L-PI-23-035, Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report

Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report

ML23354A155
Person / Time
Site:	Prairie Island
Issue date:	12/20/2023
From:	Conboy T Northern States Power Company, Minnesota, Xcel Energy
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
L-PI-23-035, EPID L-2022-LLA-0184
Download: ML23354A155 (11)

Text

(l Xcel Energy 1717 Wakonade Drive Welch, MN 55089 December 20, 2023 L-PI-23-035 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Prairie Island Nuclear Generating Plant, Units 1 and 2 Docket Nos. 50-282 and 50-306 Renewed Facility Operating License Nos. DPR-42 and DPR-60 Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) (EPID L-2022-LLA-0184)

References:

1) Letter L-PI-22-020, Application to Revise Technical Specifications Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR), dated December 2, 2022. (NRC ADAMS Accession No. ML22343A257)

2) NRC Email Re: Prairie Island Units 1 and 2 - Final Request for Additional Information re: LAR to Revise TS 5.6.6 Methods, dated December 13, 2023. (EPID l-2022-LLA-0184)

In Reference 1, Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter NSPM), submitted a license amendment request (LAR) proposing changes to the Technical Specifications (TS) for the Prairie Island Nuclear Generating Plant (PINGP). The proposed change would revise the TS 5.6.6 to allow more recent analytical methods that have been found acceptable by the NRC.

In Reference 2, the NRC identified that Section 2.4 of Enclosure 1 to Reference 1 includes,

...and Cold Overpressure Mitigating System setpoints... and that the mark-up and clean TS pages in Attachments 1 and 2, respectively, do not. Reference 2 requests additional information to clarify this discrepancy. Enclosure 1 to this letter responds to the NRC request for additional information. Enclosures 2 and 3 to this letter provides corrected TS markups and TS clean pages to replace Attachments 1 and 2, respectively, of Enclosure 1 of Reference 1.

The information provided with this letter does not alter the evaluations performed in accordance with 10 CFR 50.92 in Reference 1. In accordance with 10 CFR 50.91(b)(1), a copy of this application, with the enclosure, is being provided to the designated Minnesota official.

Document Control Desk L-Pl-23-035 Page 2 If there are any questions or if additional information is required, please contact Mr. Jeff Kivi at (612) 330-5788.

Summary of Commitments This letter makes no new commitments and no revisions to existing commitments.

I declare under penalty of perjury, that the foregoing is true and correct.

Executed on December 20, 2023.

Thomas A. Conboy Site Vice President, Prairie Island Nuc enerating Plant Northern States Power Company - Minnesota Enclosures cc: Administrator, Region Ill, USNRC Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC State of Minnesota

ENCLOSURE 1 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)

NRC Request for Additional Information Section 2.4, Description of Proposed Change, of NSPMs application states the proposed LAR would replace the current TS 5.6.6.b with (emphasis added):

The analytical methods used to determine the RCS pressure and temperature limits and Cold Overpressure Mitigating System setpoints shall be those previously reviewed and approved by the NRC, specifically those described in the following documents: , Proposed Technical Specification Changes (Mark-up), and Attachment 2, Proposed Technical Specification Changes (Clean), of the LAR state current TS 5.6.6.b would be replaced with:

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

In Section 2.4 of the LAR, the proposed change described above includes, and Cold Overpressure Mitigating System setpoints, but the mark-up and clean TS pages do not.

Please address the discrepancy between Section 2.4 and Attachments 1 and 2. If the mark-up and clean TS pages should be revised to match Section 2.4 of the LAR, please provide revised mark-up TS pages and clean TS pages with NSPMs response.

[Reference 2]

NSPM Response to RAI NSPM intended to include Cold Overpressure Mitigating System setpoints in the TS markups and clean pages included as Attachments 1 and 2, respectively, of Enclosure 1 to Reference

1. Enclosures 2 and 3 to this letter provide corrected TS markups and clean pages.

Page 1 of 2

L-PI-23-035 NSPM References

1. Letter L-PI-22-020, Application to Revise Technical Specifications Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR),

dated December 2, 2022. (NRC ADAMS Accession No. ML22343A257)

2. NRC Email Re: Prairie Island Units 1 and 2 - Final Request for Additional Information re: LAR to Revise TS 5.6.6 Methods, dated December 13, 2023. (EPID l-2022-LLA-0184)

Page 2 of 2

ENCLOSURE 2 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)

TECHNICAL SPECIFICATION PAGES (Marked-Up)

(2 pages follow)

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a. RCS pressure and temperature limits for heat-up, cooldown, low temperature operation, criticality, and hydrostatic testing, OPPS arming, PORV lift settings and Safety Injection Pump Disable Temperature as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

LCO 3.4.3, RCS Pressure and Temperature (P/T) Limits; LCO 3.4.6, RCS Loops - MODE 4; LCO 3.4.7, RCS Loops - MODE 5, Loops Filled; LCO 3.4.10, Pressurizer Safety Valves; LCO 3.4.12, Low Temperature Overpressure Protection (LTOP) -

