CNL-21-016, License Amendment Request to Revise Technical Specifications to Change the Steam Generator Secondary Side Water Level to Accommodate the Replacement Steam Generators (WBN-TS-20-05)

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License Amendment Request to Revise Technical Specifications to Change the Steam Generator Secondary Side Water Level to Accommodate the Replacement Steam Generators (WBN-TS-20-05)
ML21061A347
Person / Time
Site: Watts Bar Tennessee Valley Authority icon.png
Issue date: 03/02/2021
From: Polickoski J
Tennessee Valley Authority
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
CNL-21-016, WBN-TS-20-05
Download: ML21061A347 (30)
v • d • e


Text

1101 Market Street, Chattanooga, Tennessee 37402 CNL-21-016 March 2, 2021 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:

Watts Bar Nuclear Plant, Unit 2 - License Amendment Request to Revise Technical Specifications to Change the Steam Generator Secondary Side Water Level to Accommodate the Replacement Steam Generators (WBN-TS-20-05)

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

This license amendment request (LAR) proposes to amend the WBN Unit 2 Technical Specifications (TS) to change the steam generator (SG) water level requirement in WBN Unit 2 TS Limiting Condition for Operation (LCO) 3.4.7.b, RCS Loops - MODE 5, Loops Filled, and WBN Unit 2 Surveillance Requirements (SR) 3.4.5.2, RCS Loops - MODE 3, SR 3.4.6.3, RCS Loops - MODE 4, and SR 3.4.7.2 from greater than or equal to () 6 percent (%) to 32%. The proposed change is needed to support the WBN Unit 2 Replacement SG (RSG) project scheduled for the WBN Unit 2 Cycle 4 Refueling Outage (U2R4), which is scheduled to commence in spring 2022.

The enclosure to this submittal provides a description of the proposed change, technical evaluation of the proposed change, regulatory evaluation, and a discussion of environmental considerations. Attachment 1 to the enclosure provides the existing WBN Unit 2 TS pages marked up to show the proposed change. Attachment 2 to the enclosure provides the existing WBN Unit 2 TS pages retyped to show the proposed change. Attachment 3 to the enclosure provides the existing WBN Unit 2 TS Bases marked up to show the proposed change. Changes to the existing TS Bases are provided for information only and will be implemented under the Technical Specification Bases Control Program.

U.S. Nuclear Regulatory Commission CNL-21-016 Page 2 March 2, 2021 TVA has determined that there are no significant hazard considerations associated with the proposed change and that the change qualifies for a categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9). In accordance with 10 CFR 50.91, Notice for Public Comment; State Consultation, TVA is sending a copy of this letter and the enclosure to the Tennessee Department of Environment and Conservation.

To support the schedule for the RSG project, TVA requests NRC approval of the proposed license amendment within one year from the date of the submittal with implementation prior to entering Mode 5 upon restart from the U2R4 Refueling Outage. The milestone of Mode 5 supports the enclosed proposed change to WBN Unit 2 TS LCO 3.4.7.b, which is applicable during Mode 5 with the reactor coolant system loops filled.

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

I declare under penalty of perjury that the foregoing is true and correct. Executed on this 2nd day of March 2021.

Respectfully, James T. Polickoski Director, Nuclear Regulatory Affairs

Enclosure:

Evaluation of Proposed Change cc (Enclosure):

NRC Regional Administrator - Region II NRC Project Manager - Watts Bar Nuclear Plant NRC Senior Resident Inspector - Watts Bar Nuclear Plant Director, Division of Radiological Health - Tennessee State Department of Environment and Conservation

Enclosure Evaluation of Proposed Change

Subject:

Watts Bar Nuclear Plant, Unit 2 - License Amendment Request to Revise Technical Specifications to Change the Steam Generator Secondary Side Water Level to Accommodate the Replacement Steam Generators (WBN-TS-20-05)

CONTENTS 1.0

SUMMARY

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

3.0 TECHNICAL EVALUATION

............................................................................................. 3 3.1 System Description ...................................................................................................... 3 3.2 Background Discussion for Individual TS/TS Surveillance Requirements .................... 3 3.2.1 TS 3.4.5 RCS Loops - Mode 3 ............................................................................... 3 3.2.2 TS 3.4.6 RCS Loops - Mode 4 ............................................................................... 4 3.2.3 TS 3.4.7 RCS Loops - Mode 5, Loops Filled .......................................................... 5 3.3 Technical Analysis ....................................................................................................... 6

4.0 REGULATORY EVALUATION

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

5.0 ENVIRONMENTAL CONSIDERATION

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

6.0 REFERENCES

...............................................................................................................11 ATTACHMENTS

1. Proposed TS Changes (Mark Ups) for WBN Unit 2
2. Proposed TS Changes (Final Typed) for WBN Unit 2
3. Proposed TS Bases Changes (Mark Ups for Information Only) for WBN Unit 2 CNL-21-016 E-1 of 11

Enclosure 1.0

SUMMARY

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

TVA is planning to replace the WBN Unit 2 Westinghouse Model D3 (Alloy 600) original steam generators (OSGs) with Westinghouse Model 68AXP (Alloy 690) replacement steam generators (RSG) during the WBN Unit 2 Cycle 4 Refueling Outage (U2R4).

