CNL-18-003, Application to Revise Watts Bar Nuclear Plant Unit 2 Technical Specifications for Use of voltage-based Alternate Repair Criteria in Accordance with Generic Letter 95-05 (391-WBN2-TS-17-30)

Application to Revise Watts Bar Nuclear Plant Unit 2 Technical Specifications for Use of voltage-based Alternate Repair Criteria in Accordance with Generic Letter 95-05 (391-WBN2-TS-17-30)

ML18138A232
Person / Time
Site:	Watts Bar
Issue date:	05/14/2018
From:	Henderson E Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
391-WBN2-TS-17-30, CNL-18-003, TAC MA8635
Download: ML18138A232 (50)

Text

Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 CNL-18-003 May 14, 2018 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Watts Bar Nuclear Plant Unit 2 Facility Operating License No. NPF-96 Docket No. 50-391

Subject:

Application to Revise Watts Bar Nuclear Plant Unit 2 Technical Specifications for Use of voltage-based Alternate Repair Criteria in Accordance with Generic Letter 95-05 (391-WBN2-TS-17-30)

Reference NRC letter to TVA, Watts Bar Nuclear Plant, Unit 1 - Issuance of Amendment Regarding Steam Generator Tubing Alternate Repair Criteria (ARC) (TAC No. MA8635), dated February 26, 2002 (ML020590277)

In accordance with the provisions of Title 10 of the Code of Federal Regulations (10 CFR) 50.90, "Application for amendment of license, construction permit, or early site permit," Tennessee Valley Authority (TVA) is submitting a request for an amendment to Facility Operating License No. NPF-96 for Watts Bar Nuclear Plant (WBN), Unit 2.

The proposed license amendment request (LAR) revises WBN Unit 2 Technical Specifications (TS) 5.7.2.12, Steam Generator (SG) Program, and TS 5.9.9, Steam Generator Tube Inspection Report, to use the voltage-based alternate repair criteria (ARC) specified in the guidelines contained in Generic Letter (GL) 95-05, "Voltage-Based Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking." The use of the voltage-based ARC was previously approved by the Nuclear Regulatory Commission (NRC) for WBN Unit 1 in the referenced letter.

The purpose of GL 95-05 is to give guidance to licensees who may wish to request a license amendment to the plant technical specifications to implement alternate steam generator tube repair criteria applicable specifically to outside diameter stress corrosion cracking (ODSCC) at the tube-to-tube support plate intersections in Westinghouse designed steam generators having drilled-hole tube support plates (TSPs) and alloy 600 steam generator tubing. The implementation of the ARC in GL 95-05 ensures tube integrity and prevents unnecessary plugging of steam generator tubes with indications of ODSCC at tube support plate intersections due to the uncertainty related to depth estimation of ODSCC at

U.S. Nuclear Regulatory Commission CNL-18-003 Page 2 May 14, 2018 tube support plates. As noted in Section 3.4.1 of Enclosure 1, WBN Unit 2 will be the last alloy 600 mill annealed (MA) plant to implement these voltage-based repair criteria; therefore, there is adequate experience and data to use the ARC without a tube pull. provides a description and technical evaluation of the proposed change, a regulatory evaluation, and a discussion of environmental considerations. Attachment 1 to provides a summary of the main steam line break dose calculation that supports this LAR. Attachment 2 to Enclosure 1 provides the existing WBN Unit 2 TS pages marked-up to show the proposed changes. Attachment 3 to Enclosure 1 provides the existing WBN Unit 2 TS pages retyped to show the proposed changes. In support of, Enclosure 2 contains Westinghouse Electric Company (WEC)

SG-SGMP-13-16-NP, Revision 1, "Watts Bar Nuclear Plant Unit 2 Applicability of GL 95-05 Voltage Base Alternate Repair Criteria."

The WBN Plant Operations Review Committee has reviewed this proposed change and determined that operation of WBN Unit 2 in accordance with the proposed change will not endanger the health and safety of the public.

TVA has determined that there are no significant hazard considerations associated with the proposed change and that the change qualifies for a categorical exclusion from environmental review pursuant to the provisions of 1 O CFR 51.22(c)(9). Additionally, in accordance with 10 CFR 50.91 (b)(1 }, TVA is sending a copy of this letter and the enclosures to the Tennessee Department of Environment and Conservation.

TVA requests approval of the proposed license amendment within one year from the date of this letter with implementation within 60 days following NRC approval to support planning activities for the WBN Unit 2 Fall 2020 Refueling Outage (U2R3). contains the new regulatory commitment associated with this submittal. Please address any questions regarding this request to Edward D. Schrull at (423) 751-3850.

I declare under penalty of perjury that the foregoing is true and correct. Executed on this 14th day of May 2018.

Respectfully,

~ct/~

Erin K. Henderson Director, Nuclear Regulatory Affairs Enclosures cc: See Page 3

U.S. Nuclear Regulatory Commission CNL-18-003 Page 3 May 14, 2018

Enclosures:

1. Evaluation of Proposed Change

2. SG-SGMP-13-16-NP, Revision 1, Watts Bar Nuclear Plant Unit 2 Applicability of GL 95-05 Voltage Base Alternate Repair Criteria, May 2015

3. Regulatory Commitment cc (Enclosures):

NRC Regional Administrator - Region II NRC Project Manager - Watts Bar Nuclear Plant NRC Senior Resident Inspector - Watts Bar Nuclear Plant Director, Division of Radiological Health - Tennessee State Department of Environment and Conservation CNL-18-003 E1-1 of 31 Evaluation of Proposed Change

Subject:

Application to Revise Watts Bar Nuclear Plant Unit 2 Technical Specifications for Use of voltage-based Alternate Repair Criteria in Accordance with Generic Letter 95-05 (391-WBN2-TS-17-30)

CONTENTS 1.0

SUMMARY

DESCRIPTION............................................................................................. 2 2.0 DETAILED DESCRIPTION.............................................................................................. 2 2.1 Proposed Changes...................................................................................................... 2 2.2 Condition Intended to Resolve..................................................................................... 5 2.3 Basis for the Proposed Technical Specification Changes............................................. 6

3.0 TECHNICAL EVALUATION

............................................................................................. 6 3.1 System Description...................................................................................................... 6 3.2 Summary of voltage-based Repair Criteria.................................................................. 7 3.3 Acknowledgment of GL 95-05 Performance Requirements.......................................... 8 3.4 Application................................................................................................................... 8 3.4.1 Alternative to Section 4 of Attachment 1 of GL 95-05...........................................11 3.5 Tube Integrity Evaluation............................................................................................12 3.6 Reporting Requirements.............................................................................................16 3.7 Conclusion..................................................................................................................16

4.0 REGULATORY EVALUATION

.......................................................................................16 4.1 Applicable Regulatory Requirements and Criteria.......................................................16 4.2 Precedent...................................................................................................................18 4.3 Significant Hazards Consideration..............................................................................18 4.4 Conclusion..................................................................................................................21

5.0 ENVIRONMENTAL CONSIDERATION

..........................................................................21

6.0 REFERENCES

...............................................................................................................22 ATTACHMENTS

1.

Main Steam Line Break Dose Calculation

2.

Proposed TS Changes (Mark-Ups) for WBN Unit 2

3.

Proposed TS Changes (Final Typed) for WBN Unit 2 CNL-18-003 E1-2 of 31 1.0

SUMMARY

DESCRIPTION Pursuant to Title 10 of the Code of Federal Regulations (CFR) §50.90, Tennessee Valley Authority (TVA) is submitting a request for a change to Facility Operating License (OL) No. NPF-96 for Watts Bar Nuclear Plant (WBN) Unit 2.

Specifically, TVA is requesting a license amendment to amend the WBN Unit 2 Technical Specifications (TS) 5.7.2.12, Steam Generator (SG) Program, and TS 5.9.9, Steam Generator Tube Inspection Report, to allow the use of voltage-based alternate repair criteria (ARC) specified in the guidelines contained in Generic Letter (GL) 95-05 (Reference 1). This license amendment request (LAR) is technically based on a previous LAR for WBN Unit 1 (Reference 2) to use the voltage-based ARC that was approved by the NRC in Reference 3. The technical justification for Reference 2 was based on excerpts from Westinghouse Electric Company (WEC) WAT-D-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 ODSCC at TSPs (Reference 4).

The proposed amendment revises the WBN Unit 2 TS 5.7.2.12 and TS 5.9.9 tube repair criteria to use the voltage-based ARC specified in the guidelines contained in GL 95-05.

Reference 4 has been supplemented by WEC SG-SGMP-13-16, Revision 1, Watts Bar Nuclear Plant Unit 2 Applicability of GL 95-05 Voltage-Based Alternate Repair Criteria.

A copy of SG-SGMP-13-16, Revision 1 is included in Enclosure 2. The offsite dose calculations are consistent with GL 95-05 recommendations. A summary of the main steam line break (MSLB) dose analysis inputs and results are provided in Attachment 1 to this enclosure.

2.0 DETAILED DESCRIPTION

2.1 PROPOSED CHANGE

S The following is a detailed description of the proposed WBN Unit 2 TS changes. The proposed TS changes are consistent with the guidance in GL 95-05.

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 CNL-18-003 E1-3 of 31 axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates and flow distribution baffles (FDB). At tube support plate intersections and FDB, 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 FDB 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 FDB 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 FDB 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-D-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 ODSCC at TSPs,"

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 + 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 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 FDB 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.

