CNL-20-053, Application to Revise Watts Bar Nuclear Plant, Unit 1, Technical Specifications for Steam Generator Tube Inspection Frequency & to Adopt TSTF-510, Revision to Steam Generator Program Inspection Frequencies & Tube Sample Selection, (WBN-39

Application to Revise Watts Bar Nuclear Plant, Unit 1, Technical Specifications for Steam Generator Tube Inspection Frequency & to Adopt TSTF-510, Revision to Steam Generator Program Inspection Frequencies & Tube Sample Selection, (WBN-390-

ML20199M346
Person / Time
Site:	Watts Bar
Issue date:	07/17/2020
From:	Jim Barstow Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CNL-20-053, SQN-TS-20-01
Download: ML20199M346 (53)

Text

Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 CNL-20-053 July 17, 2020 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Watts Bar Nuclear Plant Unit 1 Facility Operating License No. NPF-90 NRC Docket No. 50-390

Subject:

Application to Revise Watts Bar Nuclear Plant (WBN), Unit 1 Technical Specifications for Steam Generator Tube Inspection Frequency and to Adopt TSTF-510, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," (WBN-390-TS-20-012)

Reference:

TVA letter to NRC, CNL-20-010, Application to Revise Sequoyah Nuclear Plant (SQN) Unit 1 Technical Specifications for Steam Generator Tube Inspection Frequency (SQN-TS-20-01), dated February 24, 2020 (ML20056C857)

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 Renewed Facility Operating License No. NPF-90 for the Watts Bar Nuclear Plant (WBN) Unit 1.

The proposed license amendment request (LAR) revises WBN, Unit 1 Technical Specifications (TS) 5.7.2.12, Steam Generator (SG) Program, and WBN Unit, 1 TS 5.9.9, Steam Generator Tube Inspection Report, to reflect a proposed change to the required SG tube inspection frequency from every 72 effective full power months (EFPM), or at least every third refueling outage, to every 96 EFPM consistent with the referenced letter. Because WBN, Unit 1 has an 18-month operating cycle, 96 EFPM normally equates to every fifth refueling outage. The LAR also reflects changes consistent with Nuclear Regulatory Commission (NRC) approved Technical Specifications Task Force (TSTF) Technical Change Traveler 510, Revision 2, Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection.

TVA is scheduled to perform the next WBN, Unit 1 SG tube inspection during the WBN, Unit 1 Cycle 17 Refueling Outage (U1R17), which is scheduled to commence in October 2021. TVA has been involved in the development of Technical Specification Task Force (TSTF)-577, Performance Based Frequencies for Steam Generator Tube Inspections, along with meetings between the industry and the Nuclear Regulatory Commission (NRC). However, the schedule for the development and NRC review and approval of TSTF-577 does not support TVAs schedule for the proposed license amendment.

U.S. Nuclear Regulatory Commission CNL-20-053 Page 2 July 17, 2020 The operational experience of the WBN replacement SGs (RSGs), as described in the enclosure to this submittal, demonstrates that the proposed change to the schedule for the SG inspections is appropriate and will result in a reduction of dose to personnel and risk to the plant. Furthermore, the WBN, Unit 1 RSG operational assessment and experience supports the proposed TS changes. to this submittal provides a description and technical evaluation of the proposed change, a regulatory evaluation, and a discussion of environmental considerations for the proposed change to the required SG tube inspection frequency. Enclosure 2 to this submittal provides a description and assessment of the proposed changes, the requested confirmation of applicability, and plant-specific verifications associated with TSTF-510. Enclosure 3 to this submittal provides the existing WBN, Unit 1 TS pages marked up to show the proposed changes. Enclosure 4 to this submittal provides the existing WBN, Unit 1 TS pages retyped to show the proposed changes. Enclosure 5 to this submittal provides the existing WBN, Unit 1 TS Bases pages marked-up to show the proposed changes to meet TSTF-510. Changes to the existing TS Bases are provided for information only and will be implemented under the TS Bases Control Program.

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 10 CFR 51.22(c)(9). In accordance with 10 CFR 50.91, Notice for Public Comment; State Consultation, TVA is sending a copy of this letter and the enclosure to the Tennessee Department of Environment and Conservation.

TVA requests approval of the proposed license amendment within one year from the date of this submittal, with implementation within 30 days following NRC approval in order to support the WBN U1R17 outage, during which the WBN SG tubes are scheduled to be inspected. There are no new regulatory commitments associated with this submittal. If you have any questions about this proposed change, please contact Gordon R. Williams, Senior Manager, Fleet Licensing (Acting) at (423) 751-2687.

I declare under penalty of perjury that the foregoing is true and correct. Executed on this 17th day of July 2020.

Respectfully, James Barstow Vice President, Nuclear Regulatory Affairs & Support Services Enclosures cc (Page 3)

U.S. Nuclear Regulatory Commission CNL-20-053 Page 3 July 17, 2020

Enclosures:

1. Application to Revise Watts Bar Nuclear Plant (WBN), Unit 1 Technical Specifications for Steam Generator Tube Inspection Frequency

2. Application to Revise Watts Bar Nuclear Plant (WBN), Unit 1 Technical Specifications to Adopt TSTF-510, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection"

3. Proposed TS Changes (Mark-Ups) for WBN, Unit 1

4. Proposed TS Changes (Final Typed) for WBN, Unit 1

5. Proposed TS Bases Page Changes (Mark-Ups) for WBN, Unit 1 (For Information Only) 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

Enclosure 1 Evaluation of Proposed Change

Subject:

Application to Revise Watts Bar Nuclear Plant (WBN) Unit 1 Technical Specifications for Steam Generator Tube Inspection Frequency CONTENTS 1.0

SUMMARY

DESCRIPTION ............................................................................................. 2 2.0 DETAILED DESCRIPTION.............................................................................................. 2 2.1 Proposed Changes ...................................................................................................... 2 2.2 Condition Intended to Resolve ..................................................................................... 3

3.0 TECHNICAL EVALUATION

............................................................................................. 4 3.1 System Description ...................................................................................................... 4 3.2 Technical Analysis ....................................................................................................... 5 3.2.1 Background and Introduction ................................................................................ 6 3.2.2 Technical Specification Sequential Periods ........................................................... 7 3.2.3 WBN Unit 1 Steam Generator Inspection, Degradation, and Plugging History ...... 9 3.2.4 Trending of Existing Degradation Mechanisms ....................................................11 3.2.5 WBN Unit 1 Steam Generator Secondary Side Conditions ..................................14 3.2.6 Secondary Chemistry Control ..............................................................................16 3.2.7 Discussion of Growth Rates, OA Methods, Projections, and Results ...................17 3.2.8 Conclusion ...........................................................................................................20

4.0 REGULATORY EVALUATION

.......................................................................................21 4.1 Applicable Regulatory Requirements and Criteria .......................................................21 4.2 Precedent ...................................................................................................................22 4.3 Significant Hazards Consideration ..............................................................................23 4.4 Conclusion ..................................................................................................................24

5.0 ENVIRONMENTAL CONSIDERATION

..........................................................................24

6.0 REFERENCES

...............................................................................................................25 CNL-20-053 E1-1 of 25

Enclosure 1 1.0

SUMMARY

DESCRIPTION In accordance with the provisions of Title 10 of the Code of Federal Regulations (10 CFR) 50.90, Application for amendment of license, construction permit, or early site permit, Tennessee Valley Authority (TVA) is requesting a license amendment to Renewed Facility Operating License No. NPF-90 for the Watts Bar Nuclear Plant (WBN),

Unit 1. The proposed license amendment request (LAR) revises WBN, Unit 1 Technical Specifications (TS) 5.7.2.12, Steam Generator (SG) Program, and TS 5.9.9, Steam Generator Tube Inspection Report, to reflect a proposed change to the required SG tube inspection frequency from every 72 effective full power months (EFPM), or at least every third refueling outage, to every 96 EFPM. Because WBN Unit 1 has an 18-month operating cycle, 96 EFPM normally equates to every fifth refueling outage.

The operational experience of the WBN, Unit 1 replacement SGs (RSGs), as described in this enclosure, demonstrates that the proposed change to the schedule for the SG inspections is appropriate and will result in a reduction of dose to personnel, and risk to the plant. The WBN Unit 1 RSG operational assessment (OA) and experience supports the proposed TS change.

2.0 DETAILED DESCRIPTION

2.1 PROPOSED CHANGE

S

• TS 5.7.2.12 d is revised to add 5.7.2.12 d.2 as follows (note this reflects changes per TSTF-510 as described in Enclosure 3 to this submittal):

After the first refueling outage following SG installation, inspect each SG at least every 96 effective full power months. In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in a and b below. 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 repair criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. 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. Each inspection period defined below 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.

a) After the first refueling outage following SG installation, inspect 100% of the tubes during the next 144 effective full power months. This constitutes the first inspection period.

b) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the second and subsequent inspection periods CNL-20-053 E1-2 of 25

Enclosure 1

• TS 5.9.9 is being revised to add the following new reporting requirement

h. Discuss trending of tube degradation over the inspection interval and provide comparison of the prior operational assessment degradation projections to the as-found condition.

Enclosure 3 provides the existing WBN, Unit 1 TS pages marked-up to show the proposed changes. Enclosure 4 provides the existing WBN, Unit 1 TS pages retyped to show the proposed changes. There are no TS Bases changes associated with the proposed TS changes described in this enclosure.

