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License Amendment Request for Measurement Uncertainty Recapture Power Uprate (WBN-TS-19-06)
	ML19283G119
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Site:	Watts Bar
Issue date:	10/10/2019
From:	Polickoski J; Tennessee Valley Authority
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
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ML19283G117	List: CNL-19-082, License Amendment Request for Measurement Uncertainty Recapture Power Uprate (WBN-TS-19-06);
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Proprietary Information Withhold from Public Disclosure Under 10 CFR 2.390 This letter is decontrolled when separated from Enclosures 5 and 6 Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 CNL-19-082 October 10, 2019 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:

License Amendment Request for Measurement Uncertainty Recapture Power Uprate (WBN-TS-19-06)

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 (OL) No. NPF-96 for Watts Bar Nuclear Plant (WBN) Unit 2 to support a Measurement Uncertainty Recapture (MUR) power uprate.

This MUR license amendment request (LAR) would increase the WBN Unit 2 authorized core power level from 3411 megawatts thermal (MWt) to 3459 MWt (i.e., an increase of approximately 1.4% Rated Thermal Power), based on the use of the Caldon1 Leading Edge Flow Meter (LEFM1) CheckPlus System. provides a description and technical evaluation of the proposed change, regulatory evaluation, and discussion of environmental considerations. Enclosure 2 provides the WBN Unit 2 specific evaluation of each item outlined in Regulatory Issue Summary (RIS) 2002-03, Guidance on Content of Measurement Uncertainty Recapture Power Uprate Applications. Enclosure 3 provides the existing WBN Unit 2 OL, Technical Specifications (TS), and TS Bases pages marked-up to show the proposed changes. provides the proposed OL, TS, and TS Bases re-typed to show the changes incorporated. The changes to the TS Bases are provided for information only. Enclosure 5 (proprietary) contains the uncertainty analysis for the Caldon LEFM CheckPlus System 1 Caldon, Inc. is now part of the Measurement Systems Division of Cameron International Corporation (Cameron). Caldon and LEFM are registered trademarks of Cameron.

Proprietary Information Withhold from Public Disclosure Under 10 CFR 2.390 This letter is decontrolled when separated from Enclosures 5 and 6

Proprietary Information Withhold from Public Disclosure Under 10 CFR 2.390 This letter is decontrolled when separated from Enclosures 5 and 6 U.S. Nuclear Regulatory Commission CNL-19-082 Page 2 October 10, 2019 installed at WBN Unit 2. Enclosure 6 (proprietary) contains the WBN Unit 2 power calorimetric uncertainty calculation based on use of the Caldon LEFM CheckPlus System.

The proprietary information in Enclosure 5 is supported by an affidavit (Enclosure 7) signed by Cameron, the owner of the information. The affidavit sets forth the basis on which the information should be withheld from public disclosure by the NRC and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR 2.390. Accordingly, TVA requests that the information, which is proprietary to Cameron, be withheld from public disclosure in accordance with 10 CFR 2.390. Correspondence with respect to the copyright or proprietary aspects of the technical information listed above or the supporting Cameron affidavit should reference Cameron letters CAW 19-05 and CAW 19-06 and should be addressed to Joanna Phillips, Nuclear Sales Manager, Caldon Ultrasonics Technology Center, Cameron, 1000 McClaren Woods Drive, Coraopolis, Pennsylvania 15216.

The proprietary information in Enclosure 6 is supported by an affidavit (Enclosure 8) signed by Westinghouse Electric Company (Westinghouse), the owner of the information. The affidavit sets forth the basis on which the information should be withheld from public disclosure by the NRC and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR 2.390. Accordingly, TVA requests that the information, which is proprietary to Westinghouse, be withheld from public disclosure in accordance with 10 CFR 2.390. Correspondence with respect to the copyright or proprietary aspects of the technical support document or the supporting Westinghouse affidavit should reference CAW-19-4927 and should be addressed to Zachary Harper, Manager Licensing Engineering, Westinghouse Electric Company, 1000 Westinghouse Drive, Building 3, Suite 310, Cranberry Township, Pennsylvania 16066.

Non-proprietary versions of Enclosures 5 and 6 are provided in Enclosures 9 and 10, respectively.

TVA requests approval of the proposed license amendment within one year from the date of this submittal with implementation of the amendment prior to completion of the WBN Unit 2 fall 2020 refueling outage, but no later than December 15, 2020.

TVA has determined that there are no significant hazards considerations associated with the proposed change and that the TS 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, a copy of this application, with the Enclosure is being provided to the designated Tennessee Official.

There are no new regulatory commitments associated with this submittal. Please address any questions regarding this request to Kimberly D. Hulvey at (423) 751-3275.

Proprietary Information Withhold from Public Disclosure Under 10 CFR 2.390 This letter is decontrolled when separated from Enclosures 5 and 6

Enclosure 1 Evaluation of the Proposed Change

Subject:

License Amendment Request for Measurement Uncertainty Recapture Power Uprate (WBN-TS-19-06)

TABLE OF CONTENTS 1.0

SUMMARY

DESCRIPTION ............................................................................................. 2 1.1 Background.................................................................................................................. 2 2.0 DETAILED DESCRIPTION.............................................................................................. 3 2.1 Description of the Proposed Change............................................................................ 3 2.2 Reason for the Proposed Change ................................................................................ 4

3.0 TECHNICAL EVALUATION

............................................................................................. 4 3.1 Leading Edge Flow Meter CheckPlus System Description ........................................... 4 3.2 Evaluation .................................................................................................................... 6

4.0 REGULATORY ANALYSIS

............................................................................................. 6 4.1 Applicable Regulatory Requirement Criteria................................................................. 6 4.2 Precedent .................................................................................................................... 7 4.3 No Significant Hazards Consideration Determination ................................................... 7 4.4 Conclusion ................................................................................................................... 9

5.0 ENVIRONMENTAL CONSIDERATION

........................................................................... 9

6.0 REFERENCES

...............................................................................................................10 CNL-19-082 E1-1 of 10

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 (OL) No. NPF-96 and the associated Technical Specifications (TS) for the Watts Bar Nuclear Plant (WBN) Unit 2 to increase the authorized core power level from 3411 megawatts thermal (MWt) to 3459 MWt; an increase of approximately 1.4 percent (%) rated thermal power (RTP).

The proposed changes are based on the application of a 1.4% measurement uncertainty recapture (MUR) power uprate analysis based on the use of the existing Caldon1 Leading Edge Flow Meter (LEFM) CheckPlus System, an ultrasonic flow measurement system. This core power uprate is effectively achieved by recapturing excess uncertainty currently included in the power uncertainty allowance originally required for emergency core cooling system (ECCS) evaluation models performed in accordance with the requirements set forth in 10 CFR 50, Appendix K, ECCS Evaluation Models. Improvement in core power level measurement accuracy is possible through the reduction in feedwater flow measurement uncertainty used in the reactor power calorimetric calculation. The Caldon LEFM CheckPlus System provides more accurate measurement of feedwater flow so that core power level can be determined with a power measurement uncertainty of less than 0.6%. This will allow WBN Unit 2 to operate at a 1.4% higher reactor thermal power level, increasing core power output from 3411 MWt to 3459 MWt, similar to WBN Unit 1.

Analysis work in the support of this 1.4% MUR power uprate license amendment request (LAR) has been performed consistent with Regulatory Issue Summary (RIS) 2002-03, Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications (Reference 1),

as summarized in Enclosure 2.

1.1 BACKGROUND

The power uprate for WBN Unit 2 is based on recapturing measurement uncertainty currently included in the analytical margin originally required of ECCS evaluation models performed in accordance with the requirements set forth in 10 CFR 50, Appendix K, which provides licensees with the option of maintaining the two percent (2%) power margin between the licensed power level and the assumed power level for the ECCS evaluation, or applying an appropriately justified reduced margin for ECCS evaluation.

For the case of an appropriately justified reduced margin for ECCS evaluation, the alternative reduced margin must be shown to account for uncertainties due to power level instrumentation error. Based on the use of the Caldon LEFM instrumentation to determine core power level with a power measurement uncertainty of less than 0.6%, TVA is reducing the licensed power uncertainty required by 10 CFR 50, Appendix K, which allows for an increase of approximately 1.4% in the licensed power level, using NRC-approved methodologies identified in TS 5.9.5b.

The Caldon LEFM instrumentation provides a more accurate indication of feedwater flow (and correspondingly reactor thermal power) than available during the original development of 10 CFR 50, Appendix K requirements, which did not explicitly consider uncertainties and prescribed a required power level of 102% for accident analysis.

1 Caldon, Inc. is now part of the Measurement Systems Division of Cameron International Corporation (Cameron).

Caldon and LEFM are registered trademarks of Cameron.

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Enclosure 1 Further information is provided in Cameron Topical Report ER-80P, Revision 0 (Reference 2),

as approved in the Nuclear Regulatory Commission (NRC) Safety Evaluation for TU Electric (Reference 3); and supplemented by Cameron Engineering Report ER-157P, Revision 8 with errata (Reference 4), as approved in the NRC Safety Evaluation for Cameron (Reference 5).

The improved thermal power measurement accuracy obviates the need for the full 2% power margin originally assumed in Appendix K to 10 CFR 50, thereby increasing the thermal power available for electrical generation.

2.0 DETAILED DESCRIPTION

2.1 DESCRIPTION

OF THE PROPOSED CHANGE The following changes are being made by this LAR.

1. WBN Unit 2 OL Item 2.C.(1) is being revised to increase the maximum core power level from 3411 MWt to 3459 MWt.

2. The definition of RTP in TS 1.1, Definitions, is being changed to account for the increase in reactor core thermal power level as follows:

RTP shall be a total reactor core heat transfer rate to the reactor coolant of 3459 MWt.

3. TS 5.9.5b, CORE OPERATING LIMITS REPORT (COLR), currently states:

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

TS 5.9.5b is being revised as follows:

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC. When an initial assumed power level of 102% RTP is specified in a previously approved method, 100.6% RTP may be used only when feedwater flow measurement (used as input for reactor thermal power measurement) is provided by the leading edge flowmeter (LEFM) as described in document number 10 listed below. When feedwater flow measurements from the LEFM are unavailable, the originally approved initial power level of 102% RTP (3411 MWt) shall be used. The approved analytical methods are specifically those described in the following documents:

4. The NRC approved Caldon Topical Reports for LEFMs are being added as document number 10 in the list of documents in TS 5.9.5b as follows:

10. Caldon, Inc., Engineering Report-80P, Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM' System, Revision 0; and Caldon Ultrasonics Engineering Report ER-157P-A, Supplement to Caldon Topical Report ER-80P: Basis for Power Uprates with an LEFM Check or LEFM CheckPlus System," Revision 8 and Revision 8 errata.

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Enclosure 1

5. The TS Bases for Limiting Condition of Operation (LCO) 3.7.1, MSSVs, is being revised to reflect that the main steam safety valve (MSSV) analysis that was performed at 102% RTP (at the current power level of 3411 MWt) equates to an analysis performed at 100.6% RTP for the proposed RTP of 3459 MWt.

6. The TS Bases for LCO 3.7.6, CST, is being revised to reflect that the condensate storage tank (CST) analysis that was performed at 102% RTP (at the current power level of 3411 MWt) equates to an analysis performed at 100.6% RTP for the proposed RTP of 3459 MWt. provides the existing WBN Unit 2 OL, TS, and TS Bases pages marked-up to show the proposed changes. Enclosure 4 provides the proposed OL, TS, and TS Bases re-typed to show the changes incorporated. The changes to the TS Bases are provided for information only.

2.2 REASON FOR THE PROPOSED CHANGE The proposed change will allow TVA to increase the WBN Unit 2 RTP from 3411 MWt to 3459 MWt.

3.0 TECHNICAL EVALUATION

3.1 LEADING EDGE FLOW METER CHECKPLUS SYSTEM DESCRIPTION WBN Unit 2 has a Caldon LEFM CheckPlus ultrasonic multi-path, transit time flow meter installed in the main feedwater header as well as feedwater flow venturis, which will be used if the LEFM system is not functional. The LEFM system uses ultrasonic transit time principles to determine fluid velocity and sound velocity. This flow measurement method is described in References 2 and 4.

The WBN Unit 2 LEFM CheckPlus system was calibrated in a site-specific model test at Alden Research Laboratories. A bounding calibration factor for the WBN Unit 2 spool piece was established by these tests and is included in the Cameron report ER-732 (Reference 6), which is included in Enclosure 5. Total thermal power uncertainties for the WBN Unit 2 LEFM CheckPlus system, for both NORMAL and MAINTENANCE modes, are provided in Cameron report ER-734P (Reference 7), which is included in Enclosure 5. These values were quantified on a plant-specific basis, in accordance with the methodology of Cameron reports ER-80P and ER-157P.

In approving topical reports ER-80P and ER-157P, the NRC established criteria that each licensee referencing these topical reports must address. TVAs response to those criteria is provided in Section I of Enclosure 2.

The WBN Unit 2 LEFM CheckPlus ultrasonic flow meter system consists of an electronic cabinet located in the WBN Unit 2 auxiliary instrument room, one measurement section/spool piece located in the Turbine Building, and associated cabling. The measurement section/spool piece is installed in the 32-inch main feedwater header. The measurement spool piece contains 16 ultrasonic, multi-path, transit time transducers grouped into the two planes of eight transducers each. The LEFM spool piece is located upstream of the existing nozzle venturis, which are located in the feedwater lines to the individual steam generators (SGs).

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Enclosure 1 Validated LEFM data including calculated results, status, and signal process information is sent to the integrated computer system (ICS). The ICS provides an audible and visual alarm upon a change in LEFM system status. The LEFM system instrumentation is not safety-related. The LEFM system was designed and manufactured per Camerons Quality Assurance Program.

The LEFM CheckPlus System performs automatic continuous self-checking of the transducer signals and the LEFM calculation results. This testing provides verification that the digital circuits are operating correctly and the system is within its specified accuracy envelope. These processes can identify failure conditions that will cause the LEFM to switch from the NORMAL mode to the MAINTENANCE mode or to the FAIL mode.

In NORMAL mode, the LEFM CheckPlus System measures the average flow of two independent LEFM Check subsystems. Each LEFM Check subsystem consists of four acoustic paths (i.e., a total of eight paths) that comprise the LEFM CheckPlus system. The LEFM CheckPlus System NORMAL status is displayed when the feedwater flow, temperature, and header pressure signals are normal and operating within design limits.

In MAINTENANCE mode, only one of the two LEFM Check subsystems is fully operational, which results in flow computation based on the fully operational LEFM Check subsystem. An LEFM CheckPlus System ALERT alarm indicates a loss of system redundancy and the system shifts from the NORMAL mode to the MAINTENANCE mode of operation. Other than when performing maintenance, the shift from the NORMAL mode to the MAINTENANCE mode of operation typically occurs due to a malfunction of a single path or plane.

An LEFM CheckPlus System FAIL status indicates a loss of function. Loss of LEFM system results in reverting to the calibrated venturi-based monitoring system. If the LEFM becomes unavailable, the secondary side calorimetric is performed with inputs from the flow venturis, which requires a core power adjustment toward a lower core power based on the 2%

uncertainty associated with the venture nozzle inaccuracies. Reactor power will be reduced from 3459 MWt (i.e., the MUR power uprate level) to 3411 MWt (i.e., the pre-MUR power uprate power level) within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months if LEFM functionality cannot be restored to the NORMAL mode, consistent with WBN Unit 1, which has an LEFM Check system (see Section I.1.D.v of ). Because WBN Unit 2 is only requesting a 1.4% power uprate, WBN Unit 2 can operate at 3459 MWt with only one LEFM Check subsystem operational; however, a 72-hour allowed outage time is currently proposed for when the LEFM CheckPlus System is in MAINTENANCE mode (see Section I.1.D.v of Enclosure 2 for details).

WBN Unit 2 is requesting a 1.4% power uprate based on use of the LEFM CheckPlus System.

WCAP-18419-P (Reference 8, which is included in Enclosure 6) is the analysis of record, for WBN Unit 2, pertaining to the use of the LEFM for calculating the total thermal power uncertainty. This bounding analysis supports a power uprate of 1.4% (by using conservative inputs and rounding the total thermal power uncertainty up to 0.6%) and will align WBN Unit 2 with WBN Unit 1. Although the Westinghouse methodology used for calculation of the total thermal power uncertainty for WBN Unit 2 is not generically approved by the NRC, it was previously approved for WBN Unit 1. The Cameron engineering reports demonstrate that the thermal power uncertainty results of the Westinghouse analysis are conservative and bound use of the LEFM in both NORMAL and MAINTENANCE modes. The actual WBN Unit 2 total thermal power uncertainty will be less than or equal to the sum of the proprietary results of Cameron engineering report ER-734P (Enclosure 5) based on the inclusion of an additional 0.02% uncertainty (to account for moisture uncertainty effects as discussed in Enclosure 2,Section I.1.D.ix).

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Enclosure 1 A summary of the impacts of a 1.4% power uprate for WBN Unit 2 is provided in the next section. A more detailed evaluation, in the format of NRC RIS 2002-03, is provided in .

3.2 EVALUATION WBN Unit 2 is currently licensed for a RTP of 3411 MWt. A more accurate feedwater flow measurement supports an increase to 3459 MWt. The technical evaluation for this MUR power uprate addressed the following aspects:

The feedwater flow measurement technique and power measurement uncertainty;

Accidents and transients that remain bounded at the higher power level;

Accidents and transients not bounded at the higher power level;

Mechanical/structural/material component integrity and design;

Electrical equipment design;

System design;

Operating, emergency, and abnormal procedures including associated operator actions;

Environmental impact; and,

Any changes to the TS including protective system setpoints.

NRC RIS 2002-03 provides generic guidance for evaluating an MUR power uprate. Enclosure 2 of this LAR provides the WBN Unit 2 specific evaluation of each item outlined in RIS 2002-03, , and provides a description of the methodology used by WBN Unit 2 to complete the evaluation. A summary of the results from Enclosure 2 is provided below.

Transient and accident analyses and associated evaluations were reviewed and were confirmed to bound the MUR uprate. The applicable acceptance criteria continue to be met.

The functional capability of the reactor control systems and setpoints were confirmed to be acceptable for the uprated conditions.

The plant operability/margin to trip analyses of record were confirmed to remain applicable for the uprated conditions.

The results of the Balance of Plant (BOP) system and component evaluations demonstrate the functional requirements continue to be met for the uprated conditions.

No hardware modifications to plant systems, structures, or components are required to support the MUR uprate.

The NRC has approved WBN Unit 2 for tritium production (Reference 9), which is currently planned to commence, concurrent with the MUR power uprate, with WBN Unit 2 Cycle 4 (anticipated November 2020 startup). WBN Unit 2 has already been evaluated as a tritium production core for conditions that bound or are equivalent to MUR power uprate conditions.

4.0 REGULATORY ANALYSIS

4.1 APPLICABLE REGULATORY REQUIREMENT CRITERIA NRC RIS 2002-03 provides generic guidance for evaluating an MUR power uprate. Enclosure 2 of this LAR provides the WBN Unit 2 specific evaluation of each item outlined in RIS 2002-03, , and provides a description of the methodology used by WBN Unit 2 to complete the evaluation. Based on Enclosure 2, (1) there is reasonable assurance that the health and CNL-19-082 E1-6 of 10

Enclosure 1 safety of the public will not be endangered by operation at the uprated power level, (2) operation at the uprated power level will be in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or the health and safety of the public.

4.2 PRECEDENT The NRC has previously approved several MUR power uprate LARs, including an uprate for WBN Unit 1, which was approved on January 19, 2001 (ML010260074). This request is similar in format and content to the following, more recent, submittals:

Duke Energy submittal for MUR power uprate of Catawba Nuclear Station, Units 1 and 2, dated June 23, 2014 (ML14176A109), as supplemented by letters dated August 26 and December 15, 2014, and January 22, April 23, and November 16, 2015 (ML14245A059, ML14353A024, ML15029A417, ML15117A010, and ML15324A0B9, respectively), which was reviewed and approved by the NRC on April 29, 2016 (ML16081A333).

TVA has reviewed the above documents for applicability to the WBN Unit 2 proposed MUR amendment.

4.3 NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION Tennessee Valley Authority (TVA) proposes to change the Watts Bar Nuclear Plant (WBN)

Unit 2 Facility Operating License (OL) NPF-96 and Technical Specifications (TS) to increase the rated thermal power by 1.4 percent (%) from 3411 megawatts thermal (MWt) to 3459 MWt.

TVA has evaluated the proposed changes to the WBN Unit 2 OL and TS using the criteria in Section 50.92 to Title 10 of the Code of Federal Regulations (10 CFR) and has determined that the proposed changes do not involve a significant hazards consideration. As required by 10 CFR 50.91(a), the TVA analysis of the issue of no significant hazards consideration is presented below:

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

Response: No.

The proposed amendment changes the rated thermal power (RTP) for WBN Unit 2 from 3411 MWt to 3459 MWt; an increase of approximately 1.4 %RTP. TVA's evaluations have shown that all structures, systems, and components (SSCs) are capable of performing their design function at the uprated power of 3459 MWt. A review of station accident analyses found that all acceptance criteria are still met at the uprated power of 3459 MWt.

The radiological consequences of operation at the uprated power conditions have been assessed. The proposed power uprate does not affect release paths, frequency of release, or the analyzed reactor core fission product inventory for any accidents previously evaluated in the Updated Final Safety Analysis Report. Analyses performed to assess the effects of mass and energy releases remain valid and bounded by accident analysis values. The acceptance criteria for radiological consequences continue to be met at the uprated power level.

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Enclosure 1 The proposed change does not involve any change to the design or functional requirements of the safety and support systems. That is, the increased power level neither degrades the performance of, nor increases the challenges to any safety systems assumed to function in the plant safety analysis.

While power level is an input to accident analyses, it is not an initiator of accidents. The proposed change does not affect any accident precursors and does not introduce any accident initiators. The proposed change does not impact the usefulness of the Surveillance Requirements (SRs) in evaluating the operability of required systems and components.

Additionally, evaluation of the proposed TS changes demonstrates that the availability of equipment and systems required to prevent or mitigate the radiological consequences of an accident is not affected.

Therefore, the proposed amendment does not significantly increase the probability or consequences of an accident previously evaluated.

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

Response: No.

No new accident scenarios, failure mechanisms, or single failures are introduced because of the proposed change. The use of the previously installed Cameron (formerly Caldon, Inc.)

Leading Edge Flow LEFM (LEFM) CheckPlus System had been previously analyzed for WBN Unit 2, and failures of the system will continue to have no adverse effect on any safety-related system or any SSCs required for transient mitigation. SSCs previously required for the mitigation of a transient continue to be capable of fulfilling their intended design functions. The proposed change has no adverse effect on any safety-related system or component and does not change the performance or integrity of any safety-related system.

The proposed change does not adversely affect any current system interfaces or create any new interfaces that could result in an accident or malfunction of a different kind than previously evaluated. Operation at the uprated power level does not create any new accident initiators or precursors. Credible malfunctions remain bounded by existing accident analyses of record.

Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

Although the proposed amendment increases the WBN Unit 2 operating power level, the unit retains its margin of safety because it is only increasing power by the amount equal to the reduction in uncertainty in the heat balance calculation. The margins of safety associated with the power uprate are those pertaining to core thermal power. These include fuel cladding, reactor coolant system pressure boundary, and containment barriers.

Analyses demonstrate that the current design basis continues to be met after the CNL-19-082 E1-8 of 10

Enclosure 1 measurement uncertainty recapture (MUR) power uprate. Components associated with the reactor coolant system pressure boundary structural integrity, including pressure-temperature limits, vessel fluence, and pressurized thermal shock are bounded by the current analyses. Systems will continue to operate within their design parameters and remain capable of performing their intended safety functions.

The current WBN Unit 2 safety analyses, including the design basis radiological accident dose calculations, bound the proposed power uprate.

Therefore, the proposed amendment 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

The proposed TS 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 Part 20, and 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 TS 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 TS change.

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Enclosure 1

6.0 REFERENCES

1. NRC Regulatory Issue Summary 2002-03, Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications, dated January 31, 2002 (ML013530183)

2. Cameron Topical Report ER-80P, Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM' System, Revision 0, dated March 1997

3. NRC letter to TU Electric, Comanche Peak Steam Electric Station, Units 1 and 2 -

Review of Caldon Engineering Topical Report ER 80P, Improving Thermal Power Accuracy and Plant Safety While Increasing Power Level Using the LEFM System (TACS Nos. MA2298 and MA2299), dated March 8, 1999 (9903190065)

4. Cameron Engineering Report ER-157P-A, Supplement to Caldon Topical Report ER-80P: Basis for Power Uprates with an LEFM Check or LEFM CheckPlus System, Revision 8, dated May 2008, and Revision 8 errata

5. NRC letter to Cameron, Final Safety Evaluation for Cameron Measurement Systems Engineering Report ER-157P, Revision 8, Caldon Ultrasonics Engineering Report ER-157P, Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM Check or CheckPlus System, (TAC No. ME1321), dated August 16, 2010 (ML102160663 and ML102160694)

6. Cameron Engineering Report ER-732P, Revision 0, Meter Factor Calculation and Accuracy Assessment for Tennessee Valley Authority, Watts Bar Nuclear Plant Unit 2, dated December 2008

7. Cameron Engineering Report ER-734P, Revision 2, Bounding Uncertainty Analysis for Thermal Power Determination at Watts Bar Unit 2 Using the LEFM+ System, dated August 2019

8. Westinghouse WCAP-18419-P, Revision 1, Leading Edge Flow Meter (LEFM) Power Measurement Uncertainty for the Watts Bar Unit 2 MUR Program, 3459 MWt Core Power with LEFM

9. NRC letter to TVA, Watts Bar Nuclear Plant, Units 1 and 2 - Issuance of Amendment Regarding Revision to Watts Bar Unit 2 Technical Specification 4.2.1 Fuel Assemblies, and Watts Bar Units 1 and 2 Technical Specifications Related to Fuel Storage (EPID L-2017-LLA-0427) (ML18347B330)

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Enclosure 2 TABLE OF CONTENTS I FEEDWATER FLOW MEASUREMENT TECHNIQUE AND POWER MEASUREMENT UNCERTAINTY .......................................................................................................... E2-2 II ACCIDENTS AND TRANSIENTS FOR WHICH THE EXISTING ANALYSES OF RECORD BOUND PLANT OPERATION AT THE PROPOSED UPRATED POWER LEVEL ...................................................................................................................... E2-14 III ACCIDENTS AND TRANSIENTS FOR WHICH THE EXISTING ANALYSES OF RECORD DO NOT BOUND PLANT OPERATION AT THE PROPOSED UPRATED POWER LEVEL........................................................................................................ E2-40 IV MECHANICAL/STRUCTURAL/MATERIAL COMPONENT INTEGRITY AND DESIGN ................................................................................................................... E2-42 V ELECTRICAL EQUIPMENT DESIGN....................................................................... E2-58 VI SYSTEM DESIGN .................................................................................................... E2-64 VII OTHER..................................................................................................................... E2-72 VIII CHANGES TO TECHNICAL SPECIFICATIONS, PROTECTION SYSTEM SETTINGS, AND EMERGENCY SYSTEM SETTINGS................................................................ E2-77 CNL-19-082 E2-1 of 78

Enclosure 2 RIS 2002-03 Requested Information This enclosure provides the TVA responses to RIS 2002-03, Attachment 1, (Reference I.1) with the WBN Unit 2 information provided in response to each item.

I. FEEDWATER FLOW MEASUREMENT TECHNIQUE AND POWER MEASUREMENT UNCERTAINTY I.1 A detailed description of the plant-specific implementation of the feedwater flow measurement technique and the power increase gained as a result of implementing this technique. This description should include:

I.1.A Identification (by document title, number, and date) of the approved topical report on the feedwater flow measurement technique

RESPONSE

The feedwater flow measurement techniques at WBN Unit 2 use the Cameron (formerly Caldon, Inc.) LEFM CheckPlus System with ultrasonic multi-path transit time flowmeter as described in the following topical reports:

Cameron Topical Report ER-80P, Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM' System, Revision 0, March 1997 (Reference I.2)

Cameron Engineering Report ER-157P-A, Supplement to Caldon Topical Report ER-80P:

Basis for Power Uprates with an LEFM Check or LEFM CheckPlus System, Revision 8, May 2008, and Revision 8 errata (Reference I.3)

I.1.B A reference to the NRCs approval of the proposed feedwater flow measurement technique

RESPONSE

The Caldon LEFM Check instruments (Report ER-80P), which are also used at WBN Unit 1, were reviewed and approved by the NRC in Reference I.4. Subsequently, the LEFM CheckPlus instruments (Report ER-157P-A, Revision 8 and errata), which are used at WBN Unit 2, were reviewed and approved by the NRC in Reference I.5.