CNL-18-003 E1-4 of 31 The mid-cycle repair limits are determined from the following equations:

VMURL= ___________VSL________

1.0+NDE + Gr[(CL-t)/CL]

VMLRL = VMURL-(VURL-VLRL)[(CL-t)/CL]

where:

VMURL =mid-cycle upper voltage repair limit based on time into cycle VSL = structural limit voltage NDE = 95-percent cumulative probability allowance for nondestructive examination uncertainty (i.e., a value of 20 percent has been approved by NRC). The NDE 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 being 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 FDB) 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.

CNL-18-003 E1-5 of 31 The following information is being 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. to this enclosure provides the existing WBN Unit 2 TS pages marked-up to show the proposed changes. Attachment 3 to this enclosure provides the existing TS pages retyped to show the proposed changes.

2.2 CONDITION INTENDED TO RESOLVE TVA proposes to change the WBN Unit 2 TS to reduce the need for plugging SG tubes having indications that exceed the current TS depth-based plugging limit. TVA proposes to add alternate voltage-based tube repair criteria at tube support plates (TSP) intersections and flow distribution baffle (FDB) plate intersections that are voltage-based and maintain structural and leakage integrity of tubes with indications of outside diameter stress corrosion cracking (ODSCC) within the confines of the TSP and FDB regions.

The proposed change preserves the reactor coolant flow margin and reduces the radiation exposure incurred in the process of plugging the SG tubes (approximately 0.060 person-rem per tube of exposure would be saved for a plugging operation). Other benefits of not plugging TSP indications that meet the proposed ARC include a reduction in person-hours associated with refueling outages.

The proposed change is based on the voltage-based ARC per GL 95-05 criteria for WBN Unit 2 based on the operating experience of the WBN Unit 1 SGs. Based on the WBN 1 operating experience, axial ODSCC at TSP intersections is a potential degradation mechanism. The implementation of the ARC could prevent unnecessary plugging of SG tubes with indications of ODSCC at TSP intersections due to the uncertainty related to depth estimation of ODSCC at TSPs.

CNL-18-003 E1-6 of 31 2.3 BASIS FOR THE PROPOSED TECHNICAL SPECIFICATION CHANGES The following provides the basis for each of the proposed TS changes in Section 2.1 of this enclosure:

The basis for the proposed changes to TS 5.7.2.12.b.2 are as follows:

o The change to state that leakage is not to exceed one gpm for the faulted SG loop and 150 gpd for each unfaulted SG is consistent with the analysis in the WBN dual-unit updated final safety analysis report (UFSAR)Section 15.5.4, which states:

An acceptable primary to secondary leakage rate for the main steam line break (MSLB) accident is 1 gallon per minute (gpm) for the faulted steam generator loop and 150 gallons per day (gpd) for each unfaulted steam generator.

o The change to state that 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 four gpm for the faulted SG loop and 150 gallons per day (gpd) for each unfaulted SG is consistent with the analysis in to this enclosure. Attachment 1 to this enclosure determined the maximum permissible SG primary to secondary leak rate during a steam line break for WBN Unit 2. The analysis in Attachment 1 to this enclosure demonstrates that a four-gpm primary to secondary leakage in the faulted SG would result in site boundary doses within 10 CFR 100 guidelines and control room doses within the GDC 19 limit.

The proposed changes to TS 5.7.2.12.c.2, 5.7.2.12.d, and 5.9.9 are consistent with the recommended TS changes in GL 95-05 and the TS changes approved by the NRC for WBN Unit 1 in Reference 2.

3.0 TECHNICAL EVALUATION

3.1 SYSTEM DESCRIPTION WBN Unit 2 contains four Westinghouse Model D3 recirculating pre-heater type SGs.

Each SG contains 4674 mill annealed (MA) Alloy 600 tubes that have an outer diameter of 0.75 inches with a 0.043-inch nominal wall thickness. These SGs are the same design as the original WBN Unit 1 SGs. The GL 95-05 ARC for WBN Unit 1 was implemented prior to startup following the WBN Unit Cycle 4 outage. The WBN Unit 1 SGs were replaced during the WBN Unit 1 Cycle 7 (U1C7) refueling outage. Two tubes on WBN Unit 1 were pulled, in accordance with GL 95-05 ARC during U1C5, for removal and laboratory non-destructive and destructive examinations to support the implementation of the ARC (References 5 and 6).

The WBN Unit 2 SGs have a vertical shell and U-tube evaporator with integral moisture separating equipment. The reactor coolant flows through the inverted U-tubes, entering and leaving through the nozzles located in the hemispherical bottom head of the SG.

The head is divided into inlet and outlet chambers by a vertical partition plate extending from the head to the tubesheet. Steam is generated on the shell side and flows upward through the moisture separators to the outlet nozzle at the top of the vessel. Details of the Unit 2 SGs are described in the WBN dual-unit UFSAR Section 5.5.2.2 and UFSAR Figure 5.5-3.

CNL-18-003 E1-7 of 31 The WBN Unit 2 SGs contain an FDB plate located approximately eight inches above the top of the tube sheet. The tube holes located in the FDB design include an increased nominal tube-to-plate diametrical gap ranging from approximately 0.115 inches to 0.150 inches, compared to 0.023 inches nominal gap at the TSPs. Based on this increased tube to plate gap at the FDB, an upper voltage repair limit based on providing tube structural integrity against a pressure equivalent of three times the normal primary-to-secondary tube differential pressure is provided. The design features of the WBN Unit 2 SGs are consistent with the scope of applicability of GL 95-05.

Materials of construction for the Unit 2 SG are provided in UFSAR Table 5.2-8.

Materials are selected and fabricated in accordance with the requirements of the ASME Code Section III. The heat transfer tubes and the divider plate are Inconel and the interior surfaces of the reactor coolant channel heads and nozzles are clad with austenitic stainless steel. The primary side of the tubesheet is weld clad with Inconel.

The tubes are roller expanded for the full depth of the tubesheet after the ends are seal welded to the tubesheet cladding.

Tube and tubesheet stress analyses of the SG, which are discussed in UFSAR Section 5.2, confirm that the SG tubesheet will withstand the loading caused by loss of reactor coolant.

Information regarding the tube inspections performed on the WBN Unit 2 SGs for Cycle 1 was provided in Reference 7.

3.2

SUMMARY

OF VOLTAGE-BASED REPAIR CRITERIA The following items address the voltage-based repair criteria in GL 95-05 and the specific requirements and actions associated with the implementation of the voltage-based repair criteria at WBN Unit 2:

All tubes will be inspected using the bobbin coil. The inspection will include hot leg TSP intersections and cold leg intersections down to the lowest TSP for which ODSCC has been identified.

Bobbin coil flaw indications greater than 1.0 volt will be inspected by a rotating probe coil (or equivalent) to evaluate the presence of detectable ODSCC and to confirm that the dominant corrosion mechanism occurring is axially oriented ODSCC.

WBN Unit 2 eddy current analysis guidelines will be compatible with and satisfy GL 95-05 requirements.

Axial ODSCC indications less than or equal to the lower voltage repair limit of 1.0 volt, as measured by bobbin coil, may remain in service without further inspection or analysis.

Axial ODSCC bobbin coil indications greater than 1.0 volt and less than or equal to the upper voltage repair limit (VURL) may remain in service if rotating probe coil (RPC) inspection indicates no detectable degradation.

Axial ODSCC bobbin coil indications greater than 1.0 volt and less than or equal to VURL must be plugged if the RPC inspection (or equivalent, see Section 3.5) indicates detectable degradation.

Axial ODSCC bobbin coil indications exceeding the upper voltage repair limit must be plugged.

CNL-18-003 E1-8 of 31 Postulated faulted condition primary-to-secondary leakage through indications to which the criteria is applied will be calculated using accepted industry and NRC approved practices. Postulated leakage in the limiting steam generator will be less than the bounding faulted condition leakage necessary to ensure that offsite doses remain a small fraction of 10 CFR Part 100 reactor site criteria and that control room doses remain within 10 CFR 50 Appendix A, General Design Criterion (GDC) 19 limits.

Projected tube burst probability at a pressure differential equal to the limiting faulted condition pressure differential will be calculated using accepted industry and NRC-approved practices and compared to the reporting value of 1.0E-2 in the limiting SG.

3.3 ACKNOWLEDGMENT OF GL 95-05 PERFORMANCE REQUIREMENTS An item-by-item comparison of the individual items of GL 95-05 is provided in Table 1 to this enclosure.

3.4 APPLICATION The voltage-based repair criteria apply to axially oriented ODSCC indications at tube-to-TSP and FDB intersections (within the plate thickness) of the SG tube bundle (Section 1.a of Attachment 1 to GL 95-05). Prior to U1C3, rotating probe examination of TSP intersections at WBN Unit 1 indicated that the degradation morphology, was generally below detectable thresholds for the +Point1 probe and only signals interpreted as possible precursors to degradation were present (e.g., deposits in the tube/TSP crevice). Indications of axial ODSCC at TSP locations were detected during U1C3 and may occur in WBN Unit 2.

Section 4 of Enclosure 2 describes the tube-to-TSP intersections of the SG tube bundle where the voltage-based repair criteria will not be applied, which is in accordance with Section 1.b of Attachment 1 of GL 95-05. As noted in Enclosure 2, the list of SG tubes identified for exclusion in WBN Unit 1 (see Table 2 to Attachment 1 of Enclosure 2) are the same for WBN Unit 2.

At tube-to-TSP intersections with dent signals exceeding 5.0 (bobbin) volts (Section l.b.2 of Attachment 1 of GL 95-05), and any crack indications confirmed by RPC will be plugged. TVA plans to inspect hot leg dented intersections greater than or equal to two volts using a +Point probe or other qualified eddy current probe.

