ML22294A092

Issuance of Amendment No. 273 Regarding Adoption of TSTF-577, Revised Frequencies for Steam Generator Tube Inspections, Revision 1

ML22294A092
Person / Time
Site:	Robinson
Issue date:	12/15/2022
From:	Luke Haeg Plant Licensing Branch II
To:	Flippin N Duke Energy Progress
Hood T
References
EPID L-2021- LLA-0222
Download: ML22294A092 (20)

Text

December 15, 2022

Ms. Nicole L. Flippin H. B. Robinson Steam Electric Plant Site Vice President Duke Energy Progress, LLC 3581 West Entrance Road, RNPA11 Hartsville, SC 29550

SUBJECT:

H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 - ISSUANCE OF AMENDMENT NO. 273 REGARDING ADOPTION OF TSTF-577, REVISED FREQUENCIES FOR STEAM GENERATOR TUBE INSPECTIONS, REVISION 1 (EPID L-2021-LLA-0222)

Dear Ms. Flippin:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 273 to Renewed Facility Operating License No. DPR-23 for the H. B. Robinson Steam Electric Plant, Unit No. 2 (Robinson). This amendment changes the Robinson Technical Specifications (TSs) in response to your application dated December 9, 2021, as supplemented by letters dated January 6, 2022, April 28, 2022, July 28, 2022, and October 4, 2022. The amendment modifies TSs based on Technical Spec ifications Task Force (TSTF) Improved Standard Technical Specifications Change Traveler TSTF-577, Revision 1, Revised Frequencies for Steam Generator Tube Inspections (TSTF-577), and the associated NRC staff safety evaluation of TSTF-577.

A copy of the related Safety Evaluation is enclosed. Notice of Issuance will be included in the Commissions monthly Federal Register notice.

Sincerely,

/RA/

Lucas Haeg, Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Docket No. 50-261

Enclosures:

1. Amendment No. 273 to DPR-23

2. Safety Evaluation

cc w/encls: Listserv DUKE ENERGY PROGRESS, LLC

DOCKET NO. 50-261

H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2

AMENDMENT TO RENEWED FA CILITY OPERATING LICENSE

Amendment No. 273 Renewed License No. DPR-23

1. The U.S. Nuclear Regulatory Commission (the Commission) has found th at:

A. The application for amendment by Duke Energy Progress, LLC (the licen see),

dated December 9, 2021, as supple mented by letters dated January 6, 2 022, April 28, 2022, July 28, 2022, and October 4, 2022, complies with the st andards and requirements of the Atomic Energy Act of 1 954, as amended (the Act), and the Commissions ru les and regulations set forth in 10 CFR Chapter I;

B. The facility will operate in conformity with the application, the provision s of the Act, and the rules and regulations of the Commission;

C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the healt h and safety of the public, and (ii) that such activities will be conducted in compliance with th e Commissions regulations;

D. The issuance of this amendment will not be inimical to the common defen se and security or to the health and safety of the public; and

E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

Enclosure 1

2. Accordingly, the license is amended by changes to the Technical Specif ications, as indicated in the attachment to this license amendment; and paragraph 3.B. of Renewed Facility Operating License No. DPR-23 is hereby amended to read, in pa rt, as follows:

B. Technical Specifications

The Technical Specifications contained in Appendix A, as revised through Amendment No. 273 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of its issuance and shall be implemented within 90 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

David J. Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reac tor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications

Date of Issuance: December 15, 2022 ATTACHMENT TO LICENSE AMENDMENT NO. 273

H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2

RENEWED FACILITY OPERATING LICENSE NO. DPR-23

DOCKET NO. 50-261

Replace the following page of Renewed Facility Operating License No. DPR-23 with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

Renewed Facility Operating License No. DPR-23

Remove Insert

Page 3 Page 3

Replace the following pages of the Appendix A, Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert

5.0-12 5.0-12 5.0-13 5.0-13 5.0-14 5.0-14 5.0-28 5.0-28 5.0-28a 5.0-28a

D. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form for sample analysis or instrument and equipment calibration or associated with radioactive apparatus or components;

E. Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by operation of the facility.

3. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Section 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

A. Maximum Power Level

The licensee is authorized to operate the facility at a steady state reactor core power level not in excess of 2339 megawatts thermal.

B. Technical Specifications

The Technical Specifications contained in Appendix A, as revised through Amendment No. 273 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

(1) For Surveillance Requirements (SRs) that are new in Amendment 176 to Final Operating License DPR-23, the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment 176. For SRs that existed prior to Amendment 176, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first performance is due at the end of the first surveillance interval that begins on the date the Surveillance was last performed prior to implementation of Amendment 176.

Renewed Facility Operating License No. DPR-23 Amendment No. 273 Programs and Manuals 5.5 5.5 Programs and Manuals (continued)

5.5.9 Steam Generator (SG) Program

An SG Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the SG Program shall include the following:

a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the as found condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The as found condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes.

Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.

b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.

1. Structural integrity performance criterion: All in-service SG tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down) and all anticipated transients included in the design specification, and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

2. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed 150 gallons per day per SG.

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

(continued)

HBRSEP 5.0-12 Amendment No. 273 Programs and Manuals 5.5 5.5 Programs and Manuals

5.5.9 Steam Generator (SG) Program (continued)

c. Provisions for SG tube plugging 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.

