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Expedited Application for Approval to Use an Alternate Method of Determining Probability of Detection for (WBN TS-391-20-024)
	ML20358A141
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Site:	Watts Bar
Issue date:	12/23/2020
From:	Polickoski J; Tennessee Valley Authority
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
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ML20358A140	List: CNL-20-104, Expedited Application for Approval to Use an Alternate Method of Determining Probability of Detection for (WBN TS-391-20-024);
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Proprietary Information Withhold Under 10 CFR § 2.390 This letter is decontrolled when separated from Enclosure 2 Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 CNL-20-104 December 23, 2020 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Watts Bar Nuclear Plant Unit 2 Facility Operating License No. NPF-96 NRC Docket No. 50-391

Subject:

Expedited Application for Approval to Use an Alternate Method of Determining Probability of Detection for the Watts Bar Nuclear Plant, Unit 2 Steam Generators (WBN TS-391-20-024)

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

The proposed license amendment request revises the WBN dual-unit Updated Final Safety Analysis Report (UFSAR) to apply an eddy current probability of detection (POD) of 0.9 to indications of axial outside diameter stress corrosion cracking at tube support plates with bobbin voltage amplitudes of greater than or equal to () 3.2 volts, but less than (<) 6.0 volts and a POD of 0.95 to indications of 6.0 volts in the WBN, Unit 2 steam generators (SG), as described in the enclosures to this submittal, for the beginning of cycle (BOC) voltage distribution in support of the WBN, Unit 2 operational assessment (OA). WBN, Unit 2 Technical Specification (TS) 5.7.2.12, "Steam Generator (SG) Program," and WBN, Unit 2 TS 5.9.9, Steam Generator Tube Inspection Report, are based on Nuclear Regulatory Commission (NRC) Generic Letter (GL) 95-05, "Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking," which requires the application of a POD of 0.6 to all previous bobbin indications for the determination of the indication voltage distribution for the BOC. Therefore, the use of the proposed POD values for the BOC voltage distribution in support of the WBN, Unit 2 OA, is an exception to GL 95-05 and requires NRC approval. A POD of 0.6, in accordance with GL 95-05, will be used for indications less than 3.2 volts.

The proposed POD values will only be used until the WBN, Unit 2 SGs are replaced, which are planned for the WBN, Unit 2 Cycle 4 refueling outage (U2R4) scheduled for Spring 2022.

Based on the results of the SG inspections during U2R3, TVA plans to perform a WBN, Unit 2 mid-cycle SG inspection, at which time the proposed POD values will be reassessed.

Proprietary Information Withhold Under 10 CFR § 2.390 This letter is decontrolled when separated from Enclosure 2

Proprietary Information Withhold Under 10 CFR § 2.390 This letter is decontrolled when separated from Enclosure 2 U.S. Nuclear Regulatory Commission CNL-20-104 Page 2 December 23, 2020 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. Attachment 1 to Enclosure 1 to this submittal provides the existing WBN, UFSAR pages marked up to show the proposed changes. Attachment 2 to Enclosure 1 to this submittal provides the existing WBN, UFSAR pages retyped to show the proposed changes.

There are no corresponding TS changes required to apply the proposed alternate POD values, as the associated requirements are only discussed in GL 95-05, and because WBN, Unit 2 TS 5.7.2.12 and TS 5.9.9 do not contain requirements for the bobbin POD. In support of the technical evaluation in Enclosure 1, Enclosure 2 contains Westinghouse Electric Company LLC (Westinghouse) Letter Report, LTR-CDMP-20-41 P-Attachment, Revision 0, Watts Bar U2R3 Steam Generator Alternate Repair Criteria Generic Letter 95-05 Probability of Detection Methodology for 90-Day Report. contains information that Westinghouse considers to be proprietary in nature pursuant to 10 CFR 2.390, "Public inspections, exemptions, requests for withholding,"

paragraph (a)(4). Enclosure 3 contains a non-proprietary version of Enclosure 2. Enclosure 4 provides the Westinghouse Application for Withholding Proprietary Information from Public Disclosure CAW-20-5136 affidavit supporting this proprietary withholding request. The affidavit sets forth the basis on which the information may be withheld from public disclosure by the NRC and addresses with specificity the considerations listed in paragraph (b)(4) of Section 2.390.

