXAS 75039 o214-812-4600 VISTRAENERGY.COM

TXX-20025 Page 2 of 2 NRC staff review and approval of the proposed amendment is requested by April 19, 2020, prior to CPNPP Unit 2 entering MODE 4. Once approved, the amendment shall be implemented immediately.

This letter contains no new regulatory commitments.

If you have any questions regarding this submittal, please contact James E. Barnette at (254} 897-5866 or james.barnette@luminant.com.

I state under penalty of perjury that the foregoing is true and correct.

Executed on April 10, 2020.

Sincerely, Thomas 11:

Enclosure LICENSE AMENDMENT REQUEST (LAR} 20-003, REVISION TO TECHNICAL SPECIFICATION (TS}

5.5.9, "UNIT 1 MODEL D76 AND UNIT 2 MODEL D5 STEAM GENERATOR (SG} PROGRAM" . CPNPP TECHNICAL SPECIFICATION 5.5.9, UNIT 1 MODEL D76 AND UNIT 2 MODEL D5 STEAM GENERATOR (SG} PROGRAM TECHNICAL SPECIFICATION PAGES

2. COMANCHE PEAK UNIT 2 STEAM GENERATOR OPERATIONAL ASSESSMENT TO SUPPORT DEFERRAL OF PLANNED INSPECTIONS FROM 2RF18 TO 2RF19 c- Scott Morris, Region IV [Scott.Morris@nrc.gov]

Dennis Galvin, NRR [Dennis.Galvin@nrc.gov]

John Ellegood, Senior Resident Inspector, CPNPP [John.Ellegood@nrc.gov]

Neil Day, Resident Inspector, CPNPP [Neil.Day@nrc.gov]

Mr. Robert Free [robert.free@dshs.state.tx.us]

Environmental Monitoring & Emergency Response Manager Texas Department of State Health Services Mail Code 1986 P. 0. Box 149347 Austin TX, 78714-9347

Enclosure to TXX-20025 Page 1 of 18 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1 System Design and Operation 2.2 Current Technical Specification Requirements 2.3 Reason for Proposed Change 2.4 Description of Proposed Change

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

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

5.0 ENVIRONMENTAL CONSIDERATION

S

6.0 REFERENCES

ATTACHMENTS Attachment 1 Technical Specification Pages Attachment 2 Comanche Peak Unit 2 Steam Generator Operational Assessment to Support Deferral of Planned Inspections from 2RF18 to 2RF19

Enclosure to TXX-20025 Page 2 of 18 1.0

SUMMARY

DESCRIPTION Proposed License Amendment Request (LAR)20-003 is to revise Technical Specification 5.5.9, Unit 1 Model D76 and Unit 2 Model D5 Steam Generator (SG) Program for Comanche Peak Nuclear Power Plant (CPNPP) Unit 2.

Vistra OpCo is requesting this change to defer the 100% inspection scope currently required for the third sequential inspection period from 2RF18 (April 2020) to 2RF19 (October 2021). This change is requested to minimize COVID-19 virus exposure at CPNPP during 2RF18.

No changes to the Comanche Peak Nuclear Power Plant Final Safety Analysis Report are anticipated as a result of this License Amendment Request.

2.0 DETAILED DESCRIPTION 2.1 System Design and Operation Steam Generator (SG) tubes in pressurized water reactors have a number of important safety functions. SG tubes are an integral part of the reactor coolant pressure boundary (RCPB) and, as such, are relied on to maintain reactor coolant system pressure and inventory. As part of the RCPB, SG tubes are unique as they transfer heat from the primary (Reactor Coolant System (RCS)) to the secondary (Main Feedwater (MFW)). In addition, SG tubes isolate radioactive fission products in the RCS from MFW.

The steam generator tube rupture (SGTR) accident is the limiting design basis event for a SG. The analysis of a SGTR event assumes a bounding primary to secondary leakage rate equal to the TS operational leakage rate limit, plus the leakage rate from a double-ended rupture of a single tube. The analysis for design basis accidents and transients other than a SGTR assume the SG tubes retain their structural integrity (i.e., they are assumed not to rupture). In these analyses, the steam discharge to the atmosphere is based on the total primary to secondary leakage from all SGs or is assumed to increase to the limit as a result of accident induced conditions. For accidents that do not involve fuel damage, the reactor coolant activity level is assumed to be equal to TS limits. For accidents that assume fuel damage, reactor coolant activity is a function of the amount of activity released from damaged fuel.