The following alternate tube plugging criteria shall be applied as an alternative to the preceding criteria:

Tubes with service-induced flaws located greater than 18.11 inches below the top of the tubesheet do not require plugging. Tubes with service-induced flaws located in the portion of the tube from the top of the tubesheet to 18.11 inches below the top of the tubesheet shall be plugged upon detection.

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 except for any portions of the tube that are exempt from inspection by alternate repair criteria, and that may satisfy the applicable tube plugging 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. A degradation assessment 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 installation.

2. After the first refueling outage following SG installation, inspect 100% of the tubes in each SG at least every 54 effective full power months, which defines the inspection period. If none of the SG tubes have ever experienced cracking other than in regions that are exempt from inspection by alternate repair criteria and the SG inspection was performed with enhanced probes, the inspection period may be extended to 72 effective full power months. Additionally, the inspection

(continued)

HBRSEP 5.0-13 Amendment No. 273 Programs and Manuals 5.5 5.5 Programs and Manuals

5.5.9 Steam Generator (SG) Program (continued)

period that began December 8, 2020 may be 72 effective full power months without prior performance of a SG inspection using enhanced probes. Enhanced probes have a capability to detect flaws of any type equivalent to or better than array probe technology. The enhanced probes shall be used from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet except any portions of the tube that are exempt from inspection by alternate repair criteria. If there are regions where enhanced probes cannot be used, the tube inspection techniques shall be capable of detecting all forms of existing and potential degradation in that region.

3. If crack indications are found in any portion of a SG tube excluding any region that is exempt from inspection by alternate repair criteria, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the crack indication shall be at the next refueling outage, but may be deferred to the following refueling outage if the 100% inspection of all SGs was performed with enhanced probes as described in paragraph d.2. 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.

5.5.10 Secondary Water Chemistry Program

This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation. The program shall include:

a. Identification of critical parameters, their sampling frequency, sampling points, and control band limits;

(continued)

HBRSEP 5.0-14 Amendment No. 273 Reporting Requirements 5.6

5.6 Reporting Requirements (continued)

5.6.7 Tendon Surveillance Report

a. Notification of a pending sample tendon test, along with detailed acceptance criteria, shall be submitted to the NRC at least two months prior to the actual test.

b. A report containing the sample tendon test evaluation shall be submitted to the NRC within six months of conducting the test.

5.6.8 Steam Generator Tube Inspection Report

A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.9, Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG;

b. The nondestructive examination techniques utilized for tubes with increased degradation susceptibility;

c. For each degradation mechanism found:

1. The nondestructive examination techniques utilized;

2. The location, orientation (if linear), measured size (if available), and voltage response for each indication. For tube wear at support structures less than 20 percent through-wall, only the total number of indications needs to be reported;

3. A description of the condition monitoring assessment and results, including the margin to the tube integrity performance criteria and comparison with the margin predicted to exist at the inspection by the previous forward-looking tube integrity assessment; and

4. The number of tubes plugged during the inspection outage.

d. An analysis summary of the tube integrity conditions predicted to exist at the next scheduled inspection (the forward-looking tube integrity assessment) relative to the applicable performance criteria, including the analysis methodology, inputs, and results;

e. The number and percentage of tubes plugged to date, and the effective plugging percentage in each SG;

f. The results of any SG secondary side inspections; (continued)

HBRSEP Unit No. 2 5.0-28 Amendment No. 273XXX Reporting Requirements 5.6

5.0 ADMINISTRATIVE CONTROLS

5.6 Reporting Requirements (continued)

5.6.8 Steam Generator Tube Inspection Report (continued)

g. The primary to secondary leakage rate observed in each SG (if it is not practical to assign the leakage to an individual SG, the entire primary to secondary leakage should be conservatively assumed to be from one SG) during the cycle preceding the inspection that is the subject of this report;

h. The calculated accident induced leakage rate from the portion of the tubes below 18.11 inches from the top of the tubesheet for the most limiting accident in the most limiting SG. In addition, if the calculated accident induced leakage rate from the most limiting accident is less than 1.87 times the maximum operational primary to secondary leakage rate, the report should describe how it was determined; and

i. The results of monitoring for tube axial displacement (slippage). If slippage is discovered, the implications of the discovery and corrective action shall be provided.

HBRSEP Unit No. 2 5.0-28a Amendment No. 273XXX SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 273 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-23 DUKE ENERGY PROGRESS, LLC H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261

Application (i.e., initial and supplements) Safety Evaluation Date December 9, 2021, Agencywide Documents December 15, 2022 Access and Management System (ADAMS) Principal Contributors to Safety Accession No. ML21343A047 Evaluation January 6, 2022 (ML22006A240) Paul Klein April 28, 2022 (ML22118A336) Clinton Ashley July 28, 2022 (ML22209A156)

October 4, 2022 (ML22277A399)

1.0 PROPOSED CHANGE

S

Duke Energy Progress, LLC (Duke Energy, the licensee) requested changes to the technical specifications (TSs) for H. B. Robinson Steam Electric Plant, Unit No. 2 (Robinson) by license amendment request (LAR). In its LAR, as supplemented, the licensee proposed changes that would revise the Steam Generator (SG) Program and the Steam Generator Tube Inspection Report TSs based on Technical Specifications Task Force (TSTF) Improved Standard Technical Specifications Change Traveler TSTF-577, Revision 1, Revised Frequencies for Steam Generator Tube Inspections (TSTF-577) (ML21060B434), and the associated U.S.