NRC letter to TU Electric, Comanche Peak Steam Electric Station, Units 1 and 2 - Review of Caldon Engineering Topical Report ER 80P, Improving Thermal Power Accuracy and Plant Safety While Increasing Power Level Using the LEFM System (TACS Nos. MA2298 and MA2299), March 8, 1999 (9903190065)

NRC letter to Cameron, Final Safety Evaluation for Cameron Measurement Systems Engineering Report ER-157P, Revision 8, Caldon Ultrasonics Engineering Report ER-157P, Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM Check or CheckPlus System, (TAC No. ME1321), August 16, 2010, and its Enclosure 1, Final SE (ML102160663 and ML102160694)

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Enclosure 2 I.1.C A discussion of the plant-specific implementation of the guidelines in the topical report and the staffs letter/safety evaluation approving the topical report for the feedwater flow measurement technique

RESPONSE

The LEFM CheckPlus ultrasonic flow meter system was installed in accordance with the manufacturers requirements as described in Cameron Topical Report ER-80P and its supplement, ER-157P, Revision 5 (Reference I.6), which was approved by the NRC in (Reference I.16). Revision 8 of ER-157P was later approved by the NRC. Revision 8 corrects minor errors in Revision 5, provides clarifying text, and incorporates revised analyses of coherent noise, non-fluid delays, and transducer replacement. Revision 8 also adds two new appendices, Appendix C and Appendix D, which describe the assumptions and data that support the coherent noise and transducer replacement calculations, respectively. The latest approved version, ER-157P-A, Revision 8 with errata, was used to support this license amendment request (LAR). ER-734P (Reference I.8) in Enclosure 5 provides details on the use of ER-157P-A, Revision 8, for the calculation of LEFM mass flow uncertainty.

The LEFM CheckPlus ultrasonic flow meter system is operated in accordance with the manufacturers requirements as described in References I.2 and I.3. The output from the system is currently used to support the correction factor for nozzle-venturi fouling. The LEFM CheckPlus system will be used for the daily calorimetric power determination by direct data link with the WBN Unit 2 ICS in order to support the MUR power uprate. The system incorporates self-verification features to ensure that the hydraulic profile and signal processing requirements are met within its design basis uncertainty analysis.

The LEFM CheckPlus ultrasonic flow meter system consists of an electronic cabinet, located in the WBN Unit 2 auxiliary instrument room, one measurement section/spool piece (with 16 ultrasonic transducer assemblies), located in the Turbine Building, and associated cabling. The measurement section/spool piece is installed in the 32-inch Main Feedwater header. The LEFM spool piece is located upstream of the existing nozzle venturis, which are located in the feedwater lines to the individual SGs.

The location of the LEFM relative to existing upstream and downstream piping changes was reviewed and it was determined that the LEFM location does not affect the LEFM performance or the existing venturi performance. The LEFM transducers are located more than five pipe diameters in length from piping changes.

The WBN Unit 2 LEFM CheckPlus system was calibrated in a site-specific model test at Alden Research Laboratories with calibration standards traceable to National Institute of Standards and Technology (NIST) standards. The LEFM CheckPlus system installation and commissioning were performed according to Cameron procedures. These procedures include verification of ultrasonic signal quality and hydraulic velocity profiles as compared to those during site-specific model testing.

A comparison of the test and plant piping configuration and an explanation of the effect of any differences that could impact the LEFM calibration is provided in ER-732 (Reference I.10),

which is included in Enclosure 5.

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Enclosure 2 I.1.D The dispositions of the criteria that the NRC staff stated should be addressed (i.e., the criteria included in the staffs approval of the technique) when implementing the feedwater flow measurement technique

RESPONSE

In approving Cameron Topical Report ER-80P, the NRC established four criteria to be addressed by each licensee. In approving Cameron Topical Report ER-157P, Revision 8, the NRC established five additional criteria to be addressed by each licensee. The following presents a discussion of each of the nine criteria relative to WBN Unit 2.

I.1.D.i Criterion 1 from ER-80P - Discuss maintenance and calibration procedures that will be implemented with the incorporation of the LEFM, including processes and contingencies for unavailable LEFM instrumentation and the effect on thermal power measurements and plant operation.

RESPONSE

Maintenance and Calibration Procedures:

As part of the TVA design change process, implementation of the power uprate license amendment will include developing the necessary procedures and documents required for operation and maintenance at the uprated power level with the new LEFM CheckPlus system.

Required training materials will be developed and implementation will include training of operating and maintenance personnel.

A preventive maintenance program has already been implemented for the existing WBN Unit 2 LEFM CheckPlus system using Cameron's maintenance and troubleshooting manual and TVAs established procedure program. Preventative maintenance and calibration is performed every 18 months. WBN Unit 2 LEFM specific maintenance activities include the following checks:

Panel status verification (including Central Processing Unit (CPU) inspection)

Power supply verification and calibration

Analog input tests

Acoustic Processor Unit (APU) clock accuracy tests

Transducer removal and replacement

Wall thickness check of the LEFM spool piece

Calibration check of the feedwater header pressure transmitter

Communication link status checks.

Caldon Customer Information Bulletin 119, Tables 3 and 4, identifies certain LEFM parameters that must be monitored over time to ensure that the LEFM system is operating within the bounds of its uncertainty analysis. The LEFM CheckPlus System is designed to continuously self-monitor most of these parameters and conditions. For the WBN Unit 2 CheckPlus system, there are only two (2) LEFM parameters that require periodic manual verification:

LEFM spool piece wall thickness (measured with an ultrasonic thickness gage)

Calibration of the feedwater system pressure monitor using a traceable standard CNL-19-082 E2-4 of 78

Enclosure 2 WBN preventative maintenance tasks address the above two items in accordance with vendor recommendations. Ultrasonic wall thickness measurements of the LEFM spool piece are to be performed every third refueling outage. The WBN Unit 2 feedwater header pressure input for the LEFM CheckPlus System (from Pressure Transmitter 2-PT-3-34) has a calibration frequency requirement of every 18 months (with up to a 25% extension allowed, for a maximum interval of 22.5 months). This pressure transmitter is required to have an instrument error of 15 psi, consistent with Cameron reports ER-734P and ER-157P.

The preventative maintenance program and continuous monitoring of the LEFM ensure that the LEFM operation remains bounded by the analysis and assumptions set forth by the LEFM vendor. The continued adherence to these requirements provides assurance that the LEFM system is properly maintained and calibrated.

Operation:

Details of WBN Unit 2 proposed operation (including contingencies for LEFM unavailability) are discussed in response to Criterion 1 from ER-157P, Revision 8, below.

I.1.D.ii Criterion 2 from ER-80P - For plants that currently have LEFMs installed, provide an evaluation of the operational and maintenance history of the installed installation and confirmation that the installed instrumentation is representative of the LEFM system and bounds the analysis and assumptions set forth in Topical Report ER-80P.

RESPONSE

Criterion 2 applies to WBN Unit 2 as the LEFM CheckPlus system is installed and operational but is not currently used as input to the secondary calorimetric power measurements.

WBN Unit 2 currently uses flow venturis to measure feedwater flow to support the secondary calorimetric power measurements. The WBN LEFM system installed instrumentation is representative of the LEFM system and is bounded by the analysis and assumptions set forth in ER-80P and ER-157P. A summary of the operational and maintenance history for the WBN Unit 2 system is provided below.

The maintenance history of the LEFM system since plant startup (i.e., October 2016) has been reviewed. No significant LEFM system failures or repairs were identified. However, an issue that resulted in the LEFM system being in MAINTENANCE mode for an extended period was identified. LEFM Path 2 indicated a Fail state due to Data Rejects. If the LEFM indicates a Fail state and Data Rejects, then transducer replacement may be necessary. (Data Rejects will occur if there is total failure of the transducer input signal, a weak transducer signal, or an increase in random or coherent noise on the transducer signal.) Therefore, both Path 2 transducers are planned to be replaced. WBN Unit 2 currently operates using the venturis to support the secondary side calorimetric, so having the LEFM system in MAINTENANCE mode does not adversely affect operation of the plant.

Preventive maintenance scope and frequency is based on vendor recommendations and performance data reviews. Transducers will continue to be replaced as determined to be necessary. Other LEFM system components (e.g., CPUs, APUs) will be replaced or upgraded, as necessary, in accordance with TVA procedures.

The operational and maintenance history of these components shows that the system is currently acceptable for feedwater flow measurement and thermal power calculations. The LEFM system will be fully functional prior to MUR power uprate.

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Enclosure 2 I.1.D.iii Criterion 3 from ER-80P - Confirm that the methodology used to calculate the uncertainty of the LEFM in comparison to the current feedwater instrumentation is based on accepted plant setpoint methodology (with regard to the development of instrument uncertainty). If an alternative approach is used, the application should be justified and applied to both venturi and ultrasonic flow measurement instrumentation for comparison.

RESPONSE

The LEFM uncertainty calculation is based on the American Society of Mechanical Engineers (ASME) Performance Test Code (PTC) 19.1-1985, Measurement Uncertainty (Reference I.11); Instrument Society of America (ISA) Recommended Practice (RP)

ISA RP 67.04, Part II, Methodologies for the Determination of Setpoints for Nuclear Safety-Related Instrumentation, September 1994 (Reference I.12); and Alden Research Laboratory Inc. calibration tests. The methodology used to calculate LEFM uncertainty is described in Cameron reports ER-80P and ER-157P. This methodology has been used for instrument uncertainty calculations for multiple MUR power uprates that were accepted by the NRC. The WBN Unit 2 LEFM uncertainty analysis is provided in Cameron report ER-734P (Enclosure 5).

The feedwater flow and temperature uncertainties are combined with other plant measurement uncertainties (e.g., SG blowdown flow, steam pressure, and feedwater pressure) to calculate the overall heat balance uncertainty as described in Section I.1.E below. Both the Caldon analysis and the current analysis of record are based on a square-root-of-the-sum-of-the-squares (SRSS) calculation.

I.1.D.iv Criterion 4 from ER-80P - For plants where the ultrasonic meter (including LEFM) was not installed and flow elements calibrated to a site-specific piping configuration (flow profiles and meter factors not representative of the plant specific installation), additional justification should be provided for its use.

The justification should show that the meter installation is either independent of the plant specific flow profile for the stated accuracy, or that the installation can be shown to be equivalent to known calibrations and plant configurations for the specific installation including the propagation of flow profile effects at higher Reynolds numbers. Additionally, for previously installed calibrated elements, confirm that the piping configuration remains bounding for the original LEFM installation and calibration assumptions.

RESPONSE

This criterion does not apply to WBN Unit 2, as the flow elements were tested and calibrated in a full-scale model of the WBN hydraulic geometry at the Alden Research Laboratory. A bounding calibration factor for the WBN Unit 2 spool piece was established by these tests and is included in the Cameron engineering report (ER-734P). An Alden data report for these tests and a Cameron Meter Factor Calculation and Accuracy Assessment (ER-732) evaluating the test data were prepared. The piping configuration at WBN Unit 2 remains bounded by the original LEFM flow meter installation and calibration assumptions as analyzed in Cameron engineering reports ER-80P and ER-157P, Revision 8. Additionally, a bounding LEFM uncertainty has been used in the total thermal power uncertainty calculation described in Section I.1.E below. The site-specific uncertainty analyses are provided in Enclosures 5 and 6.

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Enclosure 2 I.1.D.v Criterion 1 from ER-157P, Rev 8 - Continued operation at the pre-failure power level for a pre-determined time and the decrease in power that must occur following that time are plant-specific and must be acceptably justified.

RESPONSE

Similar to WBN Unit 1, if a non-functional LEFM for WBN Unit 2 is not restored to functional status prior to the next performance of TS SR 3.3.1.2, which is performed every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , then Unit 2 power will be reduced to no more than 3411 MWt (i.e., the current licensed thermal power (CLTP)). This is consistent with the proposed change to TS 5.9.5b, which states When feedwater flow measurements from the LEFM are unavailable, the originally approved initial power level of 102% RTP (3411 MWt) shall be used.

The basis for the proposed completion time (CT) of prior to the next performance of TS SR 3.3.1.2, which would potentially allow for a maximum of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to reducing power, is as follows:

1. The same CT is used for WBN Unit 1.

2. When an LEFM system is non-functional, signals from the existing feedwater flow venturis will be used as input to the secondary calorimetric portion of the RTP calculation in place of the LEFM system. During normal LEFM operations, the signals from the flow venturi are calibrated to the LEFM signals, and upon LEFM failure, the flow venturi calibration is locked to the last valid LEFM value.

3. Any slight drift of the feedwater flow nozzle measurements due to fouling would result in a higher than actual indication of feedwater flow and an overestimation of the calculated calorimetric power level. This is conservative because the reactor will actually be operating below the calculated power level. A sudden de-fouling event during a 24-hour period is unlikely and any significant sudden de-fouling would be detected as a corresponding reduction in indicated thermal power that deviates from other plant parameters.

4. It is expected that minor issues resulting in a non-functional LEFM system could be resolved prior to the next performance of TS SR 3.3.1.2.

5. The NRC has approved a 72-hour CT for previous MUR power uprate applications, which bounds the maximum allowed outage time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months proposed for the LEFM system for WBN Unit 2 (References I.13 - I.16).

Additionally, for WBN Unit 2, the redundancy inherent in the two measurement planes of an LEFM CheckPlus system makes the system more tolerant to component failures, as compared to the Check system installed for WBN Unit 1, which only has one measurement plane. The LEFM CheckPlus system has three modes: NORMAL, MAINTENANCE, and FAIL. If an LEFM flow meter is in a status other than NORMAL, the uncertainty for that meter is increased. If the WBN Unit 2 LEFM is in MAINTENANCE mode then a 72-hour CT will be used prior to reducing power to 3411 MWt. In addition to the basis provided above (in Bullets 2, 3, and 5) for the CT associated with an LEFM in FAIL mode, the following considerations are used to justify a CT of up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for an LEFM in MAINTENANCE mode for WBN Unit 2:

WBN Unit 2 is only proposing a 1.4% power uprate in order for WBN Unit 2 to remain consistent with WBN Unit 1. This 1.4% power uprate is based on a bounding uncertainty analysis (Enclosure 6) which uses the same uncertainty as for the WBN Unit 1 Check system (i.e., an LEFM flow uncertainty of 0.48%). As shown in Section I.I.D.vi, and discussed in Section I.1.E, this assumed uncertainty value bounds the total thermal power uncertainty corresponding to the LEFM CheckPlus system in MAINTENANCE mode (as calculated in Enclosure 5).

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Enclosure 2

It is expected that most issues resulting in an LEFM system in MAINTENANCE mode (e.g., replacement of a transducer) could be resolved within a 72-hour CT.

I.1.D.vi Criterion 2 from ER-157P, Rev 8 - A CheckPlus system operating with a single failure is not identical to an LEFM Check system. Although the effect on hydraulic behavior is expected to be negligible, this must be acceptably quantified if a licensee wishes to operate using the degraded CheckPlus at an increased uncertainty.

RESPONSE

When an LEFM CheckPlus system has only one of its two LEFM Check subsystems fully operational, resulting in that meter computing flow from just the remaining fully operational LEFM Check subsystem, that LEFM flow meter is considered to be in the MAINTENANCE mode. This status is indicated to operators on the ICS in the control room. The total thermal power uncertainties for the WBN Unit 2 LEFM CheckPlus system in NORMAL and MAINTENANCE modes are provided in Cameron ER-734P, Revision 2 (Reference I.8) which is included in Enclosure 5. These values were quantified on a plant-specific basis, in accordance with the methodology of ER-157P, as discussed in Cameron report ER-734P.

The additional plant specific-inputs for the calculation of total thermal power uncertainty in Cameron report ER-734P (i.e., Items 20, 21, and 22 in Appendix B to ER-734P) are consistent with those used in WCAP-18419-P (Reference I.17 and Enclosure 6, Table 3 and in the equation in Section 3.1.1) but they are displayed differently in some cases. The steam enthalpy/pressure term and other gains/losses terms in ER-734P (Enclosure 5) account for the systematic and random effects of blowdown, as appropriate, but the SG blowdown flow uncertainties and net pump heat addition uncertainty are indicated separately in WCAP-18419-P (Enclosure 6, Table 3). These plant-specific uncertainty inputs are also more conservative than those used in Cameron reports ER-80P and ER-157P. The steam enthalpy/moisture term (based on the steam tables) is the same in both Enclosures 5 and 6 and is consistent with the analyses in Cameron reports ER-80P and ER-157P.

I.1.D.vii Criterion 3 from ER-157P, Rev 8 - An applicant with a comparable geometry can reference the above Section 3.2.1 finding to support a conclusion that downstream geometry does not have a significant influence on CheckPlus calibration. However, CheckPlus test results do not apply to a Check and downstream effects with the use of a CheckPlus with disabled components that make the CheckPlus comparable to a Check must be addressed. An acceptable method is to conduct applicable Alden Laboratory tests.

RESPONSE

The NRC has determined in Reference I.7 that for conditions in which the CheckPlus system is operating with one or more transducers out of service, the effect of downstream piping should be addressed if the separation distance from the meter transducers to the downstream piping change is less than five pipe diameters. At WBN Unit 2, the LEFM is installed in the feedwater header and the distance from meter transducers to downstream piping changes is greater than five pipe diameters. Therefore, the downstream geometries for WBN Unit 2 do not have a significant influence on CheckPlus calibration.

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Enclosure 2 I.1.D.viii Criterion 4 from ER-157P, Rev 8 - An applicant that requests a MUR with the upstream flow straightener configuration discussed in Section 3.2.2 should provide justification for claimed CheckPlus uncertainty that extends the justification provided in Reference 17 (Letter from Hauser, E. (Cameron), to NRC, Documentation to support the review of ER-157P, Revision 8:

Engineering Report ER-790, Revision 1, An Evaluation of the Impact of 55 Tube Permutit Flow Conditioners on the Meter Factor of an LEFM CheckPlus, March 19, 2010). Since the Reference 17 evaluation does not apply to the Check, a comparable evaluation must be accomplished if a Check is to be installed downstream of a tubular flow straightener.

RESPONSE

The installed configuration of the WBN Unit 2 LEFM in the feedwater header does not include an upstream flow straightener. Therefore, this criterion is not applicable to WBN Unit 2.

I.1.D.ix Criterion 5 from ER-157P, Rev 8 - An applicant assuming large uncertainties in steam moisture content should have an engineering basis for the distribution of the uncertainties or, alternatively, should ensure that their calculations provide margin sufficient to cover the differences shown in Figure 1 of Reference 18 (Letter from Hauser, E. (Cameron), to NRC, Documentation to support the review of ER-157P, Revision 8: Engineering Report ER-764, Revision 0, The Effect of the Distribution of the Uncertainty in Steam Moisture Content on the Total Uncertainty in Thermal Power, March 18, 2010)

RESPONSE

The internal moisture separation equipment is designed to ensure that moisture carryover does not exceed 0.25% by weight (wt%) for the WBN Unit 2 Westinghouse Model D3-2 Steam Generators (SGs). This value was used as an input in the calculation of the total power measurement uncertainty (Enclosure 6). For steam quality, the steam tables were used to determine the sensitivity at a moisture content of 0.25%. The proprietary value for the power measurement sensitivity with respect to moisture content is provided in Table 2 of Enclosure 6.

The results of the overall power measurement uncertainty (Enclosure 6) indicate that an uncertainty value (rounded up) of 0.6% RTP is bounding.

By inspection of Figure 1 of Cameron report ER-764 (Reference I.9), for a given moisture uncertainty of 0.25% (% full power), which bounds the plant-specific value determined in , the difference in the total uncertainty (% full power) between using an uniform distribution of moisture uncertainty, with other power uncertainties normally distributed, versus using the SRSS for moisture and other uncertainty values, is estimated to be less than 0.03%

full power. For the actual moisture uncertainty value identified in Table 2 of Enclosure 6, the corresponding increase in total power uncertainty is approximately 0.02% full power per Figure 1. The margin in the total uncertainty value of 0.6% RTP for WBN Unit 2 as compared to the more precise value of approximately 0.55% (not shown in Enclosure 6) calculated by using the proprietary values in the SRSS formula in Section 3.1.1 of Enclosure 6 is greater than 0.03% RTP. Therefore, there is sufficient margin to cover the differences shown in Figure 1 of Cameron report ER-764, and the use of 0.6% RTP as the bounding total uncertainty value for WBN Unit 2 remains conservative given a moisture content of 0.25 wt%. If the additional uncertainty to account for moisture uncertainty is added to the values in Cameron report ER-734P, the total thermal power uncertainties for the LEFM in NORMAL and MAINTENANCE modes remain bounded by the 0.6% total thermal power uncertainty.

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Enclosure 2 I.1.E A calculation of the total power measurement uncertainty at the plant, explicitly identifying all parameters and their individual contribution to the power uncertainty

RESPONSE

The calculation of total power measurement uncertainty has been completed for WBN Unit 2 and is included in Enclosure 6.

Cameron report ER-734P shows that the LEFM mass flow uncertainty, which is used as input to the calorimetric, is less than 0.48% for WBN Unit 2. The uncertainty (in NORMAL and MAINTENANCE modes) was determined utilizing the calculation methodology described in Cameron Engineering Reports ER-80P and ER-157P (References I.2 and I.3).

In addition to the feedwater header mass flow rate and feedwater temperature provided by the Cameron CheckPlus system, the WBN Unit 2 ICS uses the following measured inputs:

Feedwater pressure

Steam pressure

Steam generator blowdown flow An uncertainty calculation was performed for each of these process inputs to determine a bounding instrument loop uncertainty for WBN Unit 2. Enclosure 6 denotes the individual contribution to the total power measurement uncertainty for each input parameter and additional parameters such as net pump heat addition and steam moisture content. As shown in Table I.1.E-1, an assumed bounding LEFM thermal power uncertainty of 0.48% was combined with the non-LEFM uncertainties to obtain a total power uncertainty of 0.6% RTP (rounded up) for WBN Unit 2. A bounding feedwater density/pressure uncertainty term (% power/psi)

(Enclosure 6, Table 2) was used, consistent with WBN Unit 1, instead of a realistic value based on the steam tables, which provides additional conservatism.

Table I.1.E-1: Bounding Total Thermal Power Uncertainty Determination Parameter WBN Unit 2 Analysis Assumed Bounding Power Uncertainty Due to LEFM 0.48%

(See Westinghouse Report, WCAP-18419-P, in Enclosure 6)

Total Thermal Power Uncertainty 0.6%

(See Westinghouse Report, WCAP-18419-P, in Enclosure 6)

WCAP-18419-P is the portion of the analysis of record, for WBN Unit 2, pertaining to the use of the LEFM for calculating the total thermal power uncertainty. This bounding analysis supports a power uprate of 1.4% so that WBN Unit 2 will be consistent with WBN Unit 1. Although the Westinghouse methodology used for calculation of the total thermal power uncertainty for WBN Unit 2 is not generically approved by the NRC, it was previously accepted for WBN Unit 1 (Reference I.18). The Cameron engineering reports demonstrate that the thermal power uncertainty results of the Westinghouse analysis are conservative and bound use of the LEFM in both NORMAL and MAINTENANCE modes.

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Enclosure 2 I.1.F Information to specifically address the following aspects of the calibration and maintenance procedures related to all instruments that affect the power calorimetric:

I.1.F.i Maintaining calibration

RESPONSE

Calibration of the LEFM is ensured by preventative maintenance activities described in Section I.1.D, Response to Criterion 1 of ER-80P.

I.1.F.ii Controlling software and hardware configuration

RESPONSE

The Caldon LEFM CheckPlus System was procured in accordance with the appropriate TVA requirements for real time data acquisition and control computer system software requirements, electromagnetic interference testing requirements for electronics devices, and seismic qualification. The LEFM system configuration is controlled in accordance with TVA procedures.

LEFM software is classified in accordance with TVA procedures and software and digital assets are evaluated and protected in accordance with the TVA Nuclear Cyber Security Program.

Implementation and changes to the software is controlled in accordance with TVA procedures.

Instruments that affect the power calorimetric, including the Caldon LEFM CheckPlus System inputs, are monitored and maintained. Equipment issues for plant systems, including the Caldon LEFM CheckPlus System equipment, fall under site work control processes. Conditions that are adverse to quality are documented under the TVA corrective action program (CAP).

Corrective action programs and procedures ensure compliance with the requirements of 10 CFR 50, Appendix B, and include instructions for notification of deficiencies and error reporting.

I.1.F.iii Performing corrective actions

RESPONSE

Corrective actions are monitored and performed in accordance with the TVA CAP.

I.1.F.iv Reporting deficiencies to the manufacturer

RESPONSE

Reporting deficiencies to the manufacturer will be performed in accordance with the TVA procedural guidance regarding 10 CFR 21.

I.1.F.v Receiving and addressing manufacturer deficiency reports

RESPONSE

Manufacturer deficiency reports will be received and addressed in accordance with the TVA procedural guidance regarding 10 CFR 21.

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Enclosure 2 I.1.G A proposed allowed outage time for the instrument, along with the technical basis for the time selected

RESPONSE

Refer to the response in Section I.1.D, Criterion 1 from ER-157P.

I.1.H Proposed actions to reduce power level if the allowed outage time is exceeded, including a discussion of the technical basis for the proposed reduced power level

RESPONSE

The proposed actions to reduce power are stated in Section I.1.D, Criterion 1 from ER-157P.