Because of the differences between the RSGs and OSGs described in Section 2.2 to this enclosure, the narrow range measured water level point at which the steam generator (SG) tubes are covered differs from that described in the current WBN Unit 2 Technical Specifications (TS). The proposed amendment will change WBN Unit 2 TS Limiting Condition for Operation (LCO) 3.4.7.b, RCS Loops - MODE 5, Loops Filled, and WBN Unit 2 Surveillance Requirements (SR) 3.4.5.2, RCS Loops - MODE 3, SR 3.4.6.3, RCS Loops - MODE 4, and SR 3.4.7.2 from greater than or equal to ()

six (6) percent (%) to 32% to support the RSG project.

2.0 DETAILED DESCRIPTION

2.1 PROPOSED CHANGE

The proposed amendment revises the WBN Unit 2 TS as follows:

Attachment 1 to this enclosure provides the existing WBN Unit 2 TS pages marked up to show the proposed change. Attachment 2 to this enclosure provides the existing WBN Unit 2 TS pages retyped to show the proposed change. Attachment 3 to this enclosure provides the existing WBN Unit 2 TS Bases marked up to show the proposed change. Changes to the existing TS Bases are provided for information only and will be implemented under the Technical Specification Bases Control Program.

2.2 CONDITION INTENDED TO RESOLVE The existing WBN Unit 2 Alloy 600 OSGs are scheduled to be replaced with Alloy 690 RSGs, as was similarly done for WBN Unit 1 (References 1 and 2). The external envelope and interfaces with existing piping and support structures for the RSGs are the same as for the OSGs, except that the elevation of several RSG flow and level instrument taps (upper taps and lower narrow range taps) will increase by approximately five feet. The span between the upper taps and lower narrow range taps will remain approximately the same. Internally, the RSGs differ from the OSGs in several ways, including increased tube surface area, different tube material, different tube supports, and longer tube length. The tube length increases from just below the lower narrow range taps in the OSGs to more than four feet above the lower narrow range taps in the RSGs.

CNL-21-016 E-2 of 11

Enclosure Because of the differences between the RSGs and OSGs, the narrow range measured water level point at which the SG tubes are covered differs from that described in the current WBN Unit 2 TS, specifically, from 6% narrow range level (with the OSGs) to 32%

narrow range level (with the RSGs). Therefore, the proposed change revises the SG water level requirement for WBN Unit 2 TS LCO 3.4.7.b and SRs 3.4.5.2, 3.4.6.3, and 3.4.7.2 from 6% to 32%, following replacement of the SG, to resolve this condition.

The proposed change will assure that the secondary side water level in the RSGs is high enough to cover the tubes.

3.0 TECHNICAL EVALUATION

3.1 SYSTEM DESCRIPTION The WBN Unit 2 RSGs are essentially the same as the WBN Unit 1 RSGs described in Section 3.1 to Reference 3.

3.2 BACKGROUND

DISCUSSION FOR INDIVIDUAL TS/TS SURVEILLANCE REQUIREMENTS 3.2.1 TS 3.4.5 RCS Loops - Mode 3 WBN Unit 2 TS 3.4.5 requires that two Reactor Coolant System (RCS) loops be operable in Mode 3, and either:

a. Two RCS loops shall be in operation when the Rod Control System is capable of rod withdrawal; or
b. One RCS loop shall be in operation when the Rod Control System is not capable of rod withdrawal.

WBN Unit 2 SR 3.4.5.2 specifies that the SG secondary side narrow range water levels are 6% for the required RCS loops. The WBN Unit 2 TS 3.4.5 Bases state that the SGs provide a heat sink for removal of decay heat from the reactor vessel during Mode 3. Verifying that the secondary side narrow range water level of each SG is 6%

assures that the tubes will not become uncovered and that the SGs will be capable of performing their decay heat removal function.

In Mode 3, the primary function of the reactor coolant is removal of decay heat and transfer of this heat, via the SGs, to the secondary plant fluid. Whenever the reactor trip breakers (RTB) are in the closed position and the control rod drive mechanisms (CRDM) are energized, an inadvertent rod withdrawal from subcritical, resulting in a power excursion, is possible. Such a transient could be caused by a malfunction of the Rod Control System. In addition, a power excursion, due to the ejection of an inserted control rod, is possible with the RTBs closed or open. Such a transient could be caused by the mechanical failure of a CRDM.

Therefore, in Mode 3 with the RTBs in the closed position and Rod Control System capable of rod withdrawal, accidental control rod withdrawal from subcritical is postulated and requires at least two RCS loops to be operable and in operation to ensure that the accident analyses limits are met. For those conditions when the Rod Control System is not capable of rod withdrawal, two RCS loops are required to be CNL-21-016 E-3 of 11

Enclosure operable, but only one RCS loop is required to be in operation to be consistent with Mode 3 accident analyses.

SR 3.4.5.2 requires verification of SG operability. SG operability is verified by ensuring that the secondary side narrow range water level is high enough to cover the tubes. If the SG secondary side narrow range water level is too low, the tubes may become uncovered, and the associated loop may not be capable of providing the heat sink necessary for removal of decay heat.