Accordingly, TVA requests that the information, which is proprietary to Westinghouse, be withheld from public disclosure in accordance with 10 CFR Section 2.390. Correspondence with respect to the copyright or proprietary aspects of the items listed above or the supporting Westinghouse affidavit should reference CAW-20-5136 and should be addressed to Camille T. Zozula, Manager, Regulatory Compliance & Corporate Licensing, Westinghouse Electric Company, 1000 Westinghouse Drive, Suite 165, Cranberry Township, Pennsylvania 16066.

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.

WBN, Unit 2 TS 5.9.9 requires that a report be submitted to the NRC within 90 days after initial entry into MODE 4 following completion of the inspection performed in accordance with WBN, Unit 2 TS 5.7.2.12, when the voltage based alternate repair criteria have been applied.

TVA entered Mode 4 following the WBN U2R3 refueling outage on November 16, 2020, wherein a SG inspection was performed in accordance with WBN, Unit 2 TS 5.7.2.12. Because the 90-day report required by WBN, Unit 2 TS 5.9.9 will apply the proposed alternate POD values, TVA requests NRC approval of the proposed license amendment on an expedited basis by February 8, 2021, with implementation by February 14, 2021 (i.e., 90 days following entry into Mode 4).

Proprietary Information Withhold Under 10 CFR § 2.390 This letter is decontrolled when separated from Enclosure 2

Proprietary Information Withhold Under 10 CFR § 2.390 This letter is decontrolled when separated from Enclosure 2 U.S. Nuclear Regulatory Commission CNL-20-104 Page 3 December 23, 2020 There are no new regulatory commitments associated with this submittal. Please address any questions regarding this request to Kimberly D. Hulvey, Senior Manager, Fleet Licensing, at (423) 751-3275.

I declare under penalty of perjury that the foregoing is true and correct. Executed on this 23rd day of December 2020.

Respectfully, James T. Polickoski Director, Nuclear Regulatory Affairs

Enclosures:

1. Evaluation of Proposed Change

2. Westinghouse Letter Report, LTR-CDMP-20-41 P-Attachment, Revision 0 (Proprietary)

3. Westinghouse Letter Report, LTR-CDMP-20-41 NP-Attachment, Revision 0 (Non-Proprietary)

4. Westinghouse Electric Company LLC Application for Withholding Proprietary Information From Public Disclosure (Affidavit CAW-20-5136) 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 Proprietary Information Withhold Under 10 CFR § 2.390 This letter is decontrolled when separated from Enclosure 2

Enclosure 1 Evaluation of Proposed Change

Subject:

Expedited Application for Approval to Use an Alternate Method of Determining Probability of Detection for the Watts Bar Nuclear Plant (WBN),

Unit 2 Steam Generators (WBN TS-391-20-024)

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

................................................................................................. 3 3.1 Background ...................................................................................................................... 3 3.2 System Description .......................................................................................................... 4 3.3 Technical Analysis............................................................................................................ 6

4.0 REGULATORY EVALUATION

............................................................................................ 7 4.1 Applicable Regulatory Requirements and Criteria .......................................................... 7 4.2 Precedent ......................................................................................................................... 8 4.3 No Significant Hazards Consideration ............................................................................. 9 4.4 Conclusion ...................................................................................................................... 10

5.0 ENVIRONMENTAL CONSIDERATION

............................................................................ 10

6.0 REFERENCES

.................................................................................................................. 11 Attachments:

1. Proposed UFSAR Changes (Mark-Ups) for WBN, Unit 2

2. Proposed UFSAR Changes (Final Typed) for WBN, Unit 2 CNL-20-104 E1-1 of 11

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 Facility Operating License No. NPF-96 for the Watts Bar Nuclear Plant (WBN), Unit 2.

The proposed license amendment request (LAR) revises the WBN dual-unit Updated Final Safety Analysis Report (UFSAR) to apply an eddy current probability of detection (POD) of 0.9 to indications of axial outside diameter stress corrosion cracking (ODSCC) at tube support plates (TSP) with bobbin voltage amplitudes of greater than or equal to

() 3.2 volts, but less than (<) 6.0 volts and a POD of 0.95 to indications of 6.0 volts in the WBN, Unit 2 steam generators (SG) for the beginning of cycle (BOC) voltage distribution in support of the WBN, Unit 2 operational assessment (OA). WBN, Unit 2 Technical Specification (TS) 5.7.2.12, "Steam Generator (SG) Program," and WBN, Unit 2 TS 5.9.9, Steam Generator Tube Inspection Report, are based on Nuclear Regulatory Commission (NRC) Generic Letter (GL) 95-05, "Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking," which requires the application of a POD of 0.6 to all previous bobbin indications for the determination of the indication voltage distribution for the BOC.