Nuclear Regulatory Commission (NRC, the Commi ssion) staff safety evaluation (SE) of TSTF-577 (ML21098A188).

The tubes within an SG function as an integral part of the reactor coolant pressure boundary and, in addition, isolate fission products in the primary coolant from the secondary coolant and the environment. SG tube integrity means that the tubes are capable of performing this safety function in accordance with the plant design and licensing basis.

The Robinson SGs have Alloy 600 therm ally treated (Alloy 600TT) tubes.

1.1 Proposed TS Changes to Adopt TSTF-577

In accordance with NRC staff-approved TSTF-577, the licensee proposed changes that would revise the Robinson TS 5.5.9, Steam Generat or (SG) Program, and TS 5.6.8, Steam

Enclosure 2

Generator Tube Inspection Report. Specifically, the licensee proposed the following changes to adopt TSTF-577:

TS 5.5.9, Steam Generator (SG) Program:

TS 5.5.9 introductory paragraph and paragraph b.1 would be revised by replacing steam generator with SG in a few instances.

TS 5.5.9.d would be revised by adding a phrase regarding portions of the tube that are exempt from inspection by alternate repair criteria.

TS 5.5.9.d.2 would be revised by deleting the requirement to base inspection frequency on the more restrictive metric between either the effective full power months (EFPM) or refueling outage and to use just the EFPM metric.

TS 5.5.9.d.2 would be revised by changing the requirement to inspect 100 percent of the tubes at periods of 120, 96, and 72 EFPM to 54 EFPM. A 72 EFPM inspection period would be permitted if SG tubing has never experienced cracking (not including regions exempt from inspection by alternate repair criteria) and the SG inspection was performed with enhanced probes. A description of the enhanced probe inspection would be added.

TS 5.5.9.d.2 would be revised by deleting the allowance to extend the inspection period by up to 3 EFPM and by deleting the discussion of prorating inspections.

TS 5.5.9.d.3 would be revised by replacing shall not exceed 24 effective full power months or one refueling outage (whichever results in more frequent inspections) with shall be at the next refueling outage.

TS 5.5.9.d.3 would be revised by adding a phrase regarding portions of the tube that are exempt from inspection by alternate repair crit eria that replaces the phrase not excluded above. An additional phrase would be added that permits deferring SG inspections after cracking indications are found if the 100 percent inspection was performed with enhanced probes.

TS 5.6.8, Steam Generator Tube Inspection Report:

Existing reporting requirement b. would be renumbered as c. and be revised by editorial and punctuation changes.

New reporting requirement b. would be added to require the nondestructive examination (NDE) techniques utilized for tubes with increased degradation susceptibility be reported.

Existing reporting requirement c. would be renumbered as c.1. and be revised by editorial and punctuation changes.

Existing reporting requirement d. would be renumbered as c.2. and be revised to note that the location, orientation (if linear), measured size (if available), and voltage response do not need to be reported for tube wear indications at support structures that

are less than 20 percent through-wall. However, the total number of tube wear indications at support structures that are less than 20 percent through-wall would be reported.

New reporting requirement d. would be added to require an analysis summary of the tube integrity conditions predicted to exist at the next scheduled inspection relative to the applicable performance criteria, including the analysis methodology, inputs, and results.

Existing reporting requirement e. would be renumbered as c.4. and be revised by editorial and punctuation changes.

Existing reporting requirement f. would be renumbered as e. and be revised by editorial and punctuation changes.

New reporting requirement f. would be added to require the results of any SG secondary side inspections to be reported.

Existing reporting requirement g. would be renumbered as c.3. and be revised to add the requirements to report a description of the condition monitoring assessment, the margin to the tube integrity performance criteria, and a comparison with the margin predicted to exist at the inspection by the previous forward-looking tube integrity assessment. In addition, the requirement to report the results of tube pulls and in-situ testing would be deleted.

Existing reporting requirements h., i., and j. would be renumbered to g., h., and i. and be revised by editorial and punctuation changes.

1.2 Additional Proposed TS Changes

In addition to the changes proposed consistent with the traveler discussed in Section 1.1, the licensee proposed the following variations.

1.2.1 Editorial Variations

The licensee noted that the Robinson TSs have different numbering than the standard technical specifications (STS) on which TSTF-577 was based. Specifically, the Robinson TSs use TS 5.6.8 for Steam Generator Tube Inspection Report and TSTF-577 uses TS 5.6.7.

1.2.2 Other Variations

In Section 2.2 of the LAR, as supplemented, the licensee identified two other variations:

1. The licensee proposed a Robinson SG tube inspection period of 72 EFPM for the inspection period that began December 8, 2020. This is a significant variation because Robinson has not met both conditions identified in TSTF-577 for obtaining a 72 EFPM inspection period. Although no cracking has been detected in the SG tubes to date, Robinson has not previously performed a 100 percent inspection of all in-service SG tubes using an enhanced probe inspection method.

2. The licensee proposed to revise TS 5.5.9.c (i.e., provisions for SG tube plugging criteria) to remove the time-based component of the SG tube plugging criteria and modify

the 47 percent nominal tube wall thickness to a value of 40 percent. This is a variation because the SG tube plugging criteria in the SG Program TS is not modified by TSTF-577.