Revision to WBN Unit 2 SR 3.4.5.2 is required to support operation with the RSGs in Mode 3. The proposed change to WBN Unit 2 SR 3.4.5.2 is discussed in Section 2 of this enclosure.

3.2.2 TS 3.4.6 RCS Loops - Mode 4 WBN Unit 2 TS 3.4.6 requires that two RCS loops be operable in Mode 4, and consist of either:

a. Any combination of RCS loops and residual heat removal (RHR) loops, and one loop shall be in operation when the Rod Control System is not capable of rod withdrawal; or
b. Two RCS loops, and both loops shall be in operation, when the Rod Control System is capable of rod withdrawal.

WBN Unit 2 SR 3.4.6.3 specifies that the SG secondary side narrow range water levels are 6% for the required RCS loops. The WBN Unit 2 TS 3.4.6 Bases state that the SGs provide a heat sink for removal of decay heat from the reactor vessel during Mode 4. Verifying that the secondary side narrow range water level of each SG is 6%

assures that the tubes will not become uncovered and that the SGs will be capable of performing their decay heat removal function.

In Mode 4, the primary function of the reactor coolant is the removal of decay heat and the transfer of this heat to either the SG secondary side coolant or the component cooling water via the RHR heat exchangers.

With the RTBs open and the rods not capable of withdrawal, RCS circulation is considered in the determination of the time available for mitigation of the accidental boron dilution event. The RCS and RHR loops provide this circulation. Whenever the RTBs are in the closed position and the CRDMs are energized, an inadvertent rod withdrawal from subcritical, resulting in a power excursion, is possible. Such a transient could be caused by a malfunction of the Rod Control System. In addition, a power excursion, due to the ejection of an inserted control rod, is possible with the RTBs closed or open. Such a transient could be caused by the mechanical failure of a CRDM.

Therefore, in Mode 4 with the RTBs in the closed position and Rod Control System capable of rod withdrawal, at least two RCS loops are required to be operable and in operation to ensure that the accident analyses limits are met. For those conditions when the Rod Control System is not capable of rod withdrawal, any combination of two RCS or RHR loops are required to be operable, but only one loop is required to be in operation to meet decay heat removal requirements.

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Enclosure SR 3.4.6.3 requires verification of SG secondary side water level. SG operability is verified by ensuring that the secondary side narrow range water level is high enough to cover the tubes. If the SG secondary side narrow range water level is too low, the tubes may become uncovered, and the associated loop may not be capable of providing the heat sink necessary for removal of decay heat.

Revision to WBN Unit 2 SR 3.4.6.3 is required to support operation with the RSGs in Mode 4. The proposed change to WBN Unit 2 SR 3.4.6.3 is discussed in Section 2 of this Enclosure.

3.2.3 TS 3.4.7 RCS Loops - Mode 5, Loops Filled WBN Unit 2 TS 3.4.7 requires that one RHR loop be operable and in operation in Mode 5 with the RCS loops filled, and either:

a. One additional RHR loop shall be OPERABLE; or
b. The secondary side water level of at least two SGs shall be 6% narrow range.

WBN Unit 2 SR 3.4.7.2 specifies that the SG secondary side narrow range water level is 6% in the required SGs. The WBN Unit 2 TS 3.4.7 Bases state that the SGs provide a backup to the RHR System for decay heat removal during Mode 5 with the RCS loops filled. Verifying that the secondary side narrow range water level of at least two SGs is 6% ensures that an alternate decay heat removal method is available if the second RHR loop is not operable.

In Mode 5, with the RCS loops filled, the primary function of the reactor coolant is the removal of decay heat and the transfer of this heat either to the SG secondary side coolant or to the component cooling water via the RHR heat exchangers. While the principal means for decay heat removal is via the RHR system, the SGs are specified as a backup means for redundancy. Even though the SGs cannot produce steam in this mode, the SGs are capable of being a heat sink due to their large contained volume of secondary water. As long as the SG secondary side water is at a lower temperature than the reactor coolant, heat transfer will occur. The rate of heat transfer is directly proportional to the temperature difference.

During Mode 5, RCS circulation is considered in the determination of the time available for mitigation of the accidental boron dilution event. The RHR loops provide this circulation.

The purpose of SR 3.4.7.2 is to require that at least two SGs be operable with secondary side narrow range water indication high enough to cover the tubes. Therefore, the acceptance criterion is to provide an indicated level that will ensure the tubes are covered.

Revision to WBN Unit 2 LCO 3.4.7.b and SR 3.4.7.2 is required to support operation with the RSGs in Mode 5, with the RCS loops filled. The proposed change to WBN Unit 2 TS LCO 3.4.7.b and SR 3.4.7.2 is discussed in Section 2 of this Enclosure.

The technical basis for the changes to WBN Unit 2 TS LCO 3.4.7.b and SRs 3.4.5.2, 3.4.6.3, and 3.4.7.2 is provided below.