2.2 CONDITION INTENDED TO RESOLVE The operational experience of the WBN, Unit 1 RSGs, as described in Section 3.0 to this enclosure, demonstrates that the proposed change to the schedule for the SG inspections is appropriate and will result in a reduction of dose to personnel and risk to the plant. Table 1 shows the current outage schedule and the proposed schedule for SG inspections. Overall, the proposed change will result in two less instances of SG inspections through the life of the plant while still accomplishing the 100 percent (%)

inspection of the tubing within the sequential periods.

A significant reduction in dose will be achieved through less scheduled SG inspection outages. The dose reduction will be the result of less SG activities and additional shielding of personnel during refueling outages because secondary side water covering the SG tube bundle provides shielding and reduces exposure for all activities in containment. Typical dose values for SG inspections can vary depending on outage scope and activities. During the U1R8 SG inspection outage, the associated activities accounted for 11.7 person-rem, which was approximately 17% of the total outage dose.

During the U1R11 SG inspection outage, the associated activities accounted for 12.1 person-rem, which was approximately 21% of the total outage dose. During the U1R14 SG inspection outage, the associated activities accounted for 11 person-rem, which was approximately 20% of the total outage dose. The currently projected dose exposure for the U1R17 outage is 12.5 person-rem. Assuming that the same U1R17 dose occurred in future SG inspection outages, the avoided dose for WBN, Unit 1 personnel under the proposed amendment and planned SG inspection schedule would be at least 25 person-rem over the remaining life of the plant.

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

CNL-20-053 E1-3 of 25

Enclosure 1 While risk is minimized, as much as possible, through plant processes and procedures, performing 100% SG inspection scope in a 96 EFPM sequential period will reduce the total number of outages that the above associated activities are performed. This will yield a corresponding reduction of personnel dose exposure while improving plant and personnel safety.

3.0 TECHNICAL EVALUATION

3.1 SYSTEM DESCRIPTION The WBN Unit 1 RSGs 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 WBN, Unit 1 RSGs are described in the Updated Final Safety Analysis Report (UFSAR) Section 5.5.2, Steam Generator. Materials of construction for the WBN, Unit 1 RSGs are provided in UFSAR Table 5.2-8, Reactor Coolant Pressure Boundary Materials Class 1 Primary Components. Materials are selected and fabricated in accordance with the requirements of the American Society of Mechanical Engineers (ASME) Code Sections III.

The WBN Unit 1 SGs were replaced during the WBN U1R7 outage in Fall 2006.

WBN Unit 1 is a four-loop plant with recirculating Westinghouse Model 68AXP RSGs equipped with 5,128 Alloy 690 thermally treated (Alloy 690TT) tubes arranged in a triangular pitch forming a U-tube bundle. The tubes have an outer diameter of 0.75 inches with a 0.043-inch nominal wall thickness. The tube plugging limit at WBN, Unit 1 is 12% or 615 tubes per RSG. The tubesheet base metal is clad with Alloy 690 material. Tube rows 1 through 38 received a supplemental heat treat stress relief following bending. Each tube is hydraulically expanded into the tubesheet and welded to the clad at the primary face of the tubesheet. There are twelve advanced tube support grids (ATSG) consisting of a grid of slotted bars, which provide horizontal support to the straight leg of every tube. The tubes are supported in the U-bend region by ventilated flat bar support trees with varying numbers of vertical and diagonal support elements depending on the tube location.

Figure 1 provides a general arrangement view of the WBN, Unit 1 RSG. The Model 68AXP RSG designed by Westinghouse and manufactured by Doosan Heavy Industries represents an advanced second-generation design that is expected to provide enhanced margin against known SG tubing and secondary side component degradation mechanisms. There have been no changes to the design and operating parameters, such as plant uprate or physical modification, of the WBN, Unit 1 RSG since installation.

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Enclosure 1 Moisture Separators Tube Bundle U-bend Supports Main Feedwater Nozzle ATSGs Tubesheet Figure 1: WBN Unit 1 RSG - Westinghouse Model 68AXP

3.2 TECHNICAL ANALYSIS

The SG tubes in pressurized water reactors have several important safety functions. SG tubes are an integral part of the reactor coolant pressure boundary (RCPB) and, as such, are relied on to maintain the primary systems pressure and inventory. As part of the RCPB, the SG tubes are unique, in that they act as a heat transfer surface between the primary and secondary systems to remove heat from the primary system. In addition, the SG tubes isolate the radioactive fission products in the primary coolant from the secondary system.

The SG tube rupture (SGTR) accident is the limiting design basis event for SG. The analysis of an SGTR event assumes a bounding primary to secondary leakage rate equal to the operational leakage rate TS limit, plus the leakage rate from a double-ended rupture of a single tube. The accident analysis for an SGTR assumes the contaminated secondary fluid is only briefly released to the atmosphere via safety valves. The analysis for design basis accidents and transients other than an SGTR assume the SG tubes retain their structural integrity (i.e., they are assumed not to rupture). In these CNL-20-053 E1-5 of 25

Enclosure 1 analyses, the steam discharge to the atmosphere is based on the total primary to secondary leakage from all SGs or is assumed to increase to the TS limit because of accident-induced conditions. For accidents that do not involve fuel damage, the primary coolant activity level is assumed equal to the TS limits. For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel.

SG tube integrity is necessary to ensure the tubes are capable of performing their intended safety functions. Concerns relating to the integrity of the tubing stem from the fact that the SG tubing is subject to a variety of degradation mechanisms. SG tubes have experienced tube degradation related to corrosion phenomena, such as wastage, pitting, intergranular attack, and stress corrosion cracking, along with other mechanically induced phenomena such as wear. These degradation mechanisms can impair tube integrity if they are not managed effectively. The SG performance criteria are used to manage SG tube degradation.

The industry, working through the Electric Power Research Institute (EPRI) Steam Generator Management Program (SGMP), has implemented a generic approach to managing SG performance referred to as "Steam Generator Degradation Specific Management" (SGDSM). The overall program is described in Nuclear Energy Institute (NEI) 97-06, "Steam Generator Program Guidelines," which is supported by EPRI guidelines, including:

• PWR Steam Generator Examination Guidelines

• Steam Generator Integrity Assessment Guidelines

• Steam Generator In-Situ Pressure Test Guidelines

• PWR Primary-to-Secondary Leak Guidelines

• PWR Primary Water Chemistry Guidelines

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

3.2.1 Background and Introduction The design of the original WBN, Unit 1 SG had Alloy 600 mill annealed (Alloy 600MA) tube material. Alloy 600MA is susceptible to in-service stress corrosion cracking degradation under the operating conditions of most commercial nuclear sites. This form of degradation led to plugging significant numbers of tubes and shortening the useful life of the WBN, Unit 1 original SGs to the point of early replacement, which occurred in Fall 2006.

With the operating experience of Alloy 600MA SG tube material recognized throughout the industry, Alloy 690TT has emerged as the tube material of choice for both new and replacement pressurized water reactor SGs. This tube material has been found through laboratory studies and operating experience to have significantly greater resistance to corrosion induced degradation. The first commercial nuclear SGs with Alloy 690TT tubing were put into service in 1989. The Alloy 690TT tube material now has nearly 30 years of operating service experience domestically. There are approximately 50 nuclear units operated within the United States with Alloy 690TT SG tube material CNL-20-053 E1-6 of 25

Enclosure 1 and this population has an average operating age of about 17 years. The predominant degradation mechanism for SG with Alloy 690TT tubing, including the WBN, Unit 1 RSGs, has been volumetric wear due to interaction with tube support structures. There have been no reported indications of stress corrosion cracking degradation in any Alloy 690TT tube to date.

Two tubes were plugged in the WBN, Unit 1 RSGs prior to initial operation. One tube was plugged during fabrication of the RSGs due to an over expansion of the tube at the top of tubesheet. Another tube was plugged during the pre-service inspection (PSI) due to a freespan SVI (Single Volumetric Indication) measuring 20% through-wall (%TW).

This location was coincident with a dent indicating that was likely a result of tube installation.

The RSGs at WBN have undergone a pre-service inspection (PSI) and three in-service inspections (ISI), the results of which are described in Section 3.2.3 to this enclosure.

There are currently two existing in-service tube degradation mechanisms in the RSGs (i.e., wear at the tube bundle U-bend support structures and tube wear at the ATSGs).

In U1R8, six tubes were plugged preventatively due to horizontal support grid wear and one tube was preventatively plugged due to a tube expansion bulge within the tubesheet. In U1R11, 20 tubes were plugged preventatively due to wear at the horizontal tube support grids. No tubes were plugged during the U1R14 outage. The remainder of tube plugging occurred during the pre-service inspections and fabrication previously discussed for a total of 29 plugged tubes.

3.2.2 Technical Specification Sequential Periods The current WBN, Unit 1 TS reflects TSTF-449, Revision 4, Steam Generator Tube Integrity. For ISI of SGs with Alloy 690TT tubing, in accordance with TSTF-449, Revision 4, the WBN Unit 1 TS permits an initial sequential period of 144 EFPM for 100% inspection of the SG tubes with techniques qualified for detecting existing and potential degradation. The accrual of service for this requirement begins after the first ISI and does not include the PSI. Under the current TS requirements, the lengths of the sequential periods are 144 EFPM (first inspection period), 108 EFPM (second inspection period), 72 EFPM (third inspection period), and 60 EFPM (fourth and subsequent inspection periods). WBN, Unit 1 is currently in the first inspection period.