Steam generator tube integrity is necessary to ensure SG tubes are capable of performing their intended safety functions. Concerns relating to integrity of SG tubes stem from the fact that SG tubes are subject to a variety of degradation mechanisms. Steam generator tubes have experienced tube degradation related to corrosion phenomena, such as wastage, pitting, intergranular attack, and stress corrosion cracking, along with other mechanically induced phenomena such as wear. These degradation mechanisms can impair tube integrity if they are not managed effectively. When degradation of the tube wall reaches a prescribed criterion for action, the tube is considered defective and corrective action is taken, such as plugging or repair. Note not all plants have approved repair techniques. Therefore, references to "repair" are bracketed in the Standard Technical Specifications.

Enclosure to TXX-20025 Page 3 of 18 The industry, working through the Electric Power Research Institute (EPRI)

Steam Generator Management Program (SGMP), has implemented a generic approach to managing SG performance referred to as "Steam Generator Degradation Specific Management" (SGDSM).

The overall program is described in NEI 97-06, "Steam Generator Guidelines, which is supported by a number of EPRI guidelines, such as:

PWR Steam Generator Examination Guidelines

Steam Generator Integrity Assessment Guidelines

Steam Generator In-Situ Pressure Test Guidelines

PWR Primary-to-Secondary Leak Guidelines

PWR Primary Water Chemistry Guidelines

PWR Secondary Water Chemistry Guidelines NEI 97-06 and the EPRI Guidelines define a comprehensive, performance based approach to managing SG performance.

2.2 Current Technical Specification Requirements The current Technical Specification proposed to be changed is TS 5.5.9, Steam Generator (SG) Program.

TS 5.5.9.d.2 currently states, For the Unit 2 model D5 steam generators (Alloy 600 thermally treated) after the first refueling outage following SG installation, inspect each SG at least every 48 effective full power months or at least every other refueling outage (whichever results in more frequent inspections).

TS 5.5.9.d.2.c currently states, During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the third and subsequent inspection periods.

See Attachment 1 to TXX-20025 for TS 5.5.9 - Current Technical Specification.

2.3 Reason for Proposed Change Operating experience supports a longer inspection period for plants with 600TT tubes. In addition, the current inspection practices permit a simplified inspection routine, with fixed periods for 100% inspection of the SG tubes. Delaying performance of the current inspection regime for one refueling outage due to COVID-19 virus concerns is warranted when the potential for infection is weighed against the historical performance of the 600TT steam generators on CPNPP Unit 2.

Vistra OpCo initiated dialog with the NRC on a technical specification change process in March 2020 and is making a good faith effort to submit this license amendment request in a timely manner following the escalation of federal and state guidelines for COVID-19 virus response. Vistra OpCo has communicated with the NRC Staff regarding this request. Accordingly, Vistra OpCo requests this amendment be processed under exigent circumstances pursuant to 10 CFR 50.91(a)(6) to avoid potential unnecessary personnel exposure to COVID-19 virus which could also expose the general public to COVID-19 virus.

Enclosure to TXX-20025 Page 4 of 18 Consistent with the Statement of Considerations accompanying issuance of 10 CFR 50.91(a)(6), the circumstances here result in a net increase in safety or reliability (51 Federal Register 7744, 7756 (Mar. 6, 1986). Staff implementing guidance expresses a preference for a license amendment, if possible over NOED, where possible. See Inspection Manual Chapter 410, Section 6.03. In this case, the proposed balances opportunity to improve public safety and reliability with public participation in the NRCs technical specification change process.

2.4 Description of Proposed Change The proposed change will revise TS 5.5.9:

TS 5.5.9.d.2 will state, This is a one-time change to TS 5.5.9.d.2 for CPNPP Unit 2 Cycle 19, for the Unit 2 model D5 steam generators (Alloy 600 thermally treated) after the first refueling outage following SG installation, inspect each SG at least every 54 effective full power months or at least every other refueling outage (whichever results in more frequent inspections).

TS 5.5.9.d.2.c will state, This is a one-time change to TS 5.5.9.d.2.c for CPNPP Unit 2 Cycle 19, during the remaining life of the SGs, inspect 100% of the tubes every 90 effective full power months. This constitutes the third and subsequent inspection periods.

See Attachment 1 to TXX-20025 for TS 5.5.9 - Markup Technical Specification.

3.0 TECHNICAL EVALUATION

3.1 General Evaluation of Proposed Change The proposed change provides for compliance with federal recommendations regarding exposure to COVID-19 virus as low as possible while still conducting inspection and maintenance that will ensure the health and safety of the public.

Power generated by CPNPP Unit 2 is a vital national and state asset during the current pandemic.

The proposed change adds verbiage addressing the one-time deferral of CPNPP Unit 2 steam generator inspections. Based on Unit 2 600TT steam generator historical performance and current condition there is reasonable expectation that deferring the currently called for inspection for one refueling cycle will not impact public health or safety.

The current TS requirement is to inspect 100% of the tubes during each sequential period of 120 EFPM, 96 EFPM, and 72 EFPM, and each SG must be inspected at least every 48 EFPM or at least every other refueling outage (whichever results in more frequent inspections).

