ML19063B721

Issuance of Amendment Regarding Application to Revise Technical Specifications for Use of Voltage-Based Alternate Repair Criteria in Accordance with Generic Letter 95-05

ML19063B721
Person / Time
Site:	Watts Bar
Issue date:	06/03/2019
From:	Robert Schaaf Plant Licensing Branch II
To:	James Shea, Skaggs M Tennessee Valley Authority
Schaal R 415-6020
References
EPID L-2018-LLA-0143
Download: ML19063B721 (27)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555.0001 June 3, 2019 Mr. Joseph W. Shea Vice President, Nuclear Regulatory Affairs and Support Services Tennessee Valley Authority 1101 Market Street, LP 4A Chattanooga, TN 37402-2801

SUBJECT:

WATTS BAR NUCLEAR PLANT, UNIT 2- ISSUANCE OF AMENDMENT REGARDING APPLICATION TO REVISE TECHNICAL SPECIFICATIONS FOR USE OF VOLTAGE-BASED ALTERNATE REPAIR CRITERIA IN ACCORDANCE WITH GENERIC LETTER 95-05 (EPID L-2018-LLA-0143)

Dear Mr. Shea:

The U.S. Nuclear Regulatory Commission (NRC or Commission) has issued the enclosed Amendment No. 28 to Facility Operating License No. NPF-96 for the Watts Bar Nuclear Plant (Watts Bar), Unit 2. This amendment is in response to your application dated May 14, 2018, as supplemented by letter dated November 8, 2018.

This amendment implements a voltage-based alternate repair criteria (ARC) for degraded steam generator tubes in the Watts Bar Unit 2 Westinghouse Model D3 steam generators. The ARC follow the guidelines set forth in NRC Generic Letter 95-05, "Voltage-Based Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking."

A copy of our related safety evaluation is also enclosed. Notice of issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely,

!:!:! Se! f : / .anager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-391

Enclosures:

1. Amendment No. 28 to NPF-96

2. Safety Evaluation cc: Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O.C . 20555-0001 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-391 WATTS BAR NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 28 License No. NPF-96

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

A. The application for amendment by Tennessee Valley Authority {TVA, the licensee) dated May 14, 2018, as supplemented by letter dated November 8, 2018, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions 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 health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense 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 Commission's regulations and all applicable requirements have been satisfied ..

ENCLOSURE 1

2

2. Accordingly, Facility Operating License No. NPF-96 is amended as indicated in the attachment to this license amendment, and paragraph 2.C.(2) is hereby amended to read as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 28 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. TVA shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of the date of its issuance and shall be implemented within 60 days.

FOR THE NUCLEAR REGULATORY COMMISSION u~J:z Plant Licensing Branch 11-2 Division of operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Operating License And Technical Specifications Date of Issuance: June 3, 2019

ATTACHMENT TO AMENDMENT NO. 28 WATTS BAR NUCLEAR PLANT, UNIT 2 FACILITY OPERATING LICENSE NO. NPF-96 DOCKET NO. 50-391 Replace page 3 of Facility Operating License No. NPF-96 with the attached revised page 3.

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

Remove Pages Insert Pages 5.0-16 5.0-16 5.0-16a 5.0-16b 5.0-1 ?a 5.0-17a 5.0-36 5.0-36

C. The license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I 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.

(1) Maximum Power Level TVA is authorized to operate the facility at reactor core power levels not in excess of 3411 megawatts thermal.

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 28 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. TVA shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) TVA shall implement permanent modifications to prevent overtopping of the embankments of the Fort Loudon Dam due to the Probable Maximum Flood by June 30, 2018.

(4) PAD4TCD may be used to establish core operating limits until the WBN Unit 2 steam generators are replaced with steam generators equivalent to the existing steam generators at WBN Unit 1.

(5) By December 31, 2019, the licensee shall report to the NRC that the actions to resolve the issues identified in Bulletin 2012-01, "Design Vulnerability in Electrical Power System," have been implemented.

(6) The licensee shall maintain in effect the provisions of the physical security plan, security personnel training and qualification plan, and safeguards contingency plan, and all amendments made pursuant to the authority of 10 CFR 50.90 and 50.54(p).

(7) TVA shall fully implement and maintain in effect all provisions of the Commission approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The TVA approved CSP was discussed in NUREG-0847, Supplement 28, as amended by changes approved in License Amendment No. 7.

(8) TVA shall implement and maintain in effect all provisions of the approved fire protection program as described in the Fire Protection Report for the facility, as described in NUREG-0847, Supplement 29, subject to the following provision:

Facility Operating License No. NPF-96 Amendment No. 28

Procedures, Programs, and Manuals 5.7

5. 7 Procedures, Programs, and Manuals 5.7.2.12 Steam Generator (SG) Program (continued)

2. Accident induced leakage performance criterion : The primary-to-secondary accident induced leakage rate for any design basis accident, other than an 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 for all degradation mechanisms is not to exceed 150 gpd for each unfaulted SG. Leakage for all degradation mechanisms, excluding that described in Specification 5.7.2.12.c.2, is not to exceed 1 gpm in the faulted SG. Leakage for degradation mechanisms described in Specification 5. 7 .2.12.c.2 is not to exceed 4 gpm for the faulted SG.

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

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 shall be applied as an alternative to the 40% depth based criteria:

1. Tubes with service-induced flaws located in the portion of the tube from the top of the tubesheet to 1.64 inches below the top of the tubesheet, or from the bottom of the roll transition to 1.64 inches below the bottom of the roll transition, whichever is lower, shall be plugged. Tubes with service-induced flaws located below this elevation do not require plugging.

2. The voltage based methodology, in accordance with Generic Letter (GL) 95-05, shall be applied at the tube to straight leg tube support plate interface as an alternative to the 40% depth based criteria of Specification 5.7.2.12.c: Tubes shall be plugged in accordance with GL 95-05.