During implementation of the voltage-based ARC, correlations are made between the expected burst pressure, the probability of leak, and the expected leak rate to the bobbin voltage of the indication. The correlations are used with a measured or calculated end-of-cycle (EOC) distribution of indications to estimate the likelihood of a tube burst and the primary-to-secondary total leak rate for the SG during a postulated MSLB event.

If the probability of burst is sufficiently small, and if the total estimated leak rate, at a specified confidence level, is less than acceptable limits the voltage criterion may be implemented. If either of the requirements were not met, additional tubes would be plugged until both of the requirements would be projected to be met at the EOC.

The voltage-based repair criteria for WBN Unit 2 is consistent with the guidance of GL 95-05 (except as described in Section 3.4.1 of this enclosure) and is described in further detail below:

1 +Point is a registered trademark of Zetec, Inc.

CNL-18-003 E1-9 of 31 Implementation of the applicability requirements as discussed in Section 1 of to GL 95-05. The applicability requirements ensure that the repair criteria are applied only to those intersections for which the voltage-based repair criteria were developed.

Calculation of conditional burst probability according to the guidance discussed in Section 2.a of Attachment 1 to GL 95-05. This calculation assesses the voltage distribution for the next cycle of operation. The results are compared against a threshold value. If it is not practical to complete this calculation prior to returning the SGs to service, the measured EOC voltage distribution can be used (from the previous cycle of operation) as an alternative (refer to Section 2.c of Attachment 1 to GL 95-05) for the purposes of determining whether the reporting criteria of Section 6.a.3 of Attachment 1 to GL 95-05 apply.

Calculation of leakage according to the guidance discussed in Section 2.b of to GL 95-05 (see Attachment 1 to this enclosure). This calculation, in conjunction with the use of licensing basis assumptions for calculating offsite and control room doses, enables licensees to demonstrate that the applicable limits of 10 CFR Part 100 and GDC 19 continue to be met. This calculation is performed using the projected EOC voltage distribution for the next cycle of operation. If it is not practical to complete this calculation prior to returning the SGs to service, the measured EOC voltage distribution can be used (from the previous cycle of operation) as an alternative (refer to Section 2.c of Attachment 1 to GL 95-05) for the purposes of determining whether the reporting criteria of Section 6.a.1 of to GL 95-05 apply.

Implementation of the inspection guidance discussed in Section 3 of Attachment 1 to GL 95-05. The inspection guidance ensures that the techniques used to inspect the SG tubes are consistent with the techniques used to develop the voltage-based repair criteria.

Implementation of the operational leakage monitoring program according to the guidance discussed in Section 5 of Attachment 1 to GL 95-05. The operational leak rate monitoring program is a defense-in-depth measure that provides a means for identifying leaks during operation to enable plugging before such leaks result in tube failure.

Reporting of results according to the guidance discussed in Section 6 of to GL 95-05.

CNL-18-003 E1-10 of 31 TVA will implement the ARC criteria using alternate methods utilizing eddy current techniques as a substitute for tube removal for testing. Tube removals for testing and examination in accordance with the voltage-based repair criteria are performed to:

1. Confirm axial ODSCC as the dominant degradation mechanism

2. Monitor the degradation mechanism over time

3. Provide additional data to enhance the burst pressure, probability of leakage, and conditional leak rate correlations

4. Assess the inspection capability TVAs planned inspection scope for WBN Unit 2 SGs will meet the above criteria as follows:

1. To confirm axial ODSCC as the dominant degradation mechanism TVA will be utilizing bobbin probe and +Point probe or qualified eddy current inspection techniques. The latest approved Appendix I Examination Technique Specification Sheets (ETSSs) will be used for ARC inspections. Detection will be achieved using ETSSs such as Appendix I ETSS I28411 for Bobbin and Appendix I ETSS 22401.1 for +Point. Sizing will be done using Appendix I ETSSs such as +Point Appendix I ETSS I28431 and Appendix I ETSS I28432. Further interrogation is achievable as necessary using a Ghent or Array probe.

According to NP-7480-L, "Steam Generator Tubing Outside Diameter Stress Corrosion Cracking at Tube Support Plates - Data Base for Alternate Repair Limits" (Reference 8), there have been no pulled tubes for which the ODSCC crack morphology differs from that found in the initial EPRI database prior to issuance of GL 95-05. There has been one case of combined wall thinning with ODSCC. This wall thinning was identified as a volumetric indication with field eddy current. The tubes pulled with RPC axial ODSCC calls have had morphologies consistent with the EPRI database. Therefore, it is reasonable to conclude that the eddy current techniques used for WBN Unit 2 can distinguish between axial and circumferential ODSCC and determine whether axial cracks are confined within the TSP.

NP-7480-L concludes that removal of tubes specifically for morphology verifications can be a low priority for tube removal. An allowance is provided to forgo the tube pull to the end of the first cycle following ARC implementation. Based on WBN Unit 1 degradation experience, TVA projects the need to implement the ARC in the U2C3 refueling outage (U2R3 in Fall 2020). The WBN Unit 2 replacement SGs (RSG) are currently scheduled for installation in U2R4. Therefore, it is expected that WBN Unit 2 will only need the ARC for a short time period. If the SGs are not replaced at the scheduled time, and a tube pull is completed, the data would likely be of little value based on the amount of data in the database. The objective of a tube pull program is to populate a database with limited data. However, even if data were acquired through a WBN Unit 2 tube pull, no other 600 mill-annealed (MA) plants require the data.

Any other type of tube degradation or any other location in the tube bundle will continue to be evaluated in accordance with existing WBN Unit 2 TSs. The observation of circumferential cracks, or primary water stress corrosion cracking associated with TSP indications, or ODSCC beyond the TSP thickness will be reported to the NRC prior to return to power as specified in the proposed changes to WBN Unit 2 TS Section 5.9.9.

CNL-18-003 E1-11 of 31

2. To monitor the degradation over time, TVA will utilize the same eddy current inspection techniques used to confirm axial ODSCC.

Providing additional data to enhance the burst pressure, probability of leakage, and conditional leak rate correlations is not necessary with the experience and data available using NP-7480-L, Addendum 7. Additional data acquired from tube pulls have resulted in changes in the structural limit by about 0.5 volt or less. Burst pressure versus bobbin voltage correlation has not changed significantly with the additional data since before the issuance of GL95-05. Because changes in the leak rate versus voltage correlation have been more significant because of the smaller database on leaking tubes, the primary objective would be to increase the database for leak rate. This action would require a tube pull for bobbin voltages of greater than two volts for 3/4 inch tubes to ensure a significant likelihood of leakage. With the current database, only four indications are needed greater than two volts to complete the database in one-volt bins up to 12 volts. TVA may be able to pull a tube with the desired voltage, but the data will likely not be used because TVA will be replacing the SGs.

3. Assessment of the inspection capability has already been completed by previous plants using the voltage-based ARC.

Tubes pulled from other plants using both 7/8-inch and 3/4-inch outside diameter (OD) tubing for indications of ODSCC at TSPs have been shown to have crack morphology consistent with the EPRI database used for the supporting voltage correlations.

However, as noted in Section 3.4.1, TVA does not anticipate the need to remove tubes for examination for WBN Unit 2 as was done for WBN Unit 1.

3.4.1 Alternative to Section 4 of Attachment 1 of GL 95-05 GL 95-05 states, The NRC staff emphasizes that although the NRC has approved the implementation of the voltage-based repair criteria as a short term measure, this guidance should not be construed as discouraging the development and use of better acquisition techniques, eddy current technology, and eddy current data analysis techniques. Section 4 of Attachment 1 to GL95-05 states that implementation of the voltage-based repair criteria should include a program of tube removals for testing and examination. TVA does not plan to implement the guidance in Section 4 of Attachment 1 to GL95-05 based on the following:

Improved inspection techniques using approved Appendix H and I of the EPRI Examination Guidelines using the ETSSs.

WBN Unit 2 will be the last alloy 600 MA plant to implement these voltage-based repair criteria, therefore, there is adequate experience and data to use the ARC without a tube pull. In addition, no other 600 MA plants will need tube pull data from WBN Unit 2.

Based on WBN Unit 1 operational experience, TVA projects the need to implement the ARC for WBN Unit 2 in U2R3. The WBN Unit 2 replacement SGs (RSG) are currently scheduled for installation in U2R4. Therefore, it is expected that WBN Unit 2 will only need the ARC for a short time period. Therefore, TVA anticipates that the ARC for WBN Unit 2 will not be needed for long term operation.

TVA will utilize qualified eddy current techniques as an alternative to tube removals for testing.

CNL-18-003 E1-12 of 31 3.5 TUBE INTEGRITY EVALUATION The following engineering analyses will be performed during each implementation of the 1.0 volt ARC at WBN Unit 2:

Prediction of SG bobbin voltage population distribution Calculation of SG tube leakage during a postulated MSLB Calculation of SG tube burst probability during a postulated MSLB Reference 8 will be applied in the voltage correlations (burst, probability of leakage, MSLB leak rate) used for the leak rate, burst probability, and upper voltage repair limit calculations. The NRC approved industry protocol for updating the database will be followed by TVA, in the unlikely event tube pulls are performed.

The methodology to be applied by TVA at WBN Unit 2 for the performance of these analyses, including correlations that relate bobbin voltage amplitudes, free span burst pressure, probability of leakage and associated leak rates is documented in WCAP-14277, Revision 1, "SLB Leak Rate and Tube Burst Probability Analysis Methods for ODSCC at TSP Intersections" (Reference 9). Any future NRC-approved revisions to WCAP-14277 or equivalent methodology reports may be considered for implementation.