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Enclosure

3.3 TECHNICAL ANALYSIS

A demonstrated accuracy calculation was performed applicable to each SG narrow range water level indication instrument loop. The demonstrated accuracy calculation summarizes the design inputs and uncertainty analyses used to establish the 32% of span value as the measured secondary side narrow range water level needed to cover the tubes for the RSGs during Mode 3, 4, and 5 for the Technical Specifications described above in Section 3.2. A summary of the approach and conclusions of the demonstrated accuracy calculation is presented in the paragraphs below.

Liquid level measurements of closed vessels based on differential pressure are subject to errors due to density changes in the vessel contents or the reference leg. These errors are caused by (1) pressure and temperature changes in the vessel resulting in a change in the differential pressure across the vessel, or (2) a temperature change in the environment around the reference leg. These errors, which are applicable to the SG narrow range water level function, have been calculated as described below.

To quantify the effects of process pressure variations from nominal calibrated conditions, the uncertainty as a percentage of the RSG narrow range level span was evaluated for the full range of reference leg temperatures and secondary side pressures. The evaluation demonstrated that as the reference leg temperature increases, the magnitude of the uncertainty also increases. As the secondary side pressure increases, the magnitude of the uncertainty was shown to decrease. The maximum normal process pressure uncertainty was therefore established for conditions associated with (1) maximum normal reference leg temperature, and (2) minimum secondary side pressure (atmospheric).

The RSG narrow range level channel uncertainty was determined by combining the uncertainty associated with the various temperature components (e.g., sensor drift, sensor temperature, indicator drift, indicator calibration accuracy) using the square root of the sum of the squares (SRSS) methodology. The methodology combines the uncertainty components for a narrow range level channel in an appropriate combination of those groups, which are statistically and functionally independent. Those uncertainties that are not independent are conservatively treated by arithmetic summation, and then systematically combined with the independent terms.

The minimum required RSG narrow range level (without the above uncertainties) as a percentage of span was determined by subtracting the height of the lower level tap above the tube sheet from the height of the highest tube in the RSG above the tube sheet, and then dividing that result by the distance between the upper and lower level taps.

To determine the minimum required narrow range level for the RSGs (including instrument uncertainties), the process pressure uncertainty and narrow range level channel uncertainty were added to the minimum narrow range level without uncertainties. The minimum required narrow range level surveillance requirement (including instrument uncertainties) was calculated to be 31.28%. The result was rounded to 32% to define an easily readable value on the indicator and add some additional conservatism.

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Enclosure

4.0 REGULATORY EVALUATION

4.1 APPLICABLE REGULATORY REQUIREMENTS AND CRITERIA WBN, Unit 2 was designed to meet the intent of the Proposed General Design Criteria for Nuclear Power Plant Construction Permits published in July, 1967. The WBN construction permit was issued in January 1973. The UFSAR, however, addresses the General Design Criteria (GDC) published as Appendix A to 10 CFR 50 in July 1971.

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

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

Criterion 13Instrumentation and control. Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.

Compliance with GDC 13 is described in Section 3.1.2.2 of the WBN UFSAR.

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

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

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

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

Criterion 20Protection system functions. The protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.

Compliance with GDC 20 is described in Section 3.1.2.3 of the WBN UFSAR.

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Enclosure Criterion 25Protection system requirements for reactivity control malfunctions.

The protection system shall be designed to assure that specified acceptable fuel design limits are not exceeded for any single malfunction of the reactivity control systems, such as accidental withdrawal (not ejection or dropout) of control rods.

Compliance with GDC 25 is described in Section 3.1.2.3 of the WBN UFSAR.

Criterion 34Residual heat removal. A system to remove residual heat shall be provided. The system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded.

Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

Compliance with GDC 34 is described in Section 3.1.2.4 of the WBN UFSAR.

Criterion 44Cooling water. A system to transfer heat from structures, systems, and components important to safety, to an ultimate heat sink shall be provided. The system safety function shall be to transfer the combined heat load of these structures, systems, and components under normal operating and accident conditions.

Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

Compliance with GDC 44 is described in Section 3.1.2.4 of the WBN UFSAR.

4.2 PRECEDENT The proposed change is similar to a previous license amendment approved for WBN Unit 1 (Reference 2). The approved TS changes for SG water level in support of the replacement of the WBN Unit 1 SGs are identical to the changes proposed in this request.

4.3 NO SIGNIFICANT HAZARDS CONSIDERATION The proposed amendment will amend the Watts Bar Nuclear Plant (WBN), Unit 2 Technical Specifications (TS) to change the steam generator (SG) water level requirement for the WBN Unit 2 TS Limiting Condition for Operation (LCO) 3.4.7.b and WBN Unit 2 Surveillance Requirements (SRs) 3.4.5.2, 3.4.6.3 and 3.4.7.2 from greater than or equal to () 6 percent (%) to 32% following installation of the replacement SGs (RSGs).

TVA has evaluated whether or not a significant hazards consideration is involved CNL-21-016 E-8 of 11

Enclosure with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below.

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

Response: No The accidents and transients of interest are those that may occur in Modes 3, 4, or 5 and that rely upon two of the SGs to be operable to provide a heat sink for the removal of decay heat from the reactor vessel. These events include an accidental control rod withdrawal from subcritical, ejection of a control rod, and accidental boron dilution. The TS SRs provide verification of SG water level, which demonstrates that the SG is operable and able to act as a heat sink.

