MNS-16-091, License Amendment Request for Technical Specification 3.6.14, Divider Barrier Integrity.

License Amendment Request for Technical Specification 3.6.14, Divider Barrier Integrity.

ML16351A197
Person / Time
Site:	Mcguire, McGuire
Issue date:	12/08/2016
From:	Capps S Duke Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
MNS-16-091
Download: ML16351A197 (10)

Text

~\ Steven D. Capps

(~DUKE Vice President

~ ENERGY: McGuire Nuclear Station Duke Energy MG01VP 112700 Hagers Ferry Road Huntersville, NC 28078 o: 980.875.4805 f: 980.875.4809 Steven.Capps@duke-energy.com Serial No: MNS-16-091 10 CFR 50.90 December 8, 2016 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Duke Energy Carolinas, LLC (Duke Energy)

McGuire Nuclear Station, Units 1 and 2 Docket Numbers 50-369 and 50-370 Renewed Facility License Numbers NPF-9 and NPF-17 License Amendment Request for Technical Specification (TS) 3.6.14, "Divider Barrier Integrity" By letter dated June 30, 2016, Duke Energy submitted a License Amendment Request (LAR) to revise TS 3.6.14, "Divider Barrier Integrity." By email dated October 19, 2016, the NRC submitted Requests for Additional Information (RAls) related to this LAR. Enclosure 1 of this letter provides Duke Energy's response to those RAls.

This letter contains no regulatory commitments.

Pursuant to 10 CFR 50.91, a copy of this letter has been forwarded to the appropriate North Carolina state officials.

Please direct any questions you may have in this matter to Brian Richards at (980) 875-5171.

I declare under penalty of perjury that the foregoing is true and correct. Executed on December 8, 2016.

Sincerely, Enclosure 1: Response to Request for Information

• U.S. Nuclear Regulatory Commission December 8, 2016 Page 2 xc:

C. Haney, Regional Administrator U.S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Ave. NE, Suite 1200 Atlanta, GA 30303-1257 A. Hutto NRC Senior Resident Inspector McGuire Nuclear Station Electronic copies:

V. Sreenivas, Interim Project Manager U.S. Nuclear Regulatory Commission V.Sreenivas@nrc.gov W.L. Cox, Ill, Section Chief North Carolina Department of Health and Human Services ee.cox@dhhs.nc.gov

Enclosure 1 Response to Request for Information

Enclosure 1 Page 1 of 7 By letter dated June 30, 2016 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16193A656), Duke Energy Carolinas, LLC (the licensee) submitted a license amendment request to change the McGuire Nuclear Station, Units 1 and 2, Technical Specification 3.6.14, "Divider Barrier Integrity" to allow a steam generator enclosure hatch or a pressurizer enclosure hatch to be open for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to facilitate potential inspections and maintenance to enhance personnel and radiation safety.

The U.S. Nuclear Regulatory Commission staff has reviewed the licensee's submittal and determined that additional information is required in order to complete the review. The requested additional information is as follows:

RA/ 1:

While reviewing the LAR for adherence to GDC 16 and GDC 50, the staff had the following questions regarding the containment peak pressure:

a. Provide the basis for determining that the additional bypass leakage area provided by the open hatch is inconsequential to the long term containment peak pressure analysis, stated on page 4 of 8 of your LAR. Please quantify the new long term containment peak pressure.

b. Provide the basis for determining the containment spray initiation maintains the long term containment peak pressure at or below 13.87 psig. For example, it is stated that the containment spray would be manually initiated prior to the time of ice bed meltout in the limiting scenario. How is the time of containment spray initiation changed, if at all.

State the containment spray initiation time for the open pressurizer hatch as well as the open SG hatch. In this limiting scenario, are one or two trains of containment spray assumed?

c. Before the time of complete ice melt at spray initiation, because of the additional open area of 7.5 ft2, it is expected that more steam would be present in the upper compartment than for the open area of 5 ft2. Verify quantitatively that the peak containment pressure is not affected due to additional open area.

McGuire Response:

a. To quantify the impact of the additional 7.5 ft 2 of divider deck bypass area on the Peak Containment Pressure results, an analysis using the GOTHIC 4.0/DUKE code (GOTHIC) was performed with the additional bypass area added to the GOTHIC model. A junction between the nodes representing the pressurizer enclosure and the upper containment region was added with a conservatively low loss coefficient to maximize the amount of steam which could pass through this 7.5 ft 2 of additional bypass area. There were no other changes made to the GOTHIC model. As discussed in Section 3.4 of the LAR, this 7.5 ft2 represents the bounding divider deck bypass area for the proposed Pressurizer Hatch and Steam Generator Enclosure Hatch openings. The limiting large break LOCA with the break located in the cold leg discharge piping with minimum ECCS flows was chosen; this is the same scenario as that presented for the Peak Containment Pressure transient in UFSAR Section 6.2.1.3.1.1. There were no changes in other associated boundary conditions or single failure assumptions for the evaluation.

Enclosure 1 Page 2 of 7 The GOTHIC code (as well as the RELAP5 code for mass and energy release calculations) was submitted to the NRC for licensing basis containment analyses for the McGuire and Catawba Nuclear Stations in the Duke methodology report DPC-NE-3004.

Revision 0 of this report was originally submitted and approved by the NRC in September 1995. Revision 1 of this report was submitted and approved by the NRC in June 2010.