In addition, the upper voltage repair limit used to plug bobbin indications independent of RPC confirmation is determined at each outage based on the guidance of Section 2.a.2 of GL 95-05.

MSLB Tube Leak Rate The calculated maximum allowable tube leak rate for WBN Unit 2 SGs during a postulated MSLB event is four gallons per minute (gpm) in the faulted loop (calculated at standard temperature and pressure conditions). The four-gpm leak rate value in the faulted loop, with leakage in the intact loops equal to the WBN Unit 2 TS 3.4.13 leakage limit of 150 gallons per day (gpd), would not result in either control room dose exceeding the GDC 19 limit or the off-site dose exceeding 10 percent (%) of the 10 CFR 100 requirements. The establishment of the four-gpm leak rate value is controlled by the control room inhalation dose. If it is determined during the operating cycle that this leakage limit might be exceeded, the reporting guidance of GL 95-05, Section 6.a.1 as reflected in the proposed changes to WBN Unit 2 TS 5.9.9, will be followed. The calculated leak rate limit and maximum allowable leak rate values specified for WBN Unit 2 are specified as room temperature values; therefore, these values are comparable.

Consistent with the guidance of GL 95-05, Section 2.c, the WBN Unit 2 MSLB leak rate analysis would be performed, prior to returning the SGs to service, based on either the projected next EOC voltage distribution or the actual measured bobbin voltage distribution. The method to be used for the first outage when ODSCC indication growth rates are available will be based on the indications found during that outage. As noted in GL 95-05, it may not always be practical to complete these calculations prior to returning the SGs to service. Under these circumstances, it is acceptable to use the actual measured bobbin voltage distribution instead of the projected EOC voltage distribution to determine whether the reporting criteria are being satisfied. Using the previous EOC voltage distribution, if any, the projected next EOC MSLB leak rate will be calculated using a probability of detection (POD) of 0.6.

CNL-18-003 E1-13 of 31 The offsite dose calculations were performed consistent with GL 95-05 recommendations and other analysis methods or models previously approved by the NRC related to the GL 95-05 plugging criteria. A summary of the MSLB dose analysis results and inputs are provided in Attachment 1 to this Enclosure. The calculated radiological consequences of the control room, exclusion area boundary and the low population zone are larger than previously reported for the postulated steamline break event due to the increased leakage. The increase in primary to secondary side leakage for a faulted SG was in support of implementation of the ARC being requested in this amendment. However, the calculated radiological consequences remain in compliance with the guidelines in the NRC Standard Review Plan, Chapter 15 and the regulations in 10 CFR 50, Appendix A, GDC-19, and 10 CFR 100. The analysis of MSLB effects upon the reactor core, fuel, and departure from nucleate boiling (DNBR) potential described in the WBN dual-unit UFSAR indicates that no fuel failures are predicted because of this event.

Pressurizer Power Operated Relief Valve Availability WBN Unit 2 TS Section 3.4.11, Pressurizer Power Operated Relief Valves (PORVs),

requires that each PORV and associated block valves be operable during Modes 1, 2, and 3. Action statements are included in the TS that limit plant operation with one PORV or associated block valve inoperable. Compliance with TS Section 3.4.11 meets the requirements of GL 95-05 and provides the maximum pressure for the MSLB event.

With the PORV set point of 2335 psig and a 3% uncertainty on the opening pressure, the MSLB pressure differential for SG tubing is 2405 psig. This differential pressure is used for the calculation of the degradation structural limit and for calculations of tube MSLB tube burst probability per SG.

Probability of Detection GL 95-05, Section 2.b.1 recommends that the frequency distribution by voltage of bobbin indications actually found during an inspection should be scaled upward by a factor of 1/POD to specify indications considered to be present in the SG tubes for the next operating cycle. A constant POD value of 0.6 is also specified for use in evaluating the potential effect of missed indications. The POD value of 0.6 is considered conservative especially for larger voltage indications. GL 95-05 also allows the use of an alternative POD value or a distribution function may be used, subject to NRC approval. Also, because there are neither historical nor growth data distributions for indications in WBN Unit 2 SG tubes, the POD value of 0.6 will be directly applied to the distribution of detected indications for the first several cycles of implementation.

CNL-18-003 E1-14 of 31 Voltage Growth Due to Defect Progression GL 95-05 Section 2.b.2(2) provides guidance on the determination of a defect growth curve to be used in the evaluation of the SG indication population. There are three possibilities considered:

1. When there has been no prior TSP ODSCC detected

2. When there are growth data from only one prior cycle of operation

3. When there are growth data from two or more prior cycles of operation Because there are no reported indications from previous inspections for WBN Unit 2, a bounding probability distribution function of growth rates will be used for the initial cycle of application of the ODSCC ARC. The bounding curve will be based on the experiences of similarly designed and operated plants.

For subsequent inspections, TVA will continue to apply the guidance in GL 95-05. For example, if growth rates become available from more than one cycle of operation, the more conservative growth rate of the previous two cycles will be used for the projection of bobbin voltage distribution during the next operating cycle. Both beginning of cycle (BOC) and corresponding EOC bobbin indications at a TSP intersection are necessary to specify a growth data point. The following information is provided related to the establishment of a growth curve for subsequent inspections at WBN Unit 2:

Growth data from the previous two cycles may be combined if necessary to obtain at least 200 data points in the distribution; otherwise, industry data will be used.

If greater than 200 indications are unavailable on a per SG basis, the more limiting of SG specific or all SG combined growth will be used.

Negative growth rates will not be used in growth rate distributions used to make voltage projections although those rates will be used in establishing average growth for determining the upper voltage repair limit of Sections 2.a.2 and 2.a.3 of GL 95-05.

Establishment of VURL for TSP and FDB Intersections The voltage structural limit is the voltage from the regression analysis of the burst pressure on the bobbin voltage, at the 95% lower prediction interval curve reduced to account for lower tolerance limit material properties at 650°F. The upper voltage repair limit must be adjusted for flaw growth during an operating interval and to account for non-destructive examination (NDE) uncertainty. The VURL is determined from the following equation:

NDE G

SL URL V

V V

V

=

Where:

VSL is the structural limit VG represents the growth allowance VNDE represents the allowance for potential sources of error, including analyst variability.

The structural limit voltage is taken from Reference 8. The proximity of the TSP prevents tube burst during normal operating conditions, so the structural limit voltage at TSP intersections is calculated using a differential pressure of 1.4 times the bounding MSLB pressure differential (in accordance with Section 2.a.2 of Attachment 1 of CNL-18-003 E1-15 of 31 GL 95-05) of 2405 psig, or 3367 psig. The increased tube to FDB gap may not provide sufficient constraint to prevent burst at locations within the FDB. Therefore, the FDB voltage structural limit is established using a pressure loading of three times the normal operating differential pressure (PNOP) across the SG tubes in accordance with Section 2.a.3 of Attachment 1 of GL 95-05. The current maximum SG tube PNOP for WBN Unit 2 is about 1300 psig, thus the differential pressure used to establish the voltage structural limit for the FDB locations is 3900 psig. VSL for the TSP intersection locations is 6.03 volts while VSL for the FDB intersection locations is 3.81 volts. Using the above equation, with values for VNDE of 20% of VURL and VG of 30% of VURL (minimum value allowable per GL 95-05), the measurement upper voltage repair limits (VURL) for the TSP intersection and the FDB become 4.02 and 2.54 volts respectively.

Inspection Criteria The SG tubes will be inspected with the bobbin coil during each normally scheduled refueling outage at WBN Unit 2. The inspection includes hot leg side tube to TSP intersections and cold leg side tube to TSP intersections down to the lowest cold leg side TSP with identified ODSCC. Data acquisition and analysis will be performed to provide consistent methodology as that described in GL 95-05 and NDE guidelines utilized for the most recent field applications of the criteria, as updated by the clarifications listed below, which include use of the updated probe wear guidelines and new probe acceptability guidelines. The supplementary guidance of Section 3 of GL 95-05 will be applied with the clarifications noted below. Any indication with bobbin voltage exceeding 1.0 volt will be inspected with an RPC or equivalent, and will be plugged if the bobbin indication is confirmed as a flaw by RPC or equivalent. Any indication will be plugged regardless of any RPC inspection results, if the bobbin voltage exceeds the upper voltage repair limit as obtained per Section 2.a.2 of GL 95-05. For WBN Unit 2, specific upper voltage repair limits are separately developed for both TSP and FDB intersections.

New Probe Variability Criteria In Reference 10, the Nuclear Energy Institute (NEI) provided a methodology for meeting the new probe variability criteria in Section 3.c.2 of GL 95-05. In Reference 11, the NRC requested additional information regarding the information provided in Reference 10.

The NRC accepted this methodology in Reference 12. Additional industry information was provided in Reference 13, which dealt with test data for probes manufactured by Westinghouse and Zetec. Briefly summarized, this methodology requires, in part, that the primary frequency and mix frequency voltage response of a new probe be compared to the nominal response determined by the vendor to ensure that the new probe is within

+/- 10% of the nominal response for both the primary and mix channels. TVA will follow the guidance provided in Reference 12 as supplemented by the test data contained in Reference 13.

Probe Wear Criteria References 10 and 11 also discuss an alternative to the GL 95-05 probe wear criteria.

This alternative requires that when a probe does not pass the probe wear check (15%),

tube locations inspected with the worn probe having detected indications with amplitudes greater than or equal to 75% of the repair voltage limit (i.e., 1.0 volt for 3/4 inch OD tubes) will be re-inspected with a new acceptable probe. TVA will follow the guidance provided in References 10 and 11.