Tube Support Plate Plugging Limit is used for the disposition of an Alloy 600 steam generator tube for continued service that is experiencing predominantly axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates and flow distribution baffles (FOB). At tube support plate intersections and FOB, (continued)

Watts Bar - Unit 2 5.0-16 Amendment 2, 28

Procedures, Programs, and Manuals 5.7

5. 7 Procedures, Programs, and Manuals the plugging or repair limit is described below:

a) Steam generator tubes, with degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plates and FOB with bobbin voltages less than or equal to 1.0 volt will be allowed to remain in service.

b) Steam generator tubes, with degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plates and FOB with a bobbin voltage greater than 1.0 volt will be plugged or repaired, except as noted in Specification 5. 7 .2.12.c.2.c below.

c) Steam generator tubes, with indications of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plates and FOB with a bobbin voltage greater than 1.0 volts but less than or equal to the upper voltage repair limit (calculated according to the methodology in GL 95-05 as supplemented) may remain in service if a rotating pancake

• coil or acceptable alternative inspection does not detect degradation.

d) Certain intersections as identified in Attachment 2 of WAT-0-10709 ("Tennessee Valley Authority, Watts Bar Nuclear Power Plant Unit 1, Application for Implementation of Voltage Based Repair Criteria, Westinghouse Steam Generator Tubes Affected by OOSCC at TSPs,"

Revision 0, January 12, 2000) will be excluded from application of the voltage based repair criteria as it is determined that these intersection may collapse or deform following a postulated LOCA + SSE event. As noted in Section 4 of SG-SGMP-13-16-NP, "Watts Bar Nuclear Plant Unit 2 Applicability of GL 95-05 Voltage Based Alternate Repair Criteria," the list of tubes identified for exclusion for the Unit 1 original steam generators are the same as for Unit 2.

e) Steam generator tubes, with indications of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plates and FOB with a bobbin voltage greater than the upper voltage repair limit (calculated according to the methodology in GL 95-05 as supplemented) will be plugged or repaired.

(continued)

Watts Bar - Unit 2 5.0-16a Amendment 28

Procedures, Programs, and Manuals 5.7

5. 7 Procedures, Programs, and Manuals If an unscheduled mid-cycle inspection is performed , the following mid-cycle repair limits apply instead of the limits specified in Specifications 5.7.2.12.c.2.a through 5.7.2.12.c.2.d.

The mid-cycle repair limits are determined from the following equations:

Vs 1.0+NOE + Gr[(CL-tit)/CL]

VMLRL = VMuRL-(VuRL-VLRL)[(CL-tit)/CL]

where:

VMuRL =mid-cycle upper voltage repair limit based on time into cycle V sL = structural limit voltage NOE = 95-percent cumulative probability allowance for nondestructive examination uncertainty (i.e. , a value of 20 percent has been approved by NRC). The NOE is the value provided by the in GL 95-05 as supplemented.

Gr= average growth rate per cycle length CL= cycle length (the time between two scheduled steam generator inspections)

VuRL = upper voltage repair limit (Note 1)

VLRL = lower voltage repair limit VMLRL = mid-cycle lower voltage repair limit based on VMuRL and time into cycle

.M =length of time since last scheduled inspection during which VuRL and VLRL were implemented Implementation of these mid-cycle repair limits should follow the same approach as in Specifications 5.7.2.12.c.2.a through 5.7.2.12.c.2.d.

Note 1: The upper voltage repair limit is calculated according to the methodology in GL 95-05 as supplemented. VuRL will differ at the tube support plates and flow distribution baffle.

(continued)

Watts Bar - Unit 2 5.0-16b Amendment 28

Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 5.7.2.12 Steam Generator (SG) Program (continued)

The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period.

3. If crack indications are found in any SG tube, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the cr~ck indication shall not exceed 24 effective full power months or one refueling outage (whichever results in more frequent inspections). 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.

4. Indications left in service as a result of application of the tube support plate voltage-based repair criteria (Specification

5. 7 .2.12.c.2) shall be inspected by bobbin coil probe during all future refueling outages.

Implementation of the steam generator tube-to-tube support plate repair criteria requires a 100 percent bobbin coil inspection for hot-leg and cold-leg tube support plate intersections down to the lowest cold-leg tube support plate (including the FOB) with known outside diameter stress corrosion cracking (ODSCC) indications. The determination of the lowest cold-leg tube support plate intersections having ODSCC indications shall be based on the performance of at least a 20-percent random sampling of tubes inspected over their full length.

e. Provisions for monitoring operational primary-to-secondary LEAKAGE.

(continued)

Watts Bar - Unit 2 5.0-17a Amendment 2, 28

Reporting Requirements 5.9 5.9 Reporting Requirements (continued) 5.9.9 Steam Generator Tube Inspection Report (continued)

For implementation of the voltage based repair criteria , in accordance with GL 95-05, to tube support plate (and flow distribution baffle) intersections, notify the NRC prior to returning the steam generators to service should any of the following conditions arise: *

1. If estimated leakage based on the projected end-of cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution exceeds the leakage limit (determined from the licensing basis dose calculation for the postulated main steam line break) for the next operating cycle.

2. If circumferential crack-like indications are detected at the tube support plate intersection and flow distribution baffles.

3. If indications are identified that extend beyond the confines of the tube support plate and flow distribution baffles.

4. If indications are identified at the tube support plate elevations and flow distribution baffles that are attributable to primary water stress corrosion cracking.

5. If the calculated conditional burst probability based on the projected end-of-cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution exceeds 1 x 10-2 , notify the NRC and provide an assessment of the safety significance of the occurrence.

A report shall be submitted within 90 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 5. 7.2.12 , "Steam Generator (SG) Program, when voltage-based alternate repair criteria have been applied . The report shall include information described in Section 6.b of Attachment 1 to Generic Letter 95-05, "Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking .

5.10 Record Retention (removed from Technical Specifications)

Watts Bar-Unit 2 5.0-36 Amendment 28

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON , D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 28 TO FACILITY OPERATING LICENSE NO. NPF-96 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT, UNIT 2 DOCKET NO. 50-391

1.0 INTRODUCTION

By letter dated May 14, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18138A232), as supplemented by letter dated November 8, 2018 (ADAMS Accession No. ML18312A402), Tennessee Valley Authority (TVA or the licensee) submitted a request for an amendment to Facility Operating License No. NPF-96 for Watts Bar Nuclear Plant (Watts Bar), Unit 2. The proposed amendment would implement a voltage-based alternate repair criteria (ARC) for degraded steam generator (SG) tubes in the Watts Bar Unit 2 Westinghouse Model D3 SGs. The ARC follow the guidelines set forth in U.S. Nuclear Regulatory Commission (NRC or the Commission) Generic Letter (GL) 95-05, "Voltage-Based Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking." 1 The supplemental letter dated November 8, 2018, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register on November 20, 2018 (83 FR 58618).

2.0 REGULATORY EVALUATION

2.1 System Description Watts Bar Unit 2 has four Westinghouse model D3 SGs that use mill-annealed Alloy 600 tubing.

These SGs use carbon steel drilled-hole tube support plates and flow distribution baffle plates.