The proposed revisions to WBN Unit 2 TS LCO 3.4.7.b and SRs 3.4.5.2, 3.4.6.3, and 3.4.7.2 reflect the change to the required minimum SG water level necessary to demonstrate operability of the RSGs. Therefore, because no initiating event mechanisms or operability requirements are being changed, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Therefore, the proposed change does not involve a significant increase in 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.

Operation in Modes 3, 4, or 5 with a SG water level of less than 32% of span is not an initiator of any of the accidents and transients described in the WBN dual-unit Updated Final Safety Analysis Report (UFSAR). This situation puts the plant into an LCO situation and requires that the plant initiate actions within a specified timeframe if SG operability cannot be restored within the specified timeframe. The change in the value of the SG water level reflects the differences between the original steam generators (OSGs) and the RSGs. The new value will be used in the same manner as the old one to assess the operability of the SGs.

Therefore, operation in Modes 3, 4, or 5 with a SG water level of less than 32% of span will not initiate an accident nor create any new failure mechanisms. The changes to the TS and SRs do not result in any event previously deemed not credible being made credible. The change will not result in more adverse conditions and is not expected to result in any increase in the challenges to safety systems.

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

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Enclosure

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

Response: No.

The proposed change to the affected TS and SRs revise the value of SG narrow range water level that is needed to demonstrate the operability of the SG to support operation with the RSGs. The change in the value of the SG water level reflects the differences between the OSGs and the RSGs. These changes assure that the required numbers of SGs are operable with secondary side narrow range water level indication high enough to cover the tubes. Therefore, the acceptance criterion is to provide an indicated level that will ensure the tubes are covered. Because the same acceptance criterion is being used for the RSGs as was used for the OSGs, there is no reduction in the margin of safety.

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

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

4.4 Conclusion The proposed change would accommodate the Unit 2 RSGs by modifying the TS and TS SRs necessary to demonstrate operability of the SGs to transfer heat produced by the fission process in the reactor core to the secondary side of the plant, specifically, with respect to measurement of the secondary side water level within the steam generators that ensures coverage of the tops of the SG tubes.

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

5.0 ENVIRONMENTAL CONSIDERATION

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

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Enclosure

6.0 REFERENCES

1. TVA Letter to NRC, Watts Bar Nuclear Plant (WBN) - Unit 1 - Proposed License Amendment Request Change No. WBN-TS-05-06 to Change the Steam Generator Secondary Side Water Level to Greater than or Equal to 32% Of Narrow Range, dated December 13, 2005 (ML053530127)
2. NRC Letter to TVA, Watts Bar Nuclear Plant, Unit 1 - Issuance of Amendment Regarding the Change in Steam Generator Narrow Range Level Requirements to Accommodate the Replacement Steam Generators at Watts Bar Nuclear Plant, Unit 1 (TAC No. MC9235), dated May 5, 2006 (ML060960075)
3. TVA letter to NRC, CNL-20-053, Application to Revise Watts Bar Nuclear Plant (WBN), Unit 1 Technical Specifications for Steam Generator Tube Inspection Frequency and to Adopt TSTF-510, Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection, (WBN-390-TS-20-012), dated July 17, 2020 (ML20199M346)

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ATTACHMENT 1 Proposed TS Changes (Mark Ups) for WBN Unit 2 CNL-21-016

RCS Loops - MODE 3 3.4.5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. One required RCS loop not C.1 Restore required RCS 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in operation, and reactor trip loop to operation.

breakers closed and Rod Control System capable of OR rod withdrawal.

C.2 De-energize all control 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> rod drive mechanisms (CRDMs).

D. All RCS loops inoperable. D.1 De-energize all CRDMs. Immediately OR AND No RCS loop in operation. D.2 Suspend all operations Immediately involving a reduction of RCS boron concentration.

AND D.3 Initiate action to restore Immediately one RCS loop to OPERABLE status and operation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.5.1 Verify required RCS loops are in operation In accordance with the Surveillance Frequency Control Program SR 3.4.5.2 Verify steam generator secondary side water levels In accordance with are 6% narrow range for required RCS loops. the Surveillance Frequency Control Program (continued)

Change to read: "greater than or equal to 32%"

Watts Bar - Unit 2 3.4-8 Amendment 36 ,

RCS Loops - MODE 4 3.4.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.6.1 Verify two RCS loops are in operation when the rod In accordance with control system is capable of rod withdrawal. the Surveillance Frequency Control Program SR 3.4.6.2 Verify one required RHR or RCS loop is in operation In accordance with when the rod control system is not capable of rod the Surveillance withdrawal. Frequency Control Program SR 3.4.6.3 Verify SG secondary side water levels are greater In accordance with than or equal to 6% narrow range for required RCS the Surveillance loops. Frequency Control Change "6%" to "32%" Program SR 3.4.6.4 Verify correct breaker alignment and indicated power In accordance with are available to the required pump that is not in the Surveillance operation. Frequency Control Program Watts Bar - Unit 2 3.4-11 Amendment 8, 36 ,

RCS Loops - MODE 5, Loops Filled 3.4.7 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.7 RCS Loops - MODE 5, Loops Filled LCO 3.4.7 One residual heat removal (RHR) loop shall be OPERABLE and in operation, and either:

a. One additional RHR loop shall be OPERABLE; or
b. The secondary side water level of at least two steam generators (SGs) shall be greater than or equal to 6% narrow range.