2.0 REGULATORY EVALUATION

The regulations in Title 10 of the Code of Federal Regulations (10 CFR) paragraph 50.36(c)(5),

Administrative controls, state that [a]dministr ative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner. Each licensee shall submit any reports to the Commission pursuant to approved technical specifications as specified in

[10 CFR] 50.4. Technical Specification Section 5.0, Administrative Controls, requires that an SG Program be established and implemented to ensure that SG tube integrity is maintained.

Programs established by the licensee, including the SG Program, are listed in the administrative controls section of the TS to operate the facility in a safe manner.

The NRC staffs guidance for the review of TSs is in NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor]

Edition (SRP), Chapter 16.0, Technical Specifications, Revision 3, dated March 2010 (ML100351425). As described therein, as part of the regulatory standardization effort, the NRC staff has prepared STSs for each of the LWR nuclear designs. Accordingly, the NRC staffs review includes consideration of whether the proposed changes are consistent with NUREG-1431,1 as modified by NRC-approved travelers.

TSTF-577 revised the STSs related to SG tube inspections and SG tube inspection reporting requirements. The NRC approved TSTF-577 on April 14, 2021 (ML21099A086).

3.0 TECHNICAL EVALUATION

3.1 Proposed TS Changes to Adopt TSTF-577

The NRC staff compared the licensees proposed TS changes in Section 1.1 of this SE against the changes approved in TSTF-577. In accordance with SRP Chapter 16.0, the NRC staff determined that the STS changes approved in TSTF-577 are applicable because Robinson is a pressurized-water reactor (PWR) design plant and the NRC staff approved the TSTF-577 changes for PWR designs. The NRC staff finds th at the licensees proposed changes to the Robinson TSs in Section 1.1 of this SE are consistent with those found acceptable in TSTF-577.

In the SE of TSTF-577, the NRC staff concluded that the TSTF-577 changes to STS 5.5.9, Steam Generator (SG) Program, and STS 5.6.7, Steam Generator Tube Inspection Report, were acceptable because, as discussed in Section 3.0 of that SE, they continued to ensure SG tube integrity and, therefore, protected the public health and safety. In particular, the structural integrity performance criterion and accident-induc ed leakage performance criterion (explained in STS 5.5.9.b, items 1 and 2, respectively) will continue to be met with the proposed revised SG inspection intervals (maximum allowable time between SG inspections) and inspection periods (maximum allowable time between 100 percent of SG tubes inspections). Additionally, the

1 U.S. Nuclear Regulatory Commission, Standard Technical Specifications, Westinghouse Plants, NUREG-1431, Volume 1, Specifications, and Volume 2, Bases, Revision 5, September 2021 (ML21259A155 and ML21259A159, respectively).

proposed changes to the reporting requirements will provide more detailed and consistent information to the NRC. Therefore, the NRC staff found that the proposed changes to the SG program and inspection reporting requirements were acceptable because they continued to meet the requirements of 10 CFR 50.36(c)(5) by providing administrative controls necessary to assure operation of the facility in a safe manner. For these same reasons, the NRC staff concludes that the corresponding proposed changes to the Robinson TSs in Section 1.1 of this SE continue to meet the requirements of 10 CFR 50.36(c)(5).

3.2 Additional Proposed TS Changes

3.2.1 Editorial

The licensee noted that the Robinson TSs use, in part, different numbering than the STS on which TSTF-577 was based. Specifically, the Robinson TSs use TS 5.6.8 for Steam Generator Tube Inspection Report and TSTF-577 uses TS 5.6.7. The NRC staff finds that the different TS numbering is acceptable because it does not substantively alter TS requirements.

3.2.2 Other Variations - 72 EFPM Until the Next SG Tube Inspection

The licensees justification for the technical variation from TSTF-577 related to 72 EFPM until the next SG tube inspection is provided in Attachment 3 of the LAR, along with the operational assessment provided in the supplement dated January 6, 2022. The NRC staff SE of TSTF-577 concluded that for SGs with Alloy 600TT tubing, such as those at Robinson, there is reasonable assurance that tube integrity will be maintained for up to 72 EFPM if two conditions are met: (1) stress corrosion cracking (SCC) has not been detected during tube inspections (excluding tube end cracking that is already covered by an alternate repair criteria) and (2) an enhanced probe inspection method is performed at the 100 perc ent SG tube inspection entering each 72 EFPM inspection interval. TSTF-577 defines the enhanced probe inspection method as performing 100 percent SG tube inspections from tube end to tube end (except any portion exempt from inspection by an alternate repair criteria) with eddy current probes equivalent to or better than array probe technology.

Although the licensee provides that Robinson meets the first of the above conditions (i.e., no SCC), the 100 percent SG tube inspection that initiated the inspection period beginning December 8, 2020, did not satisfy the enhanced probe inspection method of the second condition. The NRC staff notes that the 100 per cent SG tube inspection performed by the licensee in late 2020 occurred while the staff was reviewing TSTF-577 and prior to the staff taking its position related to SGs with Alloy 600TT tubing to obtain a 72 EFPM inspection period.

This staff position was subsequently communicated to stakeholders in January 2021 (ML21007A111). A key consideration for the staff during the acceptance review for this LAR was that the licensee performed its SG tubing inspection just before the staffs Alloy 600 TT position for TSTF-577 was made known in January 2021. Therefore, the staff considers this variation to represent a unique situation that should not be considered as precedent for licensees with Alloy 600TT tubing, after the communication of the staff position in January 2021, to obtain a 72 EFPM inspection period without meet ing the conditions specified in the staff SE of TSTF-577. The licensees LAR stated that future eddy current inspections of Robinson SG tubing will be performed using the enhanced probe inspection method.