Therefore, the use of the proposed POD values for the BOC voltage distribution in support of the WBN, Unit 2 OA, is an exception to GL 95-05 and requires NRC approval.

2.0 DETAILED DESCRIPTION

2.1 PROPOSED CHANGE

S The Unit 2 Only section of the WBN dual-unit UFSAR Section 5.5.2.4, which refers to GL 95-05, is revised to add the following sentence:

As an alternative to the probability of detection of 0.6 required by GL 95-05, a probability of detection (POD) of 0.9 will be applied to indications of axial ODSCC at tube support plates with bobbin voltage amplitudes of greater than or equal to 3.2 volts, but less than 6.0 volts, and a POD of 0.95 will be applied to indications of axial ODSCC at tube support plates with bobbin voltage amplitudes of greater than or equal to 6.0 volts until the Unit 2 Steam Generators are replaced(26). A POD of 0.6, in accordance with GL 95-05, will be used for indications less than 3.2 volts.

Additionally, the Reference Section in UFSAR Section 5.5 will be revised to add a new Reference 26 to reflect the NRC approval of this LAR.

Attachment 1 to Enclosure 1 to this submittal provides the existing WBN, UFSAR pages marked up to show the proposed changes. Attachment 2 to Enclosure 1 to this submittal provides the existing WBN, UFSAR pages retyped to show the proposed changes.

There are no corresponding TS changes required to apply the proposed alternate POD values, as the associated requirements are only discussed in GL 95-05, and because WBN, Unit 2 TS 5.7.2.12 and 5.9.9 do not contain requirements for the bobbin POD.

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Enclosure 1 2.2 CONDITION INTENDED TO RESOLVE During the WBN, Unit 2 Cycle 3 refueling outage (U2R3), TVA determined that the WBN, U2R3 inspection results for SG3 represented a conditional burst probability of 3.0 x 10-2, which exceeds the GL 95-05 based limit of 1 x 10-2 in WBN, Unit 2 TS 5.9.9.

Accordingly, on November 11, 2020, TVA reported this information to the NRC in accordance with 10 CFR 50.72(b)(3)(ii)(A). Based on the U2R3 SG inspections, the preliminary OA evaluation has determined that that SG3 and SG4 did not meet the criteria for a full cycle OA utilizing the standard GL 95-05 evaluation methods for POD.

Accordingly, a POD of 0.9 to indications of axial ODSCC at TSP with bobbin voltage amplitudes of 3.2 volts, but < 6.0 volts, and a POD of 0.95 to indications of 6.0 volts, are needed to support continued operation of the WBN, Unit 2 SG, and will only be used until the WBN, Unit 2 SGs are replaced, which are planned for the U2R4 refueling outage scheduled for Spring 2022. A POD of 0.6, in accordance with GL 95-05, will be used for indications less than 3.2 volts.

3.0 TECHNICAL EVALUATION

3.1 BACKGROUND

The SG tubes constitute more than half of the reactor coolant pressure boundary (RCPB) area. Design of the RCPB for structural and leakage integrity is a requirement under the General Design Criteria (GDC) 10 CFR 50, Appendix A. Specific requirements governing the maintenance and inspection of SG tubes are contained in the WBN, Unit 2 TS and Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. TVA also implements the requirements of Nuclear Energy Institute (NEI) 97-06, "Steam Generator Program Guidelines." These include requirements for periodic inservice inspection of the tubing, flaw acceptance criteria (i.e., repair limits for plugging), and primary-to-secondary leakage limits. These requirements have formed the basis for assuring adequate SG tube integrity. SG tube plugging limits are specified in the WBN, Unit 2 TS. WBN, Unit 2 TS 5.7.2.12.c requires that flawed tubes be removed from service by plugging or repair if the depths of the flaws are greater than or equal to 40 percent through-wall, unless the degradation is subject to the voltage-based tube support plate axial ODSCC repair criteria or the F* repair criteria. WBN, Unit 2 TS 5.7.2.12.c repair limits ensure that tubes accepted for continued service will retain adequate structural and leakage integrity during normal operating, transient, and postulated accident conditions, consistent with GDC 14, 15, 16, 30, 31, and 32. Structural integrity refers to maintaining adequate margins against gross failure, rupture, or collapse of the SG tubing. Leakage integrity refers to limiting primary-to-secondary leakage to within acceptable limits.