Reactor Coolant System Cold Leg Temperature (RCSCLT) > Safety Injection (SI) Pump Disable Temperature; LCO 3.4.13, Low Temperature Overpressure Protection (LTOP) -

Reactor Coolant System Cold Leg Temperature (RCSCLT) < Safety Injection (SI) Pump Disable Temperature; and LCO 3.5.3, ECCS - Shutdown.

b. The analytical methods used to determine the RCS pressure and temperature limits and Cold Overpressure Mitigation System setpoints shall be those previously reviewed and approved by the NRC, specifically those described in the following document:

WCAP-14040-NP-A, Revision 2, Methodology Used to Develop Cold Insert A Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves (includes any exemption granted by NRC to ASME Code Case N-514).

c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

Changes to the curves, setpoints, or parameters in the PTLR resulting from new or additional analysis of beltline material properties shall be submitted to the NRC prior to issuance of an updated PTLR.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.6-7 Unit 2 - Amendment No. 226

Insert A The analytical methods used to determine the RCS pressure and temperature limits and Cold Overpressure Mitigation System setpoints shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. WCAP-14040-NP-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004.

2. WCAP-18124-NP-A, Revision 0, "Fluence Determination with RAPTOR-M3G and FERRET," July 2018, and WCAP-18124-NP-A, Revision 0, Supplement 1-NP-A, Revision 0, "Fluence Determination with Raptor-M3G and FERRET - Supplement for Extended Beltline Materials," May 2022, shall be used as an alternative to Section 2.2 of WCAP-14040-NP-A.

ENCLOSURE 3 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)

TECHNICAL SPECIFICATION PAGES (Re-typed)

(3 pages follow)

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a. RCS pressure and temperature limits for heat-up, cooldown, low temperature operation, criticality, and hydrostatic testing, OPPS arming, PORV lift settings and Safety Injection Pump Disable Temperature as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

LCO 3.4.3, RCS Pressure and Temperature (P/T) Limits; LCO 3.4.6, RCS Loops - MODE 4; LCO 3.4.7, RCS Loops - MODE 5, Loops Filled; LCO 3.4.10, Pressurizer Safety Valves; LCO 3.4.12, Low Temperature Overpressure Protection (LTOP) -

Reactor Coolant System Cold Leg Temperature (RCSCLT) > Safety Injection (SI) Pump Disable Temperature; LCO 3.4.13, Low Temperature Overpressure Protection (LTOP) -

Reactor Coolant System Cold Leg Temperature (RCSCLT) < Safety Injection (SI) Pump Disable Temperature; and LCO 3.5.3, ECCS - Shutdown.

b. The analytical methods used to determine the RCS pressure and temperature limits and Cold Overpressure Mitigating System setpoints shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. WCAP-14040-NP-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, May 2004.

2. WCAP-18124-NP-A, Revision 0, Fluence Determination with RAPTOR-M3G and FERRET, July 2018, and WCAP-18124-NP-A, Revision 0, Supplement 1-NP-A, Revision 0, Fluence Determination with Raptor-M3G and FERRET - Supplement for Extended Beltline Materials, May 2022, shall be used as an alternative to Section 2.2 of WCAP-14040-NP-A.

Prairie Island Unit 1 - Amendment No. 242 Units 1 and 2 5.6-7 Unit 2 - Amendment No. 230

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (continued)

c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

Changes to the curves, setpoints, or parameters in the PTLR resulting from new or additional analysis of beltline material properties shall be submitted to the NRC prior to issuance of an updated PTLR.

5.6.7 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.8, Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG;

b. The nondestructive examination techniques utilized for tubes with increased degradation susceptibility;

c. For each degradation mechanism found:

1. The nondestructive examination techniques utilized;

2. The location, orientation (if linear), measured size (if available),

and voltage response for each indication. For tube wear at support structures less than 20 percent through-wall, only the total number of indications needs to be reported;

3. A description of the condition monitoring assessment and results, including the margin to the tube integrity performance criteria and comparison with the margin predicted to exist at the inspection by the previous forward-looking tube integrity assessment; and

4. The number of tubes plugged during the inspection outage.

Prairie Island Unit 1 - Amendment No. 242 Units 1 and 2 5.6-8 Unit 2 - Amendment No. 230

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.8 Steam Generator Tube Inspection Report (continued)

d. An analysis summary of the tube integrity conditions predicted to exist at the next scheduled inspection (the forward-looking tube integrity assessment) relative to the applicable performance criteria, including the analysis methodology, inputs, and results;

e. The number and percentage of tubes plugged to date, and the effective plugging percentage in each SG; and

f. The results of any SG secondary side inspections.

5.6.8 EM Report When a report is required by Condition C or I of LCO 3.3.3, "Event Monitoring (EM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

Prairie Island Unit 1 - Amendment No. 242 Units 1 and 2 5.6-9 Unit 2 - Amendment No. 230