The NRC staffs technical evaluation of the Robins on LAR variation related to the next SG tube inspection focused on the licensees evaluation that tube integrity will be maintained for a 72 EFPM inspection period beginning in December 2020. The performance criteria for SG tube

integrity, i.e., the structural integrity performance criterion, the accident induced leakage performance criterion, and the operational leakage performance criterion, for Robinson are defined in Robinson TS 5.5.9.b. Meeting their TS 5.5.9.b performance criteria demonstrates that SG tube integrity is maintained. Hereafter, these criteria are referred to as performance criteria for SG tube integrity. Maintaining SG tube integrity results in SG tubes performing their safety function as part of the reactor coolant system pressure boundary which protects public health and safety.

The three Robinson SGs are Westinghouse Model 44F SGs each containing 3,214 tubes. The Alloy 600TT tubes have a 0.875 inch outer diameter and a nominal wall thickness of 0.050 inches. The Robinson SGs had operated for 29.6 effective full power years (EFPY) at the time of the Robinson Refueling Outage (RO) 32 (RO32) that ended in December 2020. The Robinson primary coolant hot leg temperature (T hot) is less than 605 degrees Fahrenheit. The LAR and supplements concluded that the 72 EFPM inspection period would be acceptable at Robinson based on the low amount of existing tube degradation, the low T hot, the history of cracking in the Alloy 600 TT fleet, and the plant-specific operational assessment that projects meeting the structural and leakage performance criteria for existing and potential degradation mechanisms until RO35.

The RO32 SG tube eddy current inspection scope included 100 percent full length bobbin coil inspection, except for the U-bend portion of the Row 1 and Row 2 tubes. Array probe inspection was performed in the following locations:

All hot leg tube portions from the tube end to 4 inches above the top of the tubesheet All Row 1 and Row 2 U-bends 20 percent of the Row 9 U-bends Hot leg and cold leg tube bundle periphery and tube lane, five tubes deep, from the tube end to the first tube support plate (TSP)

Full length of all potentially high stress tubes (screened as -2 sigma tubes)

All tube dents greater than or equal to 4 volts amplitude All dents at the upper TSP (designated 6H, 6C) regardless of voltage Tube locations that required array sizing in RO30, possible loose parts, surrounding known foreign object locations, at new wear indications, and all bobbin I-codes. The NRC staff notes that bobbin probe I-codes repr esent possible flaw signals that require supplemental testing.

The licensee did not need to expand the initial RO32 inspection scope because no new types of degradation were identified by the licensee during the planned inspections.

3.2.2.1 Existing SG Tube Degradation Mechanisms

The RO32 inspection results showed that the Robinson SGs have limited tube degradation from wear at tube support structures and wear from foreign objects. As used by the NRC staff below in this SE, foreign object wear and loose parts wear are interchangeable and refer to the same degradation mechanism. The overall effective tube plugging percentage was 0.58 percent upon exiting RO32. The staff reviewed the RO32 inspection results for the existing tube degradation mechanisms. The tube wear at support structures includes wear at anti-vibration bars (AVB) and at broached TSPs. A total of 25 indications of AVB wear were detected during RO32 with the deepest indication measuring 34 percent through-wall (TW). A total of 5 indications of wear at broached TSPs were detected during RO32. One tube with a 27 percent TW wear indication

that had no previously reported indication was plugged. The deepest TSP wear indication remaining in service measured 21 percent TW. The licensee reported that 54 loose parts wear indications were detected during RO32, including 20 new indications. The deepest loose parts wear indication measured 38 percent TW and was unchanged from RO30. The licensee did not identify any possible loose parts eddy current signals in the tubes with existing loose parts (i.e.,

foreign objects) indications and none of the loose parts indications required plugging. Many of the historical foreign object indications from RO30 were dispositioned as historical no change (HNC) upon conformation that the eddy current inspection signal has not changed for three successive inspections. The staff notes that if a loose part is no longer present at a tube location with prior loose parts wear, the indication would not be expected to change since the wear mechanism would no longer be active. The staff reviewed the licensees RO32 Condition Monitoring and Operational Assessment (CMOA) report for existing tube degradation mechanisms. All existing tube degradation mechanisms detected during RO32 met condition monitoring and the measured tube wear was well within the projected tube degradation from the previous inspection.

The licensees operational assessment provided deterministic analysis for tube wear at structures and from loose parts for operation of Robinson until RO35. The licensees analysis for tube wear at structures involved evaluating the progression of wear during future operation.

For wear at anti-vibration bars and tube support plates, the licensee applied a bounding length and bounding growth rates to the deepest wear indication remaining in service for a 6.0 EFPY period. Since no wear flaws were detected at the flow distribution baffle, the licensee assumed a conservative flaw length and a flaw depth equal to the 95 th percentile probability of detection and applied a bounding growth rate. The licensees operational assessment shows that the worst flaws projected to the end of the 6 EFPY operating period will meet the TS 5.5.9.b performance criteria. The NRC staff finds this to be acceptable because the analysis uses the worst wear flaw (or assumed worst undetected flaw), it applies bounding growth rates for a conservative 6.0 additional EFPY, and is projected to meet the end of cycle performance criteria for SG tube integrity. The licensees analysis for loose parts provided the location of all tubes with loose parts wear indications and indicated that no additional loose parts wear was expected at these locations since the eddy current inspection did not identify any potential loose parts remaining at these locations. The staff finds this to be acceptable because tubes returned to service with loose parts wear had no remaining loose part present and because the licensee determined that any loose parts remaining in the SGs would not cause significant tube wear that would challenge tube integrity before the next inspection. The staff also notes that the licensee performed array probe inspections in the tube regions most likely to contain loose parts and surrounding all known loose parts. Therefore, the staff concludes that the licensees evaluation for existing tube degradation mechanisms due to wear at structures and from loose parts is acceptable.

