ML24155A137

Email from K. Green to S. Hughes Request for Additional Information Related to License Amendment Request to Revise Residual Heat Removal Flow Rate

ML24155A137
Person / Time
Site:	Watts Bar
Issue date:	05/29/2024
From:	Kimberly Green Plant Licensing Branch II
To:	Tennessee Valley Authority
Green K
Shared Package
ML24155A139	List: ML24155A137
References
EPID L-2023-LLA-0152
Download: ML24155A137 (5)

Text

From: Kimberly Green To: Hughes, Shawna Marie

Subject:

Request for Additional Information Related to License Amendment Request to Revise Residual Heat Removal Flow Rate (EPID L-2024-LLA-0152)

Date: Wednesday, May 29, 2024 3:35:00 PM Attachments: Final RCIs and RAIs Clean - Redacted.pdf

Dear Shawna Hughes,

By letter dated October 30, 2023 (Agencywide Documents Access and Management System Accession No. ML23303A095), as supplemented by letter dated January 10, 2024 (ML24010A064), Tennessee Valley Authority (TVA) submitted a license amendment request (LAR) for the Watts Bar Nuclear Plant (Watts Bar), Units 1 and 2.The proposed amendments would revise Watts Bar, Units 1 and 2, Technical Specification (TS)

Surveillance Requirement 3.9.5.1 to reduce the residual heat removal (RHR) flow rate of 2,500 gallons per minute to 2,000 gallons per minute.

The U.S. Nuclear Regulatory Commission (NRC) staff is reviewing TVAs application and has identified areas where additional information is needed to complete its review. A draft request for confirmation of information (RCI) and request for additional information (RAI) was previously transmitted to you via the BOX - Enterprise File Synchronization and Sharing service (BOX-EFSS) application on May 16, 2024, because the file contained proprietary information (ML24150A155 non-proprietary version). At TVAs request, a clarification call was held on May 23rd, to clarify the NRC staffs requests. Based on the discussion during that call, the NRC staff determined that EMIB-RCI-2 can be deleted because information regarding the TVA valve numbers and their inclusion in the inservice testing program is in the LAR; EMIB-RAI-1 can be deleted because the draft audit report will be reviewed by TVA and Westinghouse for proprietary information prior to issuance; EMIB-RAI-2 can be deleted because the memos referenced in the request pertain to mid-loop operation, and the requested TS change applicability is Mode 6, during which mid-loop operation does not occur; and SNSB-RAI-1 can be deleted because the applicable TS bases explains that the conditional de-energizing of the RHR pump does not result in a challenge to the fission product barrier, and there is no specified boron concentration in the TS. Additionally, EMIB-RCI-1 is reworded to be more accurate; EMIB-RCI-3 and -4 are combined and reworded because the requests are similar; SNSB-RAI-2 is reworded to change from a quantitative to a qualitative explanation; and SNSB-RAI-3 a and b are combined and reworded to change from a quantitative to a qualitative explanation. Lastly, the requests are renumbered as appropriate. The redacted, non-proprietary version of the RAI is attached. The version containing proprietary information will be transmitted to you via the BOX-EFSS.

A response to the attached RCIs and RAIs (non-proprietary) is requested within 30 days of the date of this email. The NRC staff considers that timely responses to RAIs help ensure sufficient time is available for staff review and contribute toward the NRCs goal of efficient and effective use of staff resources. If circumstances result in the need to revise the requested response date, please me at (301)415-1627 or via email at Kimberly.Green@nrc.gov.

Kimberly Green Senior Project Manager Division of Operating Reactor Licensing U.S. Nuclear Regulatory Commission 301-415-1627

Official Use Only - Proprietary Information

REQUEST FOR ADDITIONAL INFORMATION APPLICATION TO MODIFY WATTS BAR NUCLEAR PLANT, UNIT 1 AND UNIT 2

TECHNICAL SPECIFICATION SURVEILLANCE REQUIREMENT 3.9.5.1

WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2 TENNESSEE VALLEY AUTHORITY

DOCKET NOS. 50-390 AND 50-391

=

Background===

On October 30, 2023 (Agencywide Documents Access and Management System (ADAMS) No.

23303A095), Tennessee Valley Authority (TVA) subm itted a license amendment request (LAR) for Watts Bar Nuclear Plant (Watts Bar), Units 1 and 2. The proposed amendments would revise Watts Bar, Units 1 and 2, Technical Specification (TS) Surveillance Requirement (SR) 3.9.5.1, Residual Heat Removal (RHR) and Coolant Circulation - High Water Level, to modify the required flow rate of 2500 gallons per minute (gpm) to 2000 gpm.