The outside diameter of each tube is 3/4-inch. These SGs are the same design as the original Watts Bar Unit 1 SGs. The Watts Bar Unit 1 SGs were replaced during the Watts Bar Unit 1 Cycle 7 refueling outage. Details of the Watts Bar Unit 2 SGs are provided in the Watts Bar Updated Final Safety Analysis Report (UFSAR) Section 5.5.2.2 and UFSAR Figure 5.5-3.

1 GL 95-05 is available on the NRC website at https://www.nrc.gov/reading-rm/doc-collections/gen-comm/gen-letters/1995/gl95005.html.

Enclosure 2

2.2 Proposed Changes The licensee proposed technical specification (TS) changes as described below. The licensee states that these changes are consistent with the guidance in GL 95-05. The changes are as follows:

• TS 5.7.2.12.b.2, "Steam Generator (SG) Program ," is revised as follows:

o The requirement that "Leakage is not to exceed 1 gpm per SG" is revised to state, "Leakage is not to exceed 1 gpm for the faulted SG loop and 150 gallons per day (gpd) for each unfaulted SG."

o The following sentence is added to TS 5.7.2.12.b.2, "For the specific types of degradation at specific locations as described in TS 5.7.2.12.c.2 of the Steam Generator Program , the leakage is not to exceed 4 gpm for the faulted SG loop and 150 gallons per day (gpd) for each unfaulted SG."

• A new item 2 is added to TS 5.7.2.12.c, "Steam Generator (SG) Program," as follows:

"The voltage-based methodology, in accordance with Generic Letter (GL) 95-05, shall be applied at the tube to straight leg tube support plate interface as an alternative to the 40% depth based criteria of Specification 5.7 .2.12.c: Tubes shall be plugged in accordance with GL 95-05.

Tube Support Plate Plugging Limit is used for the disposition of an Alloy 600 steam generator tube for continued service that is experiencing predominantly axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates and flow distribution baffles (FOB). At tube support plate intersections and FOB, the plugging or repair limit is described below:

a) Steam generator tubes, with degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plates and FOB with bobbin voltages less than or equal to 1.0 volt will be allowed to remain in service.

b) Steam generator tubes, with degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plates and FOB with a bobbin voltage greater than 1.0 volt will be plugged or repaired , except as noted in Specification 5. 7 .2.12.c.2.c below.

c) Steam generator tubes, with indications of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plates and FOB with a bobbin voltage greater than 1.0 volt but less than or equal to the upper voltage repair limit (calculated according to the methodology in GL 95-05 as supplemented) may remain in service if a rotating pancake coil or acceptable alternative inspection does not detect degradation.

d) Certain intersections as identified in Attachment 2 of WAT-0-10709 ("Tennessee Valley Authority, Watts Bar Nuclear Power Plant Unit 1, Application for Implementation of Voltage-Based Repair Criteria, Westinghouse Steam Generator Tubes Affected by OOSCC [outside diameter stress corrosion cracking] at TSPs

[tube support plates] ," Revision 0, January 12, 2000) will be excluded from

application of the Voltage-Based repair criteria as it is determined that these intersections may collapse or deform following a postulated LOCA [loss-of-coolant accident]+ SSE [safe shutdown earthquake] event. As noted in Section 4 of SG-SGMP-13-16-NP, "Watts Bar Nuclear Plant Unit 2 Applicability of GL 95-05 Voltage-Based Alternate Repair Criteria," the list of tubes identified for exclusion for Unit 1 are the same as for Unit 2.

e) Steam generator tubes, with indications of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plates and FOB with a bobbin voltage greater than the upper voltage repair limit (calculated according to the methodology in GL 95-05 as supplemented) will be plugged or repaired .

f) If an unscheduled mid-cycle inspection is performed, the following mid-cycle repair limits apply instead of the limits specified in Specifications 5. 7.2.12.c.2.a through

5. 7 .2.12.c.2.d.

The mid-cycle repair limits are determined from the following equations:

VsL VMURL =

1.0+NOE + Gr[(CL-~t)/CL]

where:

VMuRL = mid-cycle upper voltage repair limit based on time into cycle VsL = structural limit voltage NOE= 95-percent cumulative probability allowance for nondestructive examination uncertainty (i.e., a value of 20 percent has been approved by NRC). The NOE is the value provided in GL 95-05 as supplemented.

Gr = average growth rate per cycle length CL= cycle length {the time between two scheduled steam generator inspections)

VuRL = upper voltage repair limit (Note 1)

VLRL = lower voltage repair limit VMLRL = mid-cycle lower voltage repair limit based on VMuRL and time into cycle

~t = length of time since last scheduled inspection during which VuRL and VLRL were implemented Implementation of these mid-cycle repair limits should follow the same approach as in Specifications 5. 7.2.12.c.2.a through 5. 7.2.12.c.2.d.

Note 1: The upper voltage repair limit is calculated according to the methodology in GL 95-05 as supplemented. VuRL will differ at the tube support plates and flow distribution baffle."

• A new item 4 is added to TS 5.7.2.12.d, "Steam Generator (SG) Program," as follows:

"Indications left in service as a result of application of the tube support plate voltage-based repair criteria (Specification 5.7.2.12.c.2) shall be inspected by bobbin coil probe during all future refueling outages.

Implementation of the steam generator tube-to-tube support plate repair criteria requires a 100 percent bobbin coil inspection for hot-leg and cold-leg tube support plate intersections down to the lowest cold-leg tube support plate (including the FOB) with known ODSCC indications. The determination of the lowest cold-leg tube support plate intersections having ODSCC indications shall be based on the performance of at least a 20-percent random sampling of tubes inspected over their full length ."

• The following information is added to TS 5.9.9, "Steam Generator Tube Inspection Report":

"For implementation of the voltage-based repair criteria, in accordance with GL 95-05, to tube support plate (and flow distribution baffle) intersections, notify the NRC prior to returning the steam generators to service should any of the following conditions arise:

1. If estimated leakage based on the projected end-of cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution exceeds the leakage limit (determined from the licensing basis dose calculation for the postulated main steam line break) for the next operating cycle.

2. If circumferential crack-like indications are detected at the tube support plate intersection and flow distribution baffles.

3. If indications are identified that extend beyond the confines of the tube support plate and flow distribution baffles.

4. If indications are identified at the tube support plate elevations and flow distribution baffles that are attributable to primary water stress corrosion cracking .