The generic criteria for voltage-based limits for axial ODSCC at TSP are contained in GL 95-05 in accordance with the WBN, Unit 2 TS. These criteria rely on empirically derived correlations between a nondestructive inspection parameter, the bobbin coil voltage, and tube burst pressure and leak rate. The GL 95-05 guidance ensures structural and leakage integrity continue to be maintained at acceptable levels consistent with the requirements of 10 CFR 50 and 10 CFR 100. GL 95-05 focuses on maintaining tube structural integrity during the full range of normal, transient, and postulated accident conditions with adequate allowance for eddy current test uncertainty and flaw growth projected to occur during the next operating cycle. In order to ensure the structural and leakage integrity of the tube until the next scheduled inspection, GL 95-05 specifies CNL-20-104 E1-3 of 11

Enclosure 1 a methodology to determine the conditional burst probability and the total primary to secondary leak rate from an affected SG during a postulated main steam line break (MSLB) event. The methodology in WCAP-14277, Revision 1, "SLB Leak Rate and Tube Burst Probability Analysis Methods for ODSCC at TSP Intersections," dated December 1996, is used to implement the GL 95-05 structural integrity methodology.

A probabilistic analysis to quantify the potential for SG tube ruptures given a MSLB event is performed per WCAP-14277, Revision 1, and compared to a reporting threshold value of 1 x 10- 2 per cycle as required by GL 95-05. This threshold value provides assurance that the probability of burst (POB) is acceptable considering the assumptions of the calculation and the results of the staff's generic risk assessment for SGs contained in NUREG-0844, "NRC Integrated Program for the Resolution of Unresolved Safety Issues A-3, A-4, and A-5 Regarding Steam Generator Tube Integrity." Failure to meet this threshold value indicates axial ODSCC confined to within the thickness of the TSP could contribute a significant factor of the overall conditional probability of tube rupture for all forms of degradation assumed and evaluated as acceptable in NUREG-0844.

The calculation of conditional burst probability is, in part, a function of the POD and the resulting indication voltage distribution at BOC. The indication voltage distribution at BOC is based on consideration of all previous bobbin indications that were detected at the BOC, including those that were plugged. The POB threshold value of 1 x 10-2 is contained in WBN, Unit 2 TS 5.9.9.

Accident leakage is determined using the methodology described in Revision 1 of WCAP-14277 to calculate the SG tube leakage from the faulted SG during a postulated MSLB event. The methodology consists of the following two major components:

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

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

This methodology for calculating tube leakage is consistent with the guidance in GL 95-05.

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

The head is divided into inlet and outlet chambers by a vertical partition plate extending from the head to the tubesheet. Steam is generated on the shell side and flows upward through the moisture separators to the outlet nozzle at the top of the vessel. The WBN, Unit 2 SG have Alloy 600 mill annealed (Alloy 600MA) tube material. Details of the WBN, Unit 2 SGs are described in the UFSAR Section 5.5.2, Steam Generator.

Figure 5.5-3b of the WBN UFSAR shows the design of the WBN, Unit 2 SGs. Materials of construction for the WBN, Unit 2 SGs 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 ASME Code Section III.

CNL-20-104 E1-4 of 11

Enclosure 1 The SG tubes in pressurized water reactors have several important safety functions.

The SG tubes are an integral part of the 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 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 NEI 97-06, which is supported by EPRI guidelines, including:

Pressurized Water Reactor (PWR) Steam Generator Examination Guidelines

SG Integrity Assessment Guidelines

SG 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.

CNL-20-104 E1-5 of 11

Enclosure 1

3.3 TECHNICAL ANALYSIS

Enclosure 2 contains Westinghouse Electric Company LLC (Westinghouse) Letter Report, LTR-CDMP-20-41 P-Attachment, Revision 0, Watts Bar U2R3 Steam Generator Alternate Repair Criteria Generic Letter 95-05 Probability of Detection Methodology for 90-Day Report, which is a technical analysis performed by Westinghouse to determine the appropriateness of the alternate PODs for application specific to the WBN, Unit 2 SGs.