The percentage of SG tubes that must be inspected is dependent on the number of scheduled inspections over the sequential period. A requirement specific to TSTF-449 and therefore WBN, Unit 1, is that at least fifty percent of the tubes must be inspected by the outage nearest the midpoint of the period. Inspections must also be planned such that 100% of the tubing has been inspected by the end of the sequential period. If an active degradation mechanism associated with cracking is present, then the affected and potentially affected SG shall be inspected at the subsequent refueling outage.

Table 1 indicates the planned SG inspection schedule for the current WBN, Unit 1 TS intervals and the proposed intervals. As shown in Table 1, 100% of the tubes will be inspected in each 96 EFPM sequential period. The SG OA performed based on the results of the most recent U1R14 inspection supports the inspection interval under the proposed license amendment (see Section 3.2.3 to this enclosure).

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Enclosure 1 Table 1 WBN Unit 1 Replacement SG Sequential Periods SGs Replaced 1R8 1R9 1R10 1R11 1R12 1R13 1R14 1R15 1R16 1R17 1R18 1R19 1R20 1R21 1R22 1R23 1R24 1R25 1R26 1R7 S-08 F-09 S-11 F-12 S-14 F-15 S-17 F-18 S-20 F-21 S-23 F-24 S-26 F-27 S-29 F-30 S-32 F-33 S-35 F-06 SG EFPY 1.2 2.6 4 5.3 6.6 8 9.3 10.6 12 est. 13.4 est. 14.8 est. 16.2 est. 17.6 est 19.0 est. 20.4 est 21.8 est. 23.2 est. 24.6 est. 26.0 est.

Cumulative EFPM Within 0 16.8 33.6 49.2 64.8 81.6 97.2 112.8 129.6 est. 2.4 est. 19.2 est. 36 est. 52.8 est. 69.6 est. 86.4 est. 103.2 est. 12 est. 28.8 est. 45.6 est.

Sequential Period End of TS Sequential 1st ISI 144 EFPM Sequential Period 108 EFPM Sequential Period 72 EFPM Sequential Period WBN Period Bobbin 100% No No 57.8% No No 100% No No 33.3% No No 33.3% No No 33.3% No No 50% Unit 1 All SGs ECT ECT All SGs ECT ECT ECT ECT ECT ECT ECT ECT ECT ECT License Base Scope 1 All SGs3 All SGs2 All SGs2 All SGs2 All SGs2 EFPM Within Proposed 15.6 32.4 est. 49.2 est 66 est. 82.8 est. 3.6 est. 20.4 est. 37.2 est. 54 est. 70.8 est. 87.6 est. 8.4 est.

Sequential Period Proposed TS 3rd 96 1st 96 EFPM Sequential Period 2nd 96 EFPM Sequential Period Sequential Period EFPM Bobbin Base Scope Under No No No No 100% No No No No 100% No No Proposed Amendment4 -> ECT ECT ECT ECT All SGs3 ECT ECT ECT ECT All SGs3 ECT ECT Notes

1. The bobbin base scope is accompanied by additional special interest and diagnostic exams using RPC and Array probes.

2. The TS minimum required inspection scope determined by dividing 100% by the number of scheduled inspections within the period.

3. The WBN U1R14 inspection consisted of a 100% combination bobbin and array coil inspection of all tubes full length with the exception of the U-bend sections of tube Rows 1 through 4, which were inspected with a singular bobbin probe due to dimensional constraints. This is the planned scope for all scheduled inspections under the proposed amendment.

4. The schedule under the proposed amendment revises the allowable inspection interval for each SG to at least every 96 EFPM.

Abbreviations ECT - Eddy Current EFPY - Effective Full Power Years EFPM - Effective Full Power Months est. - estimated F - Fall Outage and Year ISI - In-Service Inspection S- Spring Outage and Year SG - Steam Generator TS - Technical Specification CNL-20-053 E1-8 of 25

Enclosure 1 3.2.3 WBN Unit 1 Steam Generator Inspection, Degradation, and Plugging History The WBN, Unit 1 RSGs have undergone a pre-service eddy current inspection and three ISIs. Table 1 shows the actual bobbin probe inspection scope for each of the WBN Unit 1 refueling outages U1R8 through U1R14, and the plan of inspection through the remaining license of the plant. The U1R8 and U1R11 bobbin inspection programs at WBN were accompanied by rotating pancake coil (RPC) probe examinations in order to satisfy the TS requirement of sampling for potential degradation with a qualified technique within the sequential period. The ISI during WBN U1R14 consisted of a 100%

inspection with a combination bobbin and array coil probe over the full length of the tubes with the exception of the low row U-bends. The low rows were inspected with a singular bobbin probe due to dimensional restrictions through the bend region.

Pre-Service Inspection A pre-service eddy current inspection took place prior to initial service of the WBN, Unit 1 RSGs. The scope of the pre-service eddy current inspection consisted of 100%

full-length bobbin coil examination in the four RSGs. The small radius U-bend locations were inspected 100% with RPC probes to complete the full tube length examinations in the low row tubes. Further, RPC examinations were also performed in 100% of the top of the tubesheet expansion transitions on the hot leg side. In addition, a special interest RPC program was performed, as needed, to characterize bobbin indications that could not be otherwise resolved. One tube was preventatively plugged as a result of the preservice inspection due to a single volumetric indication measuring 20%TW. This location was coincident with a dent indicating that was likely a result of tube installation.

WBN U1R8 First In-Service Inspection - Spring 2008 All tubes were inspected full length with the bobbin probe during the WBN U1R8 first ISI of the RSGs. The inspections also included a special interest RPC program performed of signals not able to be resolved by bobbin data review. During the U1R8 inspection, there were six tubes identified with tube wear at horizontal tube support grid locations, predominately on the cold leg side. The largest indication detected was measured to be 13%TW) in depth. The tube wear indications detected during the U1R8 inspection could have remained in-service with no challenge to tube integrity during the inspection interval. However, TVA elected to preventively plug all tubes with detected wear indications during WBN U1R8.

Top-of-tubesheet sludge lancing and foreign object search and retrieval (FOSAR) was also performed during U1R8 in the four RSGs. FOSAR was also performed of the main feedwater inlet boxes of all four RSGs. All identified foreign material identified was removed from the SG secondary side. Steam drum upper internals inspections were performed in all SGs during U1R8 and no degradation was identified.

Further details of the WBN U1R8 SG ISI results are provided in Reference 1.

WBN U1R11 Second In-Service Inspection - Fall 2007 The base scope bobbin inspection for the WBN U1R11 inspection was 52.6% of the unplugged tubes in the RSGs. The inspections included a special interest RPC program performed of signals that could not be resolved by bobbin data review. Subsequent CNL-20-053 E1-9 of 25

Enclosure 1 localized scope expansions were performed to bound support structure wear and foreign object locations. This resulted in a final bobbin inspection scope of 57.8% of in-service tubes. During the U1R11 inspection, there were seven tubes identified with tube wear at the U-bend support structure intersections. There were 51 tubes identified with tube wear at the horizontal support structure intersections. The largest indication was located at a horizontal support and was measured to be 32%TW in depth. A total of 20 tubes were preventatively plugged during U1R11. No wear was identified due to foreign objects, but six tubes were plugged and stabilized to bound a loose part location.

Top of tubesheet sludge lance cleaning and FOSAR was also performed during U1R11 in all RSGs. The foreign material with significance relative to maintenance of tube integrity was removed except for the location that was addressed by stabilizing and plugging. No associated tube wear due to foreign objects was observed visually or by eddy current inspection. No indications of possible loose parts (PLPs) were identified by the eddy current program in WBN U1R11. Steam drum inspections were performed on two RSGs and no issues or degradation were identified.

Further details of the WBN U1R11 SG ISI are provided in Reference 2.

WBN U1R14 Third In-Service Inspection - Fall 2016 The U1R14 outage included combination bobbin and array coil inspection of 100% of the in-service tubes full length except for tubes in Rows 1 through 4. Tube Rows 1 through 4 were inspected to the first support (top support) with the combination bobbin and array coil from both the hot leg tube end (HTE) and the cold leg tube end (CTE).

The remainder of the tube length in Rows 1 through 4 was inspected with a singular bobbin probe due to dimensional clearance restrictions in the U-bend region.

In addition to the eddy current inspections, visual inspections were also performed on both the primary and secondary sides. Primary side visual inspections included the previously installed tube plugs, the channel head bowl cladding, and the divider plate.

Secondary side visual inspections were performed at the top of the tubesheet for the detection of foreign objects, assessment of hard deposit buildup in the tube bundle interior tubesheet kidney region and for determining the effectiveness of the tubesheet cleaning performed in the RSGs.

Prior to the secondary side FOSAR inspections, sludge, scale, foreign objects, and other deposit accumulations at the top of the tubesheet were removed as part of the top of tubesheet cleaning process. The secondary side FOSAR inspections performed in the RSGs included visual examination of tube bundle periphery tubes from the hot leg and cold leg annulus and center no tube lane. The foreign objects remaining were small pieces of gasket, wires, and bristles. Any foreign objects not able to be retrieved were characterized and an analysis performed to demonstrate acceptability of continued operation without exceeding the performance criteria.

The only degradation mechanisms detected during the U1R14 inspection were volumetric wear at the tube U-bend support structures and the ATSGs. The maximum depth tube wear indication detected was 27%TW at a U-bend support structure. The maximum depth tube wear indication located at an ATSG was 37%TW. There were no tubes plugged during the U1R14 SG inspections.

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Enclosure 1 Further details of the WBN U1R14 SG ISI results are provided in Reference 3.