References for Section I:

I.1. NRC Regulatory Issue Summary, RIS 2002-03, Guidance on Content of Measurement Uncertainty Recapture Power Uprate Applications, dated January 31, 2002 I.2. Cameron Engineering Report ER-80P, Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM Check System, Revision 0, dated March 1997 I.3. Cameron Engineering Report ER-157P-A, Supplement to Cameron Topical Report ER-80P: Basis for Power Uprates with an LEFM Check or an LEFM CheckPlus, Revision 8, dated May 2008, with Revision 8 errata I.4. NRC letter to TU Electric, Comanche Peak Steam Electric Station, Units 1 and 2 -

Review of Caldon Engineering Topical Report ER 80P, Improving Thermal Power Accuracy and Plant Safety While Increasing Power Level Using the LEFM System (TACS Nos. MA2298 and MA2299), dated March 8, 1999 (9903190065)

I.5. NRC letter to Cameron, Final Safety Evaluation for Cameron Measurement Systems Engineering Report ER-157P, Revision 8, Caldon Ultrasonics Engineering Report ER-157P, Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM Check or CheckPlus System, (TAC No. ME1321), dated August 16, 2010 (ML102160663)

I.6. Cameron Engineering Report ER-157P, Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFMTM or LEFM CheckPlusTM System, Revision 5, dated October 2001 I.7. NRC letter to McGuire Nuclear Station, McGuire Nuclear Station. Units 1 and 2. Issuance of Amendments Regarding Measurement Uncertainty Recapture Power Uprate (TAC Nos. ME8213 and ME8214), dated May 16, 2013 (ML13073A041)

I.8. Cameron Engineering Report ER-734P, Revision 2, Bounding Uncertainty Analysis for Thermal Power Determination at Watts Bar Unit 2 Using the LEFM+ System, dated August 2019 (Enclosure 5)

I.9. Cameron Engineering Report ER-764, Revision 0, The Effect of the Distribution of the Uncertainty in Steam Moisture Content on the Total Uncertainty in Thermal Power, dated September 2009 (ML100820167)

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Enclosure 2 I.10. Cameron Engineering Report ER-732, Revision 0, Meter Factor Calculation and Accuracy Assessment for Tennessee Valley Authority, Watts Bar Nuclear Plant Unit 2, dated December 2008 I.11. American Society of Mechanical Engineers (ASME) Performance Test Code (PTC) 19.1, Measurement Uncertainty, 1985 I.12. ISA-RP 67.04, Part II, "Methodologies for the Determination of Setpoints for Nuclear Safety Related Instrumentation," September 1994 I.13. NRC letter to Shearon Harris Nuclear Power Plant, Shearon Harris Nuclear Power Plant, Unit 1 -Issuance of Amendment Re: Measurement Uncertainty Recapture Power Uprate (TAC No. ME6169), dated May 30, 2012 (ML11356A096)

I.14. NRC letter to Calvert Cliffs Nuclear Power Plant, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 Amendment Re: Measurement Uncertainty Recapture Power Uprate (TAC Nos. MD9554 and MD9555), dated July 22, 2009 (ML091820366)

I.15. NRC letter to Exelon Nuclear, Limerick Generating Station, Units 1 and 2 - Issuance of Amendments Re: Measurement Uncertainty Recapture Power Uprate and Standby Liquid Control System Changes (TAC Nos. ME3589, ME3590, ME3591, and ME3592), dated April 8, 2011 (ML110691095)

I.16. NRC letter to Entergy Operations, Inc., Waterford Stem Electric Station, Unit 3; River Bend Station; and Grand Gulf Nuclear Station - Review of Caldon, Inc. Engineering Report ER-157P, Revision 5, dated December 20, 2001 (ML013540256)

I.17. Westinghouse WCAP-18419-P, Revision 1, Leading Edge Flow Meter (LEFM) Power Measurement Uncertainty for the Watts Bar Unit 2 MUR Program, 3459 MWt Core Power with LEFM I.18. NRC letter to TVA, Watts Bar Nuclear Plant, Unit 1 -Issuance of Amendment Regarding Increase of Reactor Power to 3459 Megawatts Thermal (TAC No. MA9152), dated January 19, 2001 (ML010260074)

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Enclosure 2 II ACCIDENTS AND TRANSIENTS FOR WHICH THE EXISTING ANALYSES OF RECORD BOUND PLANT OPERATION AT THE PROPOSED UPRATED POWER LEVEL II.1 A matrix that includes information for each analysis in this category and addresses the transients and accidents included in the plants updated final safety analysis report (UFSAR) (typically Chapter 14 or 15) and other analyses that licensees are required to perform to support licensing of their plants (i.e., radiological consequences, natural circulation cooldown, containment performance, anticipated transient without scram, station blackout, analyses to determine environmental qualification parameters, safe shutdown fire analysis, spent fuel pool cooling, flooding):

II.1.A Identify the transient or accident that is the subject of the analysis II.1.B Confirm and explicitly state that II.1.B.i The requested uprate in power level continues to be bounded by the existing analyses of record for the plant II.1.B.ii The analyses of record either have been previously approved by the NRC or were conducted using methods or processes that were previously approved by the NRC II.1.C Confirm that bounding event determinations continue to be valid II.1.D Provide a reference to the NRCs previous approvals discussed in Item B. above

RESPONSE

The response to II.1 is provided in Table II.1 WBN Unit 2 Analyses. Each analysis is described briefly below and the analyses are summarized in Table II.1-1, including the assumed core power level in each analysis and whether the analysis remains bounding for the MUR power uprate. The value of 3479 MWt corresponds to 102% of RTP, which remains the bounding power level for MUR power uprate conditions when uncertainty is applied. However, some of the analyses use higher power levels (e.g., 3480 MWt or 3565 MWt) because of rounding differences or to provide additional conservatism. The WBN Unit 2 accident and safety analyses including loss of coolant accident (LOCA) events are addressed in Chapter 15 of the WBN dual-unit updated final safety analysis report (UFSAR). These analyses were performed to support the issuance of Facility Operating License (OL) No. NPF-96, for WBN Unit 2, on October 22, 2015. The results and conclusions reported in the UFSAR, which were approved by the NRC in License Condition 2.A of the OL remain applicable at the uprated conditions. The NRC approval is documented in the Safety Evaluation Report (SER),

NUREG-0847, Safety Evaluation Report Related to the Operation of Watts Bar Nuclear Plant, Units 1 and 2, as supplemented (through Supplement 29). (Specific SER supplements are referenced in Table II.1-1, as necessary.)

Thermal conductivity degradation (TCD) is addressed in WBN Unit 2 OL condition 2.C.(4). This license condition allows the use of Performance Analysis and Design (PAD) Model PAD4TCD until the WBN Unit 2 SGs are replaced (Reference II.1) at which time WBN Unit 2 will transition to the PAD5 model. The proposed power uprate does not affect this OL condition.

Additional analyses, not described in Chapter 15 and performed to support the current licensing basis, are also discussed with an assessment of whether they remain bounding for the MUR power uprate. The methodology in these analyses is found in the UFSAR, Topical Reports, and other documents as referenced in Table II.1-1. The NRC review and approval of the analyses is also referenced in Table II.1-1.

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Enclosure 2 II.1.D.i Reactor Trip System/Engineered Safeguards Features Actuation System Allowable Values The safety analyses performed for the MUR power uprate did not adjust the Reactor Trip System (RTS) or Engineered Safeguards Features Actuation System (ESFAS) nominal setpoints or allowable values from the non-uprated values. Therefore, the setpoints and allowable values remain unchanged from those in WBN Unit 2 Technical Specification Tables 3.3.1-1 and 3.3.2-1.

II.1.D.ii UFSAR Chapter 15 Analyses As described in UFSAR Section 15.1.2.2, for most accidents, which are departure from nucleate boiling (DNB), limited, nominal values of initial conditions are assumed. The allowance on power, temperature, and pressure are determined on a statistical basis and are included in the DNB limit ratio (DNBR). This procedure is known as the Revised Thermal Design Procedure (RTDP).

The minimum measured flow value is used in all RTDP transients. These analyses are performed at 101.4% of 3411 MWt (3459 MWt), plus a reactor coolant pump (RCP) net heat input of 16 MWt.

For accidents which are not DNB limited or for which the RTDP is not employed, the initial conditions are obtained by adding the bounding steady-state errors to nominal values in such a manner to maximize the impact on the limiting parameter. The thermal design flow value, which is the minimum measured flow minus measurement uncertainty, is used for such analyses.

As indicated in UFSAR Table 15.1-2, Notes 6 and 7, although several of the Chapter 15 analyses are based upon a core power of 3411 MWt and nuclear steam supply system (NSSS) power of 3425 MWt, an uprated core power of 3459 MWt and NSSS power of 3475 MWt are also supported via evaluation, based upon a redefinition of the 2% power uncertainty (i.e., from 2% to 0.6%).

II.1.D.iii Discussion of RIS 2002-03 Section II.1 Events

1. Uncontrolled Rod Cluster Control Assembly Bank Withdrawal from a Subcritical Condition (UFSAR Section 15.2.1)

The analysis documented in UFSAR Section 15.2.1 postulates an uncontrolled addition of reactivity to the reactor core caused by withdrawal of RCCAs resulting in a power excursion.

Such a transient could be caused by a malfunction of the reactor control or rod control systems. This could occur with the reactor either subcritical, hot zero power (HZP), or at power. The analysis in Section 15.2.1 was performed at HZP. The MUR power uprate has no impact on the HZP analysis.

The analysis of record (AOR) for this analysis is reflected in the WBN UFSAR and remains acceptable for the WBN Unit 2 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the reference listed in Table II.1-1.

2. Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power (UFSAR Section 15.2.2)

The analysis documented in UFSAR Section 15.2.2 postulates an uncontrolled rod cluster control assembly (RCCA) bank withdrawal at power, which results in an increase in the core heat flux. Because the heat extraction from the SG lags behind the core power generation CNL-19-082 E2-15 of 78

Enclosure 2 until the SG pressure reaches the relief or safety valve setpoint, there is a net increase in the reactor coolant temperature. Unless terminated by manual or automatic action, the power mismatch and resultant coolant temperature rise would eventually result in DNB.

Because the analysis was performed at a power level equal to the MUR uprated power level and acceptable results were obtained, the analysis is unaffected by the MUR power uprate.

The AOR for this analysis is reflected in the WBN UFSAR and remains acceptable for the WBN Unit 2 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the reference listed in Table II.1-1.

3. Rod Cluster Control Assembly Misalignment (UFSAR Section 15.2.3)

RCCA misalignment accidents evaluated in UFSAR Section 15.2.3 include: 1) one or more dropped RCCAs within the same group; 2) a dropped RCCA bank; and, 3) statically misaligned RCCA.

Because the analysis was performed at a power level equal to the MUR uprated power level and acceptable results were obtained, the analysis is unaffected by the MUR power uprate.

The AOR for this analysis is reflected in the WBN UFSAR and remains acceptable for the WBN Unit 2 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the reference listed in Table II.1-1.

4. Uncontrolled Boron Dilution (UFSAR Section 15.2.4)

The analysis documented in UFSAR Section 15.2.4 postulates inadvertent opening of the primary water control valve and failure of the blend system by either controller or mechanical failure, resulting in insertion of positive reactivity. Boron dilutions during refueling, cold shutdown, hot shutdown, hot standby, startup, and power operation are considered in the WBN Unit 2 analyses.

MUR power uprate has no impact on the analyses, except for the analysis performed at power operation. Because the analysis was performed at a power level equal to the MUR uprated power level and acceptable results were obtained, the analysis is unaffected by the MUR power uprate.

The AOR for this analysis is reflected in the WBN UFSAR and remains acceptable for the WBN Unit 2 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the reference listed in Table II.1-1.

5. Partial Loss of Forced Reactor Coolant Flow (UFSAR Section 15.2.5)

The analysis documented in UFSAR Section 15.2.5 postulates a partial loss of reactor coolant flow involving the loss of one RCP with four loops in operation.

Because the analysis was performed at a power level equal to the MUR uprated power level and acceptable results were obtained, the analysis is unaffected by the MUR power uprate.

The AOR for this analysis is reflected in the WBN UFSAR and remains acceptable for the WBN Unit 2 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the reference listed in Table II.1-1.

6. Startup of an Inactive Reactor Coolant Loop at an Incorrect Temperature (UFSAR Section 15.2.6)

Following the startup of an inactive RCP, the flow in the inactive loop will accelerate to full flow in the forward direction over a period of several seconds. As discussed in UFSAR Section 15.2.6, no analysis is required to show that the minimum DNBR limit is satisfied for CNL-19-082 E2-16 of 78

Enclosure 2 this event. Because the WBN Unit 2 TS 3.4.4 requires the RCPs to be operating while in Modes 1 and 2, the maximum initial core power level for the startup of an inactive loop transient is approximately zero MWt. Under these conditions, there can be no significant reactivity insertion because the reactor coolant system (RCS) is initially at a nearly uniform temperature. Furthermore, the reactor will initially be subcritical as required by the TS for the other modes of operation. Thus, there will be no increase in core power, and no automatic or manual protective action is required. This analysis is normally run at high power levels for (N-1) loop operation plants. WBN design does not currently include this operating configuration and, therefore, no analysis is required.

The conclusion that no analysis of this event is required for WBN Unit 2 is unaffected by the MUR power uprate.

7. Loss of External Electrical Load and/or Turbine Trip (UFSAR Section 15.2.7)

The analysis documented in UFSAR Section 15.2.7 postulates a major load loss on the plant, which can result from loss of external electrical load or from a turbine trip. For either case, offsite power is available for the continued operation of plant components such as the RCPs. This analysis, along with the loss of normal feedwater (Section 15.2.8) and complete loss of forced reactor coolant flow (Section 15.3.4) address the case of loss of offsite power to the station auxiliaries (Section 15.2.9).

Because the analysis was performed at a power level equal to the MUR uprated power level and acceptable results were obtained, the analysis is unaffected by the MUR power uprate.

The AOR for this analysis is reflected in the WBN UFSAR and remains acceptable for the WBN Unit 2 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the reference listed in Table II.1-1.

8. Loss of Normal Feedwater (UFSAR Section 15.2.8)

The analysis documented in UFSAR Section 15.2.8 postulates a loss of normal feedwater (from pump failures, valve malfunctions, or loss of offsite AC power). Two cases are examined: The first is the case where offsite AC power is maintained, and the second is the case where offsite AC power is lost, which results in RCP coastdown.

Because the analysis was performed at a power level that bounds the MUR uprated power level and acceptable results were obtained, the analysis is unaffected by the MUR power uprate.

The AOR for this analysis is reflected in the WBN UFSAR and remains acceptable for the WBN Unit 2 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table II.1-1.

9. Coincident Loss of Onsite and External (Offsite) AC Power to the Station - Loss of Offsite Power to the Station Auxiliaries (UFSAR Section 15.2.9)

A complete loss of all offsite power (LOOP) will result in the loss of offsite power to the plant auxiliaries (e.g., the RCPs, condensate pumps). The loss of power may be caused by a complete loss of the offsite grid accompanied by a turbine generator trip at the station, or by a loss of the onsite AC distribution system. The analysis is contained in UFSAR Sections 15.2.7, 15.2.8, and 15.3.4 (Sections II.I.D.iii.7, 8, and 19 of this enclosure).

The offsite dose analysis for a loss of AC power to plant auxiliaries in UFSAR Section 15.5.1 was performed using a power level of 3565 MWt for a conventional core. In Reference II.2, NRC approved WBN Unit 2 to operate up to 1,792 Tritium Producing Burnable Absorber Rods (TPBARs). As noted in Table 4.1-4 of the WBN Unit 2 Tritium Production Core (TPC)

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Enclosure 2 LAR (Reference II.3), the secondary steam inventory will be based on 3480 MWt, which bounds the MUR power level, following TPC implementation.

10. Excessive Heat Removal Due to Feedwater System Malfunctions (UFSAR Section 15.2.10)

The analysis documented in UFSAR Section 15.2.10 postulates excessive feedwater flow, which could be caused by a full opening of one or more feedwater control valves due to a feedwater control system malfunction or an operator error. At power, this excess flow causes a greater load demand on the RCS due to increased subcooling in the SG.

Because the analysis was performed at a power level that bounds the MUR uprated power level and acceptable results were obtained, the analysis is unaffected by the MUR power uprate.

The AOR for this analysis is reflected in the WBN UFSAR and remains acceptable for the WBN Unit 2 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the reference listed in Table II.1-1.

11. Excessive Load Increase Incident (UFSAR Section 15.2.11)

The analysis documented in UFSAR Section 15.2.11 postulates an excessive load increase incident, which is defined as a rapid increase in the steam flow that causes a power mismatch between the reactor core power and the SG load demand. The RCS is designed to accommodate a 10% step load increase or a 5% per minute ramp load increase in the range of 15% to 100% of full power. Any loading rate in excess of these values may cause a reactor trip actuated by the reactor protection system. An excessive load increase accident typically does not result in a reactor trip, and the plant soon reaches a new equilibrium condition at a higher power level based on the increased steam load.

Because the analysis was performed at a power level that bounds the MUR uprated power level and acceptable results were obtained, the analysis is unaffected by the MUR power uprate.

The AOR for this analysis is reflected in the WBN UFSAR and remains acceptable for the WBN Unit 2 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the reference listed in Table II.1-1.

12. Accidental Depressurization of the Reactor Coolant System (UFSAR Section 15.2.12)

The analysis documented in UFSAR Section 15.2.12 postulates an accidental depressurization of RCS due to an inadvertent opening of a pressurizer safety valve.

Because the analysis was performed at a power level that bounds the MUR uprated power level and acceptable results were obtained, the analysis is unaffected by the MUR power uprate.

The AOR for this analysis is reflected in the WBN UFSAR and remains acceptable for the WBN Unit 2 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the reference listed in Table II.1-1.

13. Accidental Depressurization of the Main Steam System (UFSAR Section 15.2.13)

The analysis documented in UFSAR Section 15.2.13 postulates the accidental depressurization of the main steam system associated with an inadvertent opening of a single steam dump, relief, or safety valve.

Because the analysis was performed at a power level that bounds the MUR uprated power level and acceptable results were obtained, the analysis is unaffected by the MUR power uprate.

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Enclosure 2 The AOR for this analysis is reflected in the WBN UFSAR and remains acceptable for the WBN Unit 2 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the reference listed in Table II.1-1.

14. Inadvertent Operation of Emergency Core Cooling System During Power Operation (UFSAR Section 15.2.14)

The analysis documented in UFSAR Section 15.2.14 postulates spurious ECCS operation at power, which could be caused by operator error or a false electrical actuating signal.

Because the analysis was performed at a power level equal to the MUR uprated power level and acceptable results were obtained, the analysis is unaffected by the MUR power uprate.

The AOR for this analysis is reflected in the WBN UFSAR and remains acceptable for the WBN Unit 2 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the reference listed in Table II.1-1.

15. Chemical and Volume Control System Malfunction During Power Operation (UFSAR Section 15.2.15)

The analysis documented in UFSAR Section 15.2.15 postulates an increase in reactor coolant inventory caused by a malfunction of the chemical and volume control system, which may result from operator error or a control signal malfunction.

Because the analysis was performed at a power level equal to the MUR uprated power level and acceptable results were obtained, the analysis is unaffected by the MUR power uprate.

The AOR for this analysis is reflected in the WBN UFSAR and remains acceptable for the WBN Unit 2 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the reference listed in Table II.1-1.

16. Loss of Reactor Coolant from Small Ruptured Pipes or from Cracks in Large Pipes Which Actuate the Emergency Core Cooling System (UFSAR Section 15.3.1)

The analysis documented in UFSAR Section 15.3.1 postulates a rupture of RCS piping. A spectrum of break sizes was analyzed to determine the limiting break size in terms of the highest peak cladding temperature. The four-inch break was determined to be the limiting break size for WBN Unit 2. A small break loss of coolant accident (SBLOCA) evaluation was performed for the MUR power uprate with the OSGs. The purpose of the SBLOCA evaluation is to demonstrate continued compliance with the 10 CFR 50.46 requirements at MUR uprate conditions. The analysis used the Westinghouse SBLOCA Evaluation Model (References II.4, II.5, and II.6).

The SBLOCA AOR and the subsequent evaluations have explicitly analyzed a core power of 3459 MWt plus a 0.6% power uncertainty for a total core power of 3480 MWt in accordance with Appendix K to 10 CFR 50. A power measurement uncertainty less than two percent is permitted under Appendix K to 10 CFR 50. For SBLOCA analyses, the exact power measurement uncertainty is not critical as long as the combination of licensed core power and calorimetric uncertainty does not exceed the analyzed core power.

Therefore, the WBN Unit 2 AOR is applicable to a rated thermal power level of 3459 MWt with a 0.6% power uncertainty and is applicable at MUR power uprate conditions. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table II.1-1.

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Enclosure 2

17. Minor Secondary System Pipe Breaks (UFSAR Section 15.3.2)

As described in UFSAR Section 15.3.2, the analyses presented in Section 15.4.2 (Major Secondary System Pipe Rupture) demonstrate that the consequences of a minor secondary system pipe break are acceptable because a DNBR of less than the limiting value does not occur even for a more critical major secondary system pipe break. A separate analysis for minor secondary system pipe breaks is not required, and this conclusion is unaffected by the MUR power uprate.

18. Inadvertent Loading of a Fuel Assembly into an Improper Position (UFSAR Section 15.3.3)

As described in UFSAR Section 15.3.3, fuel and core loading errors such as can arise from the inadvertent loading of one or more fuel assemblies into improper positions, loading a fuel rod during manufacture with one or more pellets of the wrong enrichment, or the loading of a full fuel assembly during manufacture with pellets of the wrong enrichment, will lead to increased heat fluxes if the error results in placing fuel in core positions calling for fuel of lesser enrichment. Also included among possible core loading errors is the inadvertent loading of one or more fuel assemblies requiring burnable poison rods into a new core without burnable poison rods.

Fuel assembly enrichment errors would be prevented by administrative procedures implemented in fabrication. In the event that a single pin or pellet has a higher enrichment than the nominal value, the consequences in terms of reduced DNBR and increased fuel and clad temperatures will be limited to the incorrectly loaded pin or pins.

Fuel assembly loading errors are prevented by administrative procedures implemented during core loading. In the unlikely event that a loading error occurs, the analyses presented in Section 15.3.3 confirm that resulting power distribution effects either will be readily detected by the Power Distribution Monitoring System or will cause a sufficiently small perturbation to be acceptable within the uncertainties allowed between nominal and design power shapes.

The AOR for this analysis is reflected in the WBN UFSAR and remains acceptable for the WBN Unit 2 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the reference listed in Table II.1-1.

19. Complete Loss of Forced Reactor Coolant Flow (UFSAR Section 15.3.4)

The analysis documented in UFSAR Section 15.3.4 postulates a complete loss of forced reactor coolant flow may result from a simultaneous loss of electrical supplies to the RCPs.

The analysis performed demonstrated that for the complete loss of forced reactor coolant flow, the DNBR will not decrease below the design basis limit at any time during the transient.

Because the analysis was performed at a power level equal to the MUR uprated power level and acceptable results were obtained, the analysis is unaffected by the MUR power uprate.

The AOR for this analysis is reflected in the WBN UFSAR and remains acceptable for the WBN Unit 2 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the reference listed in Table II.1-1.

20. Waste Gas Decay Tank Rupture (UFSAR Section 15.3.5)

As discussed in UFSAR Section 15.3.5, the accident is defined as an unexpected and uncontrolled release of radioactive xenon and krypton fission product gases stored in a waste decay tank as a consequence of a failure of a single gas decay tank or associated piping. The analysis and consequences are further discussed in Section 15.5.2.

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Enclosure 2 Two analyses of the postulated waste gas decay tank rupture are performed: (1) a realistic analysis, and (2) an analysis based on Regulatory Guide (RG) 1.24. It is assumed that the reactor has been operating at full power with 1% defective fuel for the RG 1.24 analysis.

Noble gas and iodine inventories for the RG 1.24 analysis are given in UFSAR Table 15.5-4.

For the realistic analysis, source terms are based on ANSI/ANS-18.1-1984 methodology.

The conventional core primary and secondary coolant source terms are based on 3582 MWt for the realistic analysis.

Table 4.1-4 of Reference II.3 provides the updated parameters used for the primary and secondary coolant concentrations. In that table, the updated thermal power parameter is listed as 3480 MWt (following TPC implementation), which bounds the MUR power uprate power level.

The core thermal power assumed in the RG 1.24 analysis is 3565 MWt. Because the RG 1.24 and realistic analyses were performed at a power level greater than the MUR uprated power level, the waste gas decay tank AOR is unaffected by the MUR power uprate.

The AOR for this analysis is reflected in the WBN UFSAR and remains acceptable for the WBN Unit 2 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table II.1-1.

21. Single Rod Cluster Control Assembly Withdrawal at Full Power (UFSAR Section 15.3.6)

The analysis for WBN Unit 2 documented in UFSAR Section 15.3.6 postulates the case of the worst rod withdrawn from bank D inserted at the insertion limit, with the reactor initially at full power. This incident is assumed to occur at beginning-of-life because this results in the minimum value of moderator temperature coefficient. This maximizes the power rise and minimizes the tendency of increased moderator temperature to flatten the power distribution.

Because the analysis was performed at a power level equal to the MUR uprated power level and acceptable results were obtained, the analysis is unaffected by the MUR power uprate.

The AOR for this analysis is reflected in the WBN UFSAR and remains acceptable for the WBN Unit 2 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the reference listed in Table II.1-1.

22. Major Reactor Coolant System Pipe Ruptures (Loss of Coolant Accident)

(UFSAR Section 15.4.1)

The analysis for WBN Unit 2 documented in UFSAR Section 15.4.1 postulates a double-ended guillotine cold leg break LOCA. The best estimate large break LOCA (BE LBLOCA) analysis (Reference II.18) performed for WBN Unit 2 uses the Automated Statistical Treatment of Uncertainty Method (ASTRUM) (Reference II.12). Per Section 5.2.1 of Reference II.18, the BE LBLOCA analysis is performed at an assumed core power of 3479.8 MWt, which was specifically done to bound operation considering MUR power uprate conditions. Therefore, the reported 10 CFR 50.46 results in UFSAR Table 15.4-18b remain applicable and meet the acceptance criteria of 10 CFR 50.46.

The post-LOCA long-term core cooling (LTCC) AOR was also reviewed. This analysis addresses three concerns relative to managing the post-LOCA recovery: subcriticality, boric acid precipitation control (i.e., hot leg switchover), and decay heat removal. Implementation of the 1.4% MUR power uprate results in a change in the allocation of the current rated thermal power (3411 MWt) and calorimetric uncertainty (2%). There are no changes to other plant operating parameters affecting the post-LOCA LTCC analysis (boron concentrations in the refueling water storage tank (RWST) and ice beds, and fluid volumes CNL-19-082 E2-21 of 78

Enclosure 2 within various RCS sub-systems for example) being introduced as a result of the power uprate.

The post-LOCA subcriticality analysis confirms that sufficient negative reactivity will remain present in the core region to ensure that a re-criticality will not occur following the initial reactor trip with the RCCAs fully withdrawn. This analysis calculates a post-LOCA sump mixed mean boron concentration required to ensure subcriticality as a function of the initial pre-trip RCS peak boron concentration. Due to potential changes in pre-trip boron concentration resulting from cycle- specific core designs, the sump mixed mean boron concentration is confirmed on a per-cycle basis as a part of the established Westinghouse reload methodology (Reference II.7).

As the core boils and make-up flow is provided by the ECCS following a LOCA, boric acid concentration in the vessel will increase. If the boron concentration increases such that the solubility limit is exceeded, the boric acid could come out of solution (precipitate) and degrade the ability to cool the core. The potential for boric acid precipitation results in the need for a mechanism by which the reactor vessel (RV) boric acid concentration can be reduced (i.e., active dilution mechanism). For a Westinghouse designed four-loop plant such as WBN, RV boric acid dilution is achieved by aligning the ECCS to inject to the hot legs at an appropriate time The post-LOCA boric acid precipitation control (BAPC) analysis calculates the latest allowable time for operators to complete the transfer to hot leg recirculation before exceeding the solubility limit. This analysis assumes a maximum pre-transient RCS boron concentration. Due to potential changes in pre- transient boron concentration that can result from cycle-specific core designs, the maximum pre-transient RCS boron concentration used in the BAPC analysis is confirmed on a per-cycle basis (Reference II.7).

The purpose of the post-LOCA decay heat removal (DHR) analysis is to ensure that the core remains cooled by a two-phase mixture following the relatively short-term time domain explicitly modeled within the LOCA peak cladding temperature (PCT) analyses. Adequate DHR is demonstrated by showing that the ECCS flow being provided during the cold leg and hot leg recirculation phases of the event are adequate to replenish boil-off when accounting for additional margin and the boric acid dilution needs during hot leg recirculation.

The AOR calculations include a thermal power of 3459 MWt and a calorimetric uncertainty of 0.6%, explicitly modeling the 1.4% MUR power uprate conditions. The MUR power level with uncertainty is equivalent to the pre-MUR uprate rated power level and uncertainty of 3411 MWt and 2%, respectively; therefore, the post-LOCA AOR calculations are applicable to pre-MUR uprate and 1.4% MUR uprate plant configuration. The allocation of total core power and uncertainty is not critical for the post-LOCA LTCC analyses provided that the licensed core power and calorimetric uncertainty does not exceed the total analyzed core power of 3459 MWt x 1.006 (3479.75 MWt).

There have been no post-LOCA LTCC evaluations performed following the completion of the AOR that would affect the applicability of the AOR to both pre-MUR power uprate and 1.4%

MU power uprate conditions.

Because the analysis was performed at a power level equal to the MUR uprate power level and acceptable results were obtained, the analysis is unaffected by the MUR power uprate.

The AOR for this analysis is reflected in the WBN UFSAR and remains acceptable for the WBN Unit 2 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table II.1-1.

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Enclosure 2 The offsite dose analysis in UFSAR Section 15.5.3 is currently performed using a power level of 3565 MWt, which is bounding for the MUR power uprate. Upon implementation of WBN Unit 2 TPC, the Unit 2 LOCA AOR will be based on 3480 MWt, which is bounding for the MUR power uprate.

23. Major Rupture of a Main Steam Line (UFSAR Section 15.4.2.1)

The analysis documented in UFSAR Section 15.4.2.1 postulates a complete severance of a main steam line with the plant initially at hot shutdown conditions. One case considers full reactor coolant flow with offsite power available, and a second case considers a loss of offsite power with RCP coastdown. In addition, the most reactive RCCA is assumed to be stuck in its fully withdrawn position. In addition, for WBN Unit 2, a SG tube plugging level of 10% is assumed. The core is initially at 0% power and therefore unaffected by the MUR power uprate.

The AOR for this analysis is reflected in the WBN UFSAR and remains acceptable for the WBN Unit 2 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the reference listed in Table II.1-1.