Significant operating experience has been gained over the course of 27 years of CPNPP Unit 2 plant operation with the 600TT steam generators and provides justification for this one-time inspection deferral. CPNPP Unit 2 operating experience has shown no propensity for rapidly increasing crack initiation rates in Alloy 600TT SG tubes.

Enclosure to TXX-20025 Page 5 of 18 The proposed one-time change that defers steam generator tube inspections for one refueling cycle can be demonstrated to meet the structural integrity and accident induced leakage performance criteria even for SGs that have experienced cracking. Based on this operating experience, the proposed TS change to defer tube inspections for one refueling cycle is acceptable.

3.2 CPNPP Unit 2 Specific Evaluation of Proposed Change Recent operational experience is summarized as follows:

1. Primary to secondary leakage The Operational Assessment performed following 2RF17 stated, Structural and leakage performance criteria were satisfied based on the inspection results. Hence, the CM requirements are satisfied. There was no primary-to-secondary operating leakage during the preceding inspection interval. An operational assessment of the current inspection results shows that the structural and leakage integrity will be satisfied for the next cycle and that it is acceptable to continue operation until the next planned inspection in spring 2020 (2RF18) for degradation mechanisms detected during 2RF17 and for existing degradation mechanisms previously detected at CPNPP Unit 2. There was no primary-to-secondary leakage observed during Cycle 17.

2. Summarize the most recent primary and secondary inspections, degradation description and location The most recent full scope SG inspection occurred during 2RF16. A smaller inspection scope was performed during 2RF17 to address circumferential cracking mechanisms found during 2RF16. The inspection scope and degradation mechanisms found during each inspection are as follows:

2RF16 SG Inspection Scope (performed in All SGs unless otherwise stated)

50% full-length bobbin o Including all previously existing flaws

100% full-length bobbin of all high stress tubes o +Point of all hot leg and cold TSP and TTS expansion transitions in all high stress tubes

100% hot leg TTS +POINT (+3 to H* depth) in SG 2-03

50% hot leg TTS +POINT (+3 to H* depth) in SG 2-01, SG 2-02, and SG 2-04

100% hot bulges/overexpansion inside tubesheet H* region

50% +Point of Row 1 and Row 2 U-bends

50% +Point of dents and dings greater than 5 volts o Dents/dings less than 5volts inspected with bobbin coil as part of bobbin program o 50% dents/dings 2-5volts at or below TSP 07C with +Point on cold leg side

100% TTS hot leg and cold leg periphery tubes with +Point

Enclosure to TXX-20025 Page 6 of 18

100% +Point of cold leg Rows 21 to 49 at PBP-B in SG 2-02 and SG 2-03

50% +Point probe of tubes expanded into Preheater Baffle Plate D

Channel Head visual inspections in all SGs (hot and cold legs)

Hot leg and cold leg tube plug visual inspections

Sludge Lance and FOSAR in all four SGs

Visual inspection of all waterboxes

Visual inspection of PBP Plate B in Rows 21-49 in SG 2-01 and SG 2-04

Upper bundle visual inspections in SG 2-01 During the 2RF16 inspections the following degradation mechanisms were reported

AVB Wear

TSP Wear

Preheater baffle plate wear

Foreign object wear above tubesheet

Circumferential PWSCC at bulges/overexpansions within hot leg tubesheet (3 indications in 2 tubes)

Circumferential PWSCC at hot leg tubesheet expansion transition (1 indication in 1 tube)

One primary channel head cladding anomaly in SG 2-01 cold leg (historical)

One primary tubesheet cladding anomaly in SG 2-02 hot leg 2RF17 SG Inspection Scope (performed in All SGs unless otherwise stated)

100% hot leg TTS +POINT (+3 to H* depth) in SG 2-03

50% hot leg TTS +POINT (+3 to H* depth) in SG 2-01, SG 2-02 and SG2-04

100% hot bulges/overexpansion inside tubesheet H* region in SG 2-03

50% hot bulges/overexpansion inside tubesheet H* region in SG 2-01, SG 2-02 and SG2-04

Channel Head visual inspections

Hot leg and cold leg tube plug visual inspections

Visual inspection of Steam Drum Upper Internals in all 4 SGs During the 2RF17 inspections the following degradation mechanisms were reported

Foreign object wear above tubesheet

One primary channel head cladding anomaly in SG 2-02 cold leg (historical)

One primary tubesheet cladding anomaly in SG 2-02 hot leg Sludge lancing and foreign object search and retrieval (FOSAR) activities were not performed during 2RF17.

Enclosure to TXX-20025 Page 7 of 18

3. Number of tubes plugged including reason for plugging During 2RF16, SG 2-03 had three tubes plugged due to circumferential primary water stress corrosion cracking (PWSCC). No tubes were plugged during 2RF17 as a result of the base scope and special interest inspections. Currently the cumulative total number of plugged SG tubes on CPNPP for Unit 2 in all four SGs is 96 which is equal to 0.53% of the total number of tubes.