5. If the calculated conditional burst probability based on the projected end-of-cycle (or if not practical , using the actual measured end-of-cycle) voltage distribution exceeds 1 x 10-2 , notify the NRC and provide an assessment of the safety significance of the occurrence."

2.3 .Applicable Regulatory Requirements General Design Criterion 14 of Appendix A to Title 10 Code of Federal Regulations (10 CFR)

Part 50 requires that the reactor coolant system pressure boundary 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. In addition, plant TSs specify the acceptance criteria (i.e., repair limits) for degraded SG tubes. The traditional strategy for achieving adequate

structural and leakage integrity of degraded tubes has been to establish a minimum wall thickness requirement in accordance with NRC Regulatory Guide 1.121, "Bases for Plugging Degraded PWR [Pressurized-Water Reactor] Steam Generator Tubes" (ADAMS Accession No. ML003739366). The minimum wall thickness requirement was developed assuming a uniform thinning of the tube wall. This assumed degradation mechanism is inherently conservative for certain forms of tube degradation. Conservative repair limits in the plant TS may lead to removing degraded tubes from service that may otherwise have adequate structural and leakage integrity for further service.

To reduce unnecessary conservatism in the minimum wall thickness requirement for certain degradation, the industry proposed voltage-based repair criteria for predominantly axially-oriented ODSCC confined within the thickness of the tube support plates. The staff published several conclusions regarding voltage-based repair criteria in draft NUREG-1477, "Voltage-Based Interim Plugging Criteria for Steam Generator Tubes," and in a draft GL titled "Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes." The latter document was published for public comment in the Federal Register on August 12, 1994 (59 FR 41520). On August 3, 1995, the staff issued GL 95-05 that considered public comments on the draft GL cited above, domestic operating experience under the voltage-based repair criteria, and additional data made available from European nuclear power plants.

GL 95-05 guidance does not set depth-based limits on predominantly axially oriented ODSCC at tube support plate locations. Instead it relies on empirically derived correlations between a nondestructive inspection parameter, the bobbin coil voltage, and tube burst pressure and leak rate. The staff recognizes that although the total tube integrity margins may be reduced following application of a voltage-based repair criteria, the guidance in GL 95-05 ensures structural and leakage integrity continue to be maintained at acceptable levels consistent with the requirements of 10 CFR Part 50 and 10 CFR Part 100. Since the voltage-based repair criteria do not require minimum tube wall thickness, tubes with through-wall cracks might remain in service. The staff included provisions for augmented SG tube inspections and restrictive operational leakage limits because of the increased likelihood of such flaws.

GL 95-05, in part, specifies the following for licensees:

(1) Repair criteria only applies to predominantly axially oriented ODSCC located within the bounds of the tube support plates.

(2) Perform an evaluation to confirm that the degraded steam generator tubes will retain adequate structural and leakage integrity from cycle to cycle.

(3) Adhere to specific inspection criteria to ensure consistency in methods between inspections.

(4) Periodically remove tubes from the steam generators for examination and destructive testing to verify the morphology of the degradation and provide burst and leakage data for structural and leakage integrity evaluations.

(5) Reduce the operational leakage limit in the plant TS.

(6) Implement an operational leakage monitoring program.

(7) Incorporate specific reporting requirements into the plant TS.

(8) Do not apply voltage-based repair criteria at locations where tubes with degradation could substantially deform or collapse during postulated loss-of-coolant-accident and safe shutdown earthquake loading.

Section 100.11 of 10 CFR, "Determination of exclusion area, low population zone, and population center distance," requires, in part, that an applicant determine:

(1) An exclusion area of such size that an individual located at any point on its boundary for two hours immediately following onset of the postulated fission product release would not receive a total radiation dose to the whole body in excess of 25 rem [roentgen equivalent man] or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.

(2) A low population zone of such size that an individual located at any point on its outer boundary who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage) would not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.

Appendix A to Part 50, General Design Criteria for Nuclear Power Plants, Criterion 19-Control room, states, in part, that adequate radiation protection shall be provided to permit access and occupancy of the control room (CR) under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident.

Section 50.36(c)(5),"Administrative controls," of 10 CFR specifies 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.

NUREG-0800, "Standard Review Plan (SRP)," Section 15.1.5, "Radiological Consequences of Main Steam Line Failures Outside Containment of a PWR," Revision 2, dated July 1981 (ADAMS Accession No. ML052350118), provides guidance to the NRC staff for the review of the radiological consequences of a main steam line break (MSLB) outside the containment.

SRP 15.1.5 states that the NRC reviewer should evaluate the proposed change such that the calculated whole-body and thyroid doses at the exclusion area and the low population zone outer boundaries shall not exceed: (a) the exposure guidelines as set forth in 10 CFR 100.11 for an MSLB with an assumed pre-accident iodine spike or for an MSLB with the highest worth control rod stuck out of the core and (b) 1O percent of these exposure guidelines, for an MSLB with an equilibrium iodine concentration in combination with an assumed accident-generated iodine spike.

3.0 TECHNICAL EVALUATION

The licensee stated in its May 14, 2018, license amendment request (LAR) that, except as described in Section 3.4.1 of the LAR, the voltage-based repair criteria for Watts Bar Unit 2 are consistent with the guidance of GL 95-05. The licensee provided responses to staff requests for additional information (RAI) and supplemented the LAR in its November 8, 2018, letter.

The staff's review of TVA's implementation of the alternate repair criteria is discussed below.

3.1 Tube Repair Limits The licensee's proposed repair criteria will do the following:

• Allow SG tubes, with degradation attributed to ODSCC within the bounds of the TSPs and FDB, with bobbin voltages less than or equal to 1.0 volt to remain in service.

• Require SG tubes, with degradation attributed to ODSCC within the bounds of the TSPs and FDB, with a bobbin voltage greater than 1.0 volt to be plugged or repaired, except for the cases described below:

o Steam generator tubes, with indications of potential degradation attributed to ODSCC within the bounds of the TSPs and FDB with a bobbin voltage greater than 1.0 volt but less than or equal to the upper voltage repair limit may remain in service if a rotating pancake coil (RPC) or acceptable alternative inspection does not detect degradation.

• Exclude certain intersections as defined in WAT-D-10709 from application of the voltage-based repair criteria as it is determined that these intersections may collapse or deform following a postulated loss of coolant accident and safe shutdown earthquake event.

• Require SG tubes, with indications of potential degradation attributed to ODSCC within the bounds of the TSPs and FDB, with a bobbin voltage greater than the upper voltage repair limit to be plugged or repaired.