Table 1 provides a listing of indications of axial ODSCC at tube support plates for which the proposed PODs will currently be applied. The indications listed in Table 1 had measured bobbin voltages 3.2 volts for the SG inspection performed during the WBN U2R3 refueling outage. These indications are located on the hot leg side of the SG tube bundle and were confirmed by rotating pancake coil as axial ODSCC within the confines of the TSP intersections. The tubes identified in Table 1 were removed from service by plugging during U2R3. The alternate POD values to be applied to each specific indication are also indicated in Table 1. A POD of 0.6 is to be applied for all indications < 3.2 volts, which are not shown in Table 1. The proposed POD values will continue to be applied for any future indications of 3.2 volts.

Table 1 - Indications of Axial ODSCC with Bobbin Voltages 3.2 Volts SG Row Column Support Indication1 Bobbin VoltsPlugged POD 3 17 54 H01 DSV 9.35 Yes 0.95 3 12 111 H02 DSV 8.03 Yes 0.95 3 17 46 H02 DSV 6.87 Yes 0.95 4 6 36 H02 DSV 6.06 Yes 0.95 3 17 47 H02 DSV 5.13 Yes 0.90 3 14 7 H02 DSV 4.83 Yes 0.90 3 7 62 H02 DSV 3.83 Yes 0.90 2 40 53 H03 DSV 3.82 Yes 0.90 3 4 109 H03 DSV 3.64 Yes 0.90 3 48 57 H02 DSV 3.21 Yes 0.90 Note:

1. Bobbin DSV indications are distorted support indications with a measured voltage in excess of the upper voltage repair limit calculated in accordance with GL 95-05.

The indications in Table 1 can be detected with a POD of 0.9 or 0.95 as described in Enclosure 2, and the WBN, Unit 2 specific noise distributions for the TSPs in the limiting SG3.

Further technical information is provided in Enclosure 2, which indicates that the highest noise levels at WBN, Unit 2 will not interfere with the detection of large voltage flaws and that the proposed alternate PODs are appropriate.

Technical Conclusions A stepped change was considered a conservative approach to develop an alternate POD. As described in Enclosure 2, there is a very high probability of detection for large signals, no new indications throughout the industry were found to have a prior inspection voltage 3.2 volts and the WBN, Unit 2 specific POD function described within is CNL-20-104 E1-6 of 11

Enclosure 1 bounded by this threshold. Therefore, applying an increased POD to 3.2 volt indications is considered appropriate. For lower voltages, the GL 95-05 specified POD of 0.6 would still be applied.

Similarly, 6.0 volts was selected as the threshold for the next step in the proposed POD because only a limited number of indications exceeded this value during U2R3 and flaws of that magnitude generate response signals, which would not be masked by the highest noise levels observed. This step change also bounds the WBN, Unit 2 specific POD function.

As discussed in Enclosure 2, flaw injection using Data Union Software for larger value signals and noise levels was performed as a supplemental justification for the proposed POD values. The results of this study further indicate that the highest noise levels at WBN, Unit 2 will not interfere with the detection of large voltage flaws and the proposed alternate PODs are appropriate.

4.0 REGULATORY EVALUATION

4.1 APPLICABLE REGULATORY REQUIREMENTS AND CRITERIA General Design Criteria The WBN, Unit 2 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 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 specific precedent for the proposed POD values in this LAR, in Reference 1, NRC issued a license amendment for Diablo Canyon Power Plant (DCCP),

Unit No. 2, which authorized revisions to the Final Safety Analysis Report Update to incorporate the NRC approval of a POD of 1.0 to one bobbin indication, contained in a DCPP Unit No. 2 SG 4 tube at row 44, column 45 at the second tube support plate on the hot leg side (R44C45-2H), for the beginning of cycle voltage distribution for the DCPP Unit No. 2 Cycle 12 operational assessment. This LAR is similar to Reference 1 in that TVA is proposing an alternate POD for a limited number of WBN, Unit 2 SG tubes.