The offsite dose analysis in UFSAR Section 15.5.4 was performed assuming primary-to-secondary leakage with reactor coolant source terms based on the two methods defined in Standard Review Plan (SRP) (NUREG-0800), Section 15.1.5, Appendix A. These are: 1) a pre-accident iodine spike and 2) an accident-initiated iodine spike. This source term is determined independent of reactor power. For both cases, the secondary side releases were determined using expected secondary side activities, based on ANSI/ANS-18.1-1984, and on 150 gallons per day (gpd) primary-to- secondary-side leakage through any one SG.

Upon implementation of the TPC for WBN Unit 2, the offsite dose analysis uses 3480 MWt to determine the secondary loop activity, which is bounding for the MUR power uprate.

24. Steam Line Break with Coincident Rod Withdrawal at Power (Not in UFSAR)

The licensing basis main steam line break event is described in UFSAR Section 15.4.2.1 and is assumed to occur at hot shutdown conditions. The LOCA analysis for WBN Unit 2, described in UFSAR Section 15.4.1.1, uses PAD4TCD for evaluating TCD. In Reference II.8, TVA performed an analysis of a steamline break at power with coincident rod withdrawal (SLB w/ RWAP) in response to an NRC request regarding an analysis of a main steam line break to confirm that the OPT reactor trip function provides adequate protection from unacceptable consequences during a steam line break event from hot full power conditions, including consideration of TCD The analysis determined that the peak fuel linear heat generation rate remained below a value that would cause fuel melting, and that the minimum DNBR remained above the applicable safety analysis limit value. Because the analysis was performed at a power level equal to the MUR uprated power level and acceptable results were obtained, the analysis is unaffected by the MUR power uprate.

The AOR for this analysis remains acceptable for the WBN Unit 2 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the reference listed in Table II.1-1.

25. Major Rupture of a Main Feedwater Pipe (UFSAR Section 15.4.2.2)

The analysis documented in UFSAR Section 15.4.2.2 postulates a major feedwater line rupture, which is defined as a break in a feedwater pipe large enough to prevent the addition of sufficient feedwater to maintain shell-side fluid inventory in the SGs. Depending upon the size of the break and the plant operating conditions at the time of the break, the break could CNL-19-082 E2-23 of 78

Enclosure 2 cause either a RCS cooldown (by excessive energy discharge through the break), or an RCS heatup. Potential RCS cooldown resulting from a secondary pipe rupture is evaluated in Section 15.4.2.1. Therefore, only the RCS heatup effects are evaluated for a feedline rupture. Two cases are analyzed. One case assumes that offsite electrical power is maintained throughout the transient. Another case assumes the loss of offsite electrical power at the time of reactor trip, and RCS flow decreases to natural circulation. Both cases assume a double-ended rupture of the largest feedwater pipe at full power.

Because the analysis was performed at a power level equal to the MUR uprated power level and acceptable results were obtained, the analysis is unaffected by the MUR power uprate.

The AOR for this analysis is reflected in the WBN UFSAR and remains acceptable for the WBN Unit 2 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the reference listed in Table II.1-1.

26. Steam Generator Tube Rupture (UFSAR 15.4.3)

The analysis documented in UFSAR Section 15.4.3 postulates the complete severance of a single SG tube. The accident is assumed to take place at full power with the reactor coolant contaminated with fission products corresponding to continuous operation with a limited amount of defective fuel rods. The accident leads to an increase in contamination of the secondary system due to leakage of radioactive coolant from the RCS. In the event of a coincident loss of offsite power, or failure of the condenser dump system, discharge of radioactivity to the atmosphere takes place via the SG power-operated relief valves (and safety valves if their setpoint is reached).

The impacts of the MUR power uprate on the SG tube rupture (SGTR) analyses (margin to overfill and input to dose) were evaluated in to determine if they remain bounding for the MUR power uprate. The power uprate impacts on plant power, hot and cold leg RCS temperatures, and SG pressure are conservatively addressed by the power uncertainty applied. The margin to overfill calculations model a nominal NSSS power level of 3475 MWt, which includes an RCP net heat input of 16 MWt with 0.6% power uncertainty. This power level accounts for a 1.4% MUR power uprate.

Because the analysis was performed at a power level equal to the MUR uprated power level and acceptable results were obtained, the analysis is unaffected by the MUR power uprate.

The AOR for this analysis is reflected in the WBN UFSAR and remains acceptable for the WBN Unit 2 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the reference listed in Table II.1-1.

The offsite dose analysis in UFSAR Section 15.5.4 analyzed two cases. Case 1: The primary side activity release uses the maximum TS limit design reactor coolant activities and an iodine spike immediately after the accident that increases the iodine activity in the reactor coolant. Case 2: The initial reactor coolant activity has a pre-accident iodine spike caused by an RCS transient. This source term is determined independent of reactor power. For both cases, the secondary side releases were determined using expected secondary side activities, based on ANSI/ANS-18.1-1984, and on a 150 gpd/SG primary-to-secondary-side leakage. Upon implementation of the TPC for WBN Unit 2, the analysis uses 3480 MWt to determine the secondary loop activity, which is bounding for the MUR power uprate.

27. Single Reactor Coolant Pump Locked Rotor (UFSAR Section 15.4.4)

The analysis documented in UFSAR Section 15.4.4 postulates an instantaneous seizure of an RCP rotor. Flow through the affected reactor coolant loop is rapidly reduced, leading to initiation of a reactor trip on a low flow signal.

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Enclosure 2 Because the analysis was performed at a power level equal to the MUR uprated power level and acceptable results were obtained, the analysis is unaffected by the MUR power uprate.

The AOR for this analysis is reflected in the WBN UFSAR and remains acceptable for the WBN Unit 2 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the reference listed in Table II.1-1.

28. Fuel Handling Accident (UFSAR Sections 15.4.5 and 15.5.6)

UFSAR Section 15.5.6 analyzes two fuel handling accident (FHA) cases. The first case is for an accident in the spent fuel pool area located in the Auxiliary Building. The second case considered is an accident inside containment where there is open communication between the containment and the Auxiliary Building. Both cases are evaluated using the alternate source term (AST) based on RG 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors. The assembly damaged is the highest-powered assembly in the core region to be discharged. The values for individual fission product inventories in the damaged assembly are calculated assuming full-power operation at the end of core life immediately preceding shutdown. The fuel assembly inventory is based on a core power level of 3565 MWt for the conventional core. Upon implementation of the TPC for WBN Unit 2, the FHA source terms are based on TPC assemblies, and these assembly inventories are determined based on a core thermal power level of 3480 MWt. Because the analysis was performed at a power level greater than the MUR uprated power level and acceptable results were obtained, the analysis is unaffected by the MUR power uprate.

The AOR for this analysis is reflected in the WBN UFSAR and remains acceptable for the WBN Unit 2 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the reference listed in Table II.1-1.

29. Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection) (UFSAR Section 15.4.6)

The analysis documented in UFSAR Section 15.4.6 postulates the mechanical failure of a control rod mechanism pressure housing resulting in the ejection of an RCCA and drive shaft. The consequence of this mechanical failure is a rapid positive reactivity insertion together with an adverse core power distribution, possibly leading to localized fuel rod damage.

In Reference II.8, TVA performed a reanalysis of the end-of-cycle-life rod ejection cases in response to an NRC request regarding an updated analysis of the hot spot rod ejection event for end-of-life conditions, for both the hot full power and hot zero power (HZP) cases, to show that the licensing basis acceptance criteria remain satisfied when accounting for the effects of TCD.

The results from the revised analysis demonstrate that the licensing basis acceptance criteria continue to be satisfied. Therefore, the effects of fuel TCD can be accommodated for the rod ejection accident for WBN Unit 2. Because the analysis was performed at a power level greater than the MUR uprated power level and acceptable results were obtained, the analysis is unaffected by the MUR power uprate.

The AOR for this analysis remains acceptable for the WBN Unit 2 MUR power uprate. The methodology by which the AOR was performed was reviewed and approved by the NRC per the references listed in Table II.1-1.

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Enclosure 2

30. Anticipated Transients Without Scram [ATWS]

As described in UFSAR Section 15.2, a series of generic studies on anticipated transients without scram (ATWS) showed acceptable consequences would result provided that the turbine trips and auxiliary feedwater flow is initiated in a timely manner. This is also documented in NS-TMA-2182 (Reference II.9). The effects of ATWS events are not considered as part of the design basis for transients analyzed in Chapter 15. The final NRC ATWS rule (10 CFR 50.62) (Reference II.10) required that Westinghouse-designed plants install ATWS mitigation system circuitry (AMSAC) to initiate a turbine trip and actuate auxiliary feedwater flow independent of the reactor protection system. The WBN AMSAC design is described in UFSAR Section 7.7.1.12 and conforms to WCAP-10858-P-A, AMSAC Generic Design Packages.

The generic ATWS analyses were performed for the various American Nuclear Society (ANS) Condition II events (i.e., anticipated transients) considering various Westinghouse PWR configurations applicable at that time. These analyses included two, three, and four-Loop PWRs with various SG models. The generic ATWS analyses documented in Reference II.9 also support the analytical basis for the NRC-approved generic AMSAC designs as documented in WCAP-10858-P-A, Revision 1. The generic ATWS analyses applicable to WBN Unit 2 are provided for a four-loop PWR with Model D SGs modeling an NSSS power of 3427 MWt (3411 MWt core power). These conditions are summarized in Table 15.3.6-1 of Reference II.11.

The results of these generic analyses indicate that the peak RCS pressure, predicted for ATWS events in Westinghouse four-loop plants equipped with Model D SGs, is 2,780 psia, which is below the ASME Service Level C limit of the weakest component in the RCS (3,200 psig). The analysis in NS-TMA-2182 also determined that the peak RCS pressure with the 2% increase in power remains below 3200 psig. This ATWS sensitivity analysis was performed assuming a 2% variation in power, consistent with the typical calorimetric measurement uncertainty on power at the time of these analyses. The proposed increase in power of 1.4% for WBN Unit 2 is within the applicable range of the 2% increase in power assumed in the sensitivity analysis.

In Reference II.11, the NRC reviewed the WBN Unit 2 ATWS design including the analytical basis and concluded that the AMSAC design and ATWS analysis are acceptable.

Therefore, operation of WBN Unit 2 MUR power uprate remains within the bounds of the generic Westinghouse ATWS analysis documented in NS-TMA-2182 and will remain in compliance with 10 CFR 50.62(c).

31. Containment Performance Analyses The WBN Unit 2 short and long-term LOCA peak containment pressure analysis is documented in UFSAR Section 6.2.1.3.1 through 6.2.1.3.9. Main steam line break (MSLB) peak containment temperature analysis is documented in UFSAR Section 6.2.1.3.10. The maximum reverse pressure differential analysis is documented in UFSAR Section 6.1.2.3.11. These analyses are performed to demonstrate peak containment pressures and temperatures are acceptable and to ensure the pressure and temperature profiles assumed in the environmental qualification (EQ) analyses are acceptable.

The containment performance analyses were reviewed for impact from MUR power uprate.

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Enclosure 2 Long-Term LOCA/Containment Integrity Analysis This analysis demonstrates the ability of the containment safeguard systems to mitigate the consequences of a large break LOCA. The methodology for the licensing basis analysis is contained in WCAP-17721-P-A (Reference II.13). Based on this methodology, the AOR presently assumes a core thermal power of 3479.75 MWt, which is approximately 2% greater than the current licensed core power of 3411 MWt.

The power measurement margin is but one of many conservative assumptions used in the analysis. For example, as noted above, a key assumption is the assumed initial power level in the current AOR of 3479.75 MWt, which is 0.6% higher than the proposed 1.4%

uprated thermal power of 3459 MWt. Taken together, the improved power measurement uncertainty and conservative assumptions provide substantial conservatism such that the margin of safety would not be reduced.

Furthermore, because the LOCA mass and energy (M&E) release are not impacted and will not change, there will be no effect on the long-term post LOCA containment response and containment integrity analysis for the current licensing basis analysis and the conclusions discussed in UFSAR Section 6.2.1.3.

Short-Term LOCA Mass and Energy Release Analysis Several evaluations are performed to support the loop subcompartment, reactor cavity and pressurizer enclosure analysis. The analysis inputs that may potentially change with the uprate are the initial RCS fluid temperatures. Because this event lasts for approximately three seconds, the single effect of power is not significant.

The short-term blowdown transients are characterized by a peak M&E release rate that occurs during a subcooled condition. The Zaloudek correlation, which models this condition, is currently used in the short-term LOCA M&E release analyses. This correlation is used to conservatively evaluate the impact of the changes in the RCS inlet and outlet temperatures from the MUR power uprate relative to those used in the current AOR. The use of lower RCS temperatures maximizes the critical mass flux in the Zaloudek correlation.

The analyses of subcompartment pressurization utilize the Transient Mass Distribution (TMD) computer code. TMD was developed by Westinghouse to calculate the short-term pressure response in ice condenser subcompartments due to a LOCA. Reference II.14 is the topical report for the use of TMD, which was approved by the NRC in Reference II.15.

The WBN Units 1 and 2 UFSAR analyses of the containment subcompartment response are supported by these methods.

Loop Subcompartment Analysis The loop subcompartment analysis is performed to ensure that the walls of the loop subcompartments, including the lower crane wall, upper crane wall, operating deck, and the containment shell, can maintain their structural integrity during the short pressure pulse (generally less than three seconds) which accompanies a LOCA. Additionally, this analysis verifies the adequacy of the ice condenser performance.

A vessel outlet temperature of 617.1°F and a vessel/core inlet temperature of 555.2°F, both conservatively bounded low for short-term M&E release considerations, were evaluated for WBN Unit 2 with the OSGs and a 15% ice blockage. The RCS temperature values used in in the analysis bound the 1.4% uprate values of 619.1°F and 557.3°F, respectively. Therefore, the current licensing basis M&E releases and loop subcompartment analyses remain bounding for the 1.4% MUR power uprate.

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Enclosure 2 Reactor Cavity Analysis The reactor cavity analysis is performed to ensure that the walls in immediate proximity of the RV can maintain their structural integrity during the short pressure pulse, which accompanies a LOCA within the reactor cavity region. Loadings on the RV are also determined.

The magnitude of the pressure pulse is based on the mass release rate from the break.

The Zaloudek correlation was used to conservatively predict that the MUR power uprate increases the mass flux by 3.56%.

The 127 square-inch (in2) RV inlet break is used in the existing sub-compartment calculation. However, based upon results of the structural analysis of the RCS, a better estimate of the break size is 45 in2 (Reference II.16). The reduced rates from this reduced break size offset the predicted increase in break flow rate due to the 1.4% MUR power uprate. For example, the releases are approximately proportional to the break size, and as such, the releases would be reduced by a factor of (127/45 = 2.8). This reduction in break size would more than offset the 3.56% increase in mass flux. Therefore, the current licensing basis mass and energy releases and reactor cavity subcompartment analyses remain bounding for the 1.4% MUR power uprate.

Pressurizer Enclosure Analysis The pressurizer enclosure analysis is performed to ensure that the walls in the immediate proximity of the pressurizer enclosure can maintain their structural integrity. Loadings acting across the pressurizer are also determined.

The limiting pipe break is a severance in the pressurizer spray line. Comparing the pipe size assumed in the current analysis versus the as-built piping, the margin in the releases just due to the currently assumed break size is greater than 25%. The break sizes used in the current analysis are 0.1963 square feet (ft2) for the cold leg spray nozzle and 0.08727 ft2 for the pressurizer spray nozzle. The as-built break sizes are 0.0645 ft2 for the cold leg spray nozzle and 0.06154 ft2 for the pressurizer spray nozzle. The difference in break sizes leads to greater than 25% margin in the M&E releases. This more than offsets the predicted increase in M&E releases due to the 1.4% MUR power uprate of approximately 3.6%. Therefore, the current M&E releases and the current pressurizer enclosure pressure analysis remain bounding for the MUR power uprate.

Maximum Reverse Pressure Differential Analysis Following a LOCA, the pressure and temperature in the lower compartment of containment increases, which forces the air in the lower compartment into the upper compartment and increases the pressure in the upper compartment. As the temperature in the lower compartment decreases with time, the pressure in the lower compartment also decreases. Eventually the pressure in the lower compartment becomes less than the pressure in the upper compartment, which creates a reverse differential pressure across the operating deck. This analysis is used to predict this reverse differential pressure and to ensure the structural adequacy of the operating deck.

The AOR is a generic and conservative analysis discussed in UFSAR Section 6.2.1.3.11.

The dead- ended compartments adjacent to the lower compartment are assumed to be swept of air during the initial blowdown. This is a very conservative assumption because this will maximize the air forced into the upper ice bed and upper compartment thus raising the compression pressure for the operating deck. In addition, it will minimize the non-condensables in the lower compartment.

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Enclosure 2 The M&E releases utilized serve only as a vehicle to initiate the event and to purge the lower and the dead-ended compartment air. Any increases in releases during the post-blowdown period would result in the lower compartment pressure remaining at a higher value, and thus would reduce the reverse differential pressure. The M&E releases are extracted from a model used to maximize the LOCA PCT and not from a model used to maximize the peak containment pressure. It is judged that the RCS temperature changes and the resulting effects would not affect the results of the maximum reverse pressure differential calculation.

The AOR shows that significant margin exists between a calculated reverse differential pressure of 0.65 psi and the design reverse differential pressure value of 6.8 psi across the operating deck and 8.6 psi across the lower inlet doors. Thus, the 1.4% power uprate will have a minimal impact, if any, on the analysis and there is significant analysis margin available. Therefore, the current AOR remains bounding for the 1.4% MUR power uprate.

Steamline Break Mass and Energy Releases Evaluation The safety analyses related to the SLB M&E releases were evaluated to determine the effect of a power uprate of 1.4% for WBN Unit 2 with the existing WBN 2 OSGs. The evaluation determined that the NSSS design parameters for WBN Unit 2 remain unchanged or are bounded by the current safety analysis values.

Long-Term Steamline Break Mass and Energy Releases Inside Containment or Outside of Containment (Equipment Qualification Input)

The critical parameters for the long-term SLB event inside or outside of containment include the following conditions on the primary and secondary sides: NSSS power level, reactivity feedback characteristics including the minimum plant shutdown margin, the initial and trip values for the SG water mass, main feedwater flow, auxiliary feedwater (AFW) flow, main and AFW enthalpies, and the times at which steamline and feedwater line isolation occur.

The input assumptions related to these critical parameters dictate the quantity and rate of the M&E releases.

The AOR applicable for the inside containment and outside containment long-term SLBs assume the 1.4% MUR power uprate on the current licensed NSSS power of 3427 MWt.

The long-term SLB M&E releases have been analyzed at the uprated NSSS power of 3475 MWt. There is no effect on either the current licensing-basis long-term SLB M&E release analysis inside containment or the conclusions related to the containment pressure/temperature response following the SLB. There is also no effect on the current licensing basis long-term SLB M&E release analysis outside containment. Based on this, it is expected that the UFSAR conclusions related to the main steam valve vault (MSVV) temperature response outside containment following the SLB remain valid.

Short-Term Steamline Break Mass and Energy Releases Inside Containment or Outside of Containment:

The critical parameters for the short-term SLB event inside or outside of containment include the following conditions on the primary and secondary sides: no-load secondary system pressure, cross-sectional flow area of the steam piping at the break location, and total mass of steam in the piping of the unfaulted lines and the header. The input assumptions related to these critical parameters dictate the quantity and rate of the M&E releases within the SG enclosure inside containment and into the MSVV outside of containment.

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Enclosure 2 The AOR applicable to the short-term SLBs within the SG enclosure and the AOR applicable for the short- term SLBs into the MSVV assume a no-load initial power. This maximizes the initial steam pressure, which is the forcing function that defines the rate at which the steam is released from the postulated break location. The no-load RCS temperature of 557°F is unchanged due to the 1.4% MUR power uprate to the NSSS power of 3475 MWt. There is no effect on either the current licensing basis short-term SLB M&E release analysis within the SG enclosure or the conclusions related to the differential pressure response in the SG enclosure. There is also no effect on the current licensing basis short-term SLB M&E release analysis into the MSVV. Based on this, the WBN Unit 2 analysis conclusions related to the MSVV pressure response following the SLB remain valid.

32. EQ parameters The review of WBN EQ Program documentation included review of both TVA EQ program-level documentation and EQ calculations and environmental data drawings for specific components installed at WBN Unit 2. This review was conducted to focus on the EQ parameters of temperature, pressure, and radiation, with respect to any potential parameter changes due to the MUR power uprate. The evaluation determined that no programmatic changes to the EQ Program are required because of the WBN Unit MUR power uprate.

The existing WBN Unit 2 LEFM CheckPlus System is not within the scope of the WBN EQ program because the LEFM system components and associated cabling are located in mild environments in WBN Unit 2.

Temperature and Pressure Temperature and pressure were evaluated as part of the engineering evaluations for the MUR power uprate. The potential changes in ambient temperatures, system temperatures, system pressures, and potential accident external pressures (e.g., high-energy line break (HELB)) and accident temperatures were considered during the review.

The potential impact of the MUR power uprate on ambient plant temperatures was addressed via the HVAC review for the Reactor Building, Auxiliary Building, Diesel Generator Building, and Control Building (see Section VI.1.F). Additionally, the evaluation for MUR power uprate conditions shows that there is no increase to the overall heat load for containment. Therefore, the temperatures used for EQ analysis of containment components are unchanged with respect to WBN Unit 2 MUR power uprate. The evaluation also shows that the MUR power uprate will not affect the HVAC in the Auxiliary Building. Therefore, the temperatures used for EQ analysis of Auxiliary Building components at WBN Unit 2 are unchanged. The potential impact of the MUR power uprate on system temperature changes were evaluated as part of the individual MUR power uprate engineering system reviews.

To summarize the evaluation of the temperature and pressure review (due to the MUR power uprate), the BOP systems were determined to show some slight parameter changes, but these minor changes were shown to have no impact on the components within the scope of the EQ program at WBN Unit 2. The evaluation of the systems inside containment and in the MSVV for accident temperature and pressure conditions showed that the current design basis analyses were performed at 102% of 3411 MWt (i.e., 3479 MWt), which bounds the MUR power uprate. There is no EQ impact with respect to temperature or pressure due to the MUR power uprate. No areas transition from mild to harsh environments because of the MUR power uprate based on temperatures.

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Enclosure 2 Radiation The WBN EQ Program establishes normal and post-accident radiation doses to areas of the plant for harsh and mild environmental conditions. The typical bounding accident for EQ purposes is the LOCA. The primary source of LOCA dose to areas in the reactor building references ORNL/M-3739 for the source term. The ORNL/M-3739 source term modeling is based on a core thermal power of 3565 MWt, which is higher than the MUR power uprate thermal power of 3459 MWt. Therefore, the EQ analyses that use this source term are bounded for the WBN Unit 2 MUR power uprate. Not all plant building areas are analyzed based on ORNL/M-3739 and not all plant building areas consider a LOCA as the bounding accident. Thus, additional analyses of normal and accident doses are used as the bases for the EQ program.

The radiological calculations referenced as input to the EQ analyses were reviewed to determine if there were any radiological impacts associated with the MUR power uprate that could affect EQ parameters or other radiological dose analyses. Based on review of these analyses, it was determined that the radiological analyses either bound the WBN Unit 2 MUR power uprate conditions or the source term for the reference analysis will not be affected by MUR power uprate. Therefore, the existing radiological analyses remain applicable for the WBN Unit 2 MUR power uprate.

33. Flooding The plant grade elevation at WBN can be exceeded by large rainfall and seismically induced dam failure floods. The plant design features and procedures used to provide assurance that WBN can be safely shut down and maintained in these extreme flood conditions is discussed in WBN UFSAR Sections 2.4.14, 3.4, 3.8.1, and 3.8.4. The MUR power uprate will not affect the external flooding sources or the protective structural design features.

Internal flooding of safety-related structures from HELBs and moderate-energy line breaks (MELBs) is addressed in UFSAR Section 3.6A.2.1.4. Internal flooding of the Turbine Building because of a condenser circulating water (CCW) system failure is addressed in UFSAR Section 10.4.5.

The high-energy systems that represent internal flood sources were reviewed for impacts of MUR power uprate. With the exception of the main feedwater system, none of the high-energy system blowdown rates assumed in the HELB analysis will increase because of increased reactor operating power. For the evaluation of flooding in the MSSV due to a feedwater line break, it was identified that that there will be a change in the assumed break flow assumed due to changes in feedwater conditions for MUR power uprate. However, the calculated flood levels have approximately 50% margin to the maximum level. Therefore, the analysis will remain valid for uprate conditions.

The moderate energy systems that represent internal flood sources were reviewed for impacts of MUR power uprate. None of the moderate energy systems will experience an increase in maximum operating pressure because of increased reactor operating power.

The bases of the existing conditions analyzed in the flooding calculation will remain valid for projected MUR power uprate operating conditions. Therefore, MELB release rates are not impacted by MUR power uprate and the analysis will remain valid.

With regard to Turbine Building flooding analysis, the CCW system will not experience a change in operating pressure because of MUR power uprate. Therefore, flooding from a CCW rupture will not increase and the analysis will remain valid.

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Enclosure 2

34. Safe Shutdown Fire Refer to Section VII.6.A for discussion of MUR power uprate on the fire protection program.

NRC acceptance of the WBN Unit 2 fire protection program and safe shutdown analysis is documented in Reference II.17.

35. Spent Fuel Pool Accidents (loss of pool cooling)

The impact of the MUR power uprate on the Spent Fuel Pool Cooling and Cleanup System (SFPCS) is discussed in Section VI.1.D. In the event of a loss of forced cooling, the large volume of water in the spent fuel pool would take several hours to heat up. Prior to a full core discharge, the spent fuel pool heat load is determined by calculation using cycle-specific heat loads. Offload requirements are procedurally established to assure that the decay heat load in the pool is less than the maximum allowable heat load. As discussed in UFSAR Section 9.1.3.3.1, in the event that cooling capability were lost for an extended period, the pool water temperature would approach boiling. A seismically qualified line is available from the common discharge of the refueling water purification pumps to the spent fuel pool cooling loop. The piping, valves, and pumps from the RWST to the common discharge of the refueling water purification pumps are seismically qualified. Other sources for makeup available are the demineralized water system and the fire protection system. By limiting the spent fuel pool heat load to less than the maximum allowable heat load, the current licensing basis is maintained.

36. Spent Fuel Pool Criticality The MUR power uprate does not affect the spent fuel pool criticality analysis. The analysis of record for the criticality safety analysis of the WBN spent fuel pool, HI-2177876 (proprietary) (and its non-proprietary version, HI-2177950), were included in the LAR to adopt a TPC for WBN Unit 2 (Reference II.3), which was approved by the NRC in Reference II.2). The analyses consider fresh fuel with a uniform enrichment (up to 5 wt%

U-235). The same bounding enrichment is considered along the entire active length for each fuel pin. A bounding fuel density is considered. Therefore, the criticality analysis is not power level dependent, is not affected by MUR power uprate conditions, and the analysis of record remains bounding.

37. Tritium Production Accident Releases References II.2 and II.3 address the accident scenarios associated with TPC production for WBN Unit 2. WBN Unit 2 TPC operation is planned to commence concurrently with WBN Unit 2 MUR power uprate. WBN Unit 2 has already been evaluated and approved to operate as a TPC for conditions that bound or are equivalent to MUR power uprate conditions. The analyzed TPC tritium inventory is not affected by the MUR power uprate.