3.2.4 Trending of Existing Degradation Mechanisms Based on ISI results of the eddy current inspection, the only existing tube degradation mechanisms affecting the RSGs are wear at U-bend supports and wear at the ATSGs.

Wear at the ATSGs has been the predominant mechanism, whereas wear at U-bend supports has only affected a small number of tubes at shallow through-wall depths.

Trends in existing degradation have been observed with three ISIs over 14 years of WBN Unit 1 operation with the RSGs.

Figure 2 presents the total numbers of tube wear indications for each mechanism over each of the three ISIs along with the maximum detected %TW depths. Also indicated in Figure 2 is the number of tubes plugged for each mechanism at each inspection. It is important to recognize the difference in inspection scope performed and eddy current techniques applied as noted in Figure 2. For example, the U1R8 and U1R14 inspections were both 100% scope while the U1R11 inspection scope was 57.8% and the most recent U1R14 inspections used the combination bobbin and array coil probe as opposed to the bobbin coil probe for U1R8 and U1R11.

Although the most recent U1R14 inspection has the largest number of ATSG wear indications detected, it is also the only inspection where a 100% inspection was performed with the combination bobbin and array coil probe. Degradation was reported by the eddy current analysis process with either coil during the U1R14 inspection and the array coil has enhanced detection capabilities for volumetric wear indications as compared to the bobbin coil. Therefore, the U1R14 inspection results are considered the most accurate and fully representative picture of the degradation conditions in the WBN Unit 1 RSGs.

Figure 3 and Figure 4 show the population versus depth distributions of wear at U-bend supports and at ATSGs for every inspection of the WBN Unit 1 RSGs. Tubes with degradation that have been plugged are indicated with a solid black circle surrounding the data point. Also shown is a vertical line representing the TS plugging limit of 40%TW for the WBN Unit 1 RSGs tubing. The 40%TW plugging limit can be used to represent a conservative SG tube structural and leakage integrity limit. These limits are typically about 10 to 20%TW higher than the TS plugging limit for WBN Unit 1. As shown in Figure 3 and Figure 4, there have been no tube wear indications at WBN Unit 1 to date that have exceeded the associated structural and leakage integrity performance criteria nor the 40%TW tube plugging limit.

There are 20,512 tubes total between the four WBN Unit 1 RSGs and 29 of them are plugged. Table 2 shows the count and percentage of tubes plugged in the RSG at WBN, Unit 1. There were two tubes plugged prior to initial service of the RSG. There have been 21 tubes preventively plugged in-service due wear at ATSGs, six preventively plugged for a foreign object and none plugged due to wear at the U-bend support structures.

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Enclosure 1 Table 2 WBN Unit 1 RSG Tube Plugging Degradation All SG1 SG2 SG3 SG4 Mechanism SGs Plugged Pre-Service 0 1 0 1 2 U-bend Support Structure 0 0 0 0 0 ATSG Wear 3 4 7 7 21 Foreign Object 0 0 0 6 6 (preventative)

Total 3 5 7 14 29 Percentage 0.06% 0.10% 0.14% 0.27% 0.14%

450 419 400 100% Combo Bobbin & Array Coil Inspection 350 Total Number of Indications in all SGs 300 250 Ma x Detect 37%TW 200 150 57.8% Bobbin Inspection 100 72 100% Bobbin 59 Inspection 50 Ma x Ma x Ma x 31 13%TW 12%TW Detect Ma x 9 9 8 32%TW Ma x RTS 27%TW 0 0 0 14%TW 0 0 0

U-bend ATSG U-bend ATSG U-bend ATSG U1R8 U1R11 U1R14 ATSG Ind Count Plugged Inds Figure 2 WBN Unit 1 RSG Support Wear Indications Summary CNL-20-053 E1-12 of 25

Enclosure 1 1

0.95 0.9 0.85 U1R14 0.8 U1R11 0.75 0.7 Cumulative Density Function (CDF) 0.65 0.6 0.55 0.5 40%TW TS 0.45 Plugging Limit 0.4 0.35 0.3 0.25 0.2 0.15 0.1 0.05 0

0 5 10 15 20 25 30 35 40 45 50 U-bend Support %TW in (all SGs)

Figure 3 WBN Unit 1 RSG Distribution of U-bend Support Wear Depths CNL-20-053 E1-13 of 25

Enclosure 1 1

0.95 0.9 0.85 U1R14 0.8 U1R11 0.75 U1R8 0.7 Plugged Tubes Cumulative Density Function (CDF) 0.65 0.6 0.55 0.5 40%TW TS 0.45 Plugging Limit 0.4 0.35 0.3 0.25 0.2 0.15 0.1 0.05 0

0 5 10 15 20 25 30 35 40 45 50 ATSG %TW in (all SGs)

Figure 4 WBN Unit 1 RSG Distribution of ATSG Support Wear Depths 3.2.5 WBN Unit 1 Steam Generator Secondary Side Conditions Secondary Side Tubesheet Cleanings TVA has performed a top of tubesheet cleaning coincident with every ISI of the WBN, Unit 1 RSGs since their installation. Most recently, a top of tubesheet deposit cleaning process was performed in the four RSGs during WBN U1R14. There are two main purposes of performing the cleaning process. The first is to remove hardened deposits that tend to form at the top of the tubesheet, and the second is to force and filter out any loose parts or foreign objects that have migrated to the SG secondary side during operation. The cumulative mass of deposit material and debris removed by the top of tubesheet cleaning process is summarized in Table 3 below.

Table 3 WBN U1R14 RSG Deposit Removal SG 1 8.5 lbs SG 2 7.5 lbs SG 3 9.5 lbs SG 4 7.0 lbs All SG 32.5 lbs CNL-20-053 E1-14 of 25

Enclosure 1 Periodic views of the cleaning systems in-line coarse grit tank screen filter were performed throughout the U1R14 RSGs tubesheet cleaning process. This confirmed that the process was successful at removing small numbers of foreign objects and material from the RSG secondary side in addition to the hardened sludge deposits.

Post-outage chemistry laboratory testing of the sludge deposit samples indicates iron to be the main elemental constituent. The percentage of iron content was within the typical industry range.

Regarding deposit buildup trending, the results of both the low frequency eddy current and secondary side top of tubesheet visual inspections indicate the presence of a small kidney region of deposit buildup near the center of the tube bundle. Tubesheet sludge pile depths from U1R14 were measured by eddy current to have a prevailing depth of about 1.5 inches and cover a surface area that spans approximately 100 tubes on the hot leg side. This region of hardened deposit has been trended between inspections both visually and by eddy current and not found to be growing unexpectedly. Further, the total of mass deposit removal from the WBN U1R14 SG tubesheet cleanings was 32.5 lbs. Therefore, the deposit removal trends at WBN Unit 1 are within an expected trend over time.

Foreign Object Inspection Results Summary Although foreign objects have been observed in the WBN, Unit 1 RSGs at previous inspections, no tube degradation associated with the presence of these objects has been detected to date. As noted in Table 2, six tubes in SG 4 were preventively plugged for a foreign object, but did not cause any actual tube wear degradation. The array probe was utilized to aid in the detection of foreign objects and foreign object wear during the most recent U1R14 eddy current inspection. During the WBN U1R14 eddy current inspections there was one signal corresponding to a PLP indication located coincident with the top of the tubesheet. There was no tube wall degradation detected by eddy current coincident with this PLP indication. The PLP location was visually inspected from the secondary side and no foreign object or contributing deposit condition was observed. Additional visual inspections performed during WBN U1R14 from the SG secondary side identified small foreign objects such as bristles, weld slag, and metal remnants. All foreign objects identified during U1R14 measured 1.25 inches or less and the majority was less than 0.375 inch. Although many of these foreign objects were removed from the RSGs, not all of them were retrieved.

The OA performed following WBN U1R14 considered the potential effects of the foreign objects remaining in the SG secondary side. The tube wear rate analysis performed in support of the OA, established that at least six fuel cycles or nine Effective Full Power Years (EFPY) of operating time would accrue before the identified objects with the greatest potential to cause actual tube wear degradation could potentially exceed the tube integrity performance criteria. The assumed flaw lengths that would be associated with the dimensions of the enveloping foreign objects are less than those applied in the tube wear rate analysis. Therefore, the results of the tube foreign object wear rate analysis are considered conservative.

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Enclosure 1 Upper Internals Inspections The most recent secondary side inspection of the WBN Unit 1 RSG upper internals occurred during the U1R11 refueling outage. Secondary side visual inspections were performed of two the four RSGs by access through the secondary manways. Steam drum inspection areas included the steam separator areas, upper dryer banks, drains, the auxiliary feedwater nozzles, the marmon clamps on the lower hatch to the tube bundle, and the various accessible instrument taps. All components inspected were found to be structurally sound.

The design of the WBN Unit 1 feedwater box is such that the intrusion of significant foreign objects or loose parts into the SG secondary side is mitigated. The feedwater box distributes the feed flow through discharge nozzles around the circumference of the tube bundle on the cold leg side. The main feedwater flow into the RSG travels through the feedwater box discharge nozzles at an elevation just above the cold leg top of the tubesheet. The discharge nozzles have many small diameter drilled holes that serve as a screen and acts to trap foreign objects or loose parts internal to the feedwater box.

Trapped foreign material collected then cannot travel to an area where it can cause tube degradation. Access to the internal of the feedwater box for periodic inspection and cleaning is provided through two separate handholes. The effectiveness of the feedwater box foreign object trapping system was confirmed when objects were observed and retrieved internal to the feedwater box during the U1R11 inspections.