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Enclosure 1 4.3 NO SIGNIFICANT HAZARDS CONSIDERATION Tennessee Valley Authority (TVA) proposes to revise the Watts Bar Nuclear Plant (WBN) dual-unit Updated Final Safety Analysis Report (UFSAR) to apply an eddy current probability of detection (POD) of 0.9 to indications of axial outside diameter stress corrosion cracking (ODSCC) at tube support plates (TSP) with bobbin voltage amplitudes of greater than or equal to () 3.2 volts, but less than (<) 6.0 volts and a POD of 0.95 to indications of 6.0 volts in the WBN, Unit 2 steam generators (SG) for the beginning of cycle (BOC) voltage distribution in support of the WBN, Unit 2 operational assessment (OA). WBN, Unit 2 Technical Specification (TS) 5.7.2.12, "Steam Generator (SG) Program," and WBN, Unit 2 TS 5.9.9, Steam Generator Tube Inspection Report, are based on Nuclear Regulatory Commission (NRC) Generic Letter (GL) 95-05, "Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking," which requires the application of a POD of 0.6 to all previous bobbin indications for the determination of the indication voltage distribution for the BOC. Therefore, the use of the proposed POD values for the BOC voltage distribution in support of the WBN, Unit 2 OA, is an exception to GL 95-05 and requires NRC approval. The proposed POD values will only be used until the WBN, Unit 2 SGs are replaced. A POD of 0.6, in accordance with GL 95-05, was used for indications of less than 3.2 volts.

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 use of the alternate POD values for the bobbin indications measuring 3.2 volts for the BOC voltage distribution for the WBN, Unit 2 OA does not pose a significant increase in the probability of a steam generator tube rupture (SGTR) event. Based on industry and plant specific bobbin detection data for ODSCC within the SG TSP region, large voltage bobbin indications can be detected with a POD greater than 0.6. Because large voltage ODSCC bobbin indications within the SG TSP can be detected, they will not be left in service; therefore, these indications should not be included in the voltage distribution for the purpose of OA. An eddy current POD of 0.9 to indications of axial ODSCC at TSP with bobbin voltage amplitudes of 3.2 volts, but < 6.0 volts and a POD of 0.95 to indications of 6.0 volts in the WBN, Unit 2 SG for the BOC voltage distribution is justified. The use of the proposed step change POD methodology offers no significant increase in steam line break (SLB) tube burst probability because it will be utilized in conjunction with the GL 95-05 methodology that predicts a conservative operational cycle in terms of effective full power days in compliance with the acceptance criteria for tube burst in the faulted SG.

Therefore, TVA concludes that this proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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Enclosure 1

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

Response: No.

The use of the alternate POD values for the limited number of bobbin indications for WBN, Unit 2 for the BOC voltage distribution for the WBN, Unit 2 OA concerns the SG tubes and can only affect the SGTR accident. Because the SGTR accident is already considered in the UFSAR, there in no possibility to create a design basis accident, which has not, been previously evaluated.

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

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

Response: No.

The use of the alternate POD values for the limited number of bobbin indications for WBN, Unit 2 for the BOC voltage distribution for the WBN, Unit 2 OA does not involve a significant reduction in a margin of safety. The applicable margin of safety potentially impacted is the WBN, Unit 2 TS 5.9.9 projected end-of-cycle leakage for a main steam line break (MSLB) accident and the projected end-of-cycle probability of burst. Based on industry and plant specific bobbin detection data for ODSCC within the SG TSP region, large voltage bobbin indications can be detected and will not be left in service. Therefore, these indications should not be included in the voltage distribution for the purpose of operational assessments. This results in a reduction in numbers of larger indications potentially left in service at the BOC and will not result in a significant increase in the actual end-of-cycle leakage for an MSLB accident or the actual end-of-cycle probability of burst.

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 CNL-20-104 E1-10 of 11

Enclosure 1 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.

6.0 REFERENCES

1. NRC letter to Pacific Gas and Electric Company, Diablo Canyon Power Plant, Unit No. 2 - Issuance of Amendment - Alternate Method of Determining Probability of Detection for Steam Generator Tubes (TAC No. MB7875), dated June 3, 2003 (ML031540535)

CNL-20-104 E1-11 of 11

Attachment 1 Proposed UFSAR Changes (Mark-Ups) for WBN, Unit 2 CNL-20-104

WBN-1 5.5.2.4 Tests and Inspections The initial steam generator quality assurance program is given in Table 5.5-4 Unit 1 (historical information) and in Table 5.5-4b for Unit 2.

Radiographic inspection and acceptance standards are in accordance with the requirements of Section III of the ASME Code.

Liquid penetrant inspection was performed on weld deposited tubesheet cladding, channel head cladding, tube to tubesheet weldments, and weld deposit cladding.

Liquid penetrant inspection and acceptance standards are in accordance with the requirements of Section III of the ASME Code.