References for Section II:

II.1. NRC letter to TVA, Watts Bar Nuclear Plant, Unit 2 - Issuance of Amendment Regarding Application to Revise License Condition 2.C.(4) PAD4TCD (EPID L-2018-LLA-0051), dated March 20, 2019 (ML19046A286)

II.2. NRC letter to TVA, Watts Bar Nuclear Plant, Units 1 and 2 - Issuance of Amendment Regarding Revision to Watts Bar Nuclear Plant, Unit 2, Technical Specification 4.2.1, Fuel Assemblies, and Watts Bar Nuclear Plant, Units 1 and 2, Technical Specifications Related to Fuel Storage (EPID L-2017-LLA-0427), dated May 22, 2019 (ML18347B330)

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Enclosure 2 II.3. TVA letter to NRC, CNL-17-144, Application to Revise Watts Bar Unit 2 Technical Specification 4.2.1, Fuel Assemblies, and Watts Bar Units 1 and 2 Technical Specifications Related to Fuel Storage (WBN-TS-17-028), dated December 20, 2017 (ML17354B282)

II.4. Westinghouse Report, WCAP-10079-P-A, Revision 0, NOTRUMP - A Nodal Transient Small Break and General Network Code, August 1985 II.5. Westinghouse Report, WCAP-10054-P-A, Revision 0, Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code, August 1985 II.6. Westinghouse Report, WCAP-10054-P-A, Addendum 2, Revision 1, Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model, July 1997 II.7. Westinghouse Report, WCAP-9272-P-A, Revision 0, Westinghouse Reload Safety Evaluation Methodology, July 1985 II.8. TVA letter to NRC, CNL-18-116, Response to Request for Additional Information Regarding the Application to Revise Watts Bar Nuclear Plant Unit 2 - License Condition 2.C(4) PAD4TCD (391-WBN-TS-18-03) (EPID L-2018-LLA-0051), dated October 11, 2018 (ML18284A450)

II.9. Westinghouse letter to NRC, NS-TMA-2182, ATWS Submittal, dated December 30, 1979 (ML041130109)

II.10. ATWS Final Rule - Code of Federal Regulations 10 CFR 50.62 and Supplementary Information Package, Reduction of Risk from Anticipated Transients Without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants II.11. NUREG-0847, Supplement 24, Safety Evaluation Report Related to the Operation of Watts Bar Nuclear Plant, Unit 2, September 2011 II.12. Westinghouse Reports, WCAP-16009-P-A, Revision 0 (Proprietary) and WCAP-16009-NP-A, Revision 0 (Non-Proprietary), Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM), January 2005 II.13. Westinghouse Report, WCAP-17721-P-A, Revision 0, Westinghouse Containment Analysis Methodology - PWR LOCA Mass and Energy Release Calculation Methodology, September 2015 II.14. Westinghouse Report, WCAP-8077, Revision 0, Ice Condenser Containment Pressure Transient Analysis Methods, August 2005 II.15. Letter from D. B. Vassallo, Chief, Light Water Reactors Project Branch 1-1, Directorate of Licensing - Regulation, United States Atomic Energy Commission, subject: Review of TMD, December 1973 II.16. Westinghouse Report, WCAP-8889, Revision 0, Dynamic Analysis of Reactor Pressure Vessel for Postulated Loss-of-Coolant Accidents: Watts Bar Nuclear Power Station Unit Numbers 1 and 2, December 1976 II.17. NUREG-0847, Supplement 29, Safety Evaluation Report Related to the Operation of Watts Bar Nuclear Plant, Unit 2, October 2015 II.18. Westinghouse Report, WCAP-17093-P, Revision 1 (Proprietary), Best-Estimate Analysis of the Large-Break Loss-of-Coolant Accident for Watts Bar Unit 2 Nuclear Power Plant Using the ASTRUM Methodology, June 2013 CNL-19-082 E2-33 of 78

Enclosure 2 Table II.1-1: WBN Unit 2 Analyses Approved by NRC or Is Power Confirm that bounding UFSAR Power Used in conducted using Reference for NRC Analysis Title Bounding event determinations Section this Analysis methods/processes approval for MUR? remain valid approved by the NRC RIS 2002-03: II.1.A II.1.B.i II.1.B.i II.1.C II.1.B.ii II.1.D (1) Uncontrolled Rod Hot zero power. N/A See discussion in See discussion in OL and SER 15.2.1 Cluster Control 3411 MWt (critical @ Section II.1.D. Section II.1.D.

Assembly Bank 0.0 fraction of Withdrawal from a nominal [FON])

Subcritical Condition (2) Uncontrolled Rod 3475 MWt Yes See discussion in See discussion in OL and SER 15.2.2 Cluster Control (101.4% of 3411 MW Section II.1.D. Section II.1.D.

Assembly Bank plus 16 MW NSSS Withdrawal at Power pump heat)

(3) Rod Cluster Control 3475 MWt Yes See discussion in See discussion in OL and SER 15.2.3 Assembly (101.4% of 3411 MW Section II.1.D. Section II.1.D.

Misalignment plus 16 MW NSSS pump heat)

(4) Uncontrolled Boron 0 and 3475 MWt Yes See discussion in See discussion in OL and SER 15.2.4 Dilution (101.4% of 3411 MW Section II.1.D. Section II.1.D.

plus 16 MW NSSS pump heat)

(5) Partial Loss of 3475 MWt Yes See discussion in See discussion in OL and SER 15.2.5 Forced Reactor (101.4% of 3411 MW Section II.1.D. Section II.1.D.

Coolant Flow plus 16 MW NSSS pump heat)

(6) Startup of an N/A N/A See discussion in N/A OL and SER 15.2.6 Inactive Reactor Section II.1.D.

Coolant Loop at an Incorrect Temperature (7) Loss of External 3475 MWt Yes See discussion in See discussion in OL and SER 15.2.7 Electrical Load (101.4% of 3411 MW Section II.1.D. Section II.1.D.

and/or Turbine Trip plus 16 MW NSSS pump heat)

(8) Loss of Normal 3479 MWt Yes See discussion in See discussion in OL and SER 15.2.8 Feedwater (102% of 3411 MWt) Section II.1.D. Section II.1.D.

with an additional 16 MW NSSS pump heat if offsite power available CNL-19-082 E2-34 of 78

Enclosure 2 Table II.1-1: WBN Unit 2 Analyses Approved by NRC or Is Power Confirm that bounding UFSAR Power Used in conducted using Reference for NRC Analysis Title Bounding event determinations Section this Analysis methods/processes approval for MUR? remain valid approved by the NRC RIS 2002-03: II.1.A II.1.B.i II.1.B.i II.1.C II.1.B.ii II.1.D (9) Coincident Loss of 3475 MWt Yes See discussion in See discussion in OL, SER, and 15.2.9 Onsite and External (101.4% of 3411 MW Section II.1.D. Section II.1.D. Reference II.2 (Offsite) AC Power plus 16 MW NSSS to the Station - Loss pump heat) of Offsite Power to the Station Dose: 3565 MWt Auxiliaries (3480 MWt used for expected secondary side activities following TPC implementation)

(10) Excessive Heat 3475 MWt Yes See discussion in See discussion in OL and SER 15.2.10 Removal Due to (101.4% of 3411 MW Section II.1.D. Section II.1.D.

Feedwater System plus 16 MW NSSS Malfunctions pump heat)

(11) Excessive Load 3475 MWt Yes See discussion in See discussion in OL and SER 15.2.11 Increase Incident (101.4% of 3411 MW Section II.1.D. Section II.1.D.

plus 16 MW NSSS pump heat)

(12) Accidental 3475 MWt Yes See discussion in See discussion in OL and SER 15.2.12 Depressurization of (101.4% of 3411 MW Section II.1.D. Section II.1.D.

the Reactor Coolant plus 16 MW NSSS System pump heat)

(13) Accidental 3475 MWt Yes See discussion in See discussion in OL and SER 15.2.13 Depressurization of (101.4% of 3411 MW Section II.1.D. Section II.1.D.

the Main Steam plus 16 MW NSSS System pump heat)

(14) Inadvertent 3475 MWt Yes See discussion in See discussion in OL and SER 15.2.14 Operation of (101.4% of 3411 MW Section II.1.D. Section II.1.D.

Emergency Core plus 16 MW NSSS Cooling System pump heat)

During Power Operation (15) Chemical and 3475 MWt Yes See discussion in See discussion in OL and SER 15.2.15 Volume Control (101.4% of 3411 MW Section II.1.D. Section II.1.D.

System Malfunction plus 16 MW NSSS During Power pump heat)

Operation CNL-19-082 E2-35 of 78

Enclosure 2 Table II.1-1: WBN Unit 2 Analyses Approved by NRC or Is Power Confirm that bounding UFSAR Power Used in conducted using Reference for NRC Analysis Title Bounding event determinations Section this Analysis methods/processes approval for MUR? remain valid approved by the NRC RIS 2002-03: II.1.A II.1.B.i II.1.B.i II.1.C II.1.B.ii II.1.D (16) Loss of Reactor 3480 MWt Yes See discussion in See discussion in OL, SER, 15.3.1 Coolant from Small Section II.1.D. Section II.1.D. References II.4, II.5, and Ruptured Pipes or II.6 from Cracks in Large Pipes Which Actuate the Emergency Core Cooling System (17) Minor Secondary N/A N/A See discussion in N/A OL and SER 15.3.2 System Pipe Breaks Section II.1.D.

(18) Inadvertent Loading 3425 MWt Yes See discussion in See discussion in OL and SER 15.3.3 of a Fuel Assembly Section II.1.D. Section II.1.D.

into an Improper Position (19) Complete Loss of 3475 MWt Yes See discussion in See discussion in OL and SER 15.3.4 Forced Reactor (101.4% of 3411 Section II.1.D. Section II.1.D.

Coolant Flow plus 16 MW NSSS pump heat)

(20) Waste Gas Decay 3565 MWt (RG 1.24) Yes See discussion in See discussion in OL, SER, and 15.3.5 Tank Rupture 3582 MWt (Realistic) Section II.1.D. Section II.1.D. Reference II.2 (Prior to TPC) 3480 MWt (Realistic)

(Following TPC implementation)

(21) Single Rod Cluster 3475 MWt Yes See discussion in See discussion in OL and SER 15.3.6 Control Assembly (101.4% of 3411 MW Section II.1.D. Section II.1.D.

Withdrawal at Full plus 16 MW NSSS Power pump heat)

(22) Major Reactor 3479.8 MWt Yes See discussion in See discussion in Section OL, SER, 15.4.1 Coolant System Pipe Section II.1.D. II.1.D. References II.2, II.12, Ruptures (Loss of and II.18 Coolant Accident) Dose: 3565 MWt (Prior to TPC) 3480 MWt (Following TPC implementation)

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Enclosure 2 Table II.1-1: WBN Unit 2 Analyses Approved by NRC or Is Power Confirm that bounding UFSAR Power Used in conducted using Reference for NRC Analysis Title Bounding event determinations Section this Analysis methods/processes approval for MUR? remain valid approved by the NRC RIS 2002-03: II.1.A II.1.B.i II.1.B.i II.1.C II.1.B.ii II.1.D (23) Major Rupture of a Hot shutdown Yes See discussion in See discussion in Section OL, SER, and 15.4.2.1 Main Steam Line conditions, Section II.1.D. II.1.D. Reference II.2 3425 MWt (critical

@ 0.0 fraction of nominal [FON])

Dose: 3582 MWt (Prior to TPC) 3480 MWt for expected secondary side activities (Following TPC implementation)

(24) Steam Line Break 3475 MWt Yes See discussion in See discussion in Reference II.1 Not in UFSAR with Coincident Rod (101.4% of 3411 Section II.1.D. Section II.1.D.

Withdrawal at Power plus 16 MW NSSS pump heat)

(25) Major Rupture of a 3475 MWt Yes See discussion in See discussion in OL and SER 15.4.2.2 Main Feedwater (101.4% of 3411 Section II.1.D. Section II.1.D.

Pipe plus 16 MW NSSS pump heat)

(26) Steam Generator 3475 MWt Yes See discussion in See discussion in OL, SER, and 15.4.3 Tube Rupture (101.4% of 3411 MW Section II.1.D. Section II.1.D. Reference II.2 plus 16 MW NSSS pump heat)

Dose: 3582 MWt (Prior to TPC) 3480 MWt for expected secondary side activities (Following TPC implementation)

(27) Single Reactor 3475 MWt Yes See discussion in See discussion in OL and SER 15.4.4 Coolant Pump (101.4% of 3411 MW Section II.1.D. Section II.1.D.

Locked Rotor plus 16 MW NSSS pump heat)

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Enclosure 2 Table II.1-1: WBN Unit 2 Analyses Approved by NRC or Is Power Confirm that bounding UFSAR Power Used in conducted using Reference for NRC Analysis Title Bounding event determinations Section this Analysis methods/processes approval for MUR? remain valid approved by the NRC RIS 2002-03: II.1.A II.1.B.i II.1.B.i II.1.C II.1.B.ii II.1.D (28) Fuel Handling 3565 MWt (Dose) Yes See discussion in See discussion in OL, SER, and 15.4.5 Accident (Prior to TPC) Section II.1.D. Section II.1.D. Reference II.2 3480 MWt (Dose)

(Following TPC implementation)

(29) Rupture of a Control Hot zero power Yes See discussion in See discussion in OL, SER, and 15.4.6 Rod Drive and Section II.1.D. Section II.1.D. Reference II.1 Mechanism Housing 3475 MWt (101.4%

(Rod Cluster Control of 3411 MWt plus Assembly Ejection) 16 MW NSSS pump heat)

(30) Anticipated 3479 MWt Yes See discussion in See discussion in OL and SER Not in UFSAR Transients Without (102% of 3411) Section II.1.D. Section II.1.D. (Specifically Scram [ATWS] Reference II.11)

(31) Containment LOCA Long Term Yes See discussion in See discussion in See discussion in 6.2.1.3 Performance Containment Section II.1.D. Section II.1.D. Section II.1.D.

Analyses integrity analysis:

3479.75 MWt Steamline break analysis: 3475 MWt (32) EQ Parameters Ranges from Yes See discussion in See discussion in See discussion in 3480 - 4100 MWt Section II.1.D. Section II.1.D. Section II.1.D.

(33) Flooding N/A N/A See discussion in See discussion in See discussion in Section II.1.D. Section II.1.D. Section II.1.D.

(34) Safe Shutdown Fire 3475 MWt (101.4% Yes See discussion in See discussion in SER (Specifically of 3411 MWt plus Sections II.1.D and Sections II.1.D.and VII.6.A. Reference II.17) 16 MW NSSS pump VII.6.A.

heat)

(35) Spent Fuel Pool Decay Heat Yes See discussion in See discussion in See discussion in Accidents (loss of Section II.1.D. Section II.1.D. Section II.1.D.

pool cooling)

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Enclosure 2 Table II.1-1: WBN Unit 2 Analyses Approved by NRC or Is Power Confirm that bounding UFSAR Power Used in conducted using Reference for NRC Analysis Title Bounding event determinations Section this Analysis methods/processes approval for MUR? remain valid approved by the NRC RIS 2002-03: II.1.A II.1.B.i II.1.B.i II.1.C II.1.B.ii II.1.D (36) Spent Fuel Pool N/A - The criticality N/A See discussion in See discussion in Reference II.2 Criticality calculations assume Section II.1.D. Section II.1.D.

storage racks are uniformly loaded with fresh fuel assemblies with an initial enrichment up to 5 weight percent U-235.

(37) Tritium Production N/A - TPC tritium N/A - TPC See discussion in See discussion in Reference II.2 15.5.8 Accident Releases inventory (2.6E7 Ci) tritium Section II.1.D. Section II.1.D.

(Dual unit based on 1.2 grams inventory is applicability with of tritium and a based on a to WBN Unit 2 bounding maximum TPC 2304 TPBARs. design implementation) tritium inventory of 1.2 grams tritium per rod and the number of TPBARs in the core.

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Enclosure 2 III ACCIDENTS AND TRANSIENTS FOR WHICH THE EXISTING ANALYSES OF RECORD DO NOT BOUND PLANT OPERATION AT THE PROPOSED UPRATED POWER LEVEL III.1 This section covers the transient and accident analyses that are included in the plants UFSAR (typically Chapter 14 or 15) and other analyses that are required to be performed by licensees to support licensing of their plants (i.e., radiological consequences, natural circulation cooldown, containment performance, anticipated transient without scrams, station blackout, analyses for determination of environmental qualification parameters, safe shutdown fire analysis, spent fuel pool cooling, flooding).

RESPONSE

The WBN Unit 2 analyses of record for the UFSAR Chapter 15 analyses support the MUR power uprate as described in Section II.

III.2 For analyses that are covered by the NRC approved reload methodology for the plant, the licensee should:

III.2.A Identify the transient/accident that is the subject of the analysis III.2.B Provide an explicit commitment to re-analyze the transient/accident, consistent with the reload methodology, prior to implementation of the power uprate III.2.C Provide an explicit commitment to submit the analysis for NRC review, prior to operation at the uprated power level, if NRC review is deemed necessary by the criteria in 10 CFR 50.59 III.2.D Provide a reference to the NRCs approval of the plants reload methodology

RESPONSE

WBN Unit 2 has no reload analyses that require re-evaluation for the MUR power uprate.

Various reload analyses are performed for each fuel cycle in accordance with normal cycle design practice and included in the COLR (WBN Unit 2 TS 5.9.5), but there will be no change to those analyses or their methodology based on the MUR power uprate. The proposed change to TS 5.9.5b in Enclosure 1, Section 2.2, addresses the change with respect to the assumed allowable uncertainty (i.e., 100.6% RTP versus 102% RTP) to support the MUR power uprate.

III.3 For analyses that are not covered by the reload methodology for the plant, the licensee should provide a detailed discussion for each analysis. The discussion should:

III.3.A Identify the transient or accident that is the subject of the analysis III.3.B Identify the important analysis inputs and assumptions (including their values),

and explicitly identify those that changed as a result of the power uprate III.3.C Confirm that the limiting event determination is still valid for the transient or accident being analyzed III.3.D Identify the methodologies used to perform the analyses, and describe any changes in those methodologies III.3.E Provide references to staff approvals of the methodologies in Item D. above III.3.F Confirm that the analyses were performed in accordance with all limitations and restrictions included in the NRCs approval of the methodology CNL-19-082 E2-40 of 78

Enclosure 2 III.3.G Describe the sequence of events and explicitly identify those that would change as a result of the power uprate III.3.H Describe and justify the chosen single-failure assumption III.3.I Provide plots of important parameters and explicitly identify those that would change as a result of the power uprate III.3.J Discuss any change in equipment capacities (e.g., water supply volumes, valve relief capacities, pump pumping flow rates, developed head, required and available net positive suction head (NPSH), valve isolation capabilities) required to support the analysis III.3.K Discuss the results and acceptance criteria for the analysis, including any changes from the previous analysis

RESPONSE

The WBN Unit 2 analyses of record for the UFSAR Chapter 15 analyses support the MUR power uprate as described in Section II.

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Enclosure 2 IV MECHANICAL/STRUCTURAL/MATERIAL COMPONENT INTEGRITY AND DESIGN IV.1 A discussion of the effect of the power uprate on the structural integrity of major plant components. For components that are bounded by existing analyses of record, the discussion should cover the type of confirmatory information identified in Section II, above. For components that are not bounded by existing analyses of record, a detailed discussion should be provided.

RESPONSE

Table IV.1-1 presents a summary of the relevant NSSS design parameters for both current and WBN Unit 2 MUR power uprate conditions. The WBN Unit 2 maximum analyzed thermal power of 3479 MWt (i.e., 102% of 3411 MWt, the CLTP) will not change as a result of the MUR power uprate.

As shown in Table IV.1-1, steam generator tube plugging (SGTP) values of 0% and 10%

(maximum) were evaluated and a feedwater temperature range of 440.2 - 441.8°F was used for the analyses. The core bypass flow value accounts for thimble plugs installed, intermediate flow mixing vanes, upflow conversion, and Tcold of the upper head. Case 1, which is based on an average 0% SGTP, yields the maximum secondary side steam pressure and temperature.

Case 2, which is based on an average 10% SGTP level, yields the minimum secondary side SG pressure and temperature. The primary side temperatures are identical for Cases 1 and 2 in Table IV-1. The data provided in Note 1 of Table IV.1-1 were used in those NSSS analyses and evaluations that require an absolute upper limit steam pressure. These more limiting secondary side data are based on the Case 1 parameters with an assumed SG fouling factor of zero.

The following sections provide confirmation that the plant components and piping continue to be bounded by the existing analyses of record.

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Enclosure 2 Table IV.1-1: WBN Unit 2 Nuclear Steam Supply System Design Parameters for Current and MUR Power Uprate Conditions MUR MUR Power Current Current Power Thermal Design Parameters Uprate Case 1 Case 2 Uprate Case 1 Case 2 NSSS Power (MWt) 3427 3427 3475 3475 NSSS Power (106 Btu/hr) 11,693 11,693 11,857 11,857 Reactor Power (MWt) 3411 3411 3459 3459 Reactor Power (106 Btu/hr) 11,639 11,639 11,803 11,803 Thermal Design Flow (gpm/loop) 93,100 93,100 93,100 93,100 Reactor Flow (106 lb/hr) 138.5 138.5 138.5 138.5 Reactor Coolant Pressure (psia) 2250 2250 2250 2250 Core Bypass Flow (%) 9.6 9.6 9.6 9.6 Reactor Coolant Temperature (°F)

Core Outlet 624.3 624.3 624.8 624.8 Vessel Outlet (i.e., Thot) 618.6 618.6 619.1 619.1 Core Average 593.0 593.0 593.1 593.1 Vessel Average (i.e., Tavg) 588.2 588.2 588.2 588.2 Vessel/Core Inlet (i.e., Tcold) 557.8 557.8 557.3 557.3 Steam Generator Outlet 557.4 557.4 557.0 557.0 Steam Generator Steam Outlet Temperature (°F) 542.9/542.7(1) 538.9/538.8 542.1/541.9(2) 538.1/537.9 Steam Outlet Pressure (psia) 986/985(1) 954/953(2) 979/978(2) 947/946 Total Steam Outlet Flow 15.11/15.14(1) 15.08/15.12 15.35/15.39(2) 15.33/15.36 (106 lb/hr)

Feed Temperature (°F) (i.e., Tfeed) 438.4/440.0 438.4/440.0 440.2/441.8 440.2/441.8 Steam Outlet Moisture (% max.)

0.25 0.25 0.25 0.25 (i.e., moisture carryover or MCO)

Tube Plugging Level (%) 0 10 0 10 Zero Load Temperature (°F) 557 557 557 557 Hydraulic Design Parameters Mechanical Design Flow 105,000 105,000 105,000 105,000 (gpm/loop)

Minimum Measured Flow 379,100 379,100 379,100 379,100 (gpm total)

Note 1: If a high steam pressure is more limiting for analysis purposes, a greater steam pressure of 1021 psia, steam temperature of 547.2ºF, and total steam flow of 15.13 x 106 lb/hr should be assumed. This is to envelop the possibility that the plant could operate with better than expected SG performance.

Note 2: If a high steam pressure is more limiting for analysis purposes, a greater steam pressure of 1015 psia, steam temperature of 546.4ºF, and total steam flow of 15.38 x 106 lb/hr should be assumed. This is to envelop the possibility that the plant could operate with better than expected SG performance.

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Enclosure 2 IV.1.A This discussion should address the following components:

IV.1.A.i Reactor vessel, nozzles, and supports

RESPONSE

The operating conditions for the proposed WBN Unit 2 MUR power uprate were reviewed for impact on the existing design basis analyses for the reactor vessel. As shown in Table IV.1-1, no changes in RCS design or operating pressure will be made as part of the power uprate. The mechanical design flow rate is unchanged due to the MUR power uprate and, therefore, there is no increase in flow velocity. The effects of operating temperature changes (i.e., a 0.5°F increase in Thot and a 0.5°F decrease in Tcold) are within design limits. The MUR power uprate conditions are bounded by the design conditions for the NSSS primary side design transients and the various design interface loads. Because the operating transients will not change because of the power uprate and no additional transients have been proposed, the existing loads, stresses, and fatigue values remain valid. Thus, the existing stress reports for the reactor vessel remain applicable for the uprated power conditions.

IV.1.A.ii Reactor core support structures and vessel internals

RESPONSE

The reactor internals support and orient the reactor core fuel assemblies and control rod assemblies, absorb control rod assembly dynamic loads, and transmit these and other loads to the reactor vessel. The reactor vessel internal components support in-core instrumentation and direct coolant flow through the fuel assemblies (core) to provide adequate cooling flow to the various internals structures. The internals are designed to withstand forces due to structure deadweight, fuel assembly pre-load, control rod assembly dynamic loads, vibratory loads, and earthquake accelerations. The seismic reactor internals system dynamic analysis is not affected by the operating conditions for MUR power uprate. Westinghouse has performed a thermal-hydraulic analysis to assess the effect of the MUR power uprate on the reactor pressure vessel and internals for WBN Unit 2.

The slight increase in Thot and a slight decrease in Tcold offset and, therefore, Tavg remains unchanged. The core delta temperature will experience a nominal increase of approximately 1°F to account for the MUR power increase but the revised core parameters are bounded by the design values plus uncertainty that were used in the current analyses. Therefore, the reactor vessel internals system performance after the MUR power increase is bounded by the current analyses of record.

IV.1.A.iii Control rod drive mechanisms

RESPONSE

The operating conditions for the proposed WBN Unit 2 MUR power uprate were reviewed for impact on the existing design basis analyses for the control rod drive mechanisms (CRDMs).

The MUR power uprate conditions are bounded by the design conditions, as discussed above.

Because the operating transients will not change because of the power uprate and no additional transients have been proposed, the existing loads, stresses, and fatigue values remain valid.

Thus, the existing stress reports for the CRDM remain applicable for the uprated power conditions.

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Enclosure 2 IV.1.A.iv Nuclear Steam Supply System piping, pipe supports, branch nozzles

RESPONSE

The operating conditions for the proposed WBN Unit 2 MUR power uprate were reviewed for impact on the existing design basis analyses for the reactor coolant piping and supports. The MUR power uprate conditions are bounded by the design conditions. A review determined that the effect of the MUR power uprate has been accounted for in the existing piping stress analysis of the reactor coolant loop piping, which includes the in-line piping and the branch connections welded onto the surface of the in-line piping, and associated support system, and the piping stress analysis of the pressurizer surge line. Thus, the existing stress reports for the reactor coolant piping and supports remain applicable for the WBN Unit 2 MUR power uprate conditions.

A separate discussion of thermal stratification and NRC Bulletin 88-11, Pressurizer Surge Line Thermal Stratification, is provided in Section IV.I.B.iv.

IV.1.A.v Balance-of-plant (BOP) piping (NSSS interface systems, safety related cooling water systems, and containment systems)

RESPONSE

Operation of interfacing and BOP systems at MUR power uprate conditions could result in increased piping stress levels, piping support loads, nozzle loads, etc., due to higher system operating temperatures, pressures, and flowrates. The following interfacing and BOP fluid systems were reviewed to ensure that they remain within their design basis:

AFW

Chemical and Volume Control System (CVCS)

Condensate and Feedwater Systems

Extraction Steam System

Heater and Moisture Separator Drains

Main Steam System

Process Sampling System

Safety Injection System

SG Blowdown System As shown in Table IV.1-1, there will be an increase in main steam flow at the uprated power (i.e., 3459 MWt) compared to the CLTP. This corresponding increase in steam and mass flow is seen in the remainder of the BOP systems (e.g., condensate, feedwater, extraction steam).

An evaluation of the structural integrity of these systems demonstrated that the BOP piping systems design bases do not change as a result of the MUR power uprate. The BOP piping design pressure and temperature limits do not change as a result of the MUR power uprate and the BOP piping is acceptable for MUR power uprate conditions. There are minor increases in mass flow rate (typically less than 3%) to various components throughout the BOP cycle, which is expected for an MUR power uprate and will not adversely affect BOP systems.

NSSS interface systems such as CVCS, Process Sampling, and Safety Injection are not expected to see any significant change in operating conditions. An evaluation of these systems demonstrated that the interfacing piping systems and auxiliary equipment (e.g., pumps, heat exchangers, valves, and tanks) will continue to meet their design basis under MUR power uprate conditions.

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Enclosure 2 As described in Section II of this enclosure, the accident analyses, which dictate AFW flow requirements, were evaluated to a power level that bounds the MUR power uprate. The CST inventory and associated TS requirements are not impacted by MUR power uprate.

The SG Blowdown System will remain within its design basis at MUR power uprate conditions.

The blowdown flow rate required to control chemistry and the buildup of solids in the SGs is tied to allowable condenser in-leakage, total dissolved solids in the plant service water, and the allowable primary to secondary leakage. Because these variables are not impacted by the MUR power uprate, the blowdown required to control secondary chemistry and SG solids will not be impacted by the MUR power uprate. The no-load steam pressure (1200 psia) remains the same, and the minimum full-load steam pressure decreases a maximum of seven psi as a result of the MUR power uprate. This small decrease in blowdown system inlet pressure (seven psi) is not considered significant with respect to blowdown flow control. Therefore, the range of NSSS design parameters approved for the MUR power uprate will not affect blowdown flow control.

Containment Systems are discussed in Section VI.1.B. Safety-related cooling water systems are discussed in Section VI.1.C. Additional discussion related to flow accelerated corrosion is provided in Section IV.1.E.iii.