3.2.6 Secondary Chemistry Control The overall objective of the WBN Unit 1 secondary chemistry strategy is to optimize chemistry with consideration of the relative risks and expected benefits of different chemistry control approaches. This approach was developed based upon the WBN secondary cycle design and metallurgy. The WBN Chemistry Program is designed to preserve the RSGs and secondary system integrity.

The secondary chemistry plan is continuously reviewed and updated as the secondary chemistry program is optimized. Secondary chemistry specifications include the required secondary chemistry parameters, limits, and sampling frequencies, which are based on the EPRI PWR Secondary Water Chemistry Guidelines. Corrosion degradation mechanisms are not considered potential mechanisms for the WBN Unit 1 RSGs.

SG control parameters are consistently kept below limits. WBN injects hydrazine as an oxygen scavenger to ensure low oxygenated water is fed to the RSGs. Hydrazine breaks down to ammonia at elevated temperatures, which increases the pH of water entering the SG. WBN injects ethanolamine (ETA) at a target value of 4-7 parts per million (ppm) to establish an elevated pH in wet steam areas on the secondary side.

This chemistry regime has allowed feedwater iron to routinely remain less than one part per billion (ppb) while online.

The RSGs are placed in chemical wet lay-up with ammonia and hydrazine once conditions are allowed shortly after shutdown. WBN utilizes elevated SG blowdown flow during startups from refueling outages to minimize contaminants sent to the RSGs.

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Enclosure 1 An evaluation of the WBN Unit 1 RSG secondary side contaminant hideout return is performed each outage. These evaluations assess the tendency for SG tubing and tube support structure corrosion. These assessments are also used to consistently evaluate the effectiveness of the secondary water chemistry control program.

3.2.7 Discussion of Growth Rates, OA Methods, Projections, and Results Degradation Growth Rates The WBN Unit 1 RSGs have been inspected in-service by eddy current techniques three times over an operating period of greater than 12.0 EFPY since installation. Every tube in each RSG has been tested full length at least two times in this operating span with techniques qualified for detection of all existing and potential degradation mechanisms.

The nature of the RSG degradation growth rates are considered relatively predictable based on this operating experience.

Wear at tube intersections with the ATSGs had been the predominant degradation mechanism for the WBN Unit 1 RSGs. Figure 4 shows the ATSG wear growth rate distributions for the WBN Unit 1 RSG inspections for the U1R11 and U1R14 inspections.

The degradation growth rates in this figure have been normalized to a %TW/EFPY basis. A distribution fit was used to represent the growth rates for all RSGs observed during U1R11 as opposed to a plot of all individual data points for each individual RSG.

All indications of tube wear degradation were plugged at the U1R8 inspection eliminating growth rate data on existing wear indications at the subsequent inspection. ATSG support wear indications were detected between the four RSGs at the U1R14 inspection after a three-cycle interval; therefore, mature ATSG wear growth rate data has been achieved by the U1R14 inspection. However, only one distribution of degradation growth is available for volumetric wear at the U-bend supports given that the mechanism was only first detected at the U1R11 inspection. The observed behavior of ATSG wear growth rates are considered encompassing of that for tube wear occurring at U-bend supports.

The largest indication of tube wear at ATSGs remaining in-service following U1R14 is 37%TW and for wear at U-bend supports is 27%TW. The population of ATSG wear growth rates had a 95th percentile of 5.4%TW/EFPY and the population of wear at U-bend supports had a 95th percentile of 5.7%TW/EFPY during the U1R14 inspection.

The in-service plugging performed to date has been primarily for indications of tube wear at ATSGs. Plugging strategies have been proactive for the purposes of providing both margin and conservatism in the OA process and not based on degradation growth rates which would cause an inspection interval of less than the current TS limit of not more than three refueling outages.

As a result, reasonably accurate and conservative OA projection results of existing degradation mechanisms are anticipated over the proposed revised inspection intervals.

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Enclosure 1 1

0.95 0.9 0.85 0.8 0.75 0.7 Cumulative Density Function (CDF) 0.65 0.6 0.55 0.5 U1R14 ATSGs - SG1 0.45 U1R14 ATSGs - SG2 0.4 U1R14 ATSGs - SG3 0.35 U1R14 ATSGs - SG4 U1R11 ATSGs - All SGs 0.3 0.25 0.2 0.15 0.1 0.05 0

0 1 2 3 4 5 6 7 8 9 10 ATSG Wear Growth Rates (%TW/EFPY)

Figure 4 WBN Unit 1 RSG Distribution of Support Wear Growth Rates CNL-20-053 E1-18 of 25

Enclosure 1 Comparison of Historical OA Projections to As-Found Indications The growth rates of existing degradation mechanisms are used to project future inspection results in the OA. The method of projection can involve varying levels of conservatisms based on the available inspection data and history of degradation growth rates. Using the arithmetic methods is the most conservative OA projection approach as described in EPRI guidelines. As degradation mechanisms mature, statistical methods such as Monte Carlo simulations, become a viable option based on the increased populations of available growth rate data points.

Table 4 shows the different OA strategies that have been applied during the operation of WBN Unit 1. In early WBN, Unit 1 inspections, acceptability over the three-cycle inspection interval was demonstrated through the arithmetic projection method. The results of this simple and conservative method were such that there was no need to apply more complex OA projection methods. Given the increased number of degradation growth rate data points available, Monte Carlo simulations have been applied at the most recent U1R14 inspection in order to make OA degradation projections. The methods used are described in the EPRI Guidelines and justify the RSG inspection interval.

Table 4 further provides the actual WBN Unit 1 inspection results for the maximum %TW detected and the maximum %TW indication left remaining in-service following the performance of tube plugging at each inspection. Also shown is the OA projected maximum %TW that could be encountered at the subsequent SG inspection. The margin to maximum %TW detected was determined by subtracting the maximum detected %TW indication from the OA projected maximum %TW. The margins to max %TW detected shown in Table 4 are based on arithmetic OA methods in all cases and additional margin is anticipated with application of statistical methods.

Conservative assumptions are inherent in the OAs performed of the WBN, Unit 1 RSGs.

Examples of the conservatisms applied in the WBN, Unit 1 OAs are provided below:

• Tube wear degradation lengths are assumed the full length of the tube support intersections. The actual axial extent of the flaws detected and measured by eddy current have been just a fraction of the support intersection typically at the top or bottom edge. This results in overly conservative tube integrity limits as longer flaws have lower %TW tube integrity limits.

• Each WBN Unit 1 operating cycle is assumed to be 1.5 EFPY. The average operating cycle duration since SG replacement is 1.33 EFPY. Assuming longer duration operating cycles leads to conservative degradation projections.

• The structural integrity performance criteria (SIPC) of three times normal operating pressure differential is conservatively determined. The value applied ignores pressure drops within the system between the measurement point and the secondary side of the tubing. The estimated pressure drop from the RSG normal water level to the steam pressure measurement point is 12.3 psi, which would add about 36.7 psi of conservatism to the SIPC.

Despite the above conservatisms, the WBN, Unit 1 OAs have consistently demonstrated margin to the tube integrity limits.

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Enclosure 1 Table 4 provides the OA projected maximum %TW degradation for each of the existing degradation mechanisms at the U1R19 refueling outage. The degradation projections are based on the Monte Carlo statistical method. Acceptable OA degradation projections are demonstrated at the U1R19 refueling outage, which would be the next WBN, Unit 1 RSG inspection based on the proposed TS amendment.

Table 4 Comparison of WBN Unit 1 RSG Inspection Results and OA Projections WBN Unit 1 Operational Assessment Projections Inspection Results Max %TW Wear Inds Inds Max %TW OA Projected Margin to OA Outage Remaining in Location Detected Plugged Detected Max %TW Max %TW Detected Method Service U-bend 0 0 0 0 U1R8 Note 1 ATSG 9 9 13 0 U1R9 No SG inspection U1R10 U-bend 8 0 12 12 Note 1 U1R11 ATSG 72 31 32 14 64.23 32.23 Arithmetic U1R12 No SG inspection U1R13 U-bend 59 0 27 27 42 15 Arithmetic U1R14 ATSG 419 0 37 37 42 5 Arithmetic U1R15 No SG inspection U1R16 U1R17 No SG Inspection Planned U1R18 U-bend 60.6 U1R19 Note 2 Note 2 Monte Carlo ATSG 59.9 Notes

1. This was the inspection when the degradation mechanism was first detected. OA projections are only performed for mechanisms that are existing at the prior inspection.

2. Inspection data points are pending results of the next SG inspection.

As the current TS limitation is three operating cycles between inspections, previously performed OA projections did not make degradation growth predictions over five operating cycles. However, an assessment of prior ability to support five-cycle operation can be made by comparing the maximum %TW indications remaining in-service between refueling outages. The largest indications remaining in-service following the most recent U1R14 inspection for both U-bend and ATSG wear have a %TW depth greater than any other prior refueling outage. Therefore, using the same degradation growth rates applied at U1R14 for a five operating cycle projection would have demonstrated acceptability against the SG performance criteria if performed based on results of the U1R8, U1R11, or U1R14 inspection.

3.2.8 Conclusion The current TS SG inspection interval requirements were developed when there was uncertainty around the performance of the Alloy 690TT tube material. Significant experience has been gained over the course of 30 years of service of this tube material type. The WBN Unit 1 RSGs are a second-generation evolutionary design, which CNL-20-053 E1-20 of 25

Enclosure 1 incorporate industry-operating experience. The RSGs have been inspected in-service by eddy current techniques four separate times over an operating period of greater than 12.0 EFPY since installation. Through these inspections, every tube in each RSG has been tested full length at least two times with techniques qualified for detection of all existing and potential degradation mechanisms.