Magnetic particle inspection was performed on the tubesheet forging, nozzle forgings, channel head casting (Unit 2 Only), and the following weldments:

1. Nozzle to shell

2. Support brackets

3. Instrument connections (primary and secondary)

4. Temporary attachments after removal

5. Accessible pressure containing welds after hydrostatic test.

Magnetic particle inspection and acceptance standards are in accordance with requirements of Section III of the ASME Code.

An ultrasonic test was performed on the tube sheet forging, tube sheet cladding, secondary shell and heat plate and nozzle forgings.

The heat transfer tubing was subjected to eddy current test.

Hydrostatic tests are performed in accordance with Section III of the ASME Code.

In addition, the heat transfer tubes were subjected to a hydrostatic test pressure, prior to installation into the vessel, which is not less than 1.25 times the primary side design pressure, as required by Section III of the ASME Code.

Manways are provided for access to both the primary and secondary sides.

Inservice inspection of steam generator tubes is to be performed in accordance with the Technical Specifications.

Unit 2 Only Steam Generator tubing Alternate Repair Criteria for F* (F star) at the top of the tubesheet was approved by NRC. (19) Implementation of F* is in accordance with Technical Specification inservice examination requirements and Reference 19.

5.5-19

WBN Steam Generator Tubing voltage-based Alternate Repair Criteria (ARC) for Axial Outside Diameter Stress Corrosion Cracking (ODSCC) at tube support plate intersections was approved by NRC (23). Implementation of ODSCC ARC using GL 95-05 (24) as guidance is in accordance with Technical Specification inservice examination requirements and Reference 25. As an alternative to the probability of detection of 0.6 required by GL 95-05, a probability of detection (POD) of 0.9 will be applied to indications of axial ODSCC at tube support plates with bobbin voltage amplitudes of greater than or equal to 3.2 volts, but less than 6.0 volts, and a POD of 0.95 will be applied to indications of axial ODSCC at tube support plates with bobbin voltage amplitudes of greater than or equal to 6.0 volts until the Unit 2 Steam Generators are replaced(26). A POD of 0.6, in accordance with GL 95-05, will be used for indications less than 3.2 volts.

5.5.3 Reactor Coolant Piping 5.5.3.1 Design Bases The RCS piping is designed and fabricated to accommodate the system pressures and temperatures attained under all expected modes of plant operation or anticipated system interactions. Stresses are maintained within the limits of Section III of the ASME Nuclear Power Plant Components Code. Code and material requirements are provided in Section 5.2.

Materials of construction are specified to minimize corrosion/erosion and ensure compatibility with the operating environment.

The piping in the RCS is Safety Class 1 and is designed and fabricated in accordance with ASME Section III, Class 1 requirements.

Stainless steel pipe conforms to ANSI B36.19 for sizes 1/2-inch through 12 inches and wall thickness Schedules 40S through 80S. Stainless steel pipe outside of the scope of ANSI B36.19 conforms to ANSI B36.10.

The minimum wall thicknesses of the loop pipe and fittings are not less than that calculated using the ASME III Class 1 formula of Paragraph NB-3641.1 (3), with an allowable stress value of 17,550 psi. The pipe wall thickness for the pressurizer surge line is Schedule 160. The minimum pipe bend radius is 5 nominal pipe diameters; ovalness does not exceed 6%.

Butt welds, branch connection nozzle welds, and boss welds are of a full-penetration design.

Processing and minimization of sensitization are discussed in Sections 5.2.3 and 5.2.5.

Flanges conform to ANSI B16.5.

Socket weld fittings and socket joints conform to ANSI B16.11.

Inservice inspection is discussed in Section 5.2.8.

5.5.3.2 Design Description Principal design data for the reactor coolant piping are given in Table 5.5-5.

Pipe and fittings are cast, seamless without longitudinal welds and electroslag welds, and comply with the requirements of the ASME Code,Section II, Parts A and C,Section III, and 5.5-20

WBN-3

23. NRC Safety Evaluation for Watts Bar Nuclear Plant Unit 2, Amendment 28, for Steam Generator Tubing Voltage Based Alternate Repair Criteria for Outside Diameter Stress Corrosion Cracking (ODSCC) dated June 3, 2019.

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

25. TVA Letter to NRC 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 and as supplemented by letter CNL-18-128 dated November 8, 2018.

26. NRC letter to TVA, XXXXX (TAC No. XXXX), dated MM/DD/YY (MLXXXX).

5.5-52

Attachment 2 Proposed UFSAR Changes (Final Typed) for WBN, Unit 2 CNL-20-104