The NRC staff also acknowledges that predicting future loose part generation is not possible since past fleet-wide operating experience has shown that new loose part generation, transport to the SG tube bundle, and interactions with the tubes cannot be reliably predicted. However, plants can reduce the probability of loose parts by maintaining robust foreign material exclusion programs and applying lessons learned from previous industry operating experience with respect to loose parts. Plants in general have demonstrated the ability to conservatively manage loose parts once they are detected by eddy current examinations and by secondary-side foreign object search and retrieval activities. If unanticipated aggressive tube wear from a new loose part should occur, industry operating experience has shown that a primary-to-secondary leak is much more probable than a loss of tube integrity. In the event of a primary-to-secondary leak, the staff will interact with the licensee in accordance with established procedures in Inspection

Manual Chapter 0327, Steam Generator Tube Primary-to-Secondary Leakage, dated January 1, 2019 (ML18093B067), to ensure that the licensee is responding to the leakage in a conservative manner.

3.2.2.2 Potential SG Tube Degradation Mechanisms

Although the licensee has not detected any SCC in Robinson SG tubing, the operational assessment evaluated for potential tube cracking in the expansion transition region (at the top of the tubesheet) and at the TSP intersections. The NRC staff reviewed the licensees choice of potential cracking locations and found them to be acceptable because these are the most probable locations for SCC based on Alloy 600TT operating experience. The operational assessment provided the results from full bundle probabilistic analysis for both axial and circumferential outer diameter stress corrosion cracking (ODSCC) at the top of the tubesheet and axial cracking at the TSPs. The analysis results were provided for these cracking mechanisms and locations after 2 cycles and 3 cycles of operation (until the next planned inspections at RO35) relative to the performance criteria for SG tube integrity. Analysis results were shown in terms of probability of burst (POB), probability of leakage (POL), and rate of accident induced leakage (AIL). The analysis also provides the lower fifth percentile burst pressure in comparison to the performance criteria for SG tube integrity, in this case the structural integrity performance criterion of three times normal operating differential pressure.

For each of the potential degradation mechanisms, the operational assessment analysis evaluates the behavior of postulated flaws that could have been present but were not detected by the RO32 SG inspection, along with flaws that could initiate during the 3-cycle operating period until RO35. The analysis assumed that two undetected cracks were present for each SCC mechanism evaluated. Flaw distributions were based on the relevant site-specific eddy current inspection technique probability of detection (POD) function and the undetected flaw total length distribution was derived from industry Alloy 600 TT crack data. Since no cracking has been detected at Robinson, site-specific crack growth rates do not exist. Therefore, flaw growth rate distributions were developed from Electric Power Research Institute (EPRI) Report 3002020909, Steam Generator Management Program: Steam Generator Integrity Assessment GuidelinesRevision 5, default crack growth rate distributions, adjusted for Robinsons temperatures.

The operational assessment probabilistic analysis results for both axial cracking and circumferential ODSCC cracking located at the top of tubesheet were shown to satisfy all SG performance criteria for burst, leakage, and accident leakage. The NRC staff finds the evaluation of potential cracks at the top of the tubesheet acceptable because the licensee assumed that cracks were present in RO32 that were undetected, used industry data to support the probabilistic analysis including for flaw gr owth, and projected that all tube integrity performance criteria will be satisfied until the next inspection during RO35.

During its review, the NRC staff issued to the lic ensee requests for additional information (RAIs) concerning several topics, including the quantity of secondary-side deposits in the Robinson SGs relative to another Duke Energy unit with Alloy 600TT tubing that experienced cracking under deposits. In its response (ML22118A336), the licensee identified that Robinson had significantly less deposits compared to the other Duke Energy unit. In contrast to the other unit, which had not had a chemical cleaning before cracking was detected, Robinson performed a chemical cleaning at the end-of-cycle 28 (i.e., during RO28) in 2013. Based on differences in design and deposit loading, the licensee stated that it does not expect Robinson SG tubing to have the same conditions that resulted in cracking under deposits at the other unit. The staff

notes that the Robinson hot leg temperature is approximately 10°F lower than the other unit that experienced cracking under deposits. The staff finds the RAI response acceptable because Robinson has had a chemical cleaning, has less than half the deposit mass of the other unit, and operates at a significantly lower temperature.

The NRC staff also issued an RAI requesting more details about the analysis of axial ODSCC cracking at TSPs, since this potential mechanism had the smallest margin to the SG performance acceptance criteria. In its res ponse (ML22118A336), the licensee stated that two of the tube integrity vendors standard practices were not followed in the operational assessment analysis for axial ODSCC at TSPs. Therefore, the licensee provided a recalculation that included NDE measurement uncertainty and allowance for an additional cycle of flaw growth. The operational assessment also updated the array probe POD curve for cracking at TSPs, relying on a recently published proprietary EPRI report Plus Point to X-probe Amplitude Transfer Function and Probability of Detection (EPRI Transfer Function report). The EPRI Transfer Function report was made available to the staff. The staff and the licensee subsequently discussed the use of the EPRI Transfer Function report during a closed meeting (ML22166A018). The staff questioned whether additional validation of the EPRI Transfer Function report was needed. After the meeting, the licensee submitted a supplemental RAI response (ML22209A156) that provided the probabilis tic operational assessment results using four different cases related to POD and other assumptions related to axial ODSCC at the TSPs.