The U.S. Nuclear Regulatory Commission (NRC) staff conducted an audit of the proprietary documents supporting the LAR made available by the licensee in the Westinghouse electronic reading room (ML24071A098).

The NRC staff is reviewing the request and has identified areas where it needs confirmation of information and additional information to support its review.

Regulatory Requirements

The NRC regulations in 10 CFR 50.36, Technical spec ifications, require that each licensee of a nuclear power plant prepare technical specifications as part of its license in accordance with the requirements of this section of the NRC regulations.

The NRC regulations in 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, specify principal design criteria for nuclear power plants to establish the necessary design, fabrication, constructi on, testing, and performance requirements for structures, systems, and components important to safety. That is, structures, systems, and components that provide reasonable assurance that the facility can be operated without undue risk to the health and safety of the public. Watts Bar, Units 1 and 2, were designed to meet the intent of the Proposed General Design Criteria for Nuclear Power Plant Construction Permits published in July 1967, with the Watts Bar construction permits issued in January 1973. The Watts Bar Dual-Unit Updated Final Safety Analysis Report (UFSAR) addresses the General Design Criteria (GDC) specified in Appendix A to 10 CFR Part 50. The specific GDC applicable to this LAR are GDC 14, Reactor coolant pressure boundary, GDC 15, Reactor coolant system design, and GDC 34, Residual heat removal.

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Requests for Confirmation of Information (RCI)

EMIB-RCI-1

In LAR Enclosure 3 (Westinghouse Letter Report, LTR-SEE-23-NP, Revision 1), when discussing control valve cavitation, it states that the RHR flowrate is reduced during MODE 6 operation by fully closing the RHR bypass flow co ntrol valve (HCV-618); and then slowly closing the associated hand control valve (HCV-606 or 607). The report states that cavitation of the reactor coolant could result when the pressure drop across the control valve (HCV-618) increases as flow is reduced. The report notes that severe cavitation could cause excessive wear and vibration in the piping downstream of the control valve.

Confirm that this operational recommendation has been addressed in Watts Bar plant procedures for the hand control valve (HCV-606 or 607) to avoid cavitation and its potential consequences when reducing the RHR flowrate.

EMIB-RCI-2

LAR Enclosure 3 states that during the first operation with reduced RHR flowrates, the 10-inch check valve 8948 was locally monitored for chatter noise. Furthermore, in LAR Enclosure 2 (Westinghouse Letter Report, LTR-SEE-23-P, Revisi on 1), the Check Valve Chattering section states that ((

))

In that system operation and wear can cause changes in check valve performance, confirm that potential changes in the performance of check valves (such as chatter) during reduced RHR flow rates (i.e., 2000 gpm) will be addressed for (( )) RHRS operation.

Requests for Additional Information (RAI)

SNSB RAI-1

In LAR Enclosure 2 (Westinghouse Letter Report, LTR-SEE-23-4-P, Revision 1) under the heading Thermal Stratification, the last sentence of the first paragraph states:

Note that increased cavity levels 23 ft. will increase pressure in the core and DNB margin.

The report asserts that the higher water level 23 ft. in the cavity will increase pressure and DNB margin without further detail or elaboration.

Provide the basis for the quoted statement.

SNSB RAI-2

In the proprietary enclosure to the LAR (Westinghouse Letter Report, LTR-SEE-23-4-P, Revision

1) under the heading Thermal Stratification, the second paragraph states:

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((

)) Thus, it is possible for the local coolant temperature to exceed 200 °F and approach the point of nucleate boiling. However, for the worst-case scenario evaluated, it was concluded that DNB would not be a concern at the Watts Bar Units 1 and 2 at a reduced RHR flowrate during MODE 6 operation. The reactor coolant enters the reactor vessel from two cold leg nozzles, passes through the downcomer region and enters the lower plenum region. It is expected that the coolant is adequately mixed from the flow of two branch lines and therefore, the temperature across the core entrance is uniform. Thus, thermal stratification is minimized.

If the local coolant temperature exceeds 200 °F and approaches the point of nucleate boiling in the core, it could lead to both a loss of coolant through boil off which would cause boron to plate out and a reduction in overall boron concentration. The report asserts this is not a concern for reduced flow RHR operation at Watts Bar without providing any further supporting details.

Provide a qualitative explanation of how ((

)) As part of this explanation, for the reduced RHR flowrate in Mode 6, describe how the worst-case analyzed condition in which the reactor coolant locally could exceed 200 °F has sufficient departure from nucleate boiling margin.

Official Use Only - Proprietary Information