Change "6%" to "32%"

NOTES------------------------------------------

1. One required RHR loop may be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing provided that the other RHR loop is OPERABLE and in operation.
2. No reactor coolant pump shall be started with one or more RCS cold leg temperatures less than or equal to the COMS arming temperature specified in the PTLR unless the secondary side water temperature of each SG is 50°F above each of the RCS cold leg temperatures.
3. All RHR loops may be removed from operation during planned heatup to MODE 4 when at least one RCS loop is in operation.

APPLICABILITY: MODE 5 with RCS loops filled.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RHR loop inoperable. A.1 Initiate action to restore a Immediately second RHR loop to AND OPERABLE status.

Required SGs secondary OR side water levels not within limits. A.2 Initiate action to restore Immediately required SG secondary side water levels to within limits.

(continued)

Amendment Watts Bar - Unit 2 3.4-12

RCS Loops - MODE 5, Loops Filled 3.4.7 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Required RHR loops B.1 Suspend all operations Immediately inoperable. involving a reduction of RCS boron concentration.

OR AND No RHR loop in operation.

B.2 Initiate action to restore Immediately one RHR loop to OPERABLE status and operation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.7.1 Verify one RHR loop is in operation. In accordance with the Surveillance Frequency Control Program SR 3.4.7.2 Verify SG secondary side water level is greater than In accordance with or equal to 6% narrow range in required SGs. the Surveillance Frequency Control Change "6%" to "32%" Program SR 3.4.7.3 Verify correct breaker alignment and indicated power In accordance with are available to the required RHR pump that is not in the Surveillance operation. Frequency Control Program Watts Bar - Unit 2 3.4-13 Amendment 36 ,

ATTACHMENT 2 Proposed TS Changes (Final Typed) for WBN Unit 2 CNL-21-016

RCS Loops - MODE 3 3.4.5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. One required RCS loop not C.1 Restore required RCS 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in operation, and reactor trip loop to operation.

breakers closed and Rod Control System capable of OR rod withdrawal.

C.2 De-energize all control 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> rod drive mechanisms (CRDMs).

D. All RCS loops inoperable. D.1 De-energize all CRDMs. Immediately OR AND No RCS loop in operation. D.2 Suspend all operations Immediately involving a reduction of RCS boron concentration.

AND D.3 Initiate action to restore Immediately one RCS loop to OPERABLE status and operation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.5.1 Verify required RCS loops are in operation In accordance with the Surveillance Frequency Control Program SR 3.4.5.2 Verify steam generator secondary side water levels In accordance with the are greater than or equal to 32% narrow range for Surveillance required RCS loops. Frequency Control Program (continued)

Watts Bar - Unit 2 3.4-8 Amendment 36,

RCS Loops - MODE 4 3.4.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.6.1 Verify two RCS loops are in operation when the rod In accordance with the control system is capable of rod withdrawal. Surveillance Frequency Control Program SR 3.4.6.2 Verify one required RHR or RCS loop is in operation In accordance with the when the rod control system is not capable of rod Surveillance withdrawal. Frequency Control Program SR 3.4.6.3 Verify SG secondary side water levels are greater In accordance with the than or equal to 32% narrow range for required RCS Surveillance loops. Frequency Control Program SR 3.4.6.4 Verify correct breaker alignment and indicated power In accordance with the are available to the required pump that is not in Surveillance operation. Frequency Control Program Watts Bar - Unit 2 3.4-11 Amendment 8, 36,

RCS Loops - MODE 5, Loops Filled 3.4.7 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.7 RCS Loops - MODE 5, Loops Filled LCO 3.4.7 One residual heat removal (RHR) loop shall be OPERABLE and in operation, and either:

a. One additional RHR loop shall be OPERABLE; or
b. The secondary side water level of at least two steam generators (SGs) shall be greater than or equal to 32% narrow range.

NOTES------------------------------------------

1. One required RHR loop may be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing provided that the other RHR loop is OPERABLE and in operation.
2. No reactor coolant pump shall be started with one or more RCS cold leg temperatures less than or equal to the COMS arming temperature specified in the PTLR unless the secondary side water temperature of each SG is 50F above each of the RCS cold leg temperatures.
3. All RHR loops may be removed from operation during planned heatup to MODE 4 when at least one RCS loop is in operation.

APPLICABILITY: MODE 5 with RCS loops filled.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RHR loop inoperable. A.1 Initiate action to restore a Immediately second RHR loop to AND OPERABLE status.

Required SGs secondary OR side water levels not within limits. A.2 Initiate action to restore Immediately required SG secondary side water levels to within limits.

(continued)

Watts Bar - Unit 2 3.4-12 Amendment

RCS Loops - MODE 5, Loops Filled 3.4.7 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Required RHR loops B.1 Suspend all operations Immediately inoperable. involving a reduction of RCS boron concentration.

OR AND No RHR loop in operation.