The WBN RSG degradation mechanisms are well understood and exhibit predictable behavior. As such, the OAs performed to date have been both accurate and appropriately conservative.

TVA considers the proposed TS change to the SG tube inspection frequency to be a more effective and efficient way to collect data on all SG tubes over the 96 EFPM period.

Full bundle data collection, or 100% scope, can be achieved in an overall shorter period with no impact to safety or operation. Over the life of the plant, this results in more frequent inspections per tube. This change minimizes person-hours and dose, and reduces risk to the plant and personnel.

4.0 REGULATORY EVALUATION

4.1 APPLICABLE REGULATORY REQUIREMENTS AND CRITERIA General Design Criteria WBN, Unit 1 was designed to meet the intent of the "Proposed General Design Criteria for Nuclear Power Plant Construction Permits" published in July, 1967. The WBN construction permit was issued in January 1973. The UFSAR, however, addresses the General Design Criteria (GDC) published as Appendix A to 10 CFR 50 in July 1971.

Conformance with the GDCs is described in Section 3.1.2 of the UFSAR.

Each criterion listed below is followed by a discussion of the design features and procedures that meet the intent of the criteria. Any exception to the 1971 GDC resulting from the earlier commitments is identified in the discussion of the corresponding criterion.

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

Compliance with GDC 14 is described in Section 3.1.2.2 of the WBN UFSAR.

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

Compliance with GDC 15 is described in Section 3.1.2.2 of the WBN UFSAR.

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Enclosure 1 Criterion 16, Containment design 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.

Compliance with GDC 16 is described in Section 3.1.2.2 of the WBN UFSAR.

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

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

Compliance with GDC 30 is described in Section 3.1.2.4 of the WBN UFSAR.

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

Compliance with GDC 31 is described in Section 3.1.2.4 of the WBN UFSAR.

Criterion 32, Inspection of reactor coolant pressure boundary 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.

Compliance with GDC 32 is described in Section 3.1.2.4 of the WBN UFSAR.

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

• In Reference 4, NRC issued a license amendment for Arkansas Nuclear One, Unit 2, which granted a one-time change to revise the steam generator in-service inspection frequency TS requirements to allow a 40-month inspection interval after one inspection, rather than after two consecutive inspections.

• In Reference 5, NRC issued a license amendment for South Texas Project (STP),

Unit 1, which granted a one-time change to extend the steam generator in-service inspection frequency TS requirements from 40 months to 44 months.

• In Reference 6, NRC issued a license amendment for Virgil C. Summer Nuclear Station, Unit No. 1, which granted a one-time change to extend the steam generator in-service inspection frequency TS requirements from 40 months to 58 months.

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Enclosure 1 4.3 Significant Hazards Consideration Tennessee Valley Authority (TVA) proposes to revise the Watts Bar Nuclear Plant (WBN), Unit 1 Technical Specifications (TS) to allow a change to the steam generator (SG) inspection frequency. The proposed license amendment request (LAR) revises WBN Unit 1 TS 5.7.2.12, Steam Generator (SG) Program, and TS 5.9.9, Steam Generator Tube Inspection Report, to revise the required SG tube inspection frequency from every 72 effective full power months (EFPM) to every 96 EFPM and to incorporate Technical Specifications Task Force (TSTF) Technical Change Traveler 510, Revision 2, Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection. WBN operating experience supports this TS revision. No adverse impact to safety and reliability is expected as a result of such a change.

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

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

Response: No.

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

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Enclosure 1

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

Response: No.

The steam generator tubes are an integral part of the reactor coolant pressure boundary and, as such, are relied upon to maintain the primary system pressure and inventory. Revising the SG tube in-service inspection frequency will not alter their function or design. Inspections of the RSGs demonstrate that the RSGs do not have an active damage mechanism. The improved design of the RSGs [Alloy 90 thermally treated (Alloy 690TT) tubes], the ISI data and OAs also provide reasonable assurance that significant tube degradation is not likely to occur. Therefore, TVA concludes that this proposed change does not involve a significant reduction in a margin of safety.

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.

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Enclosure 1

6.0 REFERENCES

1. TVA letter to NRC, Watts Bar Nuclear Plant (WBN) - Unit 1 - Steam Generator Tube Inspection Report-Cycle 8, dated September 11, 2008 (ML082600068)

2. TVA Letter to NRC, Watts Bar Nuclear Plant (WBN) Unit 1-Cycle 11 Steam Generator Tube Inspection Report, dated April 19, 2013 (ML13114A252)

3. TVA Letter to NRC, Watts Bar Nuclear Plant (WBN) Unit 1-Cycle 14 Steam Generator Tube Inspection Report, dated July 28, 2017 (ML17209A554)

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

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

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

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

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

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Enclosure 2 Description and Assessment

Subject:

Application to Revise Watts Bar Nuclear Plant (WBN) Unit 1 Technical Specifications to Adopt TSTF-510, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection" CONTENTS

1.0 DESCRIPTION

................................................................................................................ 2 2.0 ASSESSMENT ................................................................................................................ 2 2.1 Applicability of Published Safety Evaluation ................................................................. 2 2.2 Optional Changes and Variations................................................................................. 2

3.0 REGULATORY ANALYSIS

............................................................................................. 3 3.1 No Significant Hazards Consideration Determination ................................................... 3 4.0 ENVIRONMENTAL EVALUATION .................................................................................. 4

5.0 REFERENCES

................................................................................................................ 5 CNL-20-053 E2-1 of 5

Enclosure 2 Description and Assessment

1.0 DESCRIPTION

The proposed change revises the Watts Bar Nuclear Plant (WBN), Unit 1 Technical Specifications (TS) 3.4.17, Steam Generator (SG) Tube Integrity, 5.7.2.12 Steam Generator (SG) Program, and 5.9.9, Steam Generator Tube Inspection Report. The proposed changes are needed to address implementation issues associated with the inspection periods, and address other administrative changes and clarifications.

The proposed amendment is consistent with TSTF-510, Revision 2, Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection.

2.0 ASSESSMENT 2.1 APPLICABILITY OF PUBLISHED SAFETY EVALUATION Tennessee Valley Authority (TVA) has reviewed TSTF-510, Revision 2, Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection, and the model safety evaluation dated October 19, 2011 (ML112101513) as announced in the Federal Register Notice dated October 27, 2011 (76 FR 66763). As described in the subsequent paragraphs, TVA has concluded that the justifications presented in TSTF-510 and the model safety evaluation prepared by the NRC staff are applicable to WBN Unit 1 and justify this amendment for the incorporation of the changes to the WBN Unit 1 TS.

2.2 OPTIONAL CHANGES AND VARIATIONS TVA is proposing the following variations from the TS changes described in the TSTF-510, Revision 2, or the applicable parts of the NRC staffs model safety evaluation dated October 19, 2011:

• The WBN Unit 1 TS utilize different numbering than the Standard Technical Specifications (STSs) on which TSTF-510 was based. Specifically, STS 5.5.9, Steam Generator (SG)

Program, is numbered as TS 5.7.2.12 in the WBN Unit 1 TS and STS 5.6.7, Steam Generator Tube Inspection Report, is numbered as TS 5.9.9 in the WBN Unit 1 TS. These differences are administrative and do not affect the applicability of TSTF-510 to the WBN, Unit 1 TS.

• The proposed change to WBN, Unit 1 TS 5.7.2.12.b.2 also revises the following verbiage Current verbiage: For design basis accidents that have a faulted steam generator, accident induced leakage is not to exceed 1.0 gallon per minute (gpm) for the faulted steam generator and 150 gallons per day (gpd) for the non-faulted steam generators. For design basis accidents that do not have a faulted steam generator, accident induced leakage is not to exceed 150 gpd per steam generator. This verbiage originated in References 1 and 2.

Revised verbiage: Leakage for all degradation mechanisms is not to exceed 150 gpd for each unfaulted SG. Leakage for all degradation mechanisms is not to exceed 1 gpm in the faulted SG. The proposed verbiage is consistent with the WBN Unit 2 TS and WBN dual-unit Updated Final Safety Analysis Report (UFSAR) Section 15.5.4 as described in References 3, 4, and 5.

• The proposed changes to WBN, Unit 1 TS 5.7.2.12 and 5.9.9 also reflect the changes described in Enclosure 1 to this submittal.

CNL-20-053 E2-2 of 5

Enclosure 2

3.0 REGULATORY ANALYSIS

3.1 NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION TVA requests adoption of an approved change to the standard technical specifications (STS) into the Watts Bar Nuclear Plant (WBN) Unit 1 plant specific Technical Specifications (TS), to revise TS 3.4.17, Steam Generator (SG) Tube Integrity, 5.7.2.12 Steam Generator (SG)

Program, and 5.9.9, Steam Generator Tube Inspection Report, to address inspection periods and other administrative changes and clarifications.

As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below:

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

Response: No The proposed change revises the Steam Generator (SG) Program to modify the frequency of verification of SG tube integrity and SG tube sample selection. A steam generator tube rupture (SGTR) event is one of the design basis accidents that are analyzed as part of a plants licensing basis. The proposed SG tube inspection frequency and sample selection criteria will continue to ensure that the SG tubes are inspected such that the probability of a SGTR is not increased. The consequences of a SGTR are bounded by the conservative assumptions in the design basis accident analysis. The proposed change will not cause the consequences of a SGTR to exceed those assumptions.