CNL-18-003 E1-16 of 31 Alternatives to the RPC Inspections GL 95-05 permits use of alternates to the rotating pancake coil for RPC inspections.

Currently, the +Point probes and Array Probes are considered by the NRC and industry as acceptable alternatives to rotating pancake coil. TVA will utilize the +Point or equivalent probe for TSP indication confirmation.

3.6 REPORTING REQUIREMENTS The proposed changes to WBN Unit 2 TS 5.9.9 are consistent with the reporting guidance in Section 6 of GL 95-05 and Reference 3.

3.7 CONCLUSION

The proposed amendment request permits degraded tubes with axial ODSCC indications confined to within the thickness of the tube support plates and FDBs with bobbin voltages 1.0 volt to remain in service. Further, the approved criteria permit tubes to remain in service that have indications confined to within the thickness of the tube support plates and FDPs with bobbin voltages > 1.0 volt, but less than or equal to the upper voltage repair limit, if a motorized rotating pancake coil probe, or acceptable alternative inspection, does not detect degradation. Finally, the approved criteria require degraded tubes to be plugged that have indications confined to within the thickness of the tube support plates with bobbin voltages greater than the upper voltage limit. The proposed TS changes (including the reporting requirements) and commitments, as listed in Enclosure 3, are consistent with the guidance in GL 95-05.

4.0 REGULATORY EVALUATION

4.1 APPLICABLE REGULATORY REQUIREMENTS AND CRITERIA A review of 10 CFR Part 50, Appendix A, "General Design Criteria (GDC) for Nuclear Power Plants," was conducted to assess the potential impact associated with the proposed changes. The SGs at WBN Unit 2 are designed to comply with the following applicable regulations and requirements.

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

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

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

CNL-18-003 E1-17 of 31 GDC 19, Control room, specifies that a control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents. Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of five rem whole body, or its equivalent to any part of the body, for the duration of the accident. Equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.

GDC 30, Quality of reactor coolant pressure boundary, specifies that components, which are part of the reactor coolant pressure boundary, shall be designed, fabricated, erected, and tested to the highest quality standards practical. Means shall be provided for detecting and, to the extent practical, identifying the location of the source of reactor coolant leakage.

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

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

The reactor coolant pressure boundary, containment boundary, and tube-bundle integrity will not be adversely affected by the application of the voltage-based ARC (GL 95-05) to the tube inspection scope. SG tube integrity is inherently satisfied during normal operating conditions due to the proximity of the tube support plate. Test data indicates that tube burst cannot occur within the TSP, even for tubes that have 100% through-wall electric discharge machining (EDM) notches, 0.75 inch long provided the tube support plate is adjacent to the notched area. Because tube-to-tube support plate proximity precludes tube burst during normal operating conditions, use of the criteria must retain tube integrity characteristics, which maintain a margin of safety of 1.4 times the bounding faulted condition (i.e., MSLB) differential pressure of 2405 psi. GL 95-05 recommends that maintenance of a safety factor of 1.4 times the MSLB pressure differential be determined consistent with the structural limits in RG 1.121 on tube burst.

This GL 95-05 recommendation is satisfied by using 3/4-inch diameter tubing, with bobbin coil indications with signal amplitudes less than the tube structural limit (VSL) of 6.03 volts, regardless of the indicated depth measurement. At the FDB, a safety factor of three against the normal operating condition P is applied. A voltage of VSL = 3.81 volts satisfies the burst capability recommendation at the FDB.

CNL-18-003 E1-18 of 31 This criterion is satisfied by limiting the upper voltage repair limit for TSPs. At FDBs, a safety factor of three times the normal operating condition P is applied and limits the upper voltage repair limit for FDBs. Any degradation below the F* length in accordance with WBN Unit 2 TS 5.7.2.12 (Reference 14) is shown by analyses and test results to be acceptable, thereby precluding an event with consequences similar to a postulated tube rupture event. SG tube surveillance requirements continue to ensure that degraded tubes will be plugged or removed from service upon detection. Postulated leakage in the limiting SG shall be less than the bounding faulted condition leakage necessary to ensure that offsite doses remain a small fraction of the 10 CFR Part 100 reactor site criteria and that control room doses remain within GDC 19 limits. Therefore, conformance with all applicable GDCs remains valid.

In conclusion, with the implementation of the proposed changes, WBN Unit 2 continues to meet the applicable regulations and requirements.

4.2 PRECEDENT In Reference 2, WBN Unit 1 submitted an amendment request to use the voltage-based ARC as specified in the guidelines contained in GL 95-05; the NRC approved this amendment request in Reference 3. This LAR is similar in nature to References 2 and 3 in that TVA is requesting NRC approval to use the voltage-based ARC for WBN Unit 2.

The proposed changes to WBN Unit 2 TS 5.9.9 are similar to those approved in Reference 3. However, the proposed WBN Unit 2 TS 5.7.2.12 changes in this LAR differ significantly from those in References 2 and 3, because the TS changes in References 2 and 3 were pre-Standard TS (NUREG-1431, Revision 4). The current WBN Unit 1 TS 5.7.2.12 does not make reference to voltage-based ARC, because WBN Unit 1 has RSGs. TVA has also reviewed the supplemental information and responses to NRC requests for additional information related to the WBN Unit 1 voltage-based ARC LAR (References 15 through 18) and incorporated into this LAR, as applicable, the information in References 15 through 18.

The proposed WBN Unit 2 TS 5.7.2.12 change is similar to that approved by the NRC for the Beaver Valley Power Station (BVPS) Units 1 and 2 (Reference 19) in that BVPS Units 1 and 2 TS 5.5.5.2.c.4 also allows use of the voltage-based ARC as an alternative to the 40% depth based criteria for SG tube plugging. The proposed changes to WBN Unit 2 TS 5.9.9 are also similar to the reporting criteria in BVPS Units 1 and 2 TS 5.6.6.2.3. Reference 19 also reflects changes to the BVPS Units 1 and 2 TS that were approved by the NRC in Reference 20 to adopt TSTF-449, Steam Generator Tube Integrity, Revision 4, and the changes to the BVPS Unit 2 TS that were approved by the NRC in Reference 21 that adopted the voltage-based ARC in accordance with GL 95-05.

4.3 SIGNIFICANT HAZARDS CONSIDERATION The Tennessee Valley Authority (TVA) proposes to revise the Watts Bar Nuclear Plant (WBN) Unit 2 Technical Specifications (TS) to incorporate the voltage-based alternate repair criteria (ARC) guidance in Generic Letter (GL) 95-05 associated with steam generator (SG) tube inspection and plugging. That guidance established an alternate voltage-based SG tube repair criteria at tube support plate (TSP) and flow distribution baffle (FDB) plate intersections. This proposed change is consistent with GL 95-05 and is based on a similar request that was approved by the Nuclear Regulatory Commission (NRC) for WBN Unit 1.

CNL-18-003 E1-19 of 31 he proposed changes affect WBN Unit 2 TS 5.7.2.12, Steam Generator (SG) Program, and WBN Unit 2 TS 5.9.9, Steam Generator Tube Inspection Report. These TS are being revised to incorporate voltage-based repair criteria for SG tubes affected by outside diameter stress corrosion cracking (ODSCC) at non-dented tube support intersections in accordance with the guidelines of GL 95-05.

TVA has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

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

Response: No Allowing the use of alternate repair criteria as proposed in this amendment request does not involve a significant increase in the probability or consequence of an accident previously evaluated.

Tube burst criteria are inherently satisfied during normal operating conditions due to the proximity of the TSP. Test data indicates that tube burst cannot occur within the TSP, even for tubes, which have 100% through-wall electric discharge machining (EDM) notches, 0.75 inches long, provided that the TSP is adjacent to the notched area. Because tube-to-tube support plate proximity precludes tube burst during normal operating conditions, use of the criteria must retain tube integrity characteristics, which maintain a margin of safety of 1.4 times the bounding faulted condition [i.e., main steam line break (MSLB)] differential pressure of 2405 psig.

GL 95-05 recommends that maintenance of a safety factor of 1.4 times the MSLB pressure differential, consistent with the structural limits in Regulatory Guide (RG) 1.121, on tube burst is satisfied by 3/4-inch diameter tubing with bobbin coil indications with signal amplitudes less than the tube structural limit (VSL) of 6.03 volts, regardless of the indicated depth measurement. At the FDB, a safety factor of three against the normal operating condition P is applied. A voltage of VSL = 3.81 volts satisfies the burst capability recommendation at the FDB.

The upper voltage repair limit (VURL) will be determined prior to each outage using the most recently approved NRC database to determine the VSL. The structural limit is reduced by allowances for nondestructive examination (NDE) uncertainty (VNDE) and growth (VG) to establish VURL.

Relative to the expected leakage during accident condition loadings, it has been previously established that a postulated MSLB outside of containment but upstream of the main steam isolation valves (MSIVs) represents the most limiting radiological condition relative to the alternate voltage-based repair criteria. In support of implementation of the revised repair limit, TVA will determine whether the distribution of cracking indications at the tube support plate intersections during future cycles are projected to be such that primary to secondary leakage would result in site boundary doses within a fraction of the 10 CFR 100 guidelines or control room doses within the 10 CFR 50, Appendix A, General Design Criterion (GDC) 19 limit. A separate calculation has determined this allowable MSLB leakage limit to be four gallons per minute (gpm) in the faulted loop.

CNL-18-003 E1-20 of 31 The methods for calculating the radiological dose consequences for this postulated MSLB are consistent with the WBN dual-unit Updated Final Safety Analysis Report (UFSAR) Chapter 15.