B.2 Initiate action to restore Immediately one RHR loop to OPERABLE status and operation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.7.1 Verify one RHR loop is in operation. In accordance with the Surveillance Frequency Control Program SR 3.4.7.2 Verify SG secondary side water level is greater than In accordance with the or equal to 32% narrow range in required SGs. Surveillance Frequency Control Program In accordance with the SR 3.4.7.3 Verify correct breaker alignment and indicated power Surveillance are available to the required RHR pump that is not in Frequency Control operation. Program Watts Bar - Unit 2 3.4-13 Amendment 36,

ATTACHMENT 3 Proposed TS Bases Changes (Mark Ups for Information Only) for WBN Unit 2 CNL-21-016

RCS Loops - MODE 3 B 3.4.5 BASES (continued)

ACTIONS D.1, D.2, and D.3 (continued)

If all RCS loops are inoperable or no RCS loop is in operation, except as during conditions permitted by the Note in the LCO section, all CRDMs must be de-energized by opening the RTBs or de-energizing the MG sets. All operations involving a reduction of RCS boron concentration must be suspended, and action to restore one of the RCS loops to OPERABLE status and operation must be initiated. Boron dilution requires forced circulation for proper mixing, and opening the RTBs or de-energizing the MG sets removes the possibility of an inadvertent rod withdrawal. The immediate Completion Time reflects the importance of maintaining operation for heat removal. The action to restore must be continued until one loop is restored to OPERABLE status and operation.

SURVEILLANCE SR 3.4.5.1 REQUIREMENTS This SR requires verification that the required loops are in operation.

Verification includes flow rate, temperature, and pump status monitoring, which help ensure that forced flow is providing heat removal. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Replace with: "greater than or equal to 32% (value SR 3.4.5.2 accounts for instrument error, Ref. 1)"

SR 3.4.5.2 requires verification of SG OPERABILITY. SG OPERABILITY is verified by ensuring that the secondary side narrow range water level is 6 % (value does not account for instrument error) for required RCS loops. If the SG secondary side narrow range water level is less than 6 %, the tubes may become uncovered and the associated loop may not be capable of providing the heat sink for removal of the decay heat. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.4.5.3 Change "6%" to "32%"

Verification that the required RCPs are OPERABLE ensures that safety analyses limits are met. The requirement also ensures that an additional RCP can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power availability to the required RCPs. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES None Watts Bar - Unit 2 B 3.4-24 Revision 34 Amendment 36

1. Watts Bar Drawing 2-47W605-242, "Electrical Tech Spec Compliance Tables"

RCS Loops - MODE 4 B 3.4.6 BASES SURVEILLANCE SR 3.4.6.2 REQUIREMENTS (continued) This SR requires verification that one required RCS or RHR loop is in operation when the rod control system is not capable of rod withdrawal.

Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Replace with: "32% (value accounts for SR 3.4.6.3 instrument error,"

Change "6%" to "32%"

SR 3.4.6.3 requires verification of SG OPERABILITY. SG OPERABILITY is verified by ensuring that the secondary side narrow range water level is greater than or equal to 6% (value does not account for instrument error, Ref. 1). If the SG secondary side narrow range water level is less than 6%, the tubes may become uncovered and the associated loop may not be capable of providing the heat sink necessary for removal of decay heat. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.4.6.4 Verification that the required pump is OPERABLE ensures that an additional RCS or RHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to the required pump. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES None

1. Watts Bar Drawing 2-47W605-242, "Electrical Tech Spec Compliance Tables" Watts Bar - Unit 2 B 3.4-30 Revision 8, 34 Amendment 8, 36

RCS Loops - MODE 5, Loops Filled B 3.4.7 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.7 RCS Loops - MODE 5, Loops Filled BASES BACKGROUND In MODE 5 with the RCS loops filled, the primary function of the reactor coolant is the removal of decay heat and the transfer of this heat to either the steam generator (SG) secondary side coolant or the component cooling water via the residual heat removal (RHR) heat exchangers.

While the principal means for decay heat removal is via the RHR System, the SGs are specified as a backup means for redundancy. Even though the SGs cannot produce steam in this MODE, they are capable of being a heat sink due to their large contained volume of secondary water. As long as the SG secondary side water is at a lower temperature than the reactor coolant, heat transfer will occur. The rate of heat transfer is directly proportional to the temperature difference. The secondary function of the reactor coolant is to act as a carrier for soluble neutron poison, boric acid.

In MODE 5 with RCS loops filled, the reactor coolant is circulated by means of two RHR loops connected to the RCS, each loop containing an RHR heat exchanger, an RHR pump, and appropriate flow and temperature instrumentation for control, protection, and indication.

One RHR pump circulates the water through the RCS at a sufficient rate to prevent boric acid stratification.

The number of loops in operation can vary to suit the operational needs.

The intent of this LCO is to provide forced flow from at least one RHR loop for decay heat removal and transport. The flow provided by one RHR loop is adequate for decay heat removal. The other intent of this LCO is to require that a second path be available to provide redundancy for heat removal.

The LCO provides for redundant paths of decay heat removal capability.