Therefore, it is concluded that this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed changes to the SG Program will not introduce any adverse changes to the plant design basis or postulated accidents resulting from potential tube degradation. The proposed change does not affect the design of the SGs or their method of operation. In addition, the proposed change does not impact any other plant system or component.

Therefore, it is concluded that this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The SG tubes in pressurized water reactors are an integral part of the reactor coolant pressure boundary and, as such, are relied upon to maintain the primary systems pressure and inventory. As part of the reactor coolant pressure boundary, the SG tubes are unique in that they are also relied upon as a heat transfer surface between the primary and secondary CNL-20-053 E2-3 of 5

Enclosure 2 systems such that residual heat can be removed from the primary system. In addition, the SG tubes also isolate the radioactive fission products in the primary coolant from the secondary system. In summary, the safety function of a SG is maintained by ensuring the integrity of its tubes.

Steam generator tube integrity is a function of the design, environment, and the physical condition of the tube. The proposed change does not affect tube design or operating environment. The proposed change will continue to require monitoring of the physical condition of the SG tubes such that there will not be a reduction in the margin of safety compared to the current requirements.

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

Based on the above, TVA concludes that the proposed change presents no 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.

4.0 ENVIRONMENTAL EVALUATION The proposed change 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 change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c) (9).

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.

CNL-20-053 E2-4 of 5

Enclosure 2

5.0 REFERENCES

1. TVA letter to NRC, WBN-TS-05-10, Watts Bar Nuclear Plant Unit 1 - Technical Specifications (TS) Change WBN-TS-05 Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity - Request for Additional Information (TAC No. MC9271), dated September 8, 2006 (ML062550371)

2. NRC letter to TVA, Watts Bar Nuclear Plant, Unit 1 - Issuance of Amendment Regarding Steam Generator Tube Integrity (TS-05-10) (TAC No. MC9271), dated November 3, 2006 (ML062910090)

3. TVA letter to NRC, 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), dated May 14, 2018 (ML18138A232)

4. TVA letter to NRC, CNL-18-128, Response to Request for Additional Information Regarding 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) (EPID L-2018-LLA-0143), dated November 8, 2018 (ML18312A402)

5. NRC letter to TVA, Watts Bar Nuclear Plant, Unit 2- Issuance of Amendment Regarding Application to Revise Technical Specifications for Use Of Voltage-Based Alternate Repair Criteria in Accordance With Generic Letter 95-05 (EPID L-2018-LLA-0143), dated June 3, 2019 (ML190638721)

CNL-20-053 E2-5 of 5

Enclosure 3 Proposed TS Changes (Mark-Ups) for WBN Unit 1 CNL-20-053

SG Tube Integrity 3.4.17 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.17 STEAM GENERATOR (SG) TUBE INTEGRITY LCO 3.4.17 SG tube integrity shall be maintained AND All SG tubes satisfying the tube repairplugging criteria shall be plugged in accordance with the Steam Generator Program.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS

---

NOTE--------------------------------------------------------

Separate Condition entry is allowed for each SG tube.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more SG tubes A.1 Verify tube integrity of the 7 days satisfying the tube affected tube(s) is maintained pluggingrepair criteria and not until the next refueling outage plugged in accordance with or SG tube inspection.

the Steam Generator Program AND A.2 Plug the affected tube(s) in Prior to entering accordance with the Steam MODE 4 following Generator Program. the next refueling outage or SG tube inspection B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A not met. AND OR B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SG tube integrity not maintained Watts Bar-Unit 1 3.4-43 Amendment 65, 112, XXX

SG Tube Integrity 3.4.17 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.17.1 Verify steam generator tube integrity in accordance with the In accordance with Steam Generator Program. the Steam Generator Program SR 3.4.17.2 Verify that each inspected SG tube that satisfies the tube Prior to entering repairplugging criteria is plugged in accordance with the MODE 4 following a Steam Generator Program. SG tube inspection.

Watts Bar-Unit 1 3.4-44 Amendment 65, 112, XXX

Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals (continued) 5.7.2.12 Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:

a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the as found condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The as found condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged, to confirm that the performance criteria are being met.

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

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

2. Accident induced leakage performance criterion: The primary-to-secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage for all degradation mechanisms is not to exceed 150 gpd for each unfaulted SG. Leakage for all degradation mechanisms is not to exceed 1 gpm in the faulted SGFor design basis accidents that have a faulted steam generator, accident induced leakage is not to exceed 1.0 gallon per minute (gpm) for the faulted steam generator and 150 gallons per day (gpd) for the non-faulted steam generators. For design basis accidents that do not (continued)

Watts Bar-Unit 1 5.0-15 Amendment 27, 38, 44, 65, XXX

Procedures, Programs, and Manuals 5.7 have a faulted steam generator, accident induced leakage is not to exceed 150 gpd per steam generator.

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

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

c. Provisions for SG tube pluggingrepair criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube pluggingrepair criteria. The tube-to-tubesheet weld is not part of the tube.

In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation 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 replacemeninstallationt.

2. After the first refueling outage following SG installation, inspect each SG at least every 96 effective full power months. In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in a and b below. 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 repair criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. 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 (continued)

Watts Bar-Unit 1 5.0-16 Amendment 27, 38, 44, 65, XXX

Procedures, Programs, and Manuals 5.7 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. Each inspection period defined below 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.

(continued)

Watts Bar-Unit 1 5.0-17 Amendment 27, 38, 44, 65, XXX

Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals (continued) 5.7.2.12 Steam Generator (SG) Program (continued) a) After the first refueling outage following SG installation, inspect 100% of the tubes during the next 144 effective full power months.

This constitutes the first inspection period.

b) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the second and subsequent inspection periods.Inspect 100% of the tubes at sequential periods of 144, 108, 72, and thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SGs shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected.

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 inspectionsis less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

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

(continued)

Watts Bar-Unit 1 5.0-1716a Amendment 27, 38, 44, 65, XXX

Reporting Requirements 5.9 5.9 Reporting Requirements (continued) 5.9.7 EDG Failures Report If an individual emergency diesel generator (EDG) experiences four or more valid failures in the last 25 demands, these failures and any nonvalid failures experienced by that EDG in that time period shall be reported within 30 days. Reports on EDG failures shall include the information recommended in Regulatory Guide 1.9, Revision 3, Regulatory Position C.4, or existing Regulatory Guide 1.108 reporting requirement.

5.9.8 PAMS Report When a Report is required by Condition B or F of LCO 3.3.3, Post Accident Monitoring (PAM)

Instrumentation, a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.9.9 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.7.2.12, Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG,

b. Active dDegradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism,

d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e. Number of tubes plugged during the inspection outage for each active degradation mechanism,

f. The number and percentage of tubes plugged to date, and effective plugging percentage in each steam generatorTotal number and percentage of tubes plugged to date,

g. The results of condition monitoring, including the results of tube pulls and in-situ testing, and

h. Discuss trending of tube degradation over the inspection interval and provide comparison of the prior operational assessment degradation projections to the as-found condition.The effective plugging percentage for all plugging in each SG.

Watts Bar-Unit 1 5.0-32 Amendment 27, 38, 65, 96, XXX

Enclosure 4 Proposed TS Changes (Final Typed) for WBN Unit 1 CNL-20-053

SG Tube Integrity 3.4.17 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.17 STEAM GENERATOR (SG) TUBE INTEGRITY LCO 3.4.17 SG tube integrity shall be maintained AND All SG tubes satisfying the tube plugging criteria shall be plugged in accordance with the Steam Generator Program.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS

---

NOTE--------------------------------------------------------

Separate Condition entry is allowed for each SG tube.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more SG tubes A.1 Verify tube integrity of the 7 days satisfying the tube plugging affected tube(s) is maintained criteria and not plugged in until the next refueling outage accordance with the Steam or SG tube inspection.

Generator Program AND A.2 Plug the affected tube(s) in Prior to entering accordance with the Steam MODE 4 following Generator Program. the next refueling outage or SG tube inspection B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A not met. AND OR B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SG tube integrity not maintained Watts Bar-Unit 1 3.4-43 Amendment 65, 112, XXX

SG Tube Integrity 3.4.17 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.17.1 Verify steam generator tube integrity in accordance with the In accordance with Steam Generator Program. the Steam Generator Program SR 3.4.17.2 Verify that each inspected SG tube that satisfies the tube Prior to entering plugging criteria is plugged in accordance with the Steam MODE 4 following a Generator Program. SG tube inspection.

Watts Bar-Unit 1 3.4-44 Amendment 65, 112, XXX

Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals (continued) 5.7.2.12 Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following:

a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the as found condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The as found condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged, to confirm that the performance criteria are being met.

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

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

2. Accident induced leakage performance criterion: The primary-to-secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage for all degradation mechanisms is not to exceed 150 gpd for each unfaulted SG. Leakage for all degradation mechanisms is not to exceed 1 gpm in the faulted SG.

(continued)

Watts Bar-Unit 1 5.0-15 Amendment 27, 38, 44, 65, XXX

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

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.

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

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

2. After the first refueling outage following SG installation, inspect each SG at least every 96 effective full power months. In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in a and b below. 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 repair criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. 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. Each inspection period defined below 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.

(continued)

Watts Bar-Unit 1 5.0-16 Amendment 27, 38, 44, 65, XXX

Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals (continued) 5.7.2.12 Steam Generator (SG) Program (continued) a) After the first refueling outage following SG installation, inspect 100% of the tubes during the next 144 effective full power months.