• Allow the application of mid-cycle repair limits instead of the limits described above in the case of an unscheduled mid-cycle inspection.

The proposed lower voltage limit of 1.0 volt is based on the use of a correlation between the burst pressure and the bobbin coil voltage of pulled tube and model boiler data and is consistent with the recommended value specified in GL 95-05 for 3/4-inch SG tubing. The upper voltage limit is based on the lower 95 percent prediction interval of the burst pressure versus bobbin voltage correlation, adjusted for lower bound material properties evaluated at the 95-percent confidence level. The upper voltage limit is further reduced to account for uncertainty in the nondestructive examination technique and flaw growth over the next operating cycle . The industry periodically updates the database for burst pressure and bobbin voltage when the destructive test data from pulled tubes are available; therefore, the upper voltage limit may vary as additional data are incorporated into the database. The gap between the tube and tube hole in the flow distribution baffle plates (0.115 to 0.150 inches) is wider than the gap in the tube support plates (0.023 inches); therefore, the upper voltage limit for the tube indications found at the baffle plates is different than the upper repair limit for tube indications found at tube support plates. The staff finds that the proposed voltage limits are consistent with GL 95-05 and, therefore, acceptable.

3.2 Inspection Issues Section 3.c.3 of Attachment 1 to GL 95-05 gives guidance for probe wear. The licensee proposed to use an alternative to Section 3.c.3. The alternative approach (ADAMS Legacy Library Accession No. 9711100044), developed through the Nuclear Energy Institute (NEI),

specifies that if a probe does not satisfy the probe wear voltage variability criterion of plus or minus (+/-) 15 percent before its replacement, all tube locations inspected with the worn probe having measured indications with amplitudes greater than or equal to 75 percent of the lower voltage repair limit (i.e., 1.0 volt) must be reinspected with a bobbin probe that satisfies

+/- 15 percent probe wear voltage variability criterion. The voltages from the reinspection should be used as the basis for tube repair. The staff completed a review of NEl's proposed alternative method and concluded that the approach is acceptable as discussed in the NRC's letter of March 18, 1996, to Alex Marion, NEI (ADAMS Legacy Library Accession No. 9604090441 ).

Therefore, the NRC staff finds that the licensee's proposal to follow the industry approach to address probe wear is acceptable.

Industry laboratory and field studies supporting the alternative probe wear criteria showed that worn probe voltages are seldom less than 75 percent of the new probe voltage for all significant voltage levels. This is discussed in a letter from Alex Marion, NEI, to the NRC dated January 23, 1996 (ADAMS Legacy Library Accession No. 9711100044 ). However, in a 90-day inspection report for Byron Unit 1 dated September 9, 1996 (ADAMS Legacy Library Accession No. 9610070173), Commonwealth Edison compared the worn probe voltage to the new probe voltage and found that the worn probe voltage was substantially less than 75 percent of the new probe voltage for a few indications. Commonwealth Edison evaluated these indications and concluded that NEl's criteria to retest tubes with worn probe voltages above 75 percent of the repair limit is adequate and generally conservative due to the trend for worn probe voltages to exceed new probe voltages (as shown in the study). Comparison of the actual and projected end-of-cycle voltages did not show anything unusual attributable to the alternate probe wear criteria. The staff concluded that the aforementioned probe wear results did not indicate a need to modify the industry's probe wear criteria . However, the staff has and will continue to monitor probe wear in the licensees' 90-day inspection reports.

Regarding the probe variability guidance in Section 3.c.2 of GL 95-05, the licensee proposes to follow a methodology developed by NEI, which is described by letter dated October 15, 1996 (ADAMS Legacy Library Accession No. 9709050095). This methodology requires that the voltage responses from the primary frequency and mix frequency channels of new probes be within +/-10 percent of the nominal voltage responses when voltages are normalized to the 20-percent flaw values. The nominal voltage responses were established as the average voltages obtained from the American Society of Mechanical Engineers standard drilled hole flaws for at least 10 production probes. The NEI methodology was originally submitted by letter dated January 23, 1996 (ADAMS Legacy Library Accession No. 9711100044), accepted by NRC letter dated March 18, 1996 (ADAMS Legacy Library Accession No. 9604090441 ), and supplemented with additional industry information in the October 15, 1996 letter. Therefore, based in the NRC's previous acceptance of the approach, the staff finds the licensee's proposal to follow the industry approach to address probe variability acceptable.

Section 3.b.3 of GL 95-05 specifies that all dented tube support plate intersections with bobbin voltage> 5.0 volts should be inspected with RPC. If circumferential cracking or primary water stress corrosion cracking indications are detected in these dented intersections, it may be necessary to expand the RPC sampling plan to include dented intersections with bobbin voltage < 5.0 volts. The licensee clarified in its November 8, 2018, response to RAI 1 that the

dent signals exceeding 5.0 volts coincident with crack indications confirmed by RPC will be plugged. For dented intersections with bobbin voltages~ 2.0 volts, the licensee plans to inspect them using a +Point probe or other qualified eddy current probe. The NRC staff finds the licensee's response acceptable since it is consistent with Section 3.b.3 of GL 95-05 and will require any indications exceeding 5.0 volts coincident with crack indications to be plugged.

In its November 8, 2018, response to RAI 2, the licensee identified expansion criteria for inspections of dented intersections should certain criteria be met. Specifically, the licensee stated that if circumferential cracking is identified at a ding/dent of magnitude between 3.0 and 5.0 volts, a +Point probe examination (or other qualified technique) at hot leg dented intersections~ 2.0 volts (as determined by bobbin examination) will be performed. If circumferential cracking is identified at a ding/dent of magnitude between 2.0 and 3.0 volts, a

+Point probe examination (or other qualified technique) at hot leg dented intersections

~ 1.0 volts will be performed. In addition, the licensee noted that circumferential cracking at dings/dents is not expected based on operating experience with the original SGs. The NRC staff finds the licensee's inspection plan for dented intersections acceptable because it considers the structural integrity of the tubes at the low-voltage dented intersections.