The first path can be an RHR loop that must be OPERABLE and in operation. The second path can be another OPERABLE RHR loop or maintaining two SGs with secondary side water levels greater than or equal to 6% narrow range to provide an alternate method for decay heat removal.

Change "6%" to "32%"

Watts Bar - Unit 2 B 3.4-31 (continued)

RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES (continued)

APPLICABLE In MODE 5, RCS circulation is considered in the determination of the time SAFETY available for mitigation of the accidental boron dilution event. The RHR ANALYSES loops provide this circulation.

RCS Loops - MODE 5 (Loops Filled) have been identified in 10 CFR 50.36(c)(2)(ii) as important contributors to risk reduction.

LCO The purpose of this LCO is to require that at least one of the RHR loops be OPERABLE and in operation with an additional RHR loop OPERABLE or two SGs with secondary side water level greater than or equal to 6% narrow range. One RHR loop provides sufficient forced circulation to perform the safety functions of the reactor coolant under these conditions.

Change "6%" to "32%" An additional RHR loop is required to be OPERABLE to meet single failure considerations. However, if the standby RHR loop is not OPERABLE, an acceptable alternate method is two SGs with their secondary side water levels greater than or equal to 6% narrow range.

Should the operating RHR loop fail, the SGs could be used to remove the decay heat. Change "6%" to "32%"

Note 1 allows one RHR loop to be inoperable for a period of up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, provided that the other RHR loop is OPERABLE and in operation. This permits periodic surveillance tests to be performed on the inoperable loop during the only time when such testing is safe and possible.

Note 2 requires that the secondary side water temperature of each SG be 50°F above each of the RCS cold leg temperatures before the start of a reactor coolant pump (RCP) with an RCS cold leg temperature less than the COMS arming temperature specified in the PTLR. This restriction is to prevent a low temperature overpressure event due to a thermal transient when an RCP is started.

Note 3 provides for an orderly transition from MODE 5 to MODE 4 during a planned heatup by permitting removal of RHR loops from operation when at least one RCS loop is in operation. This Note provides for the transition to MODE 4 where an RCS loop is permitted to be in operation and replaces the RCS circulation function provided by the RHR loops.

RHR pumps are OPERABLE if they are capable of being powered and are able to provide flow if required. An SG can perform as a heat sink when it has an adequate water level and is OPERABLE.

Watts Bar - Unit 2 B 3.4-32 (continued)

RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES (continued)

APPLICABILITY In MODE 5 with RCS loops filled, this LCO requires forced circulation of the reactor coolant to remove decay heat from the core and to provide proper boron mixing. One loop of RHR provides sufficient circulation for these purposes. However, one additional RHR loop is required to be OPERABLE, or the secondary side water level of at least two SGs is required to be 6% narrow range.

Change to: "greater Operation in other MODES is covered by: than or equal to 32%"

LCO 3.4.4, RCS Loops - MODES 1 and 2; LCO 3.4.5, RCS Loops - MODE 3; LCO 3.4.6, RCS Loops - MODE 4; LCO 3.4.8, RCS Loops - MODE 5, Loops Not Filled; LCO 3.9.5, Residual Heat Removal (RHR) and Coolant Circulation - High Water Level (MODE 6); and LCO 3.9.6, Residual Heat Removal (RHR) and Coolant Circulation -

Low Water Level (MODE 6).

ACTIONS A.1 and A.2 Change to: "less than 32%"

If one RHR loop is inoperable and the required SGs have secondary side water levels < 6% narrow range, redundancy for heat removal is lost.

Action must be initiated immediately to restore a second RHR loop to OPERABLE status or to restore the required SG secondary side water levels. Either Required Action A.1 or Required Action A.2 will restore redundant heat removal paths. The immediate Completion Time reflects the importance of maintaining the availability of two paths for heat removal.

B.1 and B.2 If no RHR loop is in operation, except during conditions permitted by Note 1, or if no loop is OPERABLE, all operations involving a reduction of RCS boron concentration must be suspended and action to restore one RHR loop to OPERABLE status and operation must be initiated. To prevent boron dilution, forced circulation is required to provide proper mixing and preserve the margin to criticality in this type of operation. The immediate Completion Times reflect the importance of maintaining operation for heat removal.

Watts Bar - Unit 2 B 3.4-33 (continued)

RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES (continued)

SURVEILLANCE SR 3.4.7.1 REQUIREMENTS This SR requires verification that the required loop is in operation.

Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.4.7.2 Verifying that at least two SGs are OPERABLE by ensuring their secondary side narrow range water levels are greater than or equal to 6%

(value does not account for instrument error) narrow range ensures an alternate decay heat removal method in the event that the second RHR loop is not OPERABLE. If both RHR loops are OPERABLE, this Surveillance is not needed. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.4.7.3 Change "6%" to "32%"

Verification that a second RHR pump is OPERABLE ensures that an additional pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to the RHR pump.

If secondary side water level is greater than or equal to 6% narrow range in at least two SGs, this Surveillance is not needed. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES None

1. Watts Bar Drawing 2-47W605-242, "Electrical Tech Spec Compliance Tables" Replace with: "(value accounts for instrument error, Ref. 1)"

Watts Bar - Unit 2 B 3.4-34 Revision 34 Amendment 36

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