The base case in the supplemental response used the array probe POD obtained using the EPRI Transfer Function report. The other three cases did not rely on the EPRI Transfer Function report but changed other evaluation parameters related to bobbin coil or array probe POD or plant modeling assumptions (e.g., assumed less conservative but still bounding plant values).

The NRC staff reviewed the RAI responses and determined that the additional operational assessment cases results demonstrated that all per formance criteria for SG tube integrity will be satisfied for Robinson until RO35 (i.e., three cycles of operation). This determination is based on the additional cases using bounding plant input values in operational assessment analyses performed according to the current industry integrity assessment guidelines. This determination does not rely on the base case that used the EPRI Transfer Function report as part of the analysis and the staff takes no position on the acceptability of that report.

3.2.2.3 Other SG Tube Integrity Considerations

In addition to the operational assessment pr ovided by the licensee, the NRC staff also considered the operating experience from the domestic Westinghouse Model 44F SG fleet. In addition to Robinson, there have been four other Model 44F PWR units in service with Alloy 600TT tubing. One PWR with much less operating time and a lower T hot than Robinson did not detect cracking before it permanently shut down. The three other Model 44F PWR units are operating at a higher temperature (T hot 610°F) than Robinson (Thot less than 605°F). They have also operated for a slightly longer time than Robinson (approximately one additional operating cycle). One of the three other Model 44F PWR units has also not detected any cracking to date.

The other two each detected SCC for the first time during their fall 2021 and spring 2022 inspections, respectively. Each unit detected one tube with SCC in the U-bend portion of the tubing. At each unit, tube integrity was not challenged by the SCC and the tube was plugged during the refueling outage. Therefore, SCC has only been detected in two tubes in the other three PWR units with Model 44F SGs and these units operate at higher temperature and have operated for a slightly longer time (EFPY) than Robinson. Having a T hot of five degrees lower than the other Model 44F PWR units will be beneficial to Robinson both in terms of SCC

initiation and SCC growth rates. This excellent operating experience of the Model 44F PWR fleet after approximately 30 EFPY provides additional confidence to the NRC staff that the Robinson Model 44F SGs can be expected to retain tube integrity until RO35.

The current Robinson TS 3.4.13, RCS Operatio nal LEAKAGE, primary to secondary operating leakage limit through any one SG is 75 gallons per day (gpd). In accordance with the Robinson H* alternate TS repair criteria that consider potential degradation mechanisms and then project leakage, the administrative primary to secondary operational leakage limit will be reduced to 63.8 gpd for operating cycle 35. No changes are planned for the administrative limits for operating cycles 33 and 34. The administrative limit change for operating cycle 35 is acceptable to the NRC staff because it reduces the allo wable primary to secondary leakage limit in accordance with the H* alternate TS repair criteria.

3.2.3 Other Variations - SG Tube Plugging Criteria

In a supplement to the LAR dated October 4, 2022 (ML22277A399), the licensee proposed an additional variation to revise the current Robinson TS 5.5.9.c time-based plugging criterion, which states: Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding the following criteria shall be plugged: 47 [percent] of the nominal tube wall thickness if the next inspection interval of that tube is < 12 months, and a 2 [percent] reduction in the plugging criteria for each 12 month period until the next inspection of the tube. The licensee proposed to change this language to state Provisions for SG tube plugging criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40 [percent] of the nominal tube wall thickness shall be plugged.

Robinson proposed a plugging criterion change since SG B, Tube Row 3, Column 5 (R3C5) contains a 38 percent TW indication attributed to tube wear from a foreign object. Using the time-based plugging criteria described above, the R3C5 indication depth would exceed the plugging criteria during the 12 month period preceding RO35 in 2026. Therefore, the licensee proposed to modify the plugging criterion to a fixed value for tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40 percent of the nominal tube wall thickness.

The proposed change is acceptable to the NRC staff with respect to SG B, Tube R3C5, because no further tube degradation is expected at this location since both the eddy current inspection and the foreign object visual inspection performed during RO32 showed no evidence of a foreign object remaining at the wear location. The expectation of no additional tube wear is also supported by the fact that the 38 percent TW indication depth has not changed between the past two SG tube inspections in RO30 and RO32. Therefore, the staff expects that SG B, Tube R3C5 will continue to satisfy the performance criteria for SG tube integrity at the degradation location until the next SG tube inspecti on, regardless of the tube plugging criterion.