IV.1.A.vi Steam generator tubes, secondary side internal support structures, shell, and nozzles

RESPONSE

The WBN Unit 2 SGs are discussed in UFSAR Section 5.5.2. The AOR, WNET-120, Volume 1, Revision 3, with addenda (References IV.1 and IV.2), for the WBN SGs states that the structural evaluation was applicable to both the current NSSS power level of 3427 MWt and the NSSS power level of 3475 MWt, which includes a 1.4% MUR power uprate. The AOR structural evaluation applied to the manufactured SGs as well as the hardware changes that were made to the SGs after they were installed at WBN Unit 2. These hardware changes included:

Feedwater inlet modifications

Upper internal modifications for improving moisture carryover (MCO)

Replacement of bolts with studs/nuts/washers in the bolted closure openings

Alloy 690 ribbed mechanical tube plugs for replacing the existing Alloy 600 mechanical tube plugs

Additional inspection port penetrations for potential loose parts (PLP)

Cut tube remnants for tubes removed to facilitate loose parts removal

Welded plugs for tube holes where tubes were removed The proposed WBN Unit 2 MUR power uprate operating parameters were reviewed and are discussed below. As shown in Table IV.1-1 (see Notes 1 and 2), the SG outlet pressure (for bounding operating conditions based on better than expected SG performance) decreases from 1021 psia at current full-power conditions to 1015 psia and the RCS pressure remains unchanged at 2250 psia for the proposed WBN Unit 2 MUR power uprate conditions.

Therefore, the normal operating differential pressure across a steam generator tube increases slightly from 1229 psid at current conditions to 1235 psid for MUR power uprate conditions, given better than expected SG performance. As shown in Table IV.1-1, the feedwater temperature is expected to increase 1.8°F and the steam flow rate increases by approximately 2.5E5 lbm/hr because of the MUR power uprate.

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Enclosure 2 The MUR conditions were reviewed for impact on the existing design basis analyses for the steam generators. The MUR power uprate conditions are bounded by the design conditions.

Because the operating transients will not change because of the power uprate and no additional transients have been proposed, the existing loads, stresses, and fatigue values remain valid.

Thus, the existing stress reports for the SGs remain applicable for the WBN Unit 2 MUR power uprate conditions.

The tube repair criteria discussed in TS 5.7.2.12 are not changed because of MUR power uprate. Similarly, there is no change in the SGTP limits for WBN Unit 2 because of the MUR power uprate.

See Section IV.1.F for a discussion of steam generator flow induced vibration.

IV.1.A.vii Reactor Coolant Pumps (RCPs)

RESPONSE

As shown in Table IV.1-1, primary coolant pressure will remain at 2250 psia after implementation of the MUR power uprate. Primary system flow will remain at the current value of 138.5E06 lbm/hr. The only significant change affecting the RCPs is that RCS cold leg temperature will decrease from the current value of 557.8°F to 557.3°F. The 0.5°F decrease in cold leg temperature will have a negligible effect on water density, and therefore a negligible effect on power input required to operate the pumps. Because there is no change to primary coolant pressure and flow, and the decrease in cold leg temperature after MUR power uprate is within the current design requirements for the RCPs, there is no impact to the RCPs.

Additionally, the AOR already addresses operation at MUR power uprate conditions and remains acceptable.

The results of the stress analyses in the AOR show that the WBN Unit 2 RCPs meet the requirements of the ASME Boiler and Pressure Vessel Code,Section III (1971 Edition up to and including the Summer of 1972 Addenda in accordance with the UFSAR), for the loading conditions specified for both the current NSSS power level of 3427 MWt and for the proposed MUR power uprate power level of 3475 MWt. Therefore, the WBN Unit 2 RCPs are shown to be acceptable for operation at MUR power uprate conditions.

IV.1.A.viii Pressurizer shell, nozzles, and surge line

RESPONSE

The AOR for the WBN Unit 2 pressurizer (Reference IV.3) states that the structural evaluation was applicable to both the NSSS power level of 3427 MWt and the NSSS power level of 3475 MWt, which includes the 1.4% MUR power uprate.

The operating conditions for the proposed WBN Unit 2 MUR power uprate were reviewed for impact on the existing design basis analyses for the pressurizer. No changes to the pressurizer design or operating pressure were made as part of the power uprate. The effects of operating temperature changes in the spray and surge lines are within design limits. The MUR power uprate conditions are bounded by the design conditions. Because the operating transients will not change because of the power uprate and no additional transients have been proposed, the existing loads, stresses, and fatigue values remain valid. Therefore, the existing stress reports for the pressurizer remain applicable for the power uprate conditions.

A separate discussion of thermal stratification and NRC Bulletin 88-11, Pressurizer Surge Line Thermal Stratification, is provided in Section IV.I.B.iv.

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Enclosure 2 IV.1.A.ix Safety related valves

RESPONSE

The pressurizer safety valves (PSVs) and the power operated relief valves (PORVs) and their associated block valves located on top of the pressurizer provide overpressure protection for the RCS. The WBN Unit 2 MUR power uprate is bounded by the current design basis event transient analyses (as discussed in Section II of this enclosure). Therefore, there is no adverse impact on the pressurizer overpressure protection valves from the MUR power uprate and the AOR for the pressurizer overpressure protection valves remains bounding at MUR power uprate conditions.

Other safety related valves were reviewed as part of the system that contains those valves. As discussed in Sections IV.1.A.v, operating conditions for interfacing systems will see small to no change under MUR power uprate conditions. Based on these reviews, it was determined that the AOR for interfacing system valves remain bounded at MUR conditions.

IV.1.B The discussion should identify and evaluate any changes related to the power uprate in the following areas:

IV.1.B.i Stresses

RESPONSE

No changes in the RCS design or operating transient conditions were made as part of the power uprate. NSSS primary side design transients, and various design interface loads have been shown to be satisfactorily bounded by the current evaluations. Based on the margins for the primary, primary plus secondary stresses and fatigue usage factors in the RCS piping loop, it is concluded that the MUR power uprate conditions remain acceptable and are bounded by the design conditions.

IV.1.B.ii Cumulative usage factors

RESPONSE

The revised design conditions for the NSSS components, piping, and interface systems were reviewed for impact on the existing design basis analyses. For NSSS components, the evaluation showed that the operating conditions due to the MUR power uprate are bounded by those used in the existing analyses. Further, because the evaluated transients will not change because of the power uprate, the existing loads remain valid, and the stresses and fatigue values (i.e., cumulative usage factors) also remain valid.

IV.1.B.iii Flow induced vibration (FIV)

RESPONSE

FIV is discussed in UFSAR Section 5.5.2.3.6. The reactor pressure vessel (RPV) internals are subjected to vibrations induced by flow turbulences and vortex shedding. FIV amplitudes are generally proportional to the product of the coolant density and fluid velocity raised to approximately the second power. Because the mechanical design flow rate is unchanged due to the MUR power uprate (per Table IV.1-1), there is no increase in flow velocity. There is an increase in coolant density due to the corresponding reduction in coolant temperature, which tends to increase the vibratory loads and amplitudes. The change in fluid density associated with the 0.5°F (i.e., 557.8°F minus 557.3°F) change in vessel/core inlet temperature is CNL-19-082 E2-48 of 78

Enclosure 2 negligible. (Tcold remains within design limits.) Therefore, there is a negligible impact on the flow-induced response of the reactor internals due to the proposed MUR power uprate.

High frequency acoustic sources from RCPs and low frequency acoustic sources from loop oscillations can induce vibrations in the internals during steady-state operation conditions. The mechanical design flow rate remains unchanged and, therefore, the vortex shedding frequencies remain unchanged.

Thus, the stresses imparted on the RPV internals due to FIV remain unchanged because of the proposed WBN Unit 2 MUR power uprate, and the existing analyses of record remain bounding.

The main condenser was also evaluated for tube vibration and found to be acceptable for WBN Unit 2 MUR power uprate conditions.

FIV of the steam generators is discussed in Section IV.1.F consistent with the RIS 2002-03 outline.

IV.1.B.iv Changes in temperature (pre- and post-uprate)

RESPONSE

Temperature Changes The changes in operating temperatures are provided in Table Error! Reference source not found.IV.1-1. The average temperature is unchanged, and the cold leg temperature decreases 0.5°F, while the hot leg temperature increases 0.5°F. These changes, as discussed elsewhere in Sections IV.1.A and B of this enclosure, have minimal impact.

Evaluation of Potential for Thermal Stratification Thermal stratification in the lines attached to the primary side of the RCS occurs mainly during heatup and cooldown. The current 100% power hot and cold leg operating temperatures that the plant has been designed to are essentially the same as those for the MUR power uprate.

This means that the effects of thermal stratification will not change as a result of the power uprate, as discussed below.

NRC Bulletin 88-08, Thermal Stresses in Piping Connected to Reactor Coolant Systems, addresses the issue of thermal stresses in piping attached to the primary loop. The temperatures and transients considered in the thermal stratification analysis of the pressurizer surge line adequately represent the pre-MUR uprate and post-MUR uprate temperatures and transients. In addition, the design RCS flow rates are unchanged for the MUR power uprate.

Therefore, the effects of the thermal stresses will not change because of power uprate.

NRC Bulletin 88-11, Pressurizer Surge Line Thermal Stratification, addresses the issue of surge line thermal stratification. Thermal stratification in the surge line occurs mainly during plant heatup and cooldown and is driven by the temperature difference between the hot leg and the pressurizer. The current operating temperature of the hot leg will increase very slightly due to MUR power uprate. A higher hot leg temperature gives a lower temperature differential between the hot leg and the pressurizer, which in turn lessens the stratification effects. This means that stress and fatigue in the surge line that is attributed to thermal stratification is bounded by the existing analyses.

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Enclosure 2 IV.1.B.v Changes in pressure (pre- and post-uprate)

RESPONSE

The system operating pressures remain unchanged as shown in Table Error! Reference source not found.IV.1-1.

IV.1.B.vi Changes in flow rates (pre- and post-uprate)

RESPONSE

As shown in Error! Reference source not found. Table IV.1-1, there is no change in RCS flow. Therefore, there is no impact on core design and safety analyses. A detailed review of safety analyses is provided in Section II.

As discussed in Section IV.1.A.v, there are only minor increases in mass flow rate (less than 3%) to various components throughout the BOP cycle, which is expected for an MUR power uprate and will not adversely affect BOP systems.

IV.1.B.vii High and moderate energy line break (HELB and MELB) locations

RESPONSE

WBN UFSAR Appendix 3.6A defines the classification and evaluation requirements for high and moderate energy fluid systems. High-energy fluid systems are defined as those where the normal operating temperature and pressure exceed 200°F and 275 psig. Systems not classified as high energy are considered moderate energy. In addition, systems that exceed these criteria for < 1% of the normal operating lifespan of the plant or < 2% of the time the system is performing its design function may be classified as moderate energy.

There are no HELB program changes required to be implemented because of the power uprate.

The temperature and pressure conditions for high energy systems used in existing analyses remain bounding. No new postulated line break locations will be introduced. In addition, no existing segments classified as moderate energy will become high energy due to the MUR power uprate conditions. No new lines are added, no break locations are changed, and no results change to the mass and energy blowdown from any postulated break. Therefore, MUR power uprate has no impact on the HELB analyses that were originally performed and the MUR power uprate is bounded by the existing HELB AOR for WBN Unit 2.

Additionally, systems for which a moderate energy line rupture are postulated were reviewed.

The effects of a MELB are primarily related to flooding. The systems were verified not to increase in pressure due to MUR power uprate. Therefore, MELB release rates are not impacted by MUR power uprate.

IV.1.B.viii Jet impingement and thrust forces

RESPONSE

The leak-before-break (LBB) concept applies known mechanisms for flaw growth to piping designs with assumed through-wall flaws and is based on the plants ability to detect an RCS leak. The RCS pipe loads used in the LBB evaluations are various combinations of deadweight, thermal expansion, pressure loads, and seismic loads. These loads are not affected by the MUR power uprate. The LBB analyses, discussed in UFSAR Sections 3.6A.2.1.5 and 3.6B.1, justified the elimination of large primary loop pipe rupture and pressurizer surge line pipe rupture from the design basis for WBN.

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Enclosure 2 The WBN Unit 2 plant primary loop piping analysis for the application of LBB is documented in WCAP-11985, Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for Watts Bar Units 1 and 2, (Reference IV.12), and WCAP-12500, Additional Information in Support of Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for Watts Bar Units 1 and 2, (Reference IV.13). The use of LBB was approved by the NRC in Reference IV.14.

The LBB acceptance criteria are based on SRP Section 3.6.3, Leak-Before-Break Evaluation Procedures, Revision 0. The recommended margins are summarized as follows:

Margin of 10 on leak rate

Margin of 2.0 on flaw size

Margin on loads (using faulted load combinations by the absolute summation method)

A supplement to the LBB evaluations demonstrates that the analysis is also bounding for MUR power uprate conditions. The normal operating temperature and pressure at the MUR power uprate condition were used in the evaluation. The evaluation showed that the LBB recommended margins were satisfied for the MUR power uprate condition.

The WBN Unit 2 plant pressurizer surge line analyses for the application of LBB were documented in WCAP-12773, Technical Justification for Eliminating Pressurizer Surge Line Rupture as the Structural Design Basis for Watts Bar Units 1 and 2, (Reference IV.15), and the corresponding supplemental information (Reference IV.16). The pressurizer surge line analyses was approved by the NRC in Reference IV.17. The surge line evaluation was supplemented with an analysis that bounds operation under WBN Unit 2 MUR power uprate conditions. A review of this analysis determined that the conclusions of the previous LBB analyses for WBN Unit 2 for the pressurizer surge line remain valid for MUR power uprate conditions. Therefore, the dynamic effects of pressurizer surge line pipe breaks for WBN Unit 2 are not impacted by the MUR power uprate.

The pressures, temperatures, and flow rates considered in the calculation of the hydrodynamic forces acting throughout the reactor coolant loop piping for the LOCA were selected to envelop the pre-uprate and post-uprate conditions. Similarly, the pressures, temperatures, and flow rates considered in the calculation of jet impingement and thrust forces were also selected to envelop the pre-uprate and post-uprate conditions. Therefore, the WBN Unit 2 MUR power uprate conditions do not affect the analyses for the aforementioned hydrodynamic forces and the existing LBB and surge line evaluations remain acceptable and are bounded by the existing computations of record.

IV.1.C The discussion should also identify any effects of the power uprate on the integrity of the reactor vessel with respect to:

IV.1.C.i Pressurized thermal shock calculations

RESPONSE

The pressurized thermal shock (PTS) evaluation provides a means for assessing the susceptibility of reactor vessel materials to failure during a PTS event to ensure that adequate fracture toughness exists during reactor operation. 10 CFR 50.61, Fracture toughness requirements for protection against pressurized thermal shock events, provides the requirements, methods of evaluation, and safety criteria for PTS assessments. PTS calculations were performed for the WBN Unit 2 reactor vessel materials using the latest procedures specified in 10 CFR 50.61 for end-of-license (EOL) (i.e., 32 effective full-power years (EFPY)) neutron fluence values.

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Enclosure 2 The associated reactor vessel neutron fluence projections assumed an MUR power uprate of 1.4% beginning with Cycle 2 operation and TPC designs, which include TPBARs beginning with Cycle 4 operation. The results of this analysis were summarized in WCAP-18191-NP, Revision 0, Watts Bar Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and Supplemental Reactor Vessel Integrity Evaluations (Reference IV.4), which was provided to the NRC in support of the WBN Unit 2 tritium production LAR. The NRC reviewed the applicable results in WCAP-18191-NP and confirmed that the WBN Unit 2 reactor vessel materials will continue to meet the 10 CFR 50.61 PTS screening criterion (i.e., 270°F for RPV forging materials at EOL), as discussed in the associated NRC Safety Evaluation (Reference IV.7). In the Safety Evaluation, the NRC also concluded that the tritium production LAR would not have an adverse impact on the PTS analysis for the facility.

The MUR power uprate is based on the same analyses. Therefore, the MUR power uprate also has no impact on 10 CFR 50.61 compliance. The WBN Unit 2 reactor vessel will remain within its PTS limits after implementation of the MUR power uprate.

IV.1.C.ii Fluence evaluation

RESPONSE

The fast neutron (E > 1.0 MeV) fluence experienced by the materials making up the beltline region of the RPV is used as input in all RPV integrity evaluations that involve an assessment of radiation-induced degradation in material properties. In determining the projected fast neutron fluence for the reactor materials, a calculation is performed for fuel cycles that have been completed and fluence projections for future operation are generated based on an assumed mode of operation. The key parameters in choosing a future mode of operation are the assumed spatial distribution of the neutron source within the reactor core and the core power level. Analyses of this type were completed in support of the WBN Unit 2 MUR power uprate and TPC implementation; the results of those analyses were documented in WCAP-18191-NP.

Projections for future operation were based on operation with an uprated core design and power level of 3459 MWt. Fluence calculations performed for WBN Unit 2 adhered to the NRC-approved methodology described in WCAP-14040-A (Reference IV.5). This methodology follows the guidance and meets the requirements of RG 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence.

As discussed in the NRC Safety Evaluation (Reference IV.7) which approved the use of a TPC design for WBN Unit 2, the NRC staff reviewed the methodology and results for the neutron flux and neutron fluence calculations and determined that the methodology and results are acceptable for RV integrity calculations for WBN Unit 2.

IV.1.C.iii Heatup and cooldown pressure-temperature (P-T) limit curves

RESPONSE

Appendix G to 10 CFR Part 50 provides fracture toughness requirements for ferritic low alloy steel or carbon steel materials in the RCS pressure boundary. It also includes the requirements on Upper-Shelf Energy (USE) values used for assessing the safety margins of reactor vessel materials against ductile tearing, and for calculating plant pressure-temperature (P-T) limits.

These P-T limits are established to ensure the structural integrity RCS pressure boundary ferritic components during any condition of normal operation, including anticipated operational occurrences and hydrostatic tests.

The current WBN heatup and cooldown curves are provided in the Pressure and Temperature Limits Report (PTLR) (Reference IV.6). Adjusted Reference Temperature (ART) or reference CNL-19-082 E2-52 of 78

Enclosure 2 temperature for nil-ductility transition calculations have been performed per RG 1.99, Revision 2 for the WBN Unit 2 reactor vessel materials using the MUR power uprate neutron fluence values. ART is a measure of the effect of radiation on the material and it is used as an input in the calculation of the allowable P-T limits. It is equivalent to the initial reference temperature (unirradiated) plus the shift due to irradiation and a margin term.

The new 32 EFPY P-T limit curves were developed in WCAP-18191-NP taking into account the MUR power uprate fluence projections (see Figures 8-1 and 8-2 in WCAP-18191-NP). The development of the P-T limit curves was consistent with the methodologies of NRC RG 1.99, Revision 2, and WCAP-14040-A, Revision 4 and comply with 10 CFR 50, Appendix G.

As discussed in the NRC Safety Evaluation (Reference IV.7) which approved the use of a TPC design for WBN Unit 2, the NRC found that the P-T limit curves for 32 EFPY in Chapter 8 of WCAP-18191-NP, were acceptable because the NRC staff determined that the P-T limits were calculated in accordance with the P-T limit methodology in WCAP-14040-A, Revision 4. The NRC staff also found that the proposed P-T limit curves for 32 EFPY are in compliance with the requirements in 10 CFR Part 50, Appendix G, because the NRC staff determined that the curves are at least as conservative as those that would be generated if the methods of analysis in the 2010 edition of ASME Code Section XI, Appendix G were used for the calculations. In the Safety Evaluation, the NRC staff concluded that WBN Unit 2 had adequately addressed the impacts of the tritium production LAR on the acceptability of the P-T limits for the facility.

The MUR power uprate is based on the same analyses. Therefore, the MUR power uprate also has no impact on 10 CFR 50, Appendix G, compliance. The PTLR will be updated to incorporate the 32 EFPY P-T limit curves to support implementation of the MUR power uprate and TPC implementation.

IV.1.C.iv Low-temperature overpressure protection

RESPONSE

The Cold Overpressure Mitigation System (COMS), also known as the low temperature overpressure protection system (LTOPS), provides RCS pressure relief capability during low temperature operation (i.e., TS Modes 4, 5, and 6, in which RCS temperature is less than 350°F). Therefore, the full power design parameters and feedwater temperature have no impact on the COMS transients. The only potential impact on the COMS setpoint determination would be if the uprate resulted in higher reactor vessel fluence and more restrictive P-T limit curves. The applicability of the P-T limit curves from WCAP-18191-NP, which already account for MUR power uprate conditions, is described in Section IV.1.C.iii.

The review of the P-T limit and fluence calculations confirmed that these critical inputs to the COMS PORV setpoint calculation bound the WBN Unit 2 MUR power uprate conditions. The associated COMS setpoint updates will support implementation of the MUR power uprate and TPC implementation.

IV.1.C.v Upper shelf energy

RESPONSE

USE was evaluated to ensure compliance with 10 CFR 50, Appendix G. If the limiting reactor vessel materials Charpy USE is projected to fall below 50 ft-Ibs, an equivalent margins analysis must be performed. As discussed in the NRC Safety Evaluation (Reference IV.7) which approved the use of a TPC design for WBN Unit 2, the NRC reviewed the results of the USE analysis in Appendix D of WCAP-18191-NP. The NRC found that the 1/4T EOL USE analysis is CNL-19-082 E2-53 of 78

Enclosure 2 valid and provides adequate demonstration that the USE values for the RPV beltline and extended beltline forgings, rings, and weld components will remain above the 50 ft-lb lower bound acceptance criterion established in 10 CFR Part 50, Appendix G, for EOL USE values.

The projected USE values for the WBN Unit 2 reactor vessel materials meet the 50 ft-Ib acceptance criterion of 10 CFR 50, Appendix G at the end of the current 40-year license period, including the effects of MUR power uprate. Therefore, MUR power uprate has no impact on 10 CFR 50, Appendix G compliance.

IV.1.C.vi Surveillance capsule withdrawal schedule

RESPONSE

The surveillance capsule withdrawal schedule is provided in Table 4.0-1 of the WBN Unit 2 PTLR (Reference IV.6). Revisions to the PTLR are required to be submitted to the NRC in accordance with WBN Unit 2 Technical Specification 5.9.6, Reactor Coolant System (RCS)

PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR).

IV.1.D The discussion should identify the code of record being used in the associated analyses, and any changes to the code of record.

RESPONSE

The applicable codes of record (i.e., ASME Boiler and Pressure Vessel Code,Section III, code addenda) for the WBN Unit 2 RCS are shown in UFSAR Table 5.2-1. Equipment supports are addressed in UFSAR Section 5.5.14.

As discussed in UFSAR Section 4.2.2.5, the reactor internals for WBN Unit 2 were fabricated prior to Subsection NG of the ASME Code becoming a requirement. However, with the exception of the Code Stress Report and Code Stamp, the reactor internals effectively satisfy the design and fabrication requirements of Subsection NG of the ASME Code.

A listing of Class 1 code cases used for the WBN Unit 2 RCS is provided in UFSAR Section 5.2.1.4. The code design criteria for interfacing systems is identified in UFSAR Section 3.2.3 and Table 3.2-7. A similar listing of Code Cases and provisions of later Code editions and addenda used in analysis of fluid systems is given in UFSAR Section 3.7.3.8.1. No stress/fatigue analyses were revised, therefore no codes of record changed.

IV.1.E The discussion should identify any changes related to the power uprate with regard to component inspection and testing programs and erosion/corrosion programs, and discuss the significance of these changes. If the changes are insignificant, the licensee should explicitly state so.

RESPONSE

IV.1.E.i Inservice Inspection Program The Inservice Inspection (ISI) Program is discussed in UFSAR Section 5.2.8 for ASME Class 1 components and Section 6.6 for ASME Class 2 and 3 components. ASME Code Class 1, 2 and 3 components are examined in accordance with the provisions of Section XI of the ASME Boiler and Pressure Vessel Code, as required by 10 CFR 50.55a(g), except where specific written relief has been requested and approved. The WBN Unit 2 MUR power uprate conditions were reviewed for impacts on the ISI Program. The WBN ISI Program for will continue to assess the operational qualification of ASME Class 1, 2, and 3 systems. The ISI Program does not require revision as a result of the WBN Unit 2 MUR power uprate.

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Enclosure 2 IV.1.E.ii Inservice Testing Program The Inservice Testing (IST) Program is discussed in WBN Unit 2 TS 5.7.2.11 and UFSAR Section 3.9.6. The WBN Unit 2 MUR power uprate conditions were reviewed for impacts on the IST Program. The IST Program does not require revision as a result of the WBN Unit 2 MUR power uprate.

IV.1.E.iii Flow Accelerated Corrosion Program As a result of plant and industry experience with pipe degradation in process systems, a Flow Accelerated Corrosion (FAC) program was developed at WBN. The purpose of the program is to monitor piping systems that are subject to FAC degradation, and to mitigate pipe wall loss.

The FAC program is based on the latest revision of Electric Power Research Institute (EPRI)

NSAC-202L, Recommendations for an Effective Flow-Accelerated Corrosion Program.

WBN uses the EPRI CHECWORKSTM Steam/Feedwater Application (SFA) software to model operating conditions, material data, and ultrasonic testing (UT) inspection data to provide a calculated estimate of component wear. The thermodynamic changes associated with the MUR power uprate will impact corrosion rates for components located in FAC susceptible systems.

Changes required to reflect the MUR power uprate conditions were already incorporated for the applicable power level option in the CHECWORKSTM models, as part of the WBN Unit 2 plant startup effort, and the final results and databases were previously validated.

A wear rate analysis has been performed to assess the impact of the MUR power uprate on susceptible components within the scope of the FAC Program for WBN Unit 2. Sample results are shown in Table IV.1.E-1, providing the average percent increase in wear and average post-MUR wear rates. Per this analysis, the increase in wear rates due to the MUR power uprate is considered minor and the existing FAC Program is adequate to incorporate the updated predictions.

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Enclosure 2 Table IV.1.E-1: Wear Rate Analysis for Lines with an Expected Increase in Wear Post-MUR Power Uprate Piping section in WBN2 Average Percent Increase Average Wear Rate CHECWORKS SFA Model in Wear Rate Post-Uprate (mils/yr)

CD COND BP TO FWH4 1.4% 3.2 CD FWH 4 to FWH 3 1.1% 3.2 CD FWH3 TO FWH2 0.9% 2.5 CD FWH5 to COND BP 1.2% 3.0 CD FWH6 TO FWH5 1.2% 1.8 CD SG HT EX TO CBP 0.8% 2.2 ES HP Turb to FWH1 2.2% 10.4 ES HP Turb to FWH2 1.9% 19.4 ES HP Turb to FWH3 1.2% 18.0 ES HP Turb to MSR1 2.2% 6.2 ES HP Turb to MSR2 2.2% 6.9 ES LP Turb to FWH6 2.6% 7.0 FW FWH1 TO SG 2.8% 0.8 HD FWH 2 to No.3 HDT 1.3% 2.8 HD FWH 3 to No.3 HDT 0.3% 0.4 HD FWH 4 to FWH 5 2.5% 2.1 HD LP RHR to FWH2 3.4% 0.4 HD MSR to No.3 HDTank 2.2% 0.8 HD No3 HD Tnk to Cond 1.2% 2.7 IV.1.F The discussion should address whether the effect of the power uprate on steam generator tube high cycle fatigue is consistent with NRC Bulletin 88-02, Rapidly Propagating Fatigue Cracks in Steam Generator Tubes, February 5, 1988.

RESPONSE

NRC Bulletin 88-02 describes an event in which a fatigue failure occurred in a steam generator tube and applies to holders of operating licenses of specific models of Westinghouse recirculating steam generators. The bulletin discusses the need to minimize the potential for a SGTR event caused by rapidly propagating fatigue cracks such as occurred at North Anna Unit 1 on July 15, 1987. The cause of the tube rupture was high cycle fatigue. The source of loads was a combination of high mean stress level in the tube and a superimposed alternating stress. The bulletin notes that the necessary preconditions for this phenomenon include denting in the tube at the upper support plate, a high fluid-elastic stability ratio, and the absence of effective anti-vibration bar support.

The TVA response to Bulletin 88-02 for WBN (Reference IV.9), as supplemented by the TVA letter, dated July 31, 2012 (Reference IV.10), which included the WBN Unit 2 evaluation for tube vibration-induced fatigue, WCAP-17309-P, Revision 1 (Reference IV.11), addresses SG tube high cycle fatigue concerns. This analysis and the modifications identified in the associated letter were performed as part of the WBN Unit 2 plant startup effort and included consideration of the MUR power uprate operating parameters to arrive at which tubes were to be plugged to CNL-19-082 E2-56 of 78

Enclosure 2 address high cycle fatigue concerns. Therefore, the effect of high cycle fatigue has already been considered at the MUR power uprate reactor power level and addressed for WBN Unit 2.

