ML19214A222
| ML19214A222 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 09/17/2019 |
| From: | Audrey Klett Plant Licensing Branch II |
| To: | Burchfield J Duke Energy Carolinas |
| Klett A, 415-0489 | |
| References | |
| EPID L-2018-LLA-0251 | |
| Download: ML19214A222 (7) | |
Text
September 17, 2019 Mr. J. Ed Burchfield, Jr.
Site Vice President Oconee Nuclear Station Duke Energy Carolinas, LLC 7800 Rochester Highway Seneca, SC 29672-0752
SUBJECT:
OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3 -
SUMMARY
OF REGULATORY AUDIT IN SUPPORT OF LICENSE AMENDMENT REQUEST NO. 2018-02 (EPID L-2018-LLA-0251)
Dear Mr. Burchfield:
By letter RA-18-0026 dated September 14, 2018, as supplemented by letter RA-19-0086 dated January 24, 2019 (Agencywide Documents Access and Management System (ADAMS)
Accession Nos. ML18264A023 and ML19036A625, respectively), Duke Energy Carolinas, LLC (the licensee) submitted license amendment request No. 2018-02 for Oconee Nuclear Station, Units 1, 2, and 3, regarding revisions to the tornado licensing basis in the updated final safety analysis report. The U.S. Nuclear Regulatory Commissions (NRCs) Office of Nuclear Reactor Regulation staff conducted an audit to support its review of the amendment request. The audit plan is available in ADAMS at Accession No. ML19037A005.
The audit occurred at the NRCs office in Rockville, Maryland. NRC technical reviewers assigned to review the amendment request conducted the audit using an internet-based portal provided by the licensee. The purpose of the audit was to review the licensees thermal-hydraulic analyses, finite element analysis, and calculations used to support the application, which were not available on the docket in ADAMS, to acquire additional understanding about the amendment request and to determine whether additional information was needed to be docketed to complete the NRC staffs safety evaluation. On February 14, April 11, and May 14, 2019, the NRC staff held teleconferences with the licensee to discuss the documents on the portal. The topics of discussion are enclosed. The staff requested additional information from the licensee via electronic mail dated June 28, 2019 (ADAMS Accession No. ML19183A483).
E. Burchfield Any inquiries can be directed to Ms. Audrey Klett at 301-415-0489 or via e-mail at Audrey.Klett@nrc.gov.
Sincerely,
/RA Michael Mahoney for/
Audrey L. Klett, Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-269, 50-270, and 50-287 cc: via Listserv
Audit Discussion Topics for Oconee Nuclear Station, Units 1, 2, and 3 License Amendment Request No. 2018-02 Enclosure A-1 (Technical Specifications Branch (STSB))
Regarding Section 3.8 of the license amendment request (LAR), please provide a discussion of the meaning of Passive Civil Features with respect to the Oconee Nuclear Station (Oconee) licensing basis and consideration for technical specification (TS) operability.
A-2 (Containment and Plant Systems Branch (SCPB))
OSC-11622, Revision 2, titled, Oconee Tornado Strategy Utilizing the Standby Shutdown Facility (SSF), Section 5.1, Licensing Transition, provides a summary of the transition plan to deterministic strategy. This document provides a list of commitments that are defined as items that must be completed to fully qualify the SSF for tornado mitigation. Please discuss which commitments were completed and which ones are to be completed and discuss if there is any required instrumentation modeled in TORMIS that is TS-controlled and/or has not been installed.
A-3 (SCPB)
Please discuss where in the requested audit materials are the horizontal and vertical missiles velocity values and calculation of VDAM used in TORMIS to verify bounding the design basis values of velocities in Table 9-17 of the Updated Final Safety Analysis Report (UFSAR).
A-4 (SCPB)
Please discuss if the requested audit materials for the cask decontamination tank room and the west penetration room commitments related to protecting all walls modifications have been completed.
A-5 (SCPB)
The SSF building is a reinforced concrete structure and is designed for tornado differential pressure, wind, and missile loadings in accordance with the UFSAR. Please discuss if the requested audit materials can confirm that the SSF building is protected from tornado winds and missiles at 360 mph.
A-6 (SCPB)
Page 10 of the application states, Studies [emphasis added] previously submitted to the NRC assume tornado damages are on unit with an associated LOOP [loss of offsite power] on the other two units. Please provide the reference for such studies and NRCs approval of them (e.g., ADAMS Accession No. or date of the documents).
2 A-7 (SCPB)
The LAR, Appendix 4, Section 5.1, states:
Oconee has redundant systems normally available for SSDHR [secondar side decay heat removal,] which are Emergency Feedwater (EFW), SSF ASW
[Auxiliary Service Water], and the Protected Service Water (PSW) system[,]
which is an enhanced replacement for the original Station ASW system. The RCMU [reactor coolant makeup] function can be provided either by SSF RCMU or by the HPI [High Pressure Injection] system. Both systems can provide RCP
[Reactor Coolant Pump] Seal Cooling while providing RCS [Reactor Coolant System] makeup. These redundant systems are largely separated but some spatial dependencies exist. However, in order to simplify the analysis, it is conservatively assumed that all other systems besides the SSF and its support systems are unavailable.