This constitutes the first inspection period.

b) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the second and subsequent inspection periods.

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.

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

(continued)

Watts Bar-Unit 1 5.0-16a Amendment 27, 38, 44, 65, XXX

Reporting Requirements 5.9 5.9 Reporting Requirements (continued) 5.9.7 EDG Failures Report If an individual emergency diesel generator (EDG) experiences four or more valid failures in the last 25 demands, these failures and any nonvalid failures experienced by that EDG in that time period shall be reported within 30 days. Reports on EDG failures shall include the information recommended in Regulatory Guide 1.9, Revision 3, Regulatory Position C.4, or existing Regulatory Guide 1.108 reporting requirement.

5.9.8 PAMS Report When a Report is required by Condition B or F of LCO 3.3.3, Post Accident Monitoring (PAM)

Instrumentation, a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.9.9 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.7.2.12, Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG,

b. Degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism,

d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e. Number of tubes plugged during the inspection outage for each degradation mechanism,

f. The number and percentage of tubes plugged to date, and effective plugging percentage in each steam generator,

g. The results of condition monitoring, including the results of tube pulls and in-situ testing, and

h. Discuss trending of tube degradation over the inspection interval and provide comparison of the prior operational assessment degradation projections to the as-found condition.

Watts Bar-Unit 1 5.0-32 Amendment 27, 38, 65, 96, XXX

Enclosure 5 Proposed TS Bases Changes (Mark-Ups) for WBN Unit 1 (For Information Only)

CNL-20-053

SG Tube Integrity B 3.4.17 BASES (continued)

APPLICABLE The steam generator tube rupture (SGTR) accident is the limiting design SAFETY basis event for SG tubes and avoiding an SGTR is the basis for this ANALYSES Specification. The analysis of a SGTR event assumes a bounding primary to secondary LEAKAGE rate equal to the operational LEAKAGE rate limits in LCO 3.4.13, RCS Operational LEAKAGE, plus the leakage rate associated with a double-ended rupture of a single tube. The accident analysis for a SGTR assumes the contaminated secondary fluid is only briefly released to the atmosphere via safety valves and the majority is discharged to the main condenser.

The analysis for design basis accidents and transients other than a SGTR assume the SG tubes retain their structural integrity (i.e., they are assumed not to rupture). In these analyses, the steam discharge to the atmosphere is based on the total primary to secondary LEAKAGE from 150 gallons per day (gpd) per steam generator and 1 gallon per minute (gpm) in the faulted steam generator.

For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT I-131 is assumed to be equal to the LCO 3.4.16 RCS Specific Activity, limits. For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel. The dose consequences of these events are within the limits of GDC 19 (Ref. 2), and 10 CFR 100 (Ref. 3) or the NRC approved licensing basis.

Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The LCO requires that SG tube integrity be maintained. The LCO also requires that all SG tubes that satisfy the pluggingrepair criteria be plugged in accordance with the Steam Generator Program.

During an SG inspection, any inspected tube that satisfies the Steam Generator Program pluggingrepair criteria is removed from service by plugging. If a tube was determined to satisfy the pluggingrepair criteria but was not plugged, the tube may still have tube integrity.

In the context of this Specification, a SG tube is defined as the entire length of the tube, including the tube wall, between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet weld is not considered part of the tube.

A SG tube has tube integrity when it satisfies the SG performance criteria. The SG performance criteria are defined in Specification 5.7.2.12, Steam Generator Program, and describe acceptable SG tube performance. The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria (continued)

Watts Bar-Unit 1 B 3.4-87 Revision 82 Amendment 65, XXX

SG Tube Integrity B 3.4.17 BASES LCO The operational LEAKAGE performance criterion provides an observable (continued) indication of SG tube conditions during plant operation. The limit on operational LEAKAGE is contained in LCO 3.4.13, RCS Operational LEAKAGE, and limits primary to secondary LEAKAGE through any one SG to 150 gallons per day.

This limit is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or a main steam line break. If this amount of LEAKAGE is due to more than one crack, the cracks are very small, and the above assumption is conservative.

APPLICABILITY Steam generator tube integrity is challenged when the pressure differential across the tubes is large. Large differential pressures across SG tubes can only be experienced in MODE 1, 2, 3, or 4.

RCS conditions are far less challenging in MODES 5 and 6 than during MODES 1, 2, 3, and 4. In MODES 5 and 6, primary to secondary differential pressure is low, resulting in lower stresses and reduced potential for LEAKAGE.

ACTIONS The ACTIONS are modified by a Note that the Conditions may be entered independently for each SG tube. This is acceptable because the Required Actions provide appropriate compensatory actions for each affected SG tube.

Complying with the Required Actions may allow for continued operation, and subsequent affected SG tubes are governed by subsequent Condition entry, and application of associated Required Actions.

A.1 and A.2 Condition A applies if it is discovered that one or more SG tubes examined in an inservice inspection satisfy the tube pluggingrepair criteria but were not plugged in accordance with the Steam Generator Program as required by SR 3.4.17.2. An evaluation of SG tube integrity of the affected tube(s) must be made. Steam generator tube integrity is based on meeting the SG performance criteria described in the Steam Generator Program. The SG pluggingrepair criteria define limits on SG tube degradation that allow for flaw growth between inspections while still providing assurance that the SG performance criteria will continue to be met. In order to determine if a SG tube that should have been plugged, has tube integrity, an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met until the next refueling outage or SG tube inspection. The tube integrity determination is based on the estimated condition of the tube at the time the situation is discovered and the estimated growth of the degradation prior to the next SG tube inspection. If it is determined that tube integrity is not being maintained, Condition B applies.

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SG Tube Integrity B 3.4.17 BASES ACTIONS A.1 and A.2 (continued)

A Completion Time of 7 days is sufficient to complete the evaluation while minimizing the risk of plant operation with a SG tube that may not have tube integrity.

If the evaluation determines that the affected tube(s) have tube integrity, Required Action A.2 allows plant operation to continue until the next refueling outage or SG inspection provided the inspection interval continues to be supported by an operational assessment that reflects the affected tubes.

However, the affected tube(s) must be plugged prior to entering MODE 4 following the next refueling outage or SG inspection. This Completion Time is acceptable since operation until the next inspection is supported by the operational assessment.

B.1 and B.2 If the Required Actions and associated Completion Times of Condition A are not met or if SG tube integrity is not being maintained, the reactor must be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

The allowed Completion Times are reasonable, based on operating experience, to reach the desired plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.17.1 REQUIREMENTS During shutdown periods the SGs are inspected as required by this SR and the Steam Generator Program. NEI 97-06, Steam Generator Program Guidelines (Ref. 1), and its referenced EPRI Guidelines, establish the content of the Steam Generator Program. Use of the Steam Generator Program ensures that the inspection is appropriate and consistent with accepted industry practices.

During SG inspections a condition monitoring assessment of the SG tubes is performed. The condition monitoring assessment determines the as found condition of the SG tubes. The purpose of the condition monitoring assessment is to ensure that the SG performance criteria have been met for the previous operating period.

The Steam Generator Program determines the scope of the inspection and the methods used to determine whether the tubes contain flaws satisfying the tube pluggingrepair criteria. Inspection scope (i.e., which tubes or areas of tubing within the SG are to be inspected) is a function of existing and potential degradation locations. The Steam Generator Program also specifies the inspection methods to be used to find potential degradation. Inspection methods are a function of degradation morphology, nondestructive examination (NDE) technique capabilities, and inspection locations.

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SG Tube Integrity B 3.4.17 BASES SURVEILLANCE SR 3.4.17.1 (continued)

REQUIREMENTS The Steam Generator Program defines the Frequency of SR 3.4.17.1. The Frequency is determined by the operational assessment and other limits in the SG examination guidelines (Ref. 6). The Steam Generator Program uses information on existing degradations and growth rates to determine an inspection Frequency that provides reasonable assurance that the tubing will meet the SG performance criteria at the next scheduled inspection. In addition, Specification 5.7.2.12 contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections. If crack indications are found in any SG tube, the maximum inspection interval for all affected and potentially affected SGs is restricted by Specification 5.7.2.12 until subsequent inspections support extending the inspection interval.

SR 3.4.17.2 During an SG inspection, any inspected tube that satisfies the Steam Generator Program pluggingrepair criteria is removed from service by plugging. The tube pluggingrepair criteria delineated in Specification 5.7.2.12 are intended to ensure that tubes accepted for continued service satisfy the SG performance criteria with allowance for error in the flaw size measurement and for future flaw growth. In addition, the tube pluggingrepair criteria, in conjunction with other elements of the Steam Generator Program, ensure that the SG performance criteria will continue to be met until the next inspection of the subject tube(s). Reference 1 provides guidance for performing operational assessments to verify that the tubes remaining in service will continue to meet the SG performance criteria.

The Frequency of prior to entering MODE 4 following a SG inspection ensures that the Surveillance has been completed and all tubes meeting the pluggingrepair criteria are plugged prior to subjecting the SG tubes to significant primary to secondary pressure differential.

REFERENCES 1. NEI 97-06, Steam Generator Program Guidelines.

2. 10 CFR 50 Appendix A, GDC 19, Control Room.

3. 10 CFR 100, Reactor Site Criteria.

4. ASME Boiler and Pressure Vessel Code,Section III, Subsection NB.

5. Draft Regulatory Guide 1.121, Basis for Plugging Degraded Steam Generator Tubes, August 1976.

6. EPRI, Pressurized Water Reactor Steam Generator Examination Guidelines.

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