References for Section IV:

IV.1. WNET-120, Volume 1, Revision 3, Model D3 Steam Generators Summary Stress Report, Watts Bar Nuclear Plant Unit 2, dated December 2014 IV.2. WNET-120, Volume 1, Revision 3, Addendum 2, Model D3 Steam Generators Stress Report Addendum for Tennessee valley Authority Watts Bar Unit 2, dated September 2017 IV.3. WNET-130, Volume 1, Revision 4, Model 84D Pressurizer Summary Stress Report, Watts Bar Nuclear Plant Unit 2, dated October 2014 IV.4. Westinghouse Report, WCAP-18191-NP, Revision 0, Watts Bar Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and Supplemental Reactor Vessel Integrity Evaluations, dated May 2017 (ML17289A327)

IV.5. WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, dated May 2004 IV.6. TVA letter to NRC, Watts Bar Nuclear Plant Unit 2 - Revised Pressure and Temperature Limits Report (PTLR), dated February 9, 2018 (ML18040A434)

IV.7. NRC letter to TVA, Watts Bar Nuclear Plant, Units 1 and 2 - Issuance of Amendment Regarding Revision to Watts Bar Nuclear Plant, Unit 2, Technical Specification 4.2.1, Fuel Assemblies, and Watts Bar Nuclear Plant, Units 1 and 2, Technical Specifications Related to Fuel Storage (EPID L-2017-LLA-0427), dated May 22, 2019 (ML18347B330)

IV.8. Not used IV.9. TVA letter to NRC, Watts Bar Nuclear Plant (WBN) Units 1 and 2 - Nuclear Regulatory Commission (NRC) Bulletin 88-02, Rapidly Propagating Fatigue Cracks In Steam Generator Tubes, dated March 1, 1989 (ML080450615)

IV.10. TVA letter to NRC, Watts Bar Nuclear Plant (WBN) Unit 2 - WBN Unit 2 Evaluation for Tube Vibration-Induced Fatigue, dated July 31, 2012 (ML12215A337)

IV.11. WCAP-17309-P, Revision 1, Watts Bar Unit 2 Evaluation for Tube Vibration Induced Fatigue, dated February 2012 IV.12. WCAP-11985, Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for Watts Bar Units 1 and 2, dated November 1988 IV.13. WCAP-12500, Additional Information in Support of Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for Watts Bar Units 1 and 2, dated January 1990 IV.14. NRC letter to TVA, Safety Evaluation - Compliance with GDC-4 of Appendix A to 10 CFR Part 50 Requirements, Watts Bar, Units 1 and 2 Primary Loop Piping (TAC 73389), dated May 17, 1990 (ML073520336)

IV.15. WCAP-12773, Technical Justification for Eliminating Pressurizer Surge Line Rupture as the Structural Design Basis for Watts Bar Units 1 and 2, dated December 1990 IV.16. TVA letter to the NRC, transmitting supplemental information for WCAP-12773, dated March 26, 1993 (ML073230272)

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Enclosure 2 IV.17. NRC letter to TVA, Watts Bar Nuclear Plant - Leak-Before-Break Evaluation of the Pressurizer Surge Line (TAC Nos. T3837 and M83838), April 28, 1993 (ML073230286)

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Enclosure 2 V ELECTRICAL EQUIPMENT DESIGN V.1 A discussion of the effect of the power uprate on electrical equipment. For equipment that is bounded by the existing analyses of record, the discussion should cover the type of confirmatory information identified under Section II, above. For equipment that is not bounded by existing analyses of record, a detailed discussion should be included to identify and evaluate the changes related to the power uprate. Specifically, this discussion should address the following items:

RESPONSE

The electrical systems at WBN were reviewed. The WBN offsite and onsite electrical power systems are described in UFSAR Sections 8.2 and 8.3, respectively. The plant electrical power system consists of the equipment required to supply power to both safety-related and non-safety related loads in the plant. The plant electrical power system evaluated for MUR power uprate consists of the following electrical components:

Unit Station Service Transformers (USSTs)

Common Station Service Transformers (CSSTs)

6.9 kV AC System Switchgear and distribution equipment

480 V AC System Switchgear and distribution equipment

120 V AC System distribution equipment

Plant 120 V AC and 125 V DC Vital Systems Each of the electrical power system components listed were reviewed for impact due to MUR power uprate. Any potential change in electrical load was evaluated against the existing analyses. Specific RIS 2002-03 items are addressed separately in the subsections below.

During normal operation, non-safety related station auxiliary power system loads are fed through the USSTs from the main generators, and from the 161 kV system through CSSTs A and B. Class 1E safety-related loads are fed from the 161 kV system via CSSTs C and D.

During startup and normal shutdown, all auxiliary power is supplied from the 161 kV system through CSSTs A, B, C and D. The USSTs and CSSTs continue to have adequate capacity and capability for plant operation with an MUR power uprate and are bounded by the existing analysis and calculations of record for the plant.

The preferred power system supplies Class 1E circuits and is normally fed via CSSTs C and D (i.e., the offsite 161 kV system). CSSTs C and D are connected to 6.9 kV common switchgear C and D and then to the 6.9 kV shutdown boards. The CSSTs continue to have adequate capacity and capability for plant operation with an MUR power uprate and are bounded by the existing analysis and calculations of record for the plant.

The analyses of record for the Auxiliary Power System (i.e., the onsite electrical power system) bounds the resultant AC electrical load requirements due to the MUR power uprate. The existing load flow (ETAP) analysis for the plant already accounts for plant operation at MUR power uprate conditions. The post-MUR power uprate pump brake horsepowers are below the rated motor nameplate horsepowers and acceptable margin remains available in the current AC electrical power analyses of record.

The capability of the transmission system to maintain the post-trip voltage levels at the safety buses above the reset value of the degraded voltage relay on a steady-state basis has been verified.

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Enclosure 2 Plant 120 V AC and 125 V DC Vital systems are not impacted, and continue to have adequate capacity and capability for plant operation with an MUR power uprate. The existing LEFM system loads do not change and are already reflected in existing plant calculations and are bounded by the existing analysis and calculations of record for the plant.

The plant electrical power system and offsite power components required for electrical power conversion were evaluated for MUR power uprate and determined to be acceptable.

The associated generation train components consist of the following:

Main Turbine-Generator (see Section VI.1.A.ii)

Isolated-Phase Bus

Main Power Transformer

Switchyard Power Circuit Breakers (PCBs) and Switches

System Protection Relaying

Transmission Grid / Interconnections (see Section V.1.D)

The generator at 1411 MVA, for 75 psig generator hydrogen pressure, is the limiting component in the train of generation components based on their individual MVA ratings.

The WBN Unit 2 generator protection calculation was reviewed to evaluate the settings of the protective devices that could potentially be impacted by the MUR power uprate. The changes made by the power uprate have no impact to the existing settings on the protective relays. One of the key inputs to the setting calculations is the machine ratings, which are not changing.

Therefore, system protective relaying can accommodate the MUR power uprate requirements.

See Section V.1.D of this enclosure for more details related to the transmission grid/interconnections and grid stability.

Therefore, the electrical power systems (including electrical power conversion systems) continue to have adequate capacity and capability for plant operation with an MUR power uprate and are bounded by the existing analyses and calculations of record for the plant.

V.1.A Emergency Diesel Generators

RESPONSE

The emergency diesel generators units (EDGs) are part of the WBN onsite standby AC power system. The plant onsite standby AC power system is described by UFSAR Section 8.3. The onsite standby AC power system is a safety-related system that supplies continuous power for all AC-powered devices essential to safety. The onsite standby AC power system evaluated for MUR power uprate consists of the following electrical components: four Class 1E EDGs rated 4.4 MW at 0.8 pf, 6.9 kV shutdown boards and 6.9 kV shutdown relay logic panels, 6.9 kV/480 V transformers and the 480 V shutdown boards, all motor control centers (MCCs) supplied by the 480 V shutdown boards by both units.

Four separate diesel generators provide necessary power to the 6.9 kV shutdown board essential loads in the event that the normal AC power is interrupted. As discussed in Sections II and III of this enclosure, none of the UFSAR Chapter 6 or 15 analyses are being revised as a result of the MUR power uprate. The emergency loads for the EDGs are listed in WBN UFSAR Table 8.3-3. A review concluded that the MUR power uprate does not impact loads supplied by the EDGs and that the existing loading analysis remains bounding for MUR power uprate conditions. As a result, the EDG supporting systems remain unchanged, the timing and loading CNL-19-082 E2-60 of 78

Enclosure 2 sequence remains unchanged, there is no impact to load shedding activities, and the onsite diesel fuel storage capacity remains adequate under MUR power uprate conditions. Therefore, the plant onsite standby AC power system remains unchanged under WBN Unit 2 MUR power uprate operating conditions.

V.1.B Station blackout equipment

RESPONSE

10 CFR 50.63, Loss of all alternating current power, identifies the factors that must be considered in specifying the station blackout (SBO) duration and requires that the plant be capable of maintaining core cooling and appropriate containment integrity. As described in Section 8.3 of the WBN UFSAR, WBN is required to support a four-hour coping duration.

Following the guidance of RG 1.155, Revision 0, and NUMARC 87-00, this selection is based on availability and reliability of AC power sources and is not reactor power level dependent.

The coping method for WBN is AC-independent since no alternate AC power source beyond the EDGs is credited to address SBO. The systems credited for operation during the SBO coping period were evaluated and determined to be acceptable for WBN Unit 2 MUR power uprate conditions, as discussed below.

WBN station relies on batteries and nitrogen gas cylinders during SBO. Nitrogen bottles are available to support operation of the steam generator power-operated relief valves and auxiliary feedwater level control valves. The number of full valve strokes available given the existing nitrogen supply bounds the anticipated operation of the valves during an SBO under MUR power uprate conditions for the four-hour coping period. The 125V and 250V DC power system provide power to valve motors, oil pump motors, breakers, battery boards, instrumentation, and other loads required during the coping period. The batteries are sized to support four hours of operation and loads are shed as required accordingly.

The 125V DC battery system provides control and instrumentation power for the turbine-driven AFW pump, RCS indication, emergency lighting, and instrument power for AFW system level control. Since the required AFW flow was evaluated to a condition which bounds MUR power uprate, none of the associated electrical loads are impacted by reactor power. The 125V DC DG Power system must have capacity to support three attempts to start the EDG during the coping period. This requirement is not reactor power level dependent. The 250V DC batteries provide power for various oil pump motors, the Technical Support Center inverter, battery boards, and switchyard power circuit breakers. None of these power demands is dependent on reactor power.

The RCS was evaluated to ensure natural recirculation could be maintained with no makeup for the four-hour coping duration. Seal and system leakage during the event are not power level dependent. The RCS inventory was previously analyzed for Diverse and Flexible Mitigation Capability (FLEX), which bounds the SBO coping duration without makeup. The FLEX analysis considered a reactor power of 3459 MWt, which is equivalent to the targeted WBN Unit 2 MUR power uprate.

The CSTs supply water for AFW to provide steam generator makeup. Condensate makeup for decay heat removal during the four-hour SBO is provided by the turbine-driven auxiliary feedwater pump. The capability of the AFW system to provide adequate makeup is not challenged by MUR power uprate. Similarly, the CST has adequate capacity to support the four-hour coping time under WBN Unit 2 MUR power uprate conditions, as documented in the existing SBO coping analysis.

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Enclosure 2 Evaluations have been performed for the systems and components that are credited for SBO mitigation. Each was found to be acceptable for the SBO coping duration of four hours and unaffected by the WBN Unit 2 MUR power uprate.

V.1.C Environmental qualification of electrical equipment

RESPONSE

The TVA EQ Program addresses safety-related electrical equipment within the scope of 10 CFR 50.49 for WBN. The EQ program for WBN was reviewed to evaluate the impact of the MUR power uprate and it was determined that no programmatic changes are required. See Section II.1.D.iii (Item 32).

In accordance with the TVA design change process, any specific component modifications that may be required to support the MUR power uprate will be evaluated against the EQ program requirements.

V.1.D Grid stability

RESPONSE

The WBN Unit 2 MUR power uprate is expected to produce approximately 17-18 MWe uplift in generator output, based on station heat balances. At 100% MUR power uprate conditions generator output is expected to be approximately 1240 MWe at nominal backpressure. The maximum theoretical winter output may be as high as 1283 MWe at 100% MUR power and low winter backpressure, based on PEPSE3 (Performance Evaluation of Power System Efficiencies) heat balances. TVA performed an Interconnection System Impact Study (also known as a grid stability impact study) (Reference V.1) to evaluate the impact of the WBN Unit 2 MUR power uprate on the interconnection in Rhea County, Tennessee. The study addressed the following items for analysis of the grid with respect to the added generation:

1) Loadflow Analysis Study (i.e., steady-state thermal and voltage analysis)

2) Fault Study (i.e., short circuit analysis)

3) Transient Stability Study

4) Reactive Power Capability and Voltage Control Study A summary of the results for these four items is provided below.

The power flow models utilized in this study originated from the Eastern Interconnection Reliability Assessment Group (ERAG) Multi-Regional Modeling Working Group (MMWG) and the SERC Reliability Corporation (SERC) Long Term Study Group (LTSG) 2016 series of power flow base cases. These models are created as part of the ERAG and SERC regional modeling process. The transient stability model used in this study was based on the most recent SERC dynamically reduced base cases, which were created in 2013. The most up-to-date load forecast, transmission, and generation plans available at the time of case creation were considered in the cases, including prior Interconnection Requests. The analysis of the Interconnection Request was conducted using a combination of software including PTI PSS/E, PowerWorld Simulator, and PowerGEM TARA.

3 PEPSE is a registered trademark of Scientech a business unit of Curtiss-Wright Flow Control Service Corporation CNL-19-082 E2-62 of 78

Enclosure 2 The study determined that no additional interconnection facilities or system protection upgrades would be needed for the WBN Unit 2 MUR power uprate.

VI.D.1 Loadflow Analysis Study Steady-state loadflow analysis determined that the proposed uprate to the existing interconnection (i.e., WBN Unit 2) did not cause any thermal or voltage issues on the TVA transmission system.

VI.D.2 Fault Duty Study Short circuit analysis was not evaluated for the uprate to the existing interconnection because machine and transmission system impedances were not changed.

VI.D.3 Stability Study Transient stability analysis determined that the proposed interconnection has no detrimental impact on the stability of the TVA transmission system.

VI.D.4 Reactive Capability Study Reactive capability studies evaluate the capability of the generator and downstream components to generate or carry volt-amperes reactive (VARs). With the MUR power increase, the generator will have slightly less VAR capability based on the fixed generator capability curve (see Figure V.1-1). TVAs Large Generator Interconnection Procedure (LGIP) requires the facility while at its maximum power level (MWs) to be capable of continuous operation at power factors (PFs) of at least the minimum range of 0.95 lagging to 0.95 leading at the Point of Interconnection (POI) to provide optimum support to the voltage level specified in the voltage schedules. This capability allows the generator to produce or absorb reactive power as appropriate to comply with imposed grid operating requirements.

References for Section V:

V.1. Tennessee Valley Authority, Interconnection System Impact Study No. 329 - Watts Bar Nuclear Plant - Unit 2 Uprate, Revision 1, September 2019.

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Enclosure 2 Figure V.1-1: Generator Capability Curve (or D-Curve)

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Enclosure 2 VI SYSTEM DESIGN VI.1 A discussion of the effect of the power uprate on major plant systems. For systems that are bounded by existing analyses of record, the discussion should cover the type of confirmatory information identified under Section II, above. For systems that are not bounded by existing analyses of record, a detailed discussion should be included to identify and evaluate the changes related to the power uprate.

Specifically, this discussion should address the following systems:

VI.1.A NSSS interface systems for pressurized-water reactors (PWRs) (e.g., main steam, steam dump, condensate, feedwater, auxiliary/emergency feedwater) or boiling-water reactors (BWRs) (e.g., suppression pool cooling), as applicable

RESPONSE

VI.1.A.i Main Steam The Main Steam Supply System (MSS) is described in UFSAR Section 10.3.

The MSS is designed to conduct steam from the SG outlets to the high-pressure turbine and to the condenser steam dump system. This system also supplies steam to the feedwater pump turbines, an auxiliary feedwater pump turbine, moisture separator reheaters (MSRs), and the turbine seals.

The purposes of the MS system are as follows:

Provide means for plant cooldown by steam discharge to atmosphere

Provide MS System overpressure protection

Provide steam to the Turbine Driven AFW Pump as needed

Establish the containment boundary to minimize the loss of reactor coolant inventory during applicable design basis events.

Remove heat from the RCS and provide steam to the High Pressure Turbine

Provide steam dump to the condenser in the event of a load rejection System design assures that a postulated main steam line break (MSLB) coincident with a single active failure will not develop consequences outside the current plant design bases. The existing analysis for MSLB is bounding for WBN Unit 2 MUR power uprate conditions. The ability of MSIVs to close to isolate a line break is not impacted by MUR power uprate.

There are no design changes required for this system for MUR power uprate. The pressure and temperature design conditions continue to bound operating conditions with MUR power uprate.

The current analysis for a High Pressure Turbine Stop Valve closure event resulting in a fluid transient bounds MUR power uprate conditions. The capability of the Main Steam Safety Valves, Steam Generator Power Operated Atmospheric Relief Valves, Main Steam Isolation Bypass Valves, and the Steam Dump System capacity (40% of rated steam flow) were determined to be acceptable for WBN Unit 2 MUR power uprate conditions. The moisture separation performance for the MSRs may be slightly reduced because of higher steam flowrates; however, this does not affect the capability of the MSR relief valves and the MSR performance remains acceptable. Therefore, the review of the MS System for WBN Unit 2 MUR power uprate conditions shows that all system functions will continue to be performed following the MUR power uprate.

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Enclosure 2 VI.1.A.ii Main Turbine-Generator As discussed in WBN UFSAR Section 10.2, the turbine-generator converts the thermal energy of steam produced in the steam generators into mechanical shaft power and then into electrical energy. The Siemens (formerly Westinghouse) turbine-generator consists of a (single shaft) arrangement of a double-flow, high-pressure turbine and three double-flow low-pressure turbines driving a direct-coupled generator at 1800 rpm. The turbine-generator was reviewed and found to be acceptable for the MUR power uprate level and the unit design rating of 1411 MVA.

The main turbines are designed for guaranteed performance at the MUR power level and are therefore acceptable for MUR power uprate operation. The Unit 2 steam turbine-driven polyphase generator is a four-pole machine rated at 1411 MVA, 24 kV. Generator hydrogen pressure is currently required to be between 62 - 75 psig. The maximum theoretical output due to the MUR power uprate is 1283 MWe (for bounding conditions in winter with low backpressure). However, WBN Unit 2 may be restricted to operate at less than the winter maximum capability to meet transmission connection commitments.

Turbine throttle flow margin (relative to valves wide open flow capacity) will be reduced with MUR power uprate, as the required steam flow increases and the throttle pressure (i.e., steam pressure at the throttle valve inlet) is reduced. Turbine throttle flow margin at full MUR power (i.e., NSSS power of 3475 MWt) is expected to be approximately 1.4%, which is a reduction from the 3.2% margin at current full power operation (i.e., NSSS power of 3427 MWt). While the turbine-generator is designed for full output at MUR power uprate conditions, the reduced operating margin will result in turbine control (governor) valves operating further open.

VI.1.A.iii Condensate and Feedwater The Condensate and Feedwater Systems are described in UFSAR Section 10.4.7. Power uprate would increase the flow, pressure, and temperatures throughout various parts of the condensate, feedwater, and heater drains systems. The ability of the components and piping in these systems was evaluated to ensure appropriate margin exists between design parameters and operating parameters for WBN Unit 2 MUR power uprate conditions. The components and piping were evaluated and found to be acceptable at MUR power uprate operating conditions.

There are no design changes required to this system for power uprate. Therefore, the Condensate and Feedwater systems have sufficient design and operational margin to accommodate the WBN Unit 2 MUR power uprate.

VI.1.A.iv Auxiliary Feedwater The AFW System is described in UFSAR Section 10.4.9. This system provides feedwater to the steam generators in the event of loss of main feedwater. The accident analyses were evaluated at 3475 MWt, or higher, as appropriate, and bound the WBN Unit 2 MUR power uprate conditions. There are no design changes required for this system for MUR power uprate.

Therefore, this system is not impacted by the MUR power uprate.

VI.1.A.v Condenser Circulating Water The Condenser Circulating Water (CCW) system is described in UFSAR Section 10.4.5. This system provides cooling water to the condensers for the main steam turbines to condense turbine exhaust steam. The system provides an efficient means of dissipating waste heat from the power generation cycle into the ambient surroundings while meeting all applicable chemical and thermal effluent criteria.

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Enclosure 2 The heat rejected by the CCW system increases slightly as a result of the WBN Unit 2 MUR power uprate. CCW outlet temperature is expected to increase approximately 1.2°F. A bounding discharge temperature of 124.1°F was calculated using valves wide open (VWO) steam flow and summer operating conditions. The piping downstream of the condenser is rated to a design temperature of 135°F and therefore is acceptable under MUR power uprate conditions. There is no proposed change to pumps or operating flow in the CCW system for MUR power uprate. Therefore, operating pressures are not impacted. Similarly, there is no change in CCW pump horsepower requirements. The CCW pump net positive suction head available (NPSHA) is a function of cooling tower outlet temperature and CCW flow rate.

Because the change in cooling tower outlet temperature is negligible and there is no proposed change in CCW flow the CCW pump NPSHA is not adversely impacted by MUR power uprate.

Therefore, the CCW system was evaluated for WBN Unit 2 MUR power uprate conditions and found to be acceptable.

VI.1.B Containment systems

RESPONSE

The containment systems are described in UFSAR Section 6.2. The systems are provided to limit offsite releases following a design basis accident. These systems include the containment vessel and containment isolation system, ice condenser, Containment Spray, containment combustible gas control system and Emergency Gas Treatment System. As indicated in Sections II and III of this enclosure, the existing analyses are shown to remain valid. As such, these systems are not impacted by the MUR power uprate. During normal operation, air temperature in the upper and lower containment is maintained within limits. No changes to these limits are needed.

VI.1.B.i Containment Spray System The Containment Spray System (CSS) is discussed in UFSAR Section 6.2.2. CSS is not credited for post-accident fission product cleanup capability as discussed in UFSAR Section 6.5.2. The system is not in operation during normal operation. CSS is automatically actuated by a hi-hi containment pressure signal from the solid-state protection system (SSPS).

The system can also be manually actuated from the control room. The containment response analysis for design basis events is unchanged with MUR power uprate and continues to bound the MUR power uprate conditions for WBN Unit 2. The containment spray pumps are not impacted by the MUR power uprate. The credited containment spray heat exchanger performance is also not affected by the MUR power uprate. Therefore, CSS is not impacted by MUR power uprate conditions for WBN Unit 2.

VI.1.B.ii Containment Isolation Containment isolation is discussed in UFSAR Section 6.2.4.2. The containment isolation systems provide the means of isolating fluid systems that pass through containment penetrations in order to confine to the containment any radioactivity that may be released in the containment following a design basis event. WBN does not have a particular system for containment isolation, but isolation design is achieved by applying common criteria to penetrations in many different fluid systems and by using ESF signals to actuate appropriate valves.

Because the MUR power uprate does not change any of the accident analyses discussed in Section II of this enclosure, the existing setpoints for containment isolation remain the same.

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Enclosure 2 Containment isolation was reviewed as a function of individual systems and determined to remain acceptable.

VI.1.B.iii Containment Combustible Gas Control System The containment combustible gas (or hydrogen) control system is described in UFSAR Section 6.2.5. The containment combustible gas control system is designed to control the concentration of hydrogen that may be released into the containment following a beyond-design-basis accident to ensure that containment structural integrity is maintained. The combustible gas control system consists of the containment air return system, the hydrogen analyzer system (HAS) and the hydrogen mitigation system (HMS) which conform to 10 CFR 50.44 requirements.

The LOCA analysis described in UFSAR Chapter 15 used a thermal power of 3479.8 MWt.

Because hydrogen production and accumulation has been analyzed utilizing conditions bounding the MUR power uprate conditions, and because the combustible gas control system is not credited for design basis accident analyses, no further evaluation is required for the WBN Unit 2 MUR power uprate. Therefore, the containment combustible gas control system remains acceptable for MUR power uprate.

VI.1.B.iv Emergency Gas Treatment System The Emergency Gas Treatment System (EGTS) is described in UFSAR Section 6.2.3 and has two subsystems: the non-safety-related annulus vacuum control subsystem and the safety-related air cleanup subsystem.

The annulus vacuum control subsystem maintains a negative pressure within the annular space between the primary and secondary containment structures during normal operations.

The EGTS air cleanup subsystem is automatically initiated by a Phase A containment isolation signal, which is not impacted by the MUR power uprate conditions. The LOCA analysis described in Section II was analyzed using a thermal power of 3479.8 MWt, which bounds the WBN Unit 2 MUR power uprate conditions. The EGTS is not impacted by MUR power uprate operation conditions.

VI.1.B.v Ice Condenser Refrigeration System The primary purpose of the ice condenser system is to reduce the pressure that would occur in containment as a result of a rupture of the pressure boundary of the primary or secondary system by using the heat sink afforded by a large amount of ice to condense the steam that is released following the rupture. The borated ice bed, the heat sink of the ice condenser, is an enclosed annular compartment located around the perimeter of the containment upper compartment (which also penetrates the operating deck so that a portion extends into the lower compartment). The ice is held in cylindrical basket columns and a refrigeration system maintains the ice in the solid state. During a postulated LOCA, the refrigeration system is not required to provide any heat removal function. However, the refrigeration system components, which are physically located within the containment, must be structurally secured and the component materials must be compatible with the post-LOCA environment.

The operating basis earthquake (OBE) and safe shutdown earthquake (SSE) loadings applicable to the ice condenser system are not impacted by the WBN Unit 2 MUR power uprate.

There are no impacts on the blowdown loadings from long-term and short-term LOCA, as the existing analyses of record, which provide the loadings for the ice condenser system remain CNL-19-082 E2-68 of 78

Enclosure 2 bounding for the MUR power uprate. Because there are no impacts on the OBE and SSE loadings and the DBA loadings used in the existing analyses of record the ice condenser refrigeration system remains acceptable for WBN Unit 2 MUR power uprate.

A review of the ice condenser system concluded that the system would continue to maintain containment temperatures within the limits specified in TS 3.6.5. Because the average temperature of the RCS does not increase due to the MUR power uprate, heat-producing equipment inside containment remains relatively unchanged. With negligible change in the containment heat loads, the Ice Condenser Refrigeration System will remain within its design basis for MUR power uprate conditions.

VI.1.C Safety-related cooling water systems

RESPONSE

VI.1.C.i Component Cooling System The Component Cooling System (CCS) is described in UFSAR Section 9.2.2. The design analyses, which credit CCS, bound operation under the MUR power uprate. The system will continue to be able to perform its safety function of containment isolation and heat removal under accident conditions. There is no impact to this system due to the WBN Unit 2 MUR power uprate.

VI.1.C.ii Nuclear Service Water System The Essential Raw Cooling Water System (ERCW) is described in UFSAR Section 9.2.1.

ERCW provides assured cooling water for various heat exchangers during all phases of station operation. A review of its applications indicated no change in requirements for MUR power uprate. The design analyses, which credit ERCW, bound operation under the MUR power uprate. Therefore, the WBN Unit 2 MUR power uprate has no impact on ERCW or any of its major components and thus will have no impact on the system safety functions and regulatory requirements.

VI.1.C.iii Ultimate Heat Sink The Ultimate Heat Sink (UHS) is described in UFSAR Section 9.2.5. The UHS for WBN is the Tennessee River and the associated dams near the station and the plant intake channel. The Tennessee River provides at least 30 days of cooling water to dissipate the waste heat rejected during a unit LOCA plus a unit cooldown. The WBN Unit 2 MUR power uprate does not impact the design basis heat loads to the UHS, nor does it change the UHS flow rates or pumping requirements. Therefore, the UHS design requirements are not impacted by the WBN Unit 2 MUR power uprate.