Please discuss the basis for the exposed components included in Appendix 4 of TORMIS that seem to credit the new approach, old approach, OR same for both; and whether PSW is credited in the new approach of Tornado Mitigation for redundancy?
A-8 (Reactor Systems Branch (SRXB))
Regarding Calculation OSC-11638 (Feedwater (FW) line break - overheating), the calculation discusses that the transient analysis credits either the PSW systems or the SSF for mitigation.
This makes it unclear as to which systems were assumed in operation in the calculations. For example, was the SSF makeup pump flow used or the PSW HPI pump flow? Please clarify which cases were used to support the LAR and that the PSW is not credited in these analyses.
A-9 (SRXB)
Regarding Calculation OSC-11638 (FW line break - overheating), acceptance criteria in the calculation are noted as:
- 2. Mitigating systems are capable of providing sufficient decay heat removal and primary coolant makeup to assure core coverage, and maintain the RCS in MODE 3 for an extended period of time.
These are different than the acceptance criteria in the LAR (page 17 of the Enclosure):
- 1. Core must remain intact and in a coolable core geometry during the credited strategy period.
- 2. RCS must not exceed 2750 psig (110 percent of design).
- 3. Minimum departure from nucleate boiling ratio (DNBR) meets specified acceptable fuel design limits.
- 4. Steam generator (SG) tubes remain intact.
- 5. RCS remains within acceptable pressure and temperature limits.
Please discuss why the criteria are different and how was it determined that all the criteria in the LAR were met when the criteria described in the calculation are different.
3 A-10 (SRXB)
Regarding Calculation OSC-11638 (FW line break - overheating), Appendix A says that the pressurizer heater capacity from the SSF is assumed to exceed ambient heat losses from the pressurizer by at least 50 kilowatts. Please discuss the actual capacity of the SSF-controlled pressurizer heaters and how this compares to the value used in the analysis.
A-11 (SRXB)
Regarding Calculation OSC-11638 (FW line break - overheating), Section 8.2, Model Modification No. 6, states that the SSF letdown line valve area is sized to yield 5.4263 pound mass per second at 555 degree Fahrenheit (°F) and 2,420 psig. A hand calculation shows this results in a volumetric flow rate of ~52 gpm (using a density of 46.6 pound mass per cubic foot.
However, from the previous SSF LAR, the existing SSF letdown line has a capacity of ~41 gpm.
Please discuss the discrepancy.
A-12 (SRXB)
Regarding Calculation OSC-11547 (main steamline break (MSLB) - overcooling), the acceptance criteria in the calculation are given as:
- 1. Core will remain intact for effective core cooling.
- 2. Minimum DNBR remains above acceptable limits.
- 3. No core uncovery.
These are different that the acceptance criteria in the LAR (page 17 of the Enclosure):
- 1. Core must remain intact and in a coolable core geometry during the credited strategy period.
- 2. RCS must not exceed 2750 psig (110 percent of design).
- 3. Minimum DNBR meets specified acceptable fuel design limits.
- 4. SG tubes remain intact.
- 5. RCS remains within acceptable pressure and temperature limits.
Please discuss why the criteria are different and how it was determined that all the criteria in the LAR were met when the criteria described in the calculation are different.
A-13 (SRXB)
Regarding Calculation OSC-11547 (MSLB - overcooling), both the existing and new SSF letdown lines are discussed in the calculation; however, it is not clear which one is actually used in the analysis. Please clarify and discuss if the new letdown is required to meet the acceptance criteria.
A-14 (SRXB)
Regarding Calculation OSC-11547 (MSLB - overcooling), Appendix H of the calculation discusses DNBR and states that the minimum DNBR occurs around 1,500 seconds. However, the actual minimum DNBR occurs almost immediately after the event begins and appears to be near the limit. It is stated that the lower values that occur during the initial trip and associated
4 power decrease are not a DNBR concern. Please explain why the actual minimum DNBR is not a concern.
A-15 (SRXB)
Regarding Calculation OSC-4171 (SSF ASW - design inputs calculation), Assumption 7.1 on page 16 states that the SSF control room operator will throttle valves as necessary to maintain Tc at 555 °F if at least one of a Units two SGs is pressurized. This is inconsistent with of the LAR, which states the goal is to stabilize the plant to between 325 °F - 350 °F and 650 psig - 700 psig. How would this calculation be affected if the proposed guidance to bring to lower temperature/pressure were assumed?
A-16 (SCPB)
UFSAR Section 9.6.3.1, Structure, states, The design of all future changes to and/or analysis of SSF-related systems, structures, and components subject to tornado loadings will conform to the tornado wind, differential pressure, and missile criteria specified in RG 1.76, Revision 1. As stated in RG 1.76, Revision 1, the NRC considers the missiles to can strike in all directions, including horizontal and vertical. Oconee UFSAR, Table 9-17, shows applicable tornado missiles criteria for both horizontal and vertical design impact velocity. The NRC staff requests the licensee to describe how it addressed vertical missiles.
ML19214A222 OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/LA NAME AKlett (MMahoney for)
KGoldstein (SRohrer for)
DATE 09/13/2019 08/07/2019 OFFICE NRR/DORL/LPL2-1/BC NRR/DORL/LPL2-1/PM NAME MMarkley AKlett (MMahoney for)
DATE 09/17/2019 09/17/2019