4. Relevant operating experience that may have impacted SG tube integrity There have been no thermal or hydraulic events on CPNPP, Unit 2 that may have impacted SG tube integrity since 2RF16.

Previous inspection condition monitoring is as follows:

1. For each degradation mechanism detected compare the most limiting as found condition to tube performance criteria.

The following degradation mechanisms exist in CPNPP Unit 2 Steam Generators; Identified during 2RF16 inspection

a. Wear at anti-vibration bars (AVB) supports No tubes were plugged due to AVB wear prior to inspection. The maximum depth of all AVB wear was 39% through wall (TW). The Condition Monitoring (CM) wear limit is 67.1% TW.

For volumetric wear flaws with pressure-only loading condition, tube burst and ligament tearing (i.e., pop-through) are coincident, therefore, satisfaction of the tube burst criteria at 3PNO also satisfies the AILPC at SLB differential pressure. Therefore, the SG performance criteria for structural and leakage integrity were demonstrated.

b. Wear at Preheater Baffle Plates (PBP)

The maximum depth of all PBP wear was 7% TW. The CM wear limit for PBP is 51.2% TW.

For volumetric wear flaws with pressure-only loading condition, tube burst and ligament tearing (i.e., pop-through) are coincident, therefore, satisfaction of the tube burst criteria at 3PNO also satisfies the AILPC at SLB differential pressure. Therefore, the SG performance criteria for structural and leakage integrity were demonstrated.

c. Wear at Quatrefoil Tube Support Plates (TSP)

The maximum depth of all quatrefoil TSP wear was 16% TW. The CM wear limit for quatrefoil TSP is 51.2% TW.

For volumetric wear flaws with pressure-only loading condition, tube burst and ligament tearing (i.e., pop-through) are coincident, therefore, satisfaction of the tube burst criteria at 3PNO also satisfies the AILPC at SLB differential pressure. Therefore, the SG performance criteria for structural and leakage integrity were demonstrated.

Enclosure to TXX-20025 Page 8 of 18

d. Wear due to foreign objects The maximum depth of all new foreign objects wear indications was 14% TW. The CM wear limit for foreign objects is 55.8% TW. Prior to 2RF16 the largest flaw to remain in service was 28% TW. The CM wear limit for foreign objects with an axial flaw length of 0.5 inches using ETSS 21998.1 +POINT probe sizing technique is 55.8% TW.

For volumetric wear flaws with pressure-only loading condition, tube burst and ligament tearing (i.e., pop-through) are coincident, therefore, satisfaction of the tube burst criteria at 3PNO also satisfies the AILPC at SLB differential pressure. Therefore, the SG performance criteria for structural and leakage integrity were demonstrated.

e. Circumferential primary water stress corrosion cracking (PWSCC) at bulges/overexpansions with the hot leg tubesheet Three indications found in SG 2-03 within the hot leg tubesheet below the top of the tubesheet. The largest percent degraded area (PDA) was 4.2% and the CM limit was 74%.

f. Circumferential PWSCC at hot leg tubesheet expansion transitions One indication found in SG 2-03 within the top of the hot leg tubesheet expansion transition. As found PDA was 11.5% and the CM limit was 74%.

Identified during 2RF17 inspection

a. Wear due to foreign objects The maximum depth of all foreign objects wear indications was 17%

TW with axial extent of 0.32 inches. The CM wear limit for foreign objects with an axial flaw length of 0.5 inches is 49.4% using ETSS 96911.1 pancake coil sizing technique.

For volumetric wear flaws with pressure-only loading condition, tube burst and ligament tearing (i.e., pop-through) are coincident, therefore, satisfaction of the tube burst criteria at 3PNO also satisfies the AILPC at SLB differential pressure. Therefore, the SG structural and leakage performance criteria are satisfied for foreign object wear.

b. Circumferential PWSCC at bulges/overexpansions with the hot leg tubesheet and Circumferential PWSCC at hot leg tubesheet expansion transitions Because of the indications found in 2RF16 inspections for these degradation methods were required in 2RF17. No indication of circumferential PWSCC were reported, thus no CM for these degradation mechanisms was required.

Enclosure to TXX-20025 Page 9 of 18

2. For all SG tubes that required flaw profiling, demonstrate how condition monitoring was met.

Condition monitoring was met by not exceeding any through wall limits for the above degradation mechanisms. Flaw profiling was not required to demonstrate condition monitoring.