3.3 Structural and Leakage Integrity Assessments The guidance in GL 95-05 for the voltage-based repair criteria focuses on maintaining tube structural integrity during the full range of normal, transient and postulated accident conditions with adequate allowance for eddy current test uncertainty and flaw growth projected to occur during the next operating cycle. Technical Specification 5.7.2.12.b.1 specifies that a safety factor of 3.0 against burst under normal operating conditions and 1.4 against burst under accident conditions be maintained for in-service SG tubes. Because GL 95-05 addresses tubes affected with ODSCC confined to within the thickness of the tube support plate during normal operation, the NRC staff concluded that the structural constraint provided by the tube support plate ensures all tubes to which the voltage-based criteria apply will retain a margin of 3 with respect to burst under normal operating conditions . For a postulated MSLB accident, however, the tube support plate may displace axially during SG blowdown such that the ODSCC affected portion of the tubing may no longer be fully constrained by the tube support plate. Accordingly, it is appropriate to consider the ODSCC-affected regions of the tubes as freestanding tubes for the purpose of assessing burst integrity under postulated MSLB conditions.

In order to confirm the structural and leakage integrity of the tube until the next scheduled inspection, GL 95-05 specifies a methodology to determine the conditional burst probability and the total primary-to-secondary leak rate from an affected SG during a postulated MLSB event.

To complete GL 95-05 prescribed assessments, the licensee proposes to apply the methodology documented in WCAP-14277, Revision 1, "SLB [steam line break] Leak Rate and Tube Burst Probability Analysis Methods for ODSCC at TSP Intersections." The staff finds this to be acceptable because the methodology complies with GL 95-05 guidelines and the staff has approved similar voltage-based ARC amendments for several other nuclear power plant units, including Watts Bar Unit 1 Amendment No. 38, dated February 26, 2002 (ADAMS Accession No. ML020590277).

GL 95-05 specifies that structural and leakage integrity assessments should use the latest available database from destructive examinations of tubes removed from Westinghouse-designed SGs. The industry and NRC established a protocol to formalize the requirements for periodically updating the industry database. The licensee stated that it will follow the NRG-approved industry protocol for updating the database. In response to RAI 3, the

licensee revised proposed TS 5.9.9 to incorporate a requirement to submit the GL 95-05 90-day reports and confirmed that each of the GL 95-05 90-day reports will identify the latest revision of NP-7480-L, "Steam Generator Tubing ODSCC at TSP Database for Alternate Repair Limits,"

used for GL 95-05 specified calculations. The information provided in each of the 90-day reports, including the database used, will enable the staff to assess the effectiveness of the methodology and perform confirmatory calculations. The staff finds the licensee's proposal to use the NRC-approved database to perform structural and leakage assessments and to follow the NRC-approved industry protocol for updating the database acceptable because, consistent with GL 95-05, they will use the latest available database to perform structural and leakage integrity assessments and the effectiveness of the methodology will be confirmed by the staff for each of the 90-day reports.

Guidance in GL 95-05, Section 1.b.1, states that the voltage-based repair criteria will not be applied at locations where tubes with degradation could substantially deform or collapse during postulated LOCA and SSE loading . Steam generator tubes can potentially permanently deform at TSP intersections under such loading conditions. The NRC staff has previously reviewed and

~pproved the analytical methodology to determine tube deformation or collapse in previous similar applications at other plants including Watts Bar Unit 1. The licensee used an identical methodology for Watts Bar Unit 2 to determine TSP intersections where tube deformations could potentially result in tube collapse during the combined LOCA and SSE event. The tube intersections excluded for Watts Bar Unit 2 are the same as those excluded for the Watts Bar Unit 1 original SGs. The licensee provided the NRC with a summary of the number of intersections at each TSP that are excluded due to combined LOCA and SSE loadings. The licensee determined that it will not apply the voltage-based repair criteria to 466 intersections (out of approximately 100,588 total intersections) on this basis. The top TSP (H08, C14) contains the largest number of excluded intersections (256), which is approximately 2. 7 percent of the total intersections for that TSP. Excluded tubes are clustered around wedge locations. A large number of the affected tubes contain multiple excluded intersections, both at multiple plate elevations and at both hot and cold-leg intersections. Hence, the total number of tubes affected is less than the number of affected intersection locations. At each of these excluded intersections where tubes with degradation could potentially collapse during a combined LOCA and SSE event, the licensee will not be allowed to apply the voltage-based repair criteria. The NRC staff, therefore, finds the licensee's analytical methodology acceptable.

3.3.1 Conditional Probability of Burst In accordance with GL 95-05, the licensee will perform probabilistic analyses, using the methodology described in Revision 1 of WCAP-14277, to quantify the potential for SG tube ruptures given an MSLB event. The licensee will compare the results of the probabilistic analyses to a threshold value of 1x10-2 per cycle in accordance with GL 95-05 and will notify the NRC of any calculated conditional probability of burst exceeding 1x10-2 . This threshold value assures that the probability of burst is acceptable considering calculation assumptions and the results of the staff's generic risk assessment for SGs contained in NUREG-0844, "NRC Integrated P_rogram for the Resolution of Unresolved Safety Issues A-3, A-4, and A-5 Regarding Steam Generator Tube Integrity." Failure to meet the threshold value indicates that ODSCC confined to within the thickness of the TSP could contribute a significant fraction to the overall conditional probability of tube rupture from all forms of degradation assumed and evaluated as acceptable in NUREG-0844. The NRC staff concludes the licensee's proposed methodology for calculating the conditional burst probability is consistent with the guidance in GL 95-05 and is acceptable.

3.3.2 Accident Leakage The licensee will use the methodology described in Revision 1 of WCAP-14277 to calculate the SG tube leakage from the faulted SG during a postulated MSLB event. The methodology consists of the following two major components:

• a model predicting the probability that a given indication will leak as a function of voltage (i.e., the probability of leakage model)

• a model predicting leak rate as a function of voltage, given that leakage occurs (i.e., the conditional leak rate model)

The staff concludes that the licensee's proposed methodology for calculating the tube leakage is consistent with the guidance in GL 95-05 and is acceptable.

3.3.3 Primary-to-Secondary Leakage The voltage-based repair criteria would allow degraded tubes to remain in service. Therefore, the degraded tubes may develop through-wall cracks which may leak during normal operation, transients, or postulated accidents. As a defense-in-depth measure, GL 95-05 specifies that the operational leakage limits of the plant TSs be limited to 150 gpd from any one SG. The licensee's current TSs specify 150 gpd primary-to-secondary leakage through any one SG. In addition, the licensee is following the latest revision of the "PWR Primary-to-Secondary Leak Guidelines." The staff concludes that the licensee's operational leakage limit of 150 gpd for Watts Bar Unit 2 is consistent with GL 95-05 and, therefore, is acceptable.

