RA-20-0002, Response to Request for Additional Information (RAI) Regarding License Amendment Request to Revise Technical Specification 3.4.3, RCS Pressure and Temperature (P/T) Limits
| ML20044C924 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 02/13/2020 |
| From: | Simril T Duke Energy Carolinas |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RA-20-0002 | |
| Download: ML20044C924 (7) | |
Text
Tom Simril Vice President Catawba Nuclear Station Duke Energy CN01VP / 4800 Concord Road York, SC 29745 o: 803.701.3340 f: 803.701.3221 Tom.Simril@duke-energy.com 10 CFR 50.90 Serial: RA-20-0002 February 1, 2020 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 CATAWBA NUCLEAR STATION, UNIT NOS. 1 AND 2 DOCKET NOS. 50-413 AND 50-414 RENEWED LICENSE NOS. NPF-35 AND NPF-52
SUBJECT:
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
REGARDING LICENSE AMENDMENT REQUEST PROPOSING TO REVISE TECHNICAL SPECIFICATION 3.4.3, RCS PRESSURE AND TEMPERATURE (P/T) LIMITS
Reference:
- 1. Duke Energy letter, License Amendment Request Proposing to Revise Technical Specification 3.4.3, RCS Pressure and Temperature (P/T) Limits", dated July 2, 2019 (ADAMS Accession No. ML19183A038).
- 2. NRC E-Mail, Catawba PT Limits LAR - RAIs, dated January 3, 2020 (ADAMS Accession No. ML20003E537).
Ladies and Gentlemen:
By letter dated July 2, 2019 (Reference 1), Duke Energy Carolinas, LLC (Duke Energy) submitted a license amendment request (LAR) for Catawba Nuclear Station (CNS) Unit 1. The proposed change would revise CNS Technical Specification (TS) 3.4.3, RCS Pressure and Temperature (P/T) Limits, to reflect an update to the P/T limit curves in Figures 3.4.3-1 (UNIT 1 ONLY) and 3.4.3-2 (UNIT 1 ONLY). The proposed change would also reflect that the revised CNS Unit 1 P/T limit curves in TS 3.4.3 are applicable until 42.7 effective full power years (EFPY).
By correspondence dated January 3, 2020 (Reference 2), the Nuclear Regulatory Commission (NRC) staff requested additional information from Duke Energy that is needed to complete the LAR review.
( ~ DUKE ENERGY
U.S. Nuclear Regulatory Commission RA-20-0002 Page 2 The enclosure to this letter provides Duke Energy's response to the NRC RAI. Note that although the proposed change only impacts CNS Unit 1, the original amendment request and the subject RAI response are docketed for both CNS Units 1 and 2 since the TS are common to both units.
The conclusions of the original No Significant Hazards Consideration and Environmental Consideration in the original LAR are unaffected by the enclosed RAI response.
There are no new regulatory commitments contained in this submittal.
In accordance with 1 O CFR 50.91, Duke Energy is notifying the State of South Carolina of this LAR by transmitting a copy of this letter and enclosure to the designated State Official.
If there are any questions or if additional information is needed, please contact Mr. Art Zaremba, Manager - Nuclear Fleet Licensing at (980) 373-2062.
I declare under penalty of perjury that the foregoing is true and correct. Executed on February 13, 2020.
Sincerely, Tom Simril Vice President, Catawba Nuclear Station
Enclosure:
Response to Request for Additional Information
U.S. Nuclear Regulatory Commission RA-20-0002 Page 3 cc (with Enclosure):
L. Dudes, USNRC Region II - Regional Administrator J.D. Austin, USNRC Senior Resident Inspector - CNS M. Mahoney, NRR Project Manager - CNS L. Garner, Manager, Radioactive and Infectious Waste Management (SC)
A. Nair-Gimmi, Director, Nuclear Response (SC)
U.S. Nuclear Regulatory Commission RA-20-0002 Page 1 Catawba Nuclear Station, Units 1 and 2 Docket Nos. 50-413 and 50-414 Renewed License Nos. NPF-35 and NPF-52 Response to Request for Additional Information (RAI) Regarding License Amendment Request Proposing to Revise Technical Specification 3.4.3, RCS Pressure and Temperature (P/T)
Limits Enclosure Response to Request for Additional Information
U.S. Nuclear Regulatory Commission RA-20-0002 Page 2 Request for Additional Information (RAI)
By letter to the U.S. Nuclear Regulatory Commission (NRC) dated On July 2, 2019 (Agencywide Documents Access Management System (ADAMS) Accession No. ML19183A038), Duke Energy, (the licensee), requested amend the unit-specific pressure-temperature (P-T) limit curves in Technical Specification (TS) Figure 3.4.3-1, (Unit 1 Only) RCS Heatup Limitations, and TS Figure 3.4.3-2, (Unit 1 Only) RCS Cooldown Limitations, as effective to 42.7 effective full powers years (EFPY) of licensed operations. The licensees submittal enclosed supporting information in Westinghouse Electric Company (WEC) Technical Report Nos. WCAP-15448, Revision 1 (Enclosed in ADAMS Accession No. ML19183A038) and WCAP-17669-NP, Revision 0 (ADAMS Accession ML14353A029).