In summary, the calculated radiological consequences in the control room and at the exclusion area boundary and the low population zone are in compliance with the guidelines in the Standard Review Plan, Chapter 15, and the regulations in 10 CFR 50, Appendix A, GDC 19, and 10 CFR 100 reported for the postulated steamline break event. Therefore, it is concluded that the proposed changes do not result in a significant increase in the radiological consequences of an accident previously analyzed.

Consistent with the guidance of GL 95-05, Section 2.c, the WBN Unit 2 MSLB leak rate analysis would be performed, prior to returning the SGs to service, based on either the projected next end-of-cycle (EOC) voltage distribution or the actual measured bobbin voltage distribution. The method to be used for the first outage when ODSCC indication growth rates are available will be based on the indications found during that outage. As noted in GL 95-05, it may not always be practical to complete EOC calculations prior to returning the SGs to service. Under these circumstances, it is acceptable to use the actual measured bobbin voltage distribution instead of the projected EOC voltage distribution to determine whether the reporting criteria are being satisfied.

Therefore, the voltage-based ARC at WBN Unit 2 does not adversely affect SG tube integrity and implementation is shown to result in acceptable radiological dose consequences. Therefore, the proposed TS change does not result in a significant increase in the probability or consequences of an accident previously evaluated within the WBN Unit 2 UFSAR.

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

Response: No.

Implementation of the proposed SG tube voltage-based ARC does not introduce any changes to the plant design basis. Neither a single nor multiple tube rupture event would be expected in an SG in which the repair limit has been applied (during all plant conditions).

The bobbin probe voltage-based tube repair criteria of 1.0 volt is supplemented by:

enhanced eddy current inspection guidelines to provide consistency in voltage normalization, a 100 percent eddy current inspection sample size at the tube support plate elevations, and rotating probe coil (RPC) or equivalent inspection requirements for the larger indications left in service to characterize the principal degradation as ODSCC.

As SG tube integrity upon implementation of the 1.0 volt repair limit continues to be maintained through in-service inspection and primary to secondary leakage monitoring, the possibility of a new or different kind of accident from any accident previously evaluated is not created.

CNL-18-003 E1-21 of 31

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

Response: No.

The use of the voltage-based bobbin probe tube support plate elevation repair criteria at WBN Unit 2 maintains SG tube integrity commensurate with the guidance of RG 1.121. RG 1.121 describes a method acceptable to the NRC for meeting GDCs 14, 15, and 32 by reducing the probability or the consequences of SG tube rupture. This reduction is accomplished by determining the limiting conditions of degradation of steam generator tubing, as established by in-service inspection, for which tubes with unacceptable cracking should be removed from service. Upon implementation of the proposed criteria, even under the worst-case conditions, the occurrence of ODSCC at the TSP elevations is not expected to lead to an SG tube rupture event during normal or faulted plant conditions. The EOC distribution of crack indications at the tube support plate elevations is confirmed to result in acceptable primary to secondary leakage during all plant conditions and that radiological consequences are not adversely impacted.

Implementation of the TSP intersection voltage-based repair criteria will decrease the number of tubes that must be plugged. The installation of SG tube plugs reduces the reactor coolant system flow margin. Thus, implementation of the 1.0 volt repair limit will maintain the margin of flow that would otherwise be reduced in the event of increased tube plugging.

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

4.4 CONCLUSION

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

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

CNL-18-003 E1-22 of 31

6.0 REFERENCES

1. Generic Letter 95-05, Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected By Outside Diameter Stress Corrosion Cracking, dated August 3, 1995

2. TVA letter to NRC, Watts Bar Nuclear Plant (WBN) - Unit 1 - Technical Specification (TS) Change No. WBN-TS-99-014 - Steam Generator Alternate Repair Criteria for Axial Outside Diameter Stress Corrosion Cracking (ODSCC), dated April 10, 2000 (ML003703571)

3. NRC letter to TVA, Watts Bar Nuclear Plant, Unit 1 - Issuance of Amendment Regarding Steam Generator Tubing Alternate Repair Criteria (ARC)

(TAC No. MA8635), dated February 26, 2002 (ML020590277)

4. Westinghouse Report WAT-D-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 ODSCC at TSPs,"

Revision 0, dated January 12, 2000

5. TVA letter to NRC, Watts Bar Nuclear Plant (WBN) Unit 1 - Steam Generator Voltage Based Alternate Repair Criteria For Axial Outside Diameter Stress Corrosion Cracking (ODSCC) - Cycle 5 Ninety Day Report, dated January 15, 2004 (ML040220171)

6. Steam Generator Pulled Tube Examination Watts Bar Unit I - Cycle 5 RFO, SG-SGDA-04-1 Rev. 1, Final Report March 2004 (ML041610191)

7. TVA letter to NRC, Watts Bar Nuclear Plant (WBN) Unit 2 - Cycle 1 Steam Generator Tube Inspection Report, dated February 16, 2018 (ML18047A370)

8. EPRI Report 1018047, Addendum 7 to NP-7480-L Database, "Steam Generator Tubing Outside Diameter Stress Corrosion Cracking at Tube Support Plates Database for Alternate Repair Limits," September 2008

9. Westinghouse Report WCAP-14277, "SLB Leak Rate and Tube Burst Probability Analysis Methods for ODSCC at TSP Intersections, Revision 1, December 1996

10. NEI letter from Alex Marion to Brian Sheron of NRC, Eddy Current Probe Replacement Criteria for Use in ODSCC Alternate Repair Criteria, dated January 23, 1996 (9711100044)

11. NRC letter from Brian Sheron to Alex Marion of NEI dated February 9, 1996 (9602150031)

12. NRC letter from Brian Sheron to Alex Marion of NEI dated March 18, 1996 (9604090441)

13. NEI letter from Alex Marion to Brian Sheron of NRC, Response to NRC Letter Dated February 9,1996, Regarding New Probe Variability Criteria, dated October 15, 1996 (9709050095)

14. NRC letter to TVA, Watts Bar Nuclear Plant, Unit 2 - Issuance of Amendment Regarding Steam Generator Inspection Scope Using the F* Methodology (CAC No. MF7218), dated September 6, 2016 (ML16203A365)

CNL-18-003 E1-23 of 31

15. TVA letter to NRC, Watts Bar Nuclear Plant (WBN) - Unit 1 - Technical Specification (TS) Change No. WBN-TS-99-014 - Steam Generator Alternate Repair Criteria for Axial Outside Diameter Stress Corrosion Cracking (ODSCC) - Clarification (TAC No. MA8635), dated September 18, 2000 (ML003752760)

16. TVA letter to NRC, Watts Bar Nuclear Plant (WBN) - Unit 1 - Technical Specification (TS) Change No. WBN-TS-99-014 - Steam Generator Alternate Repair Criteria for Axial Outside Diameter Stress Corrosion Cracking (ODSCC) - Clarification and Additional Information - (TAC No. MA8635), dated August 22, 2001(ML012500340)

17. TVA letter to NRC, Watts Bar Nuclear Plant (WBN) - Unit 1 - Technical Specification (TS) Change No. WBN-TS-99-014 - Steam Generator Alternate Repair Criteria for Axial Outside Diameter Stress Corrosion Cracking - Clarification and Supplemental Information (TAC No. MA8635), dated January 15, 2002 (ML020180269)

18. TVA letter to NRC, Watts Bar Nuclear Plant (WBN) - Unit 1 - Technical Specification (TS) Change No. WBN-TS-99-014 - Steam Generator Alternate Repair Criteria for Axial Outside Diameter Stress Corrosion Cracking - Response to Request for Additional Information - (TAC No. MA8635), dated November 8, 2001 (ML020160013)

19. NRC letter to FirstEnergy Nuclear Operating Company, Beaver Valley Power Station, Unit Nos. 1 and 2 - Issuance of Amendments Re: License Amendment Request to Revise Steam Generator Technical Specifications (CAC Nos. MF6054 and 6055), dated December 16, 2015 (ML15294A439)

20. NRC letter to FirstEnergy Nuclear Operating Company, Beaver Valley Power Station, Unit Nos. 1 and 2 - Issuance of Amendments Re: Steam Generator Tube Integrity (TAC Nos. MC8861 and MC8862), dated September 7, 2006 (ML062260011)

21. NRC letter to Duquesne Light Company, Beaver Valley Power Station, Unit No. 2 -

Issuance of Amendment Re: Alternate Plugging Criteria for Steam Generator Tubing and Reduction of Reactor Coolant System Specific Activity Limits (TAC Nos. M95793 and M99828), dated August 18, 1999 (ML003772235)

CNL-18-003 E1-24 of 31 Table 1 Acknowledgment of Individual GL 95-05 Performance Criteria GL 95-05 Item GL 95-05 Methodology Concurrence Comments 1.a Followed TVA will abide by the specified guidance in GL 95-05 1.b Followed The exclusion criteria listed will be followed. Enclosure 1, Table 2 lists individual intersections excluded due to permanent deformation potential from a combined LOCA + SSE event. VURL for FDB intersections is defined in Enclosure 1.

2.a.1 Modified from original

version, approved by NRC The latest NRC approved database (Reference 6) will be utilized.

2.a.2 Followed See response to section 2.a.1.

2.a.3 Followed See response to section 2.a.1.

2.b.1 Request for NRC approval Distribution of bobbin indications included in the MSLB leak rate projection will be based on the voltage dependent probability of detection as described in Reference 6.

2.b.2 Followed TVA will abide by the specified guidance in GL 95-05 2.b.2(1)

Followed TVA will abide by the specified guidance in GL 95-05 2.b.2(2)

Modified initial cycle.

Followed thereafter.