3.4 Degradation Monitoring To confirm the nature of the degradation at the TSP elevations, licensees periodically remove tubes from the SGs for destructive testing. The test data from removed tubes is used to do the following:

• confirm that the nature of the degradation observed at these locations is predominantly axially oriented ODSCC

• monitor the degradation mechanism over time

• provide data to supplement existing databases

• assess inspection capability Section 4 of GL 95-05 specifies that two tube specimens be removed from SGs with the objective of retrieving as many intersections as is practical (minimum of four) during the plant SG inspection outage preceding initial application of these criteria . On an ongoing basis, additional (minimum of two) tube specimens should be removed at the first refueling outage following 34-effective full-power months of operation or at the maximum interval of three refueling outages after the previous tube pull, whichever is shorter.

The licensee proposed an alternative tube pull program to the above GL 95-05 guidelines. The licensee's May 14, 2018, submittal stated that it will implement the ARC criteria using alternate methods utilizing eddy current techniques as a substitute for tube removal testing.

3.4.1 Confirming Axially-Oriented ODSCC as the Dominant Degradation Mechanism In order to confirm axial ODSCC as the dominant degradation mechanism, the licensee proposes to utilize bobbin probe and +Point probe or qualified eddy current inspection techniques. The licensee will qualify these techniques in accordance with Examination Technique Specification Sheets Appendix I of the latest revision of Electric Power Research Institute (EPRI) Report 3002007572, "Steam Generator Management Program: Pressurized Water Reactor Steam Generator Examination Guidelines." 2 The licensee states that because there have been no pulled tubes for which the ODSCC crack morphology differs from that found in the initial Electric Power Research Institute (EPRI) database and tubes pulled with RPC axial ODSCC calls have had morphologies consistent with the EPRI database, it is reasonable to conclude that the eddy current techniques used for Watts Bar Unit 2 can distinguish between axial and circumferential ODSCC and determine whether the axial cracks are confined to within the TSP thickness. Since the licensee's eddy current inspection techniques will be qualified per Appendix I and operating experience for ODSCC at TSPs indicates that the crack morphologies have been reasonably predictable, the NRC staff concludes that there is reasonable assurance that the licensee will be able to distinguish between axial and circumferential ODSCC with the eddy current techniques used at Watts Bar Unit 2.

3.4.2 Monitoring Degradation Over Time To monitor the degradation over time, the licensee proposes to utilize the same eddy current inspection techniques used to confirm axial ODSCC. Using these eddy current techniques, the licensee will be able to track and trend degradation in the SGs between each inspection.

Therefore, the NRC staff concludes that the licensee will be able to monitor degradation over time at the TSP elevations.

3.4.3 Supplementing Existing Databases In the LAR, the licensee describes the reasons why supplementing existing databases is of little value. The licensee states that the experience and data available in NP-7480-L, Addendum 7, is such that providing additional data is not necessary. The licensee states that because the database of leaking tubes is smaller and consequently the leak-rate-to-voltage correlation has been more prone to significant changes, the primary objective should be to increase the database for leak rate. This would require a tube pull for bobbin voltages of greater than 2.0 volts for 3/4-inch tubes to ensure a significant likelihood of leakage. The licensee states that there is a possibility that a tube will be pulled with the desired voltage, but the data will likely not be utilized since there are plans to replace the SGs at Watts Bar Unit 2 after the next several years. The existing databases contain data from plants utilizing voltage-based alternate repair criteria in accordance with GL 95-05 since its issuance more than 20 years ago. Because of the abundance of data in the existing databases and the likelihood that any supplemental data will not be used due to plans to replace the Watts Bar, Unit 2 SGs, the NRC staff concludes that supplementing existing databases is of little value.

2 EPRI Report 3002007572 is proprietary. An abstract of the report is available at https://www.epri.com/.

3.4.4 Assessing Inspection Capability The licensee states that the assessment of the inspection capability has already been completed by previous plants using the voltage-based ARC. These plants also contributed to the database of voltage correlations to burst pressure, probability of leakage, and conditional leak rates. Given the size of the EPRI database used for supporting voltage correlations, the NRC staff concludes that pulling of tubes to assess reliability of inspection methods and to supplement existing databases is of little value and the licensee's proposed alternative to utilize eddy current techniques as a substitute for tube removal testing is acceptable.

3.5 Assessment of Radiological Consequences GL 95-05 has established that a postulated MSLB outside of containment represents the most limiting radiological condition relative to the alternate voltage-based repair criteria for axial ODSCC. As part of this application, the licensee revised their MSLB dose consequence analysis. The following safety evaluation is limited to the evaluation of the licensee's changes to the Watts Bar Unit 2 current licensing basis (CLB) MSLB dose consequence analysis.

The NRC staff evaluated the licensee's application to determine whether the proposed change is consistent with the guidance, regulations, and licensing information discussed in Section 2.0 of this safety evaluation. In Attachment 1, "Main Steam Line Break Dose Calculation," to of the LAR, the licensee provided a description of the revised MSLB dose consequence analysis including all the pertinent assumptions, input data and results. In determining whether an amendment to a license will be issued, the NRC staff is guided by the considerations that govern the issuance of initial licenses to the extent applicable and appropriate. Accordingly, the NRC staff evaluated the following proposed changes to the licensee's CLB MSLB dose consequence analysis as identified in Attachment 1 of the LAR:

• An increase in the primary to secondary leakage value.

• An increase in the CR isolation delay time.

• A decrease in reactor coolant system (RCS) weight [mass].

• Revisions to the primary and secondary coolant concentrations.

3.5.1 Assumed Primary to Secondary Leakage Value In support of implementation of the ARC being requested in this LAR, the licensee proposes to increase the assumed primary to secondary leakage value for a faulted SG from 1 gpm to 4 gpm for a post-accident leak through the faulted SG to the environment. This change results in an increase to the licensee's calculated MSLB dose consequence analysis. However, since the MSLB dose consequences remain below the regulatory acceptance criteria, the NRC staff concludes that this change to the CLB MSLB analyses dose consequence analysis is acceptable.

3.5.2 Control Room Isolation Delay Time In order to correct an error found in the determination of the time constant, the licensee has proposed to increase the CR isolation delay time from 40 seconds to 74 seconds. As

documented on page E1-26 of the TVA letter dated December 20, 2017, Application to Revise Watts Bar Unit 2 Technical Specification 4.2.1, "Fuel Assemblies, " and Watts Bar Units 1 and 2 Technical Specifications Related to Fuel Storage (WBN-TS-17-028) (ADAMS Accession No. ML173548282), the previous time constant of 7 .17 x 10-3 minutes that was used to determine the rate meter response time is appropriate for a count rate between 1 x 104 and 1 x 105 counts per minute (cpm). Because the setpoint for the CR intake monitors is 400 cpm, the licensee has determined that a time constant of 4.34 x 10-1 minutes should have been used.