Operational assessment during an additional operating cycle is as follows:

List of terms used in this assessment; POB = Probability of Burst (%)

POL = Probability of Leakage (%)

Pb = Burst Pressure (psi)

LR = Leakage Rate (gpm)

TW = Through wall NDE = Non-destructive examination CM = Condition Monitoring PBP = Preheater Baffle Plates FDB = Flow Distribution Baffle TSP = Tube Support Plate POD = Probability of flaw Detection

1. List mechanisms considered and reason for consideration Existing Degradation Mechanisms

a. Projected Tube Wear at AVBs Existing 55.3% TW with burst pressure of 5075 psi. Criterion is less than 69.4% TW with burst pressure greater than or equal to 3909 psi.

Undetected 38.4% TW with burst pressure of 6438 psi. Criterion is less than 69.4% TW with burst pressure greater than or equal to 3909 psi.

Prediction of existing AVB Wear at 2RF19 starts with 39% TW is projected to be no worse than 38.4% TW with a burst pressure of 5075 psi. The bounding load limit is 69.4% TW with a burst pressure greater than or equal to 3909 psi.

Prediction of undetected AVB Wear at 2RF19 starts with 17% TW is projected to be no worse than 55.3% TW with a burst pressure of 6438 psi. The bounding load limit is 69.4% TW with a burst pressure greater than or equal to 3909 psi.

Deferring AVB Wear inspections until 2RF19 will not violate SG tube integrity performance criteria.

b. Projected Tube Wear at Tube Support Plates (PBP and TSP)

Existing 51.1% TW with burst pressure of 5063 psi. Criterion is less than 63.7% TW with burst pressure greater than or equal to 3909 psi.

Undetected 57.6% TW with burst pressure of 4465 psi. Criterion is less than 63.7% TW with burst pressure greater than or equal to 3909 psi.

Enclosure to TXX-20025 Page 10 of 18 Prediction of existing PBP Wear at 2RF19 starts with 7% TW is projected to be no worse than 42.6% TW with a burst pressure of 5938 psi. The bounding load limit is 65.9% TW with a burst pressure greater than or equal to 3909 psi.

Prediction of undetected PBP Wear at 2RF19 starts with 9% TW is projected to be no worse than 53.6% TW with a burst pressure of 4981 psi. The bounding load limit is 65.9% TW with a burst pressure greater than or equal to 3909 psi.

Deferring PBP Wear inspections until 2RF19 will not violate SG tube integrity performance criteria.

Prediction of existing TSP Wear at 2RF19 starts with 16% TW is projected to be no worse than 51.1% TW with a burst pressure of 5063 psi. The bounding load limit is 63.7% TW with a burst pressure greater than or equal to 3909 psi.

Prediction of undetected TSP Wear at 2RF19 starts with 13% TW is projected to be no worse than 57.6% TW with a burst pressure of 4465 psi. The bounding load limit is 63.7% TW with a burst pressure greater than or equal to 3909 psi.

Deferring TSP Wear inspections until 2RF19 will not violate SG tube integrity performance criteria.

c. Projected Tube Wear Due to Foreign Objects Existing 28% TW (NDE). Criterion is less than 55.8% TW (CM).

No foreign objects remain within the vicinity of the affected tubes and no flaw growth has occurred since initial identification. Therefore, there is no mechanism for future flaw growth.

Deferring Tube Wear from Foreign Objects inspections until 2RF19 will not violate SG tube integrity performance criteria.

d. Projected Circumferential PWSCC Indications Existing with POB of 1.67%, POL of 2.72%, Pb of 4604 psi, and LR of 0.146 gpm. Criterion POB less than or equal to 5%, POL less than or equal to 5%, Pb greater than or equal to 3909 psi, and LR less than or equal to 0.3472 gpm.

Deferring Projected Circumferential PWSCC inspections until 2RF19 will not violate SG tube integrity performance criteria.

Potential Degradation Mechanisms

a. Axial ODSCC at Tube Support Plates for High Stress Tubes Potential with POB of 0.617%, POL of 0.847%, Pb of 5291 psi, and LR of 0 gpm. Criterion POB less than or equal to 5%, POL less than or equal to 5%, Pb greater or equal to 3909 psi, and LR less than or equal to 0.3472 gpm.

Enclosure to TXX-20025 Page 11 of 18 Deferring Axial ODSCC at Tube Support Plates for High Stress Tubes inspections until 2RF19 will not violate SG tube integrity performance criteria.

b. Circumferential ODSCC at top of Tubesheet Expansion Transitions Potential with POB of 0.63%, POL of 2.14%, Pb of 5584 psi, and LR of 0.05 gpm. Criterion POB less than or equal to 5%, POL less than or equal to 5%, Pb greater or equal to 3909 psi, and LR less than or equal to 0.3472 gpm.