Because there are no reported indications from previous inspections, a bounding probability distribution function of growth rates will be used for the initial cycle of application. For subsequent inspections, TVA will continue to abide by the specified guidance in GL 95-05.

2.b.3 Followed TVA will abide by the specified guidance in GL 95-05 2.b.3(1)

Followed TVA will abide by the specified guidance in GL 95-05 2.b.3(2)

Followed TVA will abide by the specified guidance in GL 95-05 2.b.4 Modified The MSLB leak rate calculated in Enclosure 1 Attachment 1, uses currently accepted licensing basis assumptions. The current TS allowable dose equivalent iodine (DEI) value is already below 0.35

µCi/gm (0.265 µci/gm per WBN Unit 2 TS 3.4.16).

2.c Followed TVA will abide by the specified guidance in GL 95-05 3.a Followed TVA will abide by the specified guidance in GL 95-05 3.b Followed TVA will utilize the +Point or future equivalent probe (e.g., array probes) for confirmation of bobbin indications. RPC for the purposes of the TS change also includes the use of comparable or improved nondestructive examination techniques.

3.b.1 Followed

+Point (or future equivalent) will be used for inspection of bobbin voltages > 1.0 volt.

3.b.2 Followed

+Point (or future equivalent) will be used for inspection where copper could influence bobbin signal, possibly masking a 1.0 volt indication. Any indication found at such intersections with RPC should cause the tube to be plugged.

CNL-18-003 E1-25 of 31 Table 1 Acknowledgment of Individual GL 95-05 Performance Criteria GL 95-05 Item GL 95-05 Methodology Concurrence Comments 3.b.3 Followed

+Point (or future equivalent) will be used for inspection of all dents greater than five volts, possibly masking a 1.0 volt indication. Any indications found at such intersections with RPC should cause the tube to be plugged. If circumferential cracking or primary water stress corrosion cracking indications are detected, it may be necessary to expand the RPC sampling plan to include dents less than 5.0 volts.

3.b.4 Followed

+Point (or future equivalent) will be used for inspection of large mixed residuals, possibly masking a 1.0 volt indication. TVA will inspect all intersections with large mixed residuals utilizing a RPC probe. Any indications found at such intersections with RPC should cause the tube to be plugged.

3.c.1 Followed TVA will use a bobbin coil calibrated against a reference standard used in the laboratory as part of the development of the voltage-based approach, through the use of a transfer standard or the latest industry method approved by NRC.

3.c.2

Modified, accepted by NRC The probe variability limits defined in Reference 10 as supplemented by test data contained in Reference 11 will be implemented.

3.c.3

Modified, accepted by NRC Limits on re-inspection of tubes due to out of specification probe wear will be followed according to Reference 9.

3.c.4 Followed Data analysts will be trained and qualified in the use of the analysis guidelines and procedures specific for application of the criteria.

3.c.5 Followed Data analysts will use quantitative noise criteria guidelines in the evaluation of the data. However, it is expected that these criteria will be evolving over the inspection and as a result, are subject to change. Data failing to meet these criteria should be rejected, and the tube will be re-inspected.

3.c.6 Followed TVA data analysts will review the mixed residuals on the standard itself and take action as required in the TVA analysis guidelines.

3.c.7 Followed TVA will use 0.610-inch diameter bobbin probes. Prior to TVAs use of a different bobbin probe size, TVA will demonstrate (on a plant specific or generic basis) that probes and procedures will provide (on a statistically significant basis) equivalent voltage response and detection capability when compared to the 0.610-inch diameter bobbin probe.

3.c.8 Followed Data analysts will be trained on the potential for primary water stress corrosion cracking (PWSCC) to occur at TSP intersections and sensitized to identifying indications attributable to PWSCC.

4.

Modified TVA will use the alternate methods described in Section 3.4 5.a Followed TS 3.4.13 leakage is 150 gpd primary-to-secondary leakage in any one SG.

CNL-18-003 E1-26 of 31 Table 1 Acknowledgment of Individual GL 95-05 Performance Criteria GL 95-05 Item GL 95-05 Methodology Concurrence Comments 5.b Followed TVA leakage monitoring techniques are consistent with EPRI ID# 1022832, PWR Primary-to-Secondary Leak Guidelines, and are adequate to meet GL 95-05 recommendations.

5.c Followed Known leaking tubes will be plugged.

6.

Followed Reporting requirements will be followed.

CNL-18-003 E1-27 of 31 Table 2 Number of TSP Intersections with D > 0.030 inch, subsequently excluded from application of the voltage-based plugging criteria.

WBN Unit 2 Note: Model D3 SGs nominally have 4674 tubes, therefore, the maximum number of tube support holes at a plate elevation is 9348 (4674 in each of the hot leg and cold leg portions of the plates).

Westinghouse TSP Designation WBN Unit 2 TSP Designation Number of Affected Intersections Number of Intersections in TSP1,2,3

% Affected Intersections A (FDB)

H01 0

< 4674 0 %

B (FDB)

C01 0

4674.

0 %

C H02 0

4674 0 %

D C02 0

4674 0 %

E C03 6

4354.

0.14 %

F C04 6

4354 0.14 %

G H03 6

4674 0.13 %

H C05 6

4354 0.14 %

J C06 20 4354 0.46 %

K C07 20 4354 0.46 %

L H04 20 4674 0.43 %

M C08 20 4354 0.46 %

N C09 20 4354 0.46 %

P C10 20 4674 0.43 %

Q H05, C11 18 9348.

0.19 %

R H06, C12 18 9348 0.19 %

S H07, C13 30 9348 0.32 %

T H08, C14 256 9348 2.74 %

1. The plate A (FDB) has a crescent shaped center cutout region in the hot leg side of the baffle; therefore, the number of intersections is less than twice the tube count.

2. Number of tube holes in plates E, F, H, J, K, M, and N has less than 4674 intersections to allow the flow to turn in the pre-heater section.

3. Plates Q, R, S, and T are full plates, represented by a 360° circumference. All other plates are half-plates.

CNL-18-003 E1-28 of 31 Watts Bar Nuclear Plant Unit 2 Main Steam Line Break Dose Calculation Summary 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.

A calculation determined the maximum permissible SG primary to secondary leak rate during a steam line break for WBN Unit 2. The calculation determined that four gpm (at standard temperature and pressure) primary to secondary leakage in the faulted SG would result in site boundary doses within 10 CFR 100 guidelines and control room doses within the GDC 19 limit.

The establishment of the four-gpm value is controlled by the control room inhalation dose.

The calculation used TVA computer codes STP, FENCDOSE, and COROD. The STP output is used as input to COROD (which determines control room operator dose) and FENCDOSE (which determines 30-day Low Population Zone (LPZ) and two-hour Exclusion Area Boundary (EAB) offsite doses).

Two different cases were analyzed:

A pre-accident iodine spike where the iodine level in the reactor coolant spiked to the maximum allowable limit of 14 µCi/gm I-131 equivalent (refer to WBN Unit 2 TS 3.4.16) just prior to the initiation of the accident.

The reactor coolant at the maximum steady state dose equivalent I-131 of 0.265 µCi/gm with an accident initiated iodine spike consisting of a 500 times the iodine release rate from the fuel.

In both cases, the primary-to-secondary side leak is 150 gpd in the unfaulted loops, and the secondary side activity is at the WBN Unit 2 TS 3.7.14 limit of 0.1 µCi/gm dose equivalent I-131.

The analysis used the same models and inputs used in the original licensing of WBN Unit 2 except for the following:

The assumed primary to secondary leakage value for a faulted SG was increased from one gpm to four gpm for a post-accident leak through the faulted steam generator to the environment. The increase in primary to secondary side leakage for a faulted SG was in support of implementation of the ARC being requested in this LAR.

The control room isolation delay time was increased from 40 seconds to 74 seconds. The control room isolation delay time was increased due to an error in the time constant assumed in the determination of this value as documented on page E1-26 of the referenced letter.

The reactor coolant system (RCS) weight was decreased from 2.45E8 gm to 2.14E8 gm.

The decrease in the RCS weight was due to errors found in the determination of that value as documented on page E1-23 of the referenced letter.

The primary and secondary coolant concentrations were revised as described in page E1-23 of the referenced letter.

Tables 1 through 5 of this attachment provide the input used in the models.

CNL-18-003 E1-29 of 31 Reference TVA letter to NRC, CNL-17-144, 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), dated December April 20, 2000 (ML17354B282)

Table 1 to Attachment 1 Parameter used in the MSLB analysis Source Term Pre-Accident Iodine spike (µCi/gm) 14 Accident initiated iodine spike (µCi/gm) 0.265 Secondary Side I-131 equivalent (µCi/gm) 0.1 Dose conversion factor (DCF) for I equivalence RG 1.109 DCF for I consequences ICRP-30 Noble Gases 100/E-Bar2 Concentration basis ANS-18.1-1984 Volume/ Mass Reactor Coolant Mass (gm) 2.14E+08 Faulted SG mass (gm) 4.31E+07 Intact SGs Mass (gm) 1.29E+08 Flow Rates/Mass Releases Primary to Secondary leak - Faulted SG (gpm) 4 Primary to Secondary leak - Intact SG (gpd) 150 Steam released from faulted SG (lb) 95,000 0-2 hr mass release from unfaulted SG (lb) 433,079 2-8 hr mass released from unfaulted SG (lb) 870,754 Parameters used for iodine production rate for 500x spike Max Letdown (gpm) 124.39 Primary side leakage for I removal (gpm) 11 Letdown Demineralizer Efficiency 1

Partition Factors Faulted SG 1

Unfaulted SG 100 2 E-Bar is the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95%

of the total non-iodine activity in the coolant. The noble gas inventories are maximized by scaling them up to 100/E-Bar.

