ML20247J559
| ML20247J559 | |
| Person / Time | |
|---|---|
| Issue date: | 09/03/2020 |
| From: | Michael Mahoney NRC/NRR/DORL/LPL2-1 |
| To: | Vaughan J Duke Energy Progress |
| Michael Mahoney-NRR/DORL 301-415-3867 | |
| References | |
| EPID L-2020-LLA-0111 | |
| Download: ML20247J559 (4) | |
Text
From:
Mahoney, Michael To:
Vaughan, Jordan L Cc:
Art Zaremba; Dennis Earp (dennis.Earp@duke-energy.com)
Subject:
Request for Additional Information - Shearon Harris Nuclear Power Plant, Unit 1 - Revise TS 3/4.4.9, Pressure/Temperature Limits - Reactor Coolant System (EPID L-2020-LLA-0111)
Date:
Thursday, September 03, 2020 8:39:53 AM Attachments:
image005.png image006.png
- Jordan,
By application dated May 12, 2020 (Agencywide Document Access and Management System Accession Number ML20134H888), Duke Energy Progress, LLC (Duke Energy, the licensee) submitted a license amendment request (LAR) to revise Technical Specifications (TS) 3/4.4.9, Reactor Coolant System [RCS] Pressure/Temperature [P/T] Limits of the Shearon Harris Nuclear Power Plant (Harris), Unit 1 to the U.S. Nuclear Regulatory Commission (NRC).
Specifically, the licensee proposed to revise the RCS P/T limits in Figure 3.4-2 Reactor Coolant System Cooldown Limitations - Applicable to 36 EFPY [effective full power years]
and in Figure 3.4-3 Reactor Coolant System Heatup Limitations - Applicable to 36 EFPY with curves that are applicable up to 55 EFPY. The NRC staff determined it needs additional information to complete its review of the LAR.
In order to complete its review, the U.S. Nuclear Regulatory Commission staff requests additional information. A clarification call to ensure mutual understanding was conducted on September 2, 2020. Please provide your response to the following requests for additional information (RAIs) within 30 days of the date of this correspondence.
Regulatory Basis
The NRC has established requirements in Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50) to protect the integrity of the reactor coolant pressure boundary in nuclear power plants. Specifically, the staff evaluates the acceptability of a facilitys proposed P/T limits based on Appendix G, Fracture Toughness Requirements, to 10 CFR Part 50. This regulation requires, in part, that facility P/T limits for the reactor pressure vessel (RPV) to be at least as conservative as those obtained by applying the linear elastic fracture mechanics methodology of Appendix G, Fracture Toughness Criteria for Protection Against Failure, to Section XI of the American Society for Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). The following requests for additional (RAIs) have to do with the calculations regarding the proposed 55 EFPY P/T limits.
RAI-1
Discussion
The NRC staff notes in Section 1.1.5 of the Harris Updated Final Safety Analysis Report (UFSAR), Amendment 62, that Harris has undergone a Measurement Uncertainty Recapture (MUR) uprate in 2012 that increased the reactor core thermal output from 2900 Megawatts Thermal (MWt) to 2948 MWt. However, the licensee did not state in the LAR application if the ART calculations included the effect of the MUR uprate in 2012.
Request Please confirm that the effect of the 2012 MUR and any other uprates have been included in the proposed 55 EFPY P/T limits.
RAI-2
Discussion The NRC staff notes that the last two rows of Table 3, Fast Neutron Fluence (E > 1 MeV) for the Reactor Vessel Clad-Base Metal Interface, of the enclosure to the LAR lists two fluence values at 55 EFPY of 4.04x1019 neutrons per square centimeter (n/cm2) and 9.99x1018 n/cm2 at the 1/4T at Maximum Peak Location and 3/4T at Maximum Peak Location, respectively, but did not state which RPV components these locations refer to.
Request Please clarify if the 1/4T at Maximum Peak Location and 3/4T at Maximum Peak Location refer to Intermediate Shell Plate Heat B4197-2.
Note: the NRC staffs calculated fluence values at the 1/4T and 3/4T locations for the Intermediate Shell Plate Heat B4197-2 are 4.32x1019 n/cm2 and 1.70x1019 n/cm2, respectively, based on the thickness of 7.75 inches given in Table 1 of the enclosure to the LAR, using the fluence attenuation formula (Equation 3) in NRC Regulatory Guide 1.99, Radiation Embrittlement of Reactor Vessel Materials, Revision 2.
RAI-3
Discussion The NRC staff notes that TS limiting condition for operation (LCO) 3.4.9.1 and 3.4.9.2 of TS 3/4.4.9 refer to the same heatup P/T limits in Figure 3.4-3 of TS 3/4.4.9. TS LCO 3.4.9.1 is applicable to operational modes 1, 2, 3 of the reactor (when the average coolant temperature is greater than or equal to 350°F as defined in Table 1.2, Operation Modes, of the Harris TS), and LCO 3.4.9.2 is applicable to Modes 4, 5, and 6. Modes 1 and 2 appear to be criticality modes (i.e., when the reactor core is critical); and Modes 3, 4, 5, and 6 appear to be non-criticality modes. The NRC staff notes that Figure 3.4-3 of TS 3/4.4.9 (both in the current figure and the proposed replacement figure) does not show the criticality limits required by 2.d (and 2.c for lower pressures) of Table 1, Pressure and Temperature Requirements for the Reactor Pressure Vessel, of Appendix G to 10 CFR 50 even though TS LCO 3.4.9.1, applicable to criticality Modes 1 and 2, refers to Figure 3.4-3 of TS 3/4.4.9.
The NRC staff plotted the values of temperature and pressure from Table 4 of the LAR enclosure for a heatup of 100°F per hour since 100°F per hour is the maximum heatup rate indicated in TS LCO 3.4.9.1. The NRC staff also plotted the required criticality curve (heatup of 100°F per hour plus 40°F) for temperatures greater than 350°F. The resulting plot is shown below. The below plot also indicates the 2485 pounds per square inch,
gauge (psig) setpoint of the pressurizer safety relief valves, as noted in Figure 3.4-3 of TS 3/4.4.9. As shown in the plot, for temperatures of 350°F to approximately 371°F, the allowable pressure for criticality is below the pressurizer safety relief valve setpoint of 2485 psig.
Request
Please confirm that:
- a. At indicated temperatures of 350°F to 371°F, the indicated pressure is not allowed to exceed the heatup 100°F per hour curve + 40°F shown in the plot above; and
- b. At indicated temperatures of 371°F and above, the indicated pressure does not exceed the pressurizer safety relief valve setpoint of 2485 psig.
Once this email is added to ADAMS, I will provide the accession number for your reference.
If you have any questions, please contact me.
Thanks
Mike Mahoney Project Manager, LPL2-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Desk: (301)-415-3867 Mobile: (301)-250-0450
Email: Michael.Mahoney@nrc.gov