The results of this evaluation show that there is no impact on the calculated peak pressure due to the additional divider deck bypass area. The trends for containment pressure (Figure 1) and upper containment temperature (Figures 2) show that during the first 5 to 6 minutes of the transient, there is more steam present in upper containment due to the increased divider deck bypass area. This steam results in slightly higher pressure and upper containment temperatures in the first 5 to 6 minutes compared to the Analysis of Record (AOR), although containment pressure is still well below the design limit of 15 psig. From the period between 5 minutes and 60 minutes, more steam enters upper containment as ice bays start to melt out and the steam passes through the bays without condensing onto the ice surfaces. As discussed in Section 3.4 of the LAR, the additional divider beck bypass area becomes less and less significant as more ice bays are evacuated of ice (more melted bays represents more "bypass" area, or more area for steam to pass through and enter upper containment without condensing). Furthermore, once the containment air return fan initiates at approximately 10 minutes into the transient, the air I steam mixture in upper containment is forced into lower containment and then into the ice condenser. This allows the steam which entered upper containment through the additional divider deck bypass area earlier in the transient to pass through the ice condenser and come into contact with the ice surfaces of the remaining ice in the ice condenser baskets. After about one hour, the impact of the additional divider deck bypass area has diminished to the point where the containment pressures and upper containment temperatures are virtually the same (Figures 1 and 2).

The remainder of the transient progresses in the same manner in the case with the additional divider deck bypass area as in the AOR.

Figure 1 shows that the containment pressure in the case with the additional divider deck bypass area is actually slightly lower than that for the AOR ( 13.45 psig vs. 13.87 psig). Due to slight differences in the ice melt rates and the sump temperature responses (Figures 3 and 4), further iterations for both the bypass area and the evaluation case with the mass and energy release analysis would be required to get an exact comparison. The Duke methodology report DPC-NE-3004-PA, Revision 1b contains further discussion on this iteration process between the RELAP5 mass and energy release calculation and the GOTHIC containment response calculation in Section ,,

3.3.3.3. The GOTHIC results demonstrate that the additional divider deck bypass area does not lead to an increase in the peak containment pressure calculation. The impact of the proposed increase in the divider deck bypass area is considered to be inconsequential to the long-term containment peak pressure analysis.

b. There were no changes made to the containment spray initiation time. Following the approval of the ECCS Water Management Initiative LAR for McGuire (ADAMS No.

ML101600256), one train of containment spray is manually initiated after the Refueling Water Storage Tank (FWST) low level has been reached for LOCA scenarios. The suction source for this spray flow is the containment sump. For the Peak Containment Pressure transient analysis described in UFSAR Section 6.2.1.3.1.1, containment spray initiation occurs at 3988 seconds (UFSAR page 6.2-11, item 13). This initiation time is

Enclosure 1 Page 3 of 7 not impacted by the additional divider deck bypass area, as the draindown rate of the FWST is not impacted by this bypass area.

c. As discussed in the response to part a. above, early in the transient there is more steam present in the upper containment region due to the increased divider deck bypass area (See Figures 1 and 2). However, there is no impact on the calculated peak containment pressure due to the increased bypass area.

RA/ 2:

Reference 36 stated in FSAR Section 6.2.1.3.1.1 is the Westinghouse WCAP-10325-P-A LOCA Mass and Energy (M&E) release methodology. The SATAN code in this methodology is used for the blowdown M&E release analysis during a LOCA which is an input to determine the subcompartment pressure and differential pressure responses (UFSAR Section 6.2.1.2.3) during the blowdown phase using the TMD code. Westinghouse has issued Nuclear Safety Advisory Letters (NSALs)-06-6, 5, and 2, and lnfoGram IG-14-1 reporting errors in the WCAP-10325-P-A methodology. Correction of M&E analysis should be made for the errors reported in the above. NSALs and lnfoGram. Provide the revised containment pressure and temperature responses, and subcompartment pressure, and differential pressure responses, and the revised results of any other analysis that used the WCAP-10325-P-A methodology.

McGuire Response:

The results from the SATAN code are used in the analyses of record (UFSAR Section 6.2.1.3.1) to calculate the short-term (or blowdown) mass and energy release during a LOCA. These mass and energy release calculations are then used as input to determine the subcompartment pressures and differential pressure responses using the TMD code, as .discussed in UFSAR Section 6.2.1.2.3. The TMD results presented in the McGuire UFSAR (Tables 6-2, 6-3, .6-44 and 6-45) show that the peak subcompartment pressures and differential pressures are reached between 0.009 and 3.0 seconds after the postulated break. Therefore, any issues identified with the SATAN-V code would not impact the mass and energy results within the first few seconds of the transient and would therefore have no impact on the subcompartment pressure and differential pressure results. The subcompartment pressure and differential pressure results are largely driven by the forced movement of the air/steam mixture after the postulated LOCA in the lower containment volume and/or the subcompartments being analyzed.

The issues raised by Westinghouse NSALs 6, 5, and 2 as well as lnfoGram IG 14-1 can all be characterized as having no impact on the first few seconds of the analyzed transients.

These issues include the volumetric heat capacities of stainless steel, the flow values, purge volumes, and AFW flows during the reflood and post-reflood periods, reactor vessel metal mass, secondary side pressures, and amounts of stored energy within the Reactor Coolant System metal mass. Since all of these factors either involve conduction .of energy into the Reactor Coolant System, or are associated with periods of the transient much further out than the first few seconds, it is concluded that there is no impact on the TMD results of transients where the peak subcompartment or differential pressure is reached within the first 3 seconds ..

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