The proposed 40 percent TW degradation plugging crit erion is also acceptable to the NRC staff with respect to whether all tubes in the Robinson SGs will maintain tube integrity. The time-based tube plugging criterion (i.e., 2 percent reduction for each 12 month period) was added to the Robinson TSs by Amendment No. 44 issued in 1979 (ML0205000754). At that time, the Robinson SGs had Alloy 600 mill annealed (600 MA) tubing and plants were operating with a sodium phosphate based secondary side water chemistry. Since the Alloy 600 MA tubing can be susceptible to tube wastage in this environment, especially at the top of the tubesheet sludge pile, the 2 percent reduction in the tube plugging limit per 12 month period criterion was added as a thinning allowance between inspections. The tube wastage degradation mechanism

resulting from phosphate water chemistry is no longer a concern for the Robinson SGs. The original SGs with Alloy 600 MA tubing were replaced in 1984 with SGs containing more corrosion resistant Alloy 600 TT tubing. In addition, the change in secondary side water chemistry from sodium phosphate to an all-volatile water chemistry based on ammonia and hydrazine has eliminated the concern about a tube wastage degradation mechanism. The Robinson replacement SGs have operated for 38 years and have not experienced general tube wastage or thinning degradation. In addition, Robinson is the only U.S. PWR with a time-based tube plugging criterion. Therefore, revising the TS to a fixed plugging criterion, rather than a time-based plugging criterion, is acceptable to the staff.

With respect to the proposed fixed 40 percent TW flaw degradation tube plugging criterion, the NRC staff notes that a 40 percent plugging criterion has been approved by the NRC for most of the U.S. PWR fleet and is consistent with the plugging criterion in the Westinghouse STS (i.e.,

NUREG-1431). Additionally, the other operating units with Westinghouse Model 44F SGs have the same nominal tube outside diameter and nominal tube wall thickness as Robinson, and a 40 percent plugging criterion. Therefore, the 40 percent plugging criterion proposed by the licensee is acceptable to the staff because that value has been previously approved by the NRC for most PWRs, including all operating units with the same Model 44F SGs (i.e., same nominal SG tubing dimensions).

3.2.4 Other Variations - Conclusion

The NRC staff finds the licensees variation for a Robinson SG tube inspection period of 72 EFPM for the inspection period that began December 8, 2020, acceptable since the licensees analysis acceptably demonstrates that the performanc e criteria for SG tube integrity in Robinson TS 5.5.9.b will be satisfied for active and potential degradation mechanisms for up to 72 EFPM, which bounds the next planned SG tube inspection during RO35.

The NRC staff finds the licensees variation fo r a change in the Robinson TS 5.5.9.c tube plugging criterion to tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40 percent of the nominal tube wall thickness acceptable since the degradation mechanism that resulted in the previous time based plugging criterion is no longer a concern at Robinson, the new plugging criterion has been approved by the NRC for most PWRs, and the new plugging criterion has been shown to be effective in maintaining tube integrity at other units with the same Model 44F SGs as Robinson.

3.3 TS Change Consistency

The NRC staff reviewed the proposed TS changes for technical clarity and consistency with the existing requirements for customary terminology and formatting. The NRC staff finds that the proposed changes are consistent with Chapter 16.0 of the SRP and are therefore acceptable.

4.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

NOTICES AND ENVIRONMENTAL FINDINGS RELATED TO AMENDMENT NO. 273 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-23 DUKE ENERGY PROGRESS, LLC H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261

Application (i.e., initial and supplements) Safety Evaluation Date December 9, 2021, Agencywide Documents December 15, 2022 Access and Management System (ADAMS)

Accession No. ML21343A047 January 6, 2022 (ML22006A240)

April 28, 2022 (ML22118A336)

July 28, 2022 (ML22209A156)

October 4, 2022 (ML22277A399)

1.0 INTRODUCTION

Duke Energy Progress, LLC (Duke Energy, the licensee) requested changes to the technical specifications (TSs) for H. B. Robinson Steam El ectric Plant, Unit No. 2 (Robinson) by license amendment request (application). In its applicat ion, as supplemented, the licensee proposed changes that would revise the Steam Generator (SG) Program and the Steam Generator Tube Inspection Report TSs based on Technical S pecifications Task Force (TSTF) Improved Standard Technical Specifications Change Traveler TSTF-577, Revision 1, Revised Frequencies for Steam Generator Tube Inspections (TSTF-577) (ML21060B434), and the associated U.S. Nuclear Regulatory Commission (NRC, the Commission) staff safety evaluation (SE) of TSTF-577 (ML21098A188).

The supplements did not expand the scope of the application as originally noticed and did not change the U.S. Nuclear Regulatory Commission (NRC, the Commission) staffs original proposed no significant hazards consideration determination as published in the Federal Register on February 22, 2022 (87 FR 9647).

2.0 STATE CONSULTATION

In accordance with the Commissions regulations, the South Carolina State official was notified of the proposed issuance of the amendment on October 20, 2022. The State official had no comments.

3.0 ENVIRONMENTAL CONSIDERATION

The amendment relates, in part, to changes in recordkeeping, reporting, or administrative procedures or requirements. The amendment also relates, in part, to changing requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative

occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on February 22, 2022 (87 FR 9647).

Accordingly, the amendment meets the eligibility cr iteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and 51.22(c)(10). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

ML22294A092 OFFICE NRR/DORL/LPL2-2/PM NRR/DORL/LPL2-2/LA NRR/DNRL/NCSG/BC NRR/DSS/STSB/BC NAME LHaeg RButler SBloom VCusumano DATE 10/21/2022 10/28/2022 10/14/2022 10/19/2022 OFFICE OGC - NLO NRR/DORL/LPL2-2/BC NRR/DORL/LPL2-2/PM NAME MWoods DWrona LHaeg DATE 12/05/2022 12/15/2022 12/15/2022