Deferring Circumferential ODSCC at top of Tubesheet Expansion Transitions inspections until 2RF19 will not violate SG tube integrity performance criteria.

c. Axial PWSCC at top of Tubesheet Expansion Transitions Potential with POB of 1.29%, POL of 0.64%, Pb of 4767 psi, and LR of 0 gpm. Criterion POB less than or equal to 5%, POL less than or equal to 5%, Pb greater or equal to 3909 psi, and LR less than or equal to 0.3472 gpm.

Deferring Axial PWSCC at top of Tubesheet Expansion Transitions inspections until 2RF19 will not violate SG tube integrity performance criteria.

d. Axial ODSCC at top of Tubesheet Expansion Transitions & Sludge Pile Potential with POB of 1.17%, POL of 0.42%, Pb of 4757 psi, and LR of 0.0014 gpm. Criterion POB less than or equal to 5%, POL less than or equal to 5%, Pb greater or equal to 3909 psi, and LR less than or equal to 0.3472 gpm.

Deferring Axial ODSCC at top of Tubesheet Expansion Transitions &

Sludge Pile inspections until 2RF19 will not violate SG tube integrity performance criteria.

Leakage Integrity The existing and potential stress corrosion cracking (SCC) mechanisms have been evaluated against condition monitoring criteria of less than 5%

POL. Following EPRI SG IAGL guidance the circumferential PWSCC and circumferential ODSCC POLs are combined for a total 4.76% which is bounded by the less than 5% CM limit.

The predicted primary to secondary leakage based on not performing SG inspections until 2RF19 is calculated to be no more than 68.3 gpd. This value is bounded by the TS primary to secondary leakage limit of 150 gpd and the administrative limit of 75 gpd (Action Level 2 - be in MODE 3 in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

Enclosure to TXX-20025 Page 12 of 18 Channel Head Indications Combined aerated and deaerated corrosion rate for CPNPP Unit 2 SGs is 0.00087 inch per cycle. Projected total wall loss through 2RF19 is 0.0166 inch. The allowable surface flaw depth is 0.10 inch. Predicted total wall loss of 0.0166 inch is bounded by allowable depth of 0.10 inch.

Foreign Object Evaluation CPNPP Unit 2 SG foreign object inventory with all four SGs will not reach minimum tube wall thickness at any location by deferring foreign object inspection to 2RF19. No retrieval, inspection, or plugging is necessary prior to 2RF19.

Secondary Side Integrity Steam drum inspection interval is every 12 years. The inspection was last performed during 2RF17 (fall or 2018). That inspection yielded no evidence of progressing degradation. No impact on deferring inspections to 2RF19.

2. Inspection strategy details at last inspection for each item considered above The inspection strategy is to perform a 100% tube inspection of all CPNPP Unit 2 SGs during 2RF19 in the fall of 2021.

3. CPNPP Unit 2 SG Alloy 600TT tubing;

a. Number of high stress tubes remaining in service In 2003, a total 73 high stress tubes were identified in CPNPP Unit 2 SGs. Nine of the 73 tubes have been plugged, resulting in 64 high stress tubes remaining in service.

b. How tubes that could have remained unidentified as high stress after screening are considered in inspection strategy Inspection of the 64 high stress SG tubes on CPNPP Unit 2 includes full-length bobbin coil inspections and targets +POINT probe inspection of areas more susceptible to degradation.

CPNPP Unit 2 SG tubes not identified with high residual stress are inspected in accordance with TS 5.5.9 requirements.

c. Operational experience related to the high stress tubes that remain in service Most recently, during 2RF16 in 2017 one high stress SG tube presented PWSCC indication and was plugged. No other SCC indications have presented on CPNPP Unit 2 SGs.

Enclosure to TXX-20025 Page 13 of 18 This operational assessment (Attachment 2) reviewed all existing degradation mechanisms and evaluated detected and undetected flaws. Potential degradation mechanisms were reviewed. The methodology assumes the degradations have occurred and are undetected until 2RF19. Overall, data supports deferral of CPNPP Unit 2 SG inspections until 2RF19.

Mitigating strategies are as follows:

1. Administrative limits versus Technical Specification (TS) limits TS 3.4.13, RCS Operational LEAKAGE is limited to 150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).

The current administrative limit is 75 gpd for Action Level 2 entry which requires MODE 3 entry within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Action Level 2 entry will be changed to 50 gpd.

2. Increased monitoring between operating cycles Increased monitoring can be correlated to TS 3.4.13, RCS Operational LEAKAGE, surveillance requirement (SR) 3.4.13.1, Verify RCS operational LEAKAGE is within limits by performance of RCS water inventory balance. The frequency requirement is once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in accordance with the Surveillance Frequency Control Program (SFCP). At CPNPP this SR is performed once per day during steady state operations.

This is not a new requirement it is provided to illustrate that CPNPP performs an RCS water inventory balance more frequently than what is required by the SFCP.