The NRC staff has reviewed the application and, based upon this review, determined that the following additional information is needed to complete our review:
RAI-01 (SNSB)
The NRC staff requests the licensee to justify the acceptance of the LTOP analysis assuming relief valve setpoints lower than those allowed in TS Limiting Condition for Operation (LCO) 3.4.12, to support plant operation to 42.7 EFPY for Catawba, Unit 1.
Duke Energy Response to RAI-01 In the Technical Evaluation section of the July 2, 2019 Duke Energy license amendment request (ADAMS Accession No. ML19183A038), the LTOP PORV setpoint of 400 psig and the residual heat removal (RHR) relief valve setpoint of 463 psig are cited. The NRC staff notes in the RAI above that these setpoints are lower than the allowable values for these relief devices listed in the LCO for Technical Specification 3.4.12 (i.e., 425 psig for PORVs and 509 psig for RHR suction relief valves). The 400 psig value for the LTOP PORV setpoint and 463 psig value for the RHR relief valve setpoint are the actual calibration setpoints for these devices. A Duke Energy calculation establishes the setpoints for the Pressurizer PORVs in LTOP mode and evaluates the RHR pump suction relief valve setpoints for LTOP protection. The calculation evaluated a low temperature pressure transient caused by a mass input from a combination of two high pressure injection pumps or from a heat input transient caused by a pump start and a 50 0F temperature differential between a Steam Generator and the Reactor Coolant System.
These transients are evaluated with a single PORV or single RHR relief valve opening in response to the transient. The analysis assumes a 60 psig uncertainty in the PORV setpoint and a 10% setpoint tolerance is used for the RHR relief valve to determine the peak Reactor Coolant System pressure under the LTOP transient cases analyzed. The analysis also accounts for valve accumulation at the transient flow rate, Reactor Coolant System differential pressure for different reactor coolant pump combinations and elevation correction between the relief device pressure input and the Reactor Vessel Head Flange. The calculated peak pressures for each transient include the relief device setpoint uncertainty and accounts for valve accumulation, relative elevation and differential pressure across the core as a result of the number of reactor coolant pumps in operation and verifies the peak pressure is less than the limiting Reactor Vessel Closure Head Flange limit. Therefore, the allowable as found setpoint values documented in the LCO for Technical Specification 3.4.12 are bounded by the LTOP setpoint analysis of record.
U.S. Nuclear Regulatory Commission RA-20-0002 Page 3 RAI-02 (SNSB)
On page 5 of the LAR, the licensee indicated that the calculated new limiting pressure for the heatup and cooldown curve at 42.7 EFPY is 1,089 psig for the reactor vessel beltline region and remains 621 psig for the Closure Head/Vessel Flange region. Both pressure limits include measurement uncertainty, while the pressure limits presented in Figure 3.4.3-1 and Figure 3.4.3-2 do not include margin for instrument errors.
The NRC staff requests the licensee to demonstrate that the calculated pressure limits of 1,089 psig and 621 psig are consistent with the pressure limits shown in Figures 3.4.3-1 and 3.4.3-2 in Attachment 1 to the LAR.