This resulted in an increase in the rate meter response time from 0.86 seconds to 52.08 seconds, which was rounded to 60 seconds. The isolation damper response time of 14 seconds remained unchanged resulting in a total CR isolation delay time of 74 seconds.

The correction to the time constant and the associated increase in the CR isolation delay time results in an increase to the licensee's calculated CR dose. However, since the CR dose remains below the regulatory acceptance criteria, the NRC staff concludes that this change to the CLB MSLB analyses dose consequence analysis is acceptable.

3.5.3 Decrease in Reactor Coolant System Mass As a result of several corrections in the calculation of RCS mass, the licensee decreased the RCS mass from 2.45 x 1as grams to 2.14 x 1as grams. As documented on page E1-23 of the December 20, 2017, TVA letter referenced in Section 3.5.2 above, the RCS concentrations have changed since the original review of the Watts Bar Unit 2 operating license due to the following corrections:

• The RCS volume was corrected from 11375 cubic feet to 12708.4 cubic feet.

• The specific volume previously used to determine the RCS weight was based on a temperature outside the normal operating range.

• The mass of RCS water previously included the volume of vapor space in the pressurizer.

• The mass of water in the SG previously included the mass of water in the primary side of the SG instead of just the secondary side .

.* The condensate demineralizer was previously assumed to be in operation but is not typically used and should not have been cred ited .

The NRC staff concludes that these changes are enhancements to the accuracy of the licensee's MSLB dose consequence analysis and are, therefore, acceptable.

3.5.4 Revisions to Primary and Secondary Coolant Concentrations In addition to the corrections that resulted in a reduction in RCS mass, the licensee also updated the power level used in the dose consequence analyses consistent with the CLB. The NRC staff notes that both the reduction in RCS mass discussed in Section 3.5.3 and the change in assumed power level will affect the calculation of coolant concentrations. As documented on page E1-23 of the December 20, 2017, TVA letter referenced in Section 3.5.2 above, the assumed power level for this analysis had been 3565 megawatt thermal (MWt), which is 104.5 percent of the licensed power level. The NRC staff notes that, typically, dose

consequences are analyzed using a power level that is adjusted upward to account for the uncertainty in the measurement of power level. This uncertainty is referred to as the calorimetric error, which is the error assumed in the determination of core thermal power as obtained from secondary plant measurements. Historically, licensees have been required to include a margin of 2 percent in the calculation of the reactor thermal power to account for instrumentation uncertainty. As stated in Watts Bar UFSAR Section 15.1.2.2, "Initial Conditions," for Unit 2 accident evaluations the core power allowance for calorimetric error is 2 percent. Therefore the licensee revised the power level used in the calculations of primary and secondary concentrations to 3480 MWt, which is 102 percent of the licensed power level.

The NRC staff concludes that this change aligns the assumed power level used in the licensee's dose consequence analyses with the appropriate calorimetric uncertainty and is, therefore, acceptable.

3.6 Evaluation of Technical Specification Changes The NRC staff reviewed the proposed changes to TSs 5.7.2.12 and 5.9.9 and determined that the Steam Generator Program continues to meet the requirements of 10 CFR 50.36(c)(5).

The proposed changes to the TSs are consistent with the guidance in GL 95-05. TSs 5.7.2.12 and 5.9.9 will continue to contain the provisions necessary to ensure that SG tube integrity is maintained. Therefore, the NRC staff finds the proposed TS changes are acceptable.

3.7 Summary of Technical Conclusions The staff has reviewed licensee's proposed amendment to implement the voltage-based repair criteria described in GL 95-05 for SG tubes in the TSs for Watts Bar Unit 2. The staff concludes that the proposed alternate repair criteria are consistent with GL 95-05 and are acceptable. The staff also concludes that adequate structural and leakage integrity of SG tubing can be assured, consistent with 10 CFR Part 50 requirements, for indications to which the voltage-based repair criteria will be applied. The staff approves the proposed voltage-based repair criteria based, in part, on the licensee being able to successfully

• demonstrate after each inspection outage (as shown in its 90-day SG tube inspection report) that the conditional probability of burst and the primary-to-secondary leakage during a postulated MSLB will be acceptable per the guidance in GL 95-05 or otherwise notifying the NRC per TS 5.9.9 prior to returning the steam generators to service should conditional probability of burst or the primary-to-secondary leakage during a postulated MSLB be found unacceptable. On the basis of the staff conclusions, the licensee may incorporate the proposed alternate repair criteria into the TSs for Watts Bar Unit 2.

The NRC staff has evaluated the impact of the proposed change on the design-basis MSLB radiological dose consequence analysis against the regulatory requirements and guidance identified in Section 2.0 of this SE. The NRC staff finds, with reasonable assurance that the licensee's dose consequence analysis will continue to comply with the requirements of 10 CFR 100.11 as well as the accident-specific acceptance criteria specified in SRP Section 15.1.5. Therefore, the proposed change is acceptable with regard to the dose consequences of the MSLB accident.

The proposed changes to TSs 5.7.2.12 and 5.9.9 continue to meet the requirements of 10 CFR 50.36(c)(5) and are consistent with the guidance in GL 95-05.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Tennessee State official was notified of the proposed issuance of the amendment on March 4, 2019. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component 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 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 November 20, 2018 (83 FR 58618). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51 .22(c)(9). 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.

6.0 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 Commission's 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.

Principal Contributors: Alan Huynh John Parillo Date: June 3, 2019

J. Shea

SUBJECT:

WATTS BAR NUCLEAR PLANT, UNIT 2- ISSUANCE OF AMENDMENT REGARDING APPLICATION TO REVISE TECHNICAL SPECIFICATIONS FOR USE OF VOLTAGE-BASED ALTERNATE REPAIR CRITERIA IN ACCORDANCE WITH GENERIC LETTER 95-05 (EPID L-2018-LLA-0143)

DATED JUNE 3, 2019 DISTRIBUTION:

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NAME RSchaaf BClayton SBloom DATE 3/18/2019 3/14/2019 1/23/2019 OFFICE NRR/DRA/ARCB/BC* NRR/DSS/STSB/BC** OGC-NLO**

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