Increased monitoring can also be correlated to TS 3.4.13, RCS Operational LEAKAGE, surveillance requirement (SR) 3.4.13.2, Verify primary to secondary LEAKAGE is 150 gallons per day through any one SG. The frequency requirement is once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in accordance with the Surveillance Frequency Control Program (SFCP). At CPNPP this SR is performed once per day during steady state operations. This is not a new requirement it is provided to illustrate that CPNPP performs an RCS water inventory balance more frequently than what is required by the SFCP.

CPNPP has adopted the EPRI PWR Primary-To-Secondary Leak Guidelines provide two methodologies for responding to power operation primary-to-secondary leakage: Rate of Change, and Constant Leakage.

Rate of Change methodology evaluates the leakage and the rate of leakage change. Constant Leakage methodology only considers leakage. CPNPP has determined to use the Constant Leakage methodology for primary-to-secondary leakage response. The following actions are taken in response to increasing primary-to-secondary leakage.

Increased Monitoring: Mode 1 normal operation plant condition entered when leakage has been detected but is not in a range that can be accurately monitored by most online radiation monitors, does not necessarily indicate imminent risk to steam generator tube integrity, but warrants additional attention. This condition is entered when the total primary-to-secondary leakage is greater than or equal to 5 gpd.

Enclosure to TXX-20025 Page 14 of 18

Action Level 1: The plant condition entered when leakage has increased to a condition that requires frequent monitoring by the radiation monitoring system with periodic bench marking by laboratory analyses. Action Level 1 is entered when primary-to-secondary leakage is greater than or equal to 30 gpd.

Action Level 2: Plant condition entered when primary-to-secondary leakage is greater than or equal to 75 gpd in any steam generator AND is sustained for greater than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . Be in Mode 3 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of entering Action Level 2.

Action Level 3: Plant condition entered when primary-to-secondary leakage in any steam generator is greater than or equal to 100 gpd.

Commence prompt and controlled plant shut down and be at less than or equal to 50% power within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be in Mode 3 within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (total of 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months ).

3. Procedure changes that will be implemented between operating cycles

a. ABN-106, High Secondary Activity Procedure currently uses 75 gpd to determine whether to monitor and continue operation or shutdown. The procedure will be changed to use 50 gpd as the determination leak rate.

b. ODA-102, Conduct of Operations Procedure contains a table with a list of procedures that require directing a trip or down power. This table will be revised to use 50 gpd vice 75 gpd for unit down power.

c. COP-732, Primary to Secondary Leakage Procedure contains a note listing the leakage values to be used when developing a correlation graph using the Condenser Off Gas radiation monitor to determine primary to secondary leakage. One of the values is 75 gpd, this will be changed to 50 gpd.

d. CHM-113, Primary to Secondary Leakage Procedure requires entry into Action Level 2 when primary to secondary leakage is greater than or equal to 75 gpd in any steam generator and is sustained for greater than one hour. Action Level 2 requires MODE 3 entry within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The 75 gpd value will be changed to 50 gpd.

e. CHM-113-12, Primary to Secondary Leakage Data Sheet The data sheet will replace the 75 gpd value with 50 gpd when using tritium analysis method for primary to secondary leakage.

f. CHM-101-11, Chemistry Checklist MODE 5 to MODE 4 The mode change restraint for primary to secondary leakage is 75 gpd and will be changed to 50 gpd.

g. CHM-101-12, Chemistry Checklist MODE 4 to MODE 3 The mode change restraint for primary to secondary leakage is 75 gpd and will be changed to 50 gpd.

Enclosure to TXX-20025 Page 15 of 18

h. CHM-101-13, Chemistry Checklist MODE 3 to MODE 2 The mode change restraint for primary to secondary leakage is 75 gpd and will be changed to 50 gpd.

i. RPI-621, Response to Primary to Secondary Leakage Action Level 2 entry criteria will be changed from 75 gpd to 50 gpd.

See Attachment 2 to TXX-20025 for Comanche Peak Unit 2 Steam Generator Operational Assessment to Support Deferral of Planned Inspections from 2RF18 to 2RF19 Based on this operating data, the one-time TS change to defer the CPNPP Unit 2 SG tube inspections from 2RF18 to 2RF19 (for 18 months) is acceptable.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements The regulatory basis for Technical Specifications 5.5.9, ASteam Generator (SG)

Program@ is to ensure that SG tube integrity is maintained. In addition, the SG Program shall include the following:

Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include technical specifications as part of the license.

The Commission's regulatory requirements related to the content of the technical specifications are contained in Title 10, Code of Federal Regulations (10 CFR),

Section 50.36, "Technical Specifications," of 10 CFR Part 50 "Domestic Licensing of Production and Utilization Facilities." The technical specification requirements in 10 CFR 50.36 include the following categories: (1) safety limits, limiting safety systems settings and control settings, (2) limiting conditions for operation, (3) surveillance requirements, (4) design features, and (5) administrative controls.