VI.1.C.iv Residual Heat Removal The Residual Heat Removal System (RHR) heat exchanger and RHR pump seal water heat exchanger are cooled by CCS. The current RHR cooldown AOR assumed a bounding power level (3459 MWt, core) and no changes to the analysis are required. Therefore, the RHR heat exchanger heat load is not changed from the currently analyzed value. Similarly, there is no change to the required RHR pump operation and the RHR heat exchanger loads are not changed. Therefore, there is no impact to this system due to the WBN Unit 2 MUR power uprate.

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Enclosure 2 VI.1.D Spent fuel pool storage and cooling systems

RESPONSE

The SFPCS is described in UFSAR Section 9.1.3. The current analysis for the SFP is performed to a bounding design basis heat load, which is used to determine acceptable offload times for each cycle to ensure that the heat load limits are not exceeded. Since the criteria for fuel pool operation are based on heat load and re-evaluated for each offload, the operating requirements will accommodate power uprate. The analyzed design basis heat load will bound the post-MUR power uprate heat loads based on the selected offload delay time, in accordance with plant procedures. The system will continue to perform its design functions of spent fuel decay heat removal and maintaining purity and optical clarity of SFP water after the MUR power uprate. Therefore, there is no impact to the SFPCS due to the WBN Unit 2 MUR power uprate.

VI.1.E Radioactive waste systems

RESPONSE

Radioactive waste management systems are described in UFSAR Chapter 11. These systems provide the means to sample, collect, process, temporarily hold, and discharge, as necessary, gaseous and liquid low-level effluents generated during normal operation.

As discussed in UFSAR Section 11.1, Source Terms, radioactivities in waste management systems and components are determined using ANSI/ANS-18.1-1984 standardized activities adjusted to match WBN plant parameters. A key WBN plant parameter presented in this section is the core thermal power. Prior to the implementation of the WBN Unit 2 TPC, the core thermal power parameter used is 3582 MWt. Following the implementation of the WBN Unit 2 TPC, 3480 MWt is used. Both core thermal power levels, prior to and following WBN Unit 2 TPC implementation, bound the WBN Unit 2 MUR power uprate value.

Additionally, the gaseous waste disposal system was evaluated for a gas decay tank rupture event and that event assumes a core thermal power of 3565 MWt, which also bounds the MUR power uprate. The expected liquid releases were also based on the guidance in NUREG-0017, Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors, adjusted to meet WBN parameters, including a reactor core thermal power of 3582 MWt prior to WBN Unit 2 TPC implementation and 3480 MWt following WBN Unit 2 TPC implementation. The WBN liquid waste processing system is also credited for performing containment isolation for mitigating design basis events. This function is not affected by the MUR power uprate because design basis events were analyzed at 102% of the CLTP, or higher, which bounds the MUR power uprate. Therefore, the gaseous waste disposal and liquid waste processing systems are not impacted by the MUR power uprate.

The WBN Solid Waste System (SWS), described in UFSAR Section 11.5, is designed to contain solid radioactive waste materials as they are produced in the station, and to provide for their temporary onsite storage and preparation for eventual shipment to an offsite disposal facility.

The SWS is a shared system between both units and was previously evaluated for MUR power uprate conditions as part of the WBN Unit 1 MUR power uprate. The operation of this system is primarily influenced by the volume of waste processed, which is not expected to change because of the 1.4% uprate condition, consistent with WBN Unit 1. SWS functions and design bases are unaffected by the MUR power uprate and the associated analyses are not power level dependent. The SWS is typically operated in a batch mode and the only potential effect of the uprate is a slight increase in the frequency at which the batches may be processed.

Therefore, the WBN SWS remains acceptable for MUR power uprate conditions.

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Enclosure 2 VI.1.F Engineered safety features (ESF) heating, ventilation, and air conditioning (HVAC) systems

RESPONSE

The Control Building HVAC and Air Cleanup system is described in UFSAR Section 9.4.1. The Control Building HVAC and Air Cleanup system is designed to maintain the temperature and humidity in the building for personnel comfort, protection, and operation of plant controls, and to provide safe, uninterrupted occupancy of the MCR habitability zone during normal, accident, and post-accident recovery conditions. The ventilation systems that serve the Control Building are not expected to experience any significant heat load increases as a result of MUR power uprate. The control building served by the Control Building HVAC and Air Cleanup system does not contain any reactor power level dependent piping. Additionally, design heat loads associated with lighting, electrical equipment, and personnel will not be changing with power uprate, and no new equipment will be installed. The ventilation systems, which provide clean, filtered air, are not impacted by MUR power uprate. Therefore, the Control Building HVAC and Air Cleanup system is not impacted as a result of WBN Unit 2 MUR power uprate implementation.

The Auxiliary Building ventilating system is described in UFSAR Section 9.4.3. The Auxiliary Building ESF equipment coolers are described in UFSAR Section 9.4.5.3 and the fuel handling area ventilation system, a subsystem of the Auxiliary Building HVAC system, is described in UFSAR Section 9.4.2. The Auxiliary Building HVAC serves all areas of the Auxiliary Building including the fuel handling area and the radwaste areas. Separate subsystems are utilized for the environmental control of the shutdown board rooms, auxiliary board rooms, and other miscellaneous rooms and laboratories.

Post-LOCA containment response for MUR power conditions is bounded by the existing containment analysis. Safety-related equipment, piping, and transmission heat loads consider bounding post-LOCA containment response and therefore, there is no change to these cooling loads and the Auxiliary Building ESF room coolers are unaffected by the WBN Unit 2 MUR power uprate. Similarly, there is no change to the operation of the emergency diesel generators or batteries and there are no changes to equipment in the shutdown board room and 480 V board and battery rooms due to MUR power uprate.

Therefore, MUR power uprate has no impact on the heat loads in these areas and the associated HVAC subsystem are not impacted by WBN Unit 2 MUR power uprate.

The other general areas in the Auxiliary Building are not significantly affected by MUR power uprate because there is no impact to the piping or equipment heat loads in the rooms. The transmission heat loads from adjacent structures (i.e., the Turbine Building and Reactor Building) may increase slightly for the general areas of the Auxiliary Building but the impact on the Auxiliary Building HVAC is negligible.

Turbine Building heat loads will increase slightly with MUR power uprate due to increased condensate, feedwater, heater drains, and extraction steam. However, the Turbine Building area ventilation system is not an ESF HVAC system. Additionally, process temperatures for high temperature piping/equipment increase by a maximum of about 2°F and the corresponding impact on area temperatures is expected to be minimal. Therefore, the WBN Unit 2 MUR power uprate has negligible impact on the Turbine Building ventilation system and the associated increase in heat transfer from the Turbine Building to adjacent structures is considered to be negligible.

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Enclosure 2 The DG Building ventilation system is described in UFSAR Section 9.4.5.2. The DG Building HVAC is designed to provide adequate ventilation to the DG Building spaces to maintain the required environmental conditions for safety-related equipment, and prevent hydrogen buildup in the battery area during normal operation and design basis event conditions. The heat loads in the DG building are not reactor power level dependent. Additionally, no new equipment is required for MUR power uprate that is powered from the emergency diesel generators. The LOCA analysis is performed at a power level that bounds the MUR power uprate. Therefore, there is no change to the DG Building HVAC system.

The Reactor Building Purge Ventilating System (RBPVS) and Containment Air Cooling System (CACS) are described in UFSAR Sections 9.4.6 and 9.4.7, respectively. The RBPVS is designed to maintain the environment in the primary containment and Shield Building annulus within acceptable limits for equipment operation and for personnel access during inspection, testing, maintenance, and refueling operations; and to provide a filtration path for any through-duct outleakage from the primary containment to limit the release of radioactivity to the environment. The purge function of the RBPVS is not a safety-related function. The safety-related RBPVS isolation functions are not affected by the MUR power uprate. The containment air cooling systems are designed to maintain acceptable temperatures within the Reactor Building upper and lower compartments, reactor well, control rod drive mechanism (CRDM) shroud, and instrument room for the protection of equipment and controls during normal reactor operation and normal shutdown.

Except for the main steam and feedwater piping areas, areas served by the WBN Unit 2 Reactor Building HVAC systems will not experience any increase in heat loads or temperature with MUR power uprate. The feedwater piping temperature increases by 1.9°F whereas the main steam piping temperature is expected to decrease slightly (by 1.3°F). The main steam piping temperature decreases due to the slight reduction in SG dome pressure for MUR power uprate conditions. The average RCS temperature (Tavg) is not expected to change for MUR power uprate conditions. The cold leg temperature will be slightly colder (by 0.5°F) and the hot leg temperature will be slightly hotter (by 0.5°F) for MUR uprate conditions. Therefore, the impact to heat load and area temperatures containing RCS piping, the SGs, Pressurizer, and associated components and piping is expected to be insignificant.

Therefore, the overall impact of MUR power uprate on the WBN Unit 2 Reactor Building temperature is expected to be insignificant. Similar to the Turbine Building, the potential impact to adjacent structures is also considered to be negligible.

The post-LOCA containment response for MUR power uprate conditions is bounded by the existing containment analysis. Therefore, there is no impact to the post-LOCA Reactor Building HVAC heat loads.

The EGTS is addressed in Section VI.1.B.iv and the containment combustible gas control system is addressed in Section VI.1.B.iii.

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Enclosure 2 VII OTHER VII.1 A statement confirming that the licensee has identified and evaluated operator actions that are sensitive to the power uprate, including any effects of the power uprate on the time available for operator actions.

RESPONSE

The proposed MUR power uprate will be implemented under the TVA design change process.

The design change process ensures any impacted normal, abnormal, and emergency operating procedures having operator actions are revised prior to the implementation of the MUR power uprate, if required. Time Critical Operator Actions (TCOAs) are associated with the mitigation of postulated events. These actions must be performed in a specified time in order to assure the plant complies with assumptions made during the analysis of these postulated events. No change to TCOAs are expected to be required for MUR power uprate implementation because there are no changes to the associated analyses of record, as shown in Section II of this enclosure.

VII.2 A statement confirming that the licensee has identified all modifications associated with the proposed power uprate, with respect to the following aspects of plant operations that are necessary to ensure that changes in operator actions do not adversely affect defense in depth or safety margins:

VII.2.A Emergency and abnormal operating procedures

RESPONSE

The proposed MUR power uprate will be implemented under the TVA design change process.

TVA procedures provide the controls relevant to identifying impacts to procedures, controls, displays, alarms, the ICS and other operator interfaces, the simulator, and training. The design change process ensures any impacted emergency and abnormal operating procedures are revised prior to the implementation of the power uprate.

VII.2.B Control room controls, displays (including the safety parameter display system) and alarms

RESPONSE

A review of plant systems has indicated that only minor modifications are necessary (e.g., software modification that redefines the new 100% RTP). As part of the TVA design change process, instrument loop scaling updates for the MUR service conditions will be implemented for the following functions: Steam Generator Main Feedwater Flow, Temperature and Pressure; Main Feedwater Header Pressure and Flow; Pressurizer Pressure; and Turbine Impulse Pressure.

An LEFM system status indication already exists for the control room as an ICS screen that alerts the operator when there is a problem with the LEFM. The LEFM system status will indicate NORMAL, ALERT (indicating the system is in MAINTENANCE mode), or FAIL statuses. Any status other than NORMAL will generate both a visual and audible ICS alarm.

This is consistent with alarms for various other ICS inputs (i.e., ICS points). As part of the TVA design change process, the alarm will be updated, as necessary, to support the use of the LEFM for continuous calorimetric power determination to support the MUR power uprate.

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Enclosure 2 VII.2.C Control room plant reference simulator

RESPONSE

A review of the plant simulator will be conducted, and necessary changes made, under the TVA design change process.

VII.2.D Operator training program

RESPONSE

As part of the normal TVA process for implementing license amendment requests, operator training on the plant changes required to support the MUR power uprate will be completed prior to MUR power uprate implementation.

Similarly, training on operation and maintenance of the Caldon LEFM CheckPlus System, will be updated, as necessary, prior to implementation of the MUR power uprate.

VII.3 A statement confirming licensee intent to complete the modifications identified in Item 2.

above (including the training of operators), prior to implementation of the power uprate.

RESPONSE

The changes/modifications to the simulator and the associated manuals and instructional materials will be implemented in accordance with the TVA design change process to capture the plant changes resulting from the MUR power uprate. As part of the TVA design change process, TVA will complete all modifications identified in Section VII.2.B related to the MUR power uprate and complete the training of operators, prior to implementation of the power uprate.

VII.4 A statement confirming licensee intent to revise existing plant operating procedures related to temporary operation above full steady-state licensed power levels to reduce the magnitude of the allowed deviation from the licensed power level. The magnitude should be reduced from the pre-power uprate value of 2 percent to a lower value corresponding to the uncertainty in power level credited by the proposed power uprate application.

RESPONSE

Operating procedures have been reviewed and required changes will be documented and implemented as part of the TVA design change process including the procedure related to temporary operation above full steady-state licensed power levels.

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Enclosure 2 VII.5 A discussion of the 10 CFR 51.22 criteria for categorical exclusion for environmental review including:

VII.5.A A discussion of the effect of the power uprate on the types or amounts of any effluents that may be released offsite and whether or not this effect is bounded by the final environmental statement and previous Environmental Assessments for the plant.

RESPONSE

VII.5.A.i Non-Radiological Effluents Limits for non-radiological discharges from WBN to the environment are defined in National Pollutant Discharge Elimination System (NPDES) Permit No. TN0020168 (Reference VII.a).

The NPDES Permit identifies both chemical and thermal discharge limits for the plant.

Chemical discharge: The MUR power uprate will not change chemical discharges controlled by the NPDES permit. No changes in the types or amounts of effluents released into the environment will occur due to the power uprate.

Regarding thermal discharge, cooling tower blowdown has a maximum thermal discharge limit of 95°F. The increased heat load on the cooling towers due to MUR power uprate will have a negligible impact on cooling tower basin temperature (approximately 0.1°F). Supplemental CCW taken from the Tennessee River and gravity fed to the cooling tower basins provides for further blowdown dilution, as necessary. Therefore, the discharge and mixing zone thermal limits can continue to be met under MUR power uprate conditions. Thermal discharge will continue to comply with the NPDES requirements.

VII.5.A.ii Radiological Effluents:

During normal operations, the controls for the release rates of radwaste systems do not change with operating power. Thus, no impact on routine releases is anticipated due to the MUR power uprate. Actual, measured doses due to normal effluent associated with the reactor operating at the CLTP are documented in the annual radioactive effluent release reports. A review of historical liquid and gaseous release data indicates that resultant doses are a small fraction of annual limits. The effluent doses are determined in accordance with the ODCM. The ODCM methodologies ensure that doses to the public remain within regulatory dose limits and are as low as reasonably achievable (ALARA). The MUR power uprate will not result in changes to the ODCM.

The doses after WBN Unit 2 power uprate would be similar to those for WBN Unit 1 and would not result in significant doses compared to regulatory dose limits. Therefore, based on review of the annual effluent release reports and operating experience with WBN Unit 1, the MUR power uprate for WBN Unit 2 will not cause doses from liquid and gaseous effluent releases to exceed allowable limits.

VII.5.B A discussion of the effect of the power uprate on individual or cumulative occupational radiation exposure.

RESPONSE

A significant increase in individual and cumulative occupational radiation exposure is not expected because the MUR power uprate is bounded by the existing analyses of record at 102% of the CLTP, or higher, as discussed in Sections II and III of this enclosure. Additionally, similar to WBN Unit 1, individual worker exposures will continue to be maintained within the regulatory and administrative dose limits by the WBN radiation protection and ALARA programs.

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Enclosure 2 Thus, no adverse impact on radiological doses for plant personnel is anticipated as a result of the WBN Unit 2 MUR power uprate.

VII.6 Programs and Generic Issues VII.6.A Fire Protection Program

RESPONSE

The effects of MUR power uprate on the Appendix R fire safe shutdown analysis (SSA) and fire protection program at WBN Unit 2 were reviewed. The Fire Protection Report (FPR) continues to be valid under power uprate conditions. The MUR power uprate does not alter the components or component functions used in achieving and maintaining post-fire safe shutdown, and the SSA continues to be valid. The results of the Appendix R RCS cooldown analysis remain acceptable and the plant will still be able to achieve cold shutdown within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The MUR power uprate does not cause environmental conditions in the plant to interfere with the performance of post-fire safe shutdown operator actions. No new operator actions were identified. The crediting of fire protection design features as outlined in the FPR remains valid under MUR power uprate conditions. The crediting of administrative controls as outlined in the FPR remains valid under MUR power uprate conditions. In addition, the crediting of the High Pressure Fire Protection (HPFP) system water supply to provide a backup source for non-fire protection purposes, as described in the UFSAR, is not adversely impacted under MUR power uprate conditions.

VII.6.B Containment Coatings Program

RESPONSE

Conformance to RG 1.54, Rev. 0, Quality Assurance Requirements for Protective Coatings Applied to Water-Cooled Nuclear Power Plants, for paints and coatings inside containment is discussed in WBN UFSAR Section 6.1.4.

The impact of MUR power uprate on Service Level I, II, and III coatings was reviewed, along with the impact of power uprate on the GSI-191 treatment of coatings in containment. Because the current design basis accident (DBA) pressure/temperature and dose analyses remain applicable for MUR power uprate, currently DBA-qualified Service Level I coatings remain qualified following MUR power uprate and the quantity of unqualified coatings does not increase. Furthermore, the GSI-191 treatment of coatings (i.e., the quantity of failed qualified coatings within the zone of influence for a given break size) is not power level dependent. The impact of Service Level II and III coatings on plant operation is also unchanged with MUR power uprate. The MUR power uprate does not necessitate any changes to the coatings program.

VII.6.C Maintenance Rule Program

RESPONSE

10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance in Nuclear Power Plants, provides requirements for monitoring the performance or condition of SSCs in a manner sufficient to provide reasonable assurance that applicable SSCs are capable of fulfilling their intended functions. TVA has a Maintenance Rule Program for WBN. The proposed MUR power uprate for WBN Unit 2 does not have any impact to the programmatic aspects of the Maintenance Rule Program. It does not change any of the regulatory requirements of the program or in any way change the scope of the program. It does not add or delete any systems because the LEFM system is not within the scope of the Maintenance Rule Program.

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Enclosure 2 VII.6.D Motor- and Air-Operated Valve Programs

RESPONSE

The Motor-Operated Valve (MOV) program for WBN Unit 2 is not impacted by the MUR power uprate. It does not change any of the regulatory requirements or change the scope of the program. The systems that contain MOVs within the scope of the program were evaluated and determined to remain within existing design parameters after implementation of the MUR power uprate or were determined to not be impacted by the power uprate. The MUR power uprate does not alter the basis, scope, or content of the MOV Program. No MOVs will be added or deleted from the program due to the MUR power uprate. No maintenance or material changes for any MOVs will be required.

The Air-Operated Valve (AOV) Program for WBN Unit 2 Category 1 and 2 valves is not impacted by the MUR power uprate. The systems that contain these AOVs within the scope of the program were evaluated and determined to remain within existing design parameters after implementation of the MUR power uprate or were determined to not be impacted by the power uprate. The MUR power uprate does not alter the basis, scope, or content of the AOV Program.

No AOVs will be added or deleted from the program due to the MUR power uprate. No maintenance or material changes for any AOVs will be required.

VII.6.E Containment Leakage Rate Testing Program

RESPONSE

The Containment Leakage Rate Testing Program for WBN Unit 2 is discussed in WBN Unit 2 TS 5.7.2.19. The MUR power uprate does not have any impact on the programmatic aspects of this program. It does not change any of the regulatory requirements of the program or change the scope of the program. The MUR power uprate does not impact the post-accident containment response since the associated analyses assumed an initial power level that bounds MUR power uprate conditions, as discussed in Sections II and III of this enclosure.

References for Section VII:

VII.1. Tennessee Department of Environment and Conservation (TDEC), modified NPDES Permit No. TN0020168, TVA - Watts Bar Nuclear Plant, issued September 1, 2011 (with corrected pages and clarifications from October 2011 and November 2011, respectively, as explained in the TVA notification letter to the NRC, dated December 27, 2011).

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Enclosure 2 VIII CHANGES TO TECHNICAL SPECIFICATIONS, PROTECTION SYSTEM SETTINGS, AND EMERGENCY SYSTEM SETTINGS VIII.1 A detailed discussion of each change to the plants technical specifications, protection system settings, and/or emergency system settings needed to support the power uprate:

VIII.1.A A description of the change

RESPONSE

The description of proposed WBN Unit 2 OL and TS changes is provided in Section 2.2 of .

There are no proposed changes to protection system settings or emergency system settings.

VIII.1.B Identification of analyses affected by and/or supporting the change

RESPONSE

The secondary side heat balance (or calorimetric) uncertainty has been revised to reflect the reduced uncertainty given use of the LEFMs. Site-specific calculations by Cameron for the accuracy of the installed LEFMs were compared against the assumed values used in the revised heat balance uncertainty analysis performed by Westinghouse. These analyses are discussed in Section I of this enclosure. (See Enclosures 5 and 6 for the analyses.)

VIII.1.C Justification for the change, including the type of information discussed in Section III, above, for any analyses that support and/or are affected by change.

RESPONSE

WBN Unit 2 is currently licensed for a RTP of 3411 MWt. A more accurate feedwater flow measurement supports an increase to 3459 MWt. The justification for the associated TS changes are provided in Enclosures 1 and 2 (Sections I through VII). The analyses supporting the change are provided in Enclosures 5 and 6.

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Enclosure 3 Proposed Operating License, Technical Specification, and Technical Specification Bases Changes (Unit 2 Markup)

CNL-19-082 E3-1

C. The license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act, and to the rules regulations, and orders of the Commission now or hereafter in effect, and is subject to the additional conditions specified or incorporated below.

(1) Maximum Power Level TVA is authorized to operate the facility at reactor core power levels not in excess of 345911 megawatts thermal.

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 31, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. TVA shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

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

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

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

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

(7) TVA shall fully implement and maintain in effect all provisions of the Commission approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p).

The TVA approved CSP was discussed in NUREG-0847, Supplement 28, as amended by changes approved by License Amendment No. 7.

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

Unit 2 Amendment No. 31, XX Facility Operating License No. NPF-96

Definitions 1.1 1.1 Definitions (continued)

QUADRANT POWER TILT QPTR shall be the ratio of the maximum upper excore RATIO (QPTR) detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to the (RTP) reactor coolant of 345911 MWt.

REACTOR TRIP SYSTEM The RTS RESPONSE TIME shall be that time interval from (RTS) RESPONSE TIME when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC.

SHUTDOWN MARGIN (SDM) SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

a. All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM; and

b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level.

SLAVE RELAY TEST A SLAVE RELAY TEST shall consist of energizing each slave relay and verifying the OPERABILITY of each slave relay.

The SLAVE RELAY TEST shall include, as a minimum, a continuity check of associated testable actuation devices.

(continued)

Watts Bar - Unit 2 1.1-6 Amendment XX

Reporting Requirements 5.9 5.9 Reporting Requirements (continued) 5.9.3 Radioactive Effluent Release Report

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

A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

The Radioactive Effluent Release Report covering the operation of the unit during the previous year shall be submitted prior to May 1 of each year in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1.

5.9.4 Reserved for Future Use 5.9.5 CORE OPERATING LIMITS REPORT (COLR)

a. Core operating limits shall be established prior to the initial and each reload cycle, or prior to any remaining portion of a cycle, and shall be documented in the COLR for the following:

LCO 3.1.4 Moderator Temperature Coefficient LCO 3.1.6 Shutdown Bank Insertion Limits LCO 3.1.7 Control Bank Insertion Limits LCO 3.2.1 Heat Flux Hot Channel Factor LCO 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor LCO 3.2.3 Axial Flux Difference LCO 3.9.1 Boron Concentration

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC. When an initial assumed power level of 102% RTP is specified in a previously approved method, 100.6% RTP may be used only when feedwater flow measurement (used as input for reactor thermal power measurement) is provided by the leading edge flowmeter (LEFM) as described in document number 10 listed below. When feedwater flow measurements from the LEFM are unavailable, the originally approved initial power level of 102% RTP (3411 MWt) shall be used. The approved analytical methods are, specifically those described in the following documents:

(continued)

Watts Bar - Unit 2 5.0-30 Amendment XX

Reporting Requirements 5.9 5.9 Reporting Requirements 5.9.5 CORE OPERATING LIMTS REPORT (COLR) (continued)

11. Caldon, Inc., Engineering Report-80P, Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM' System, Revision 0, March 1997; and Caldon Ultrasonics Engineering Report ER-157P-A, Supplement to Caldon Topical Report ER-80P: Basis for Power Uprates with an LEFM Check or LEFM CheckPlus System, Revision 8 and Revision 8 errata.

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

(continued)

Watts Bar - Unit 2 5.0-33 Amendment XX

MSSVs B 3.7.1 BASES APPLICABLE The MSSVs are assumed to have two active failure modes. The active SAFETY failure modes are spurious opening, and failure to reclose once opened.

ANALYSES (continued) The MSSVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO The accident analysis requires that five MSSVs per steam generator be OPERABLE to provide overpressure protection for design basis transients occurring at 100.62% RTP. The LCO requires that five MSSVs per steam generator be OPERABLE in compliance with Reference 2 and the DBA analysis.

The OPERABILITY of the MSSVs is defined as the ability to open upon demand within the setpoint tolerances to relieve steam generator overpressure, and reseat when pressure has been reduced. The OPERABILITY of the MSSVs is determined by periodic surveillance testing in accordance with the Inservice Testing Program.

This LCO provides assurance that the MSSVs will perform their designed safety functions to mitigate the consequences of accidents that could result in a challenge to the RCPB, or Main Steam System integrity.

APPLICABILITY In MODES 1, 2, and 3, five MSSVs per steam generator are required to be OPERABLE to prevent Main Steam System overpressurization.

In MODES 4 and 5, there are no credible transients requiring the MSSVs.

The steam generators are not normally used for heat removal in MODES 5 and 6, and thus cannot be overpressurized; there is no requirement for the MSSVs to be OPERABLE in these MODES.

ACTIONS The ACTIONS table is modified by a Note indicating that separate Condition entry is allowed for each MSSV.

With one or more MSSVs inoperable, action must be taken so that the available MSSV relieving capacity meets Reference 2 requirements.

(continued)

Watts Bar - Unit 2 B 3.7-3 Amendment XX

CST B 3.7.6 BASES APPLICABLE The limiting event for the condensate volume is the large feedwater line SAFETY break coincident with a loss of offsite power. Single failures that also ANALYSES affect this event include the following:

(continued)

a. Failure of the diesel generator powering the motor driven AFW pump to the unaffected steam generators (requiring additional steam to drive the remaining AFW pump turbine); and

b. Failure of the steam driven AFW pump (requiring a longer time for cooldown using only one motor driven AFW pump).

These are not usually the limiting failures in terms of consequences for these events.

A non-limiting event considered in CST inventory determinations is a break in either the main feedwater bypass line or AFW line near where the two join. This break has the potential for dumping condensate until terminated by operator action. This loss of condensate inventory is partially compensated for by the retention of steam generator inventory.

Because the CST is the preferred source of feedwater and is relied on almost exclusively for accidents and transients, the CST satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO As the preferred water source to satisfy accident analysis assumptions, the CST must contain sufficient cooling water to remove decay heat for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months following a reactor trip from 100.62% RTP, and then to cool down the RCS to RHR entry conditions, assuming a coincident loss of offsite power and the most adverse single failure. In doing this, it must retain sufficient water to ensure adequate net positive suction head for the AFW pumps during cooldown, as well as account for any losses from the steam driven AFW pump turbine, or before isolating AFW to a broken line.

The CST level required is equivalent to a usable volume of 200,000 gallons, which is based on holding the unit in MODE 3 for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , followed by a cooldown to RHR entry conditions at 50°F/hour.

This basis is established in Reference 4 and exceeds the volume required by the accident analysis.

The OPERABILITY of the CST is determined by maintaining the tank level at or above the minimum required level.

(continued)

Watts Bar - Unit 2 B 3.7-33 Amendment XX

Enclosure 4 Proposed Operating License, Technical Specification, and Technical Specification Bases Changes (Unit 2 Re-Typed)

CNL-19-082 E4-1

C. The license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act, and to the rules regulations, and orders of the Commission now or hereafter in effect, and is subject to the additional conditions specified or incorporated below.

(1) Maximum Power Level TVA is authorized to operate the facility at reactor core power levels not in excess of 3459 megawatts thermal.