Duke Energy Response to RAI-02 The RAI question above states that the 1,089 psig and 621 psig limits cited in the license amendment request (LAR) include measurements uncertainty. To clarify, the LAR (see Page 5 of the Enclosure) discussed these new limiting pressures at 42.7 EFPY with measurement uncertainty recapture (MUR). MUR was a separate project and licensing action which changed the applicability of the curves. However, the values 1,089 psig and 621 psig do not include measurement uncertainty.
The WCAP-15448 table inputs used to develop the proposed CNS Unit 1 limit curves in Figures 3.4.3-1 and 3.4.3-2 for 42.7 EFPY were not adjusted for instrument uncertainty. Thus, the LAR Figures 3.4.3-1 and 3.4.3-2 (i.e., the proposed change) do not reflect instrument uncertainty.
However, instrument uncertainty will be applied to the operating curves in the CNS Unit 1 data book utilized in the Control Room to ensure that the plant is operated within the TS 3.4.3 limits.
Summarizing, the pressure limits of 1,089 psig and 621 psig are consistent with the pressure limits shown in proposed Figures 3.4.3-1 and 3.4.3-2 in that neither reflect instrument uncertainty.
RAI-03 (SFNB)
As part of its review, NRC staff made comparisons between the post-measurement uncertainty recapture (MUR) fluence results presented in the submittal and those of the pre-MUR fluence results presented in WCAP-15448, Revision 1. The fluence results from the submittal were interpolated to 51 EFPY before comparisons were made to the WCAP-15448, Revision 1 fluence results at 51 EFPY. The comparisons indicate an approximate 16.2 percent decrease in post-MUR fluence versus pre-MUR fluence at the clad/base metal interface and the 1/4T and 3/4T depths for the same beltline materials. Given that the Catawba, Unit 1 MUR increased the rated thermal power by approximately 1.7 percent, which will result in a linear increase in core flux, a decrease in post-MUR fluence of 16.2 percent for the same EFPY is unexpected. The results suggest there may be an inaccuracy in the fluence calculation or an unaccounted-for uncertainty, which must be taken into consideration for an accurate fast neutron fluence calculation in order to estimate the fracture toughness of the reactor vessel.
10 CFR Part 50, Appendix G, Fracture Toughness Requirements and 10 CFR 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events set forth requirements for fracture toughness on the reactor pressure vessel materials. In order to estimate the fracture toughness, it is necessary to accurately determine the fast neutron fluence (E > 1MeV).
U.S. Nuclear Regulatory Commission RA-20-0002 Page 4 The NRC staffs request the licensee to provide justification that the post-MUR fluence results presented in the submittal are accurate in light of the 16.2 percent reduction seen in comparison to the pre-MUR results. Conversely, if an inaccuracy or uncertainty is present, provide updated fluence results and a discussion of the identified inaccuracy or uncertainty.
Duke Energy Response to RAI-03 Low-leakage loading pattern strategies were implemented in the CNS Unit 1 reactor cores post-1998, which is justification that the post-MUR fluence results presented in the LAR submittal are accurate considering the 16.2 percent reduction in comparison to the pre-MUR results. Note that the fluence values used to develop the 51 EFPY P-T limit curves were calculated in 1998.
Such low-leakage loading pattern strategies account for the decrease in projected EOL vessel fluence. Duke Energy has also removed and tested ex-vessel neutron dosimetry since 1998.
The corresponding technical reports (WCAP-16869, Ex-Vessel Neutron Dosimetry Program for Catawba Unit 1 Cycles 15 and 16, Revision 1, dated May 2009 and WCAP-18162 Ex-Vessel Neutron Dosimetry Program for Catawba Unit 1 Cycle 22," dated July 2016) both indicate a drop in the max EOL RV fluence. This drop in fluence further illustrates the effect of switching CNS Unit 1 to a low-leakage core.
It is also noted that the reactor vessel (RV) fluence values that were used to develop the 51 EFPY P-T limit curves (i.e., WCAP-15117, Analysis of Capsule V and the Dosimeters from Capsules U and X from the Duke Power Company Catawba Unit 1 Reactor Vessel Radiation Surveillance Program, October 1998) and the fluence values calculated as part of the CNS Unit 1 MUR licensing action (i.e., WCAP-17669, Catawba Unit 1 Measurement Uncertainty Recapture (MUR) Power Uprate: Reactor Vessel Integrity and Neutron Fluence Evaluations, June 2013) were calculated approximately 15 years apart.