As stated in 10 CFR 50.59(c)(1)(i), a licensee is required to submit a license amendment pursuant to 10 CFR 50.90 if a change to the technical specifications is required. Furthermore, the requirements of 10 CFR 50.59 necessitate that the NRC approve technical specification changes before the changes are implemented. Vistra OpCos submittal meets the requirements of 10 CFR 50.59(c)(1)(i) and 10 CFR 50.90.

General Design Criterion (GDC) 14 "Reactor Coolant Pressure Boundary (RCPB)," of Appendix A "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50 requires, among other things, that the reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture. The proposed change continues to provide steam generator tube inspections that will contribute to a robust RCPB.

GDC 15, AReactor Coolant System (RCS) 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

Enclosure to TXX-20025 Page 16 of 18 pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences. The proposed change does not negatively impact steam generator tube integrity during normal operation or anticipated operational occurrences.

GDC 30, AQuality of reactor coolant pressure boundary,@ requires that components which are part of the RCPB shall be designed, fabricated, erected, and tested to the highest quality standards practical. The proposed change does not reduce quality standards for steam generator design, fabrication, or testing.

GDC 31, Fracture prevention of reactor coolant pressure boundary, requires the RCPB 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 proposed change does not alter the fracture prevention design of the steam generator tubes.

GDC 32, Inspection of reactor coolant pressure boundary, the steam generator tubes are designed to permit periodic inspection and testing to assess their structural and leaktight integrity. The proposed change does not eliminate periodic inspection or testing of the steam generator tubes.

4.2 Precedent None cited 4.3 No Significant Hazards Consideration Determination Vistra OpCo has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, AIssuance of amendment,@ as discussed below:

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

Response: No The proposed change calls for a one-time change in inspection frequencies for steam generator tube inspections and associated reporting requirements. Inspection frequencies are not an initiator to a steam generator tube rupture accident, or any other accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased. The steam generator tubes inspected by the Steam Generator (SG) Program continue to be required to meet the SG Program performance criteria and to be capable of performing any functions assumed in the accident analysis. As a result, the consequences of any accident previously evaluated are not significantly increased.

Enclosure to TXX-20025 Page 17 of 18 Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No The proposed change calls for a one-time change in inspection frequencies for steam generator tube inspections and associated reporting requirements. The proposed change does not alter the design function or operation of the steam generators or the ability of a steam generator to perform the design function. The steam generator tubes continue to be required to meet the Steam Generator (SG)

Program performance criteria. The proposed change does not create the possibility of a new or different kind of accident due to credible new failure mechanisms, malfunctions, or accident initiators that not considered in the design and licensing bases.

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

3. Do the proposed changes involve a significant reduction in a margin of safety?

Response: No The proposed change calls for a one-time change in inspection frequencies for steam generator tube inspections and associated reporting requirements. The proposed change does not change any of the controlling values of parameters used to avoid exceeding regulatory or licensing limits. The proposed change does not affect a design basis or safety limit, or any controlling value for a parameter established in the FSAR or the license.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above evaluations, Vistra OpCo concludes that the propose amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c) and, accordingly, a finding of Ano 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 Commission=s regulations, and (3) the issuance of the amendment will not be adverse to the common defense and security or to the health and safety of the public.

Enclosure to TXX-20025 Page 18 of 18

5.0 ENVIRONMENTAL CONSIDERATION

S Vistra OpCo has determined that the proposed amendment would change requirements with respect to the 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 amount of effluent that may be released offsite, or (iii) a significant increase in the individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), an environmental assessment of the proposed change is not required.

6.0 REFERENCES

6.1 TSTF-449, Revision 4, "Steam Generator Tube Integrity," April 14, 2005, (ADAMS Accession No. ML051090200).

6.2 TSTF-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," March 1, 2011, (ADAMS Accession No. ML110610350).

6.3 TSTF-577, Revision 0a, DRAFT " Performance Based Frequencies for Steam Generator Tube Inspections," September 18, 2019, (ADAMS Accession No. ML19301A001).

6.4 UNIT 2 SIXTEENTH REFUELING OUTAGE (2RF16) STEAM GENERATOR 180 DAY REPORT (ML17313A447) 6.5 UNIT 2 SEVENTEENTH REFUELING OUTAGE (2RF17) STEAM GENERATOR 180 DAY REPORT (ML19171A190)

Attachment 1 to TXX-20025 Page 1 of 12 Attachment 1 to TXX-20025 Technical Specifications

1. Technical Specification 5.5.9, Unit 1 Model D76 and Unit 2 Model D5 Steam Generator (SG) Program - Current

2. Technical Specification 5.5.9, Unit 1 Model D76 and Unit 2 Model D5 Steam Generator (SG) Program - Markup

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