CNL-18-003 E1-30 of 31 Table 2 to Attachment 1 RCS and Secondary Side Concentrations (µCi/gm)

Nuclide Reactor Coolant Secondary Side Kr-85m 1.90E-01 4.04E-08 Kr-85 2.59E-01 5.36E-08 Kr-87 1.79E-01 3.58E-08 Kr-88 3.33E-01 7.03E-08 Xe-131m 6.89E-01 1.42E-07 Xe-133m 7.87E-02 1.69E-08 Xe-133 2.73E+00 5.66E-07 Xe-135m 1.55E-01 3.23E-08 Xe-135 1.00E+00 2.13E-07 Xe-138 1.43E-01 2.99E-08 I-131 4.67E-02 4.65E-06 I-132 2.44E-01 5.28E-06 I-133 1.51E-01 1.11E-05 I-134 4.01E-01 3.71E-06 I-135 2.92E-01 1.31E-05 Table 3 to Attachment 1 Atmospheric Dispersion Factors based on 1991-2010 Meteorology Time EAB (sec/m3)

LPZ (sec/m3)

CR (sec/m3) 0-2 hr 6.382E-04 1.784E-04 2.59E-03 2-8 hr N/A 8.835E-05 2.12E-03 8-24 hr N/A 6.217E-05 1.77E-04 1-4 day N/A 2.900E-05 1.33E-04 4-30 day N/A 9.811E-06 1.12E-04 CNL-18-003 E1-31 of 31 Table 4 to Attachment 1 Control Room Parameters Volume (ft3) 257,198 Emergency pressurization flow (cfm) 711 Normal pressurization flow (cfm) 3200 Emergency recirc flow (cfm) 2889 Normal recirc flow (cfm) 3200 Unfiltered inleakage (cfm) 51 Charcoal Filter efficiency first pass 95%

2nd pass 70%

Occupancy Factors 0-24 hr 100%

1-4 days 60%

4-30 days 40%

Delay in isolation (sec) 74 Table 5 to Attachment 1 Main Steam Line Break Dose Calculation Results Control Room (rem)

Limit (rem) 30-Day LPZ (rem) 2-Hour EAB (rem)

Limit (rem)

Accident Initiated Iodine Spike Case Gamma:

Beta:

Inhalation:

2.57E-02 2.05E-01 2.31E+01 5

30 30 5.46E-01 1.30E-01 2.02E+01 4.28E-01 1.02E-01 1.02E+01 2.5 30 30 Pre-Accident Iodine Spike Case Gamma:

Beta:

Inhalation:

9.66E-03 1.02E-01 9.83E+00 5

30 30 3.60E-02 1.36E-02 3.87E+00 8.22E-02 2.67E-02 6.51E+00 25 300 300 CNL-18-003 Proposed TS Changes (Mark-Ups) for WBN Unit 2

Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals (continued)

Watts Bar - Unit 2 5.0-16 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 is not to exceed 1 gpm per SGfor the faulted SG loop and 150 gallons per day (gpd) for each unfaulted SG. 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 or the faulted SG loop and 150 gallons per day (gpd) for each unfaulted 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 (FDB). At tube support plate intersections and FDB, the plugging or repair limit is described below:

Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals (continued)

Watts Bar - Unit 2 5.0-16a Amendment 2 a) Steam generator tubes, with degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plates and FDB 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 FDB 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 FDB 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-D-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 ODSCC 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 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 FDB 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.

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.

Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals (continued)

Watts Bar - Unit 2 5.0-16b Amendment 2 The mid-cycle repair limits are determined from the following equations:

VMURL= VSL 1.0+NDE + Gr[(CL-t)/CL]

VMLRL = VMURL-(VURL-VLRL)[(CL-t)/CL]

where:

VMURL =mid-cycle upper voltage repair limit based on time into cycle VSL = structural limit voltage NDE = 95-percent cumulative probability allowance for nondestructive examination uncertainty (i.e., a value of 20 percent has been approved by NRC). The NDE 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 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.

Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals (continued)

Watts Bar - Unit 2 5.0-17 Amendment 2 5.7.2.12 Steam Generator (SG) Program (continued)

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 1.64 inches below the bottom of the roll transition or 1.64 inches below the top of the tubesheet, whichever is lower at the tube inlet, to 1.64 inches below the bottom of the roll transition or 1.64 inches below the top of the tubesheet, whichever is lower at the tube outlet, and that may satisfy the applicable tube plugging criteria. 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 each SG at least every 24 effective full power months or at least every refueling outage (whichever results in more frequent inspections). In addition, inspect 100% of the tubes at sequential periods of 60 effective full power months beginning after the first refueling outage inspection following SG installation. Each 60 effective full power month inspection period may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated.

Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals (continued)

Watts Bar - Unit 2 5.0-18 Amendment 2 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 crack 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 FDB) 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.

Reporting Requirements 5.9 5.9 Reporting Requirements (continued)

Watts Bar-Unit 2 5.0-36 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.

5.10 Record Retention (removed from Technical Specifications)

CNL-18-003 Proposed TS Changes (Final Typed) for WBN Unit 2

Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals (continued)

Watts Bar - Unit 2 5.0-16 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 is not to exceed 1 gpm for the faulted SG loop and 150 gallons per day (gpd) for each unfaulted SG. 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 or the faulted SG loop and 150 gallons per day (gpd) for each unfaulted 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 (FDB). At tube support plate intersections and FDB, the plugging or repair limit is described below:

Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals (continued)

Watts Bar - Unit 2 5.0-16a Amendment 2 a) Steam generator tubes, with degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plates and FDB 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 FDB 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 FDB 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-D-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 ODSCC 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 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 FDB 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.

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.

Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals (continued)

Watts Bar - Unit 2 5.0-16b Amendment 2 The mid-cycle repair limits are determined from the following equations:

VMURL= VSL 1.0+NDE + Gr[(CL-t)/CL]

VMLRL = VMURL-(VURL-VLRL)[(CL-t)/CL]

where:

VMURL =mid-cycle upper voltage repair limit based on time into cycle VSL = structural limit voltage NDE = 95-percent cumulative probability allowance for nondestructive examination uncertainty (i.e., a value of 20 percent has been approved by NRC). The NDE 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 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.

Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals (continued)

Watts Bar - Unit 2 5.0-17 Amendment 2 5.7.2.12 Steam Generator (SG) Program (continued)

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 1.64 inches below the bottom of the roll transition or 1.64 inches below the top of the tubesheet, whichever is lower at the tube inlet, to 1.64 inches below the bottom of the roll transition or 1.64 inches below the top of the tubesheet, whichever is lower at the tube outlet, and that may satisfy the applicable tube plugging criteria. 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 each SG at least every 24 effective full power months or at least every refueling outage (whichever results in more frequent inspections). In addition, inspect 100% of the tubes at sequential periods of 60 effective full power months beginning after the first refueling outage inspection following SG installation. Each 60 effective full power month inspection period may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated.

Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals (continued)

Watts Bar - Unit 2 5.0-18 Amendment 2 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 crack 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 FDB) 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.

Reporting Requirements 5.9 5.9 Reporting Requirements (continued)

Watts Bar-Unit 2 5.0-36 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.

5.10 Record Retention (removed from Technical Specifications)

CNL-18-003 E2-1 SG-SGMP-13-16-NP, Revision 1 Watts Bar Nuclear Plant Unit 2 Applicability of GL 95-05 Voltage Based Alternate Repair Criteria Regulatory Commitment CNL-18-003 E3-1 Commitment Due Date/Event The following criteria will added to the WBN Unit 2 Steam Generator Repair Procedures following NRC approval of the voltage-based ARC at WBN Unit 2:

All tubes will be inspected using the bobbin coil. The inspection will include hot leg TSP intersections and cold leg intersections down to the lowest TSP for which ODSCC has been identified.

Bobbin coil flaw indications greater than 1.0 volt will be plugged and inspected by a rotating probe coil to evaluate the presence of detectable ODSCC and to confirm that the dominant corrosion mechanism occurring is axially oriented ODSCC.

Eddy current analysis guidelines will be compatible with and satisfy GL 95-05 requirements.

Axial ODSCC indications less than or equal to the lower voltage repair limit of 1.0 volt, as measured by bobbin coil, may remain in service without further inspection or analysis.

Axial ODSCC bobbin coil indications greater than 1.0 volt and less than or equal to the upper voltage repair limit (VURL) can remain in service if rotating probe coil (RPC) (or equivalent) inspection indicates no detectable degradation.

Axial ODSCC bobbin coil indications greater than 1.0 volt and less than or equal to the upper voltage repair limit (VURL) must be plugged if RPC inspection (or equivalent) indicates detectable degradation.

Axial ODSCC bobbin coil indications exceeding the upper voltage repair limit must be plugged.

Postulated faulted condition primary-to-secondary leakage through indications to which the criteria is applied will be calculated using accepted industry and NRC approved practices. Postulated leakage in the limiting steam generator will be less than the bounding faulted condition leakage necessary to ensure that offsite doses remain a small fraction of 10 CFR Part 100 reactor site criteria and that control room doses remain within 10 CFR 50 Appendix A, General Design Criterion (GDC) 19 limits.

Projected tube burst probability at a pressure differential equal to the limiting faulted condition pressure differential will be calculated using accepted industry and NRC approved practices and compared to the reporting value of 1.0E-2 in the limiting SG.

Within 60 days following NRC approval of License Amendment Request 391 WBN2-TS-17-30 (TVA letter CNL-18-003)