ML13281A809
| ML13281A809 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 10/02/2013 |
| From: | Mark D. Sartain Dominion Nuclear Connecticut |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 13-524 | |
| Download: ML13281A809 (18) | |
Text
'JDominilon Dominion Nuclear Connecticut, Inc.
5000 Dominion Boulevard, Glen Allen, VA 23060 Web Address: www.dom.com October 2, 2013 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Serial No.
NSSL/MLC Docket No.
License No.13-524 RO 50-336 DPR-65 DOMINION NUCLEAR CONNECTICUT, INC.
MILLSTONE POWER STATION UNIT 2 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE AMENDMENT REQUEST FOR CHANGES TO TECHNICAL SPECIFICATION 3/4.7.11, "ULTIMATE HEAT SINK" By letter dated May 3, 2013, Dominion Nuclear Connecticut, Inc. (DNC) submitted a license amendment request (LAR) for Millstone Power Station Unit 2 (MPS2).
The proposed amendment would modify Technical Specification (TS) 3/4.7.11, "Ultimate Heat Sink," to increase the current ultimate heat sink (UHS) water temperature limit and change the TS Action to state, "With the ultimate heat sink water temperature greater than 80'F, be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />."
In a letter dated June 26, 2013, the Nuclear Regulatory Commission (NRC) provided DNC an opportunity to supplement the LAR identified above. Supplemental information was provided to the NRC in a letter dated June 27, 2013.
In a letter dated July 18, 2013, the NRC transmitted a request for additional information (RAI) related to the LAR.
DNC responded to the RAI in a letter dated July 19, 2013. In an e-mail dated July 23, 2013, the NRC transmitted a second RAI to DNC. DNC responded to the second RAI in a letter dated July 30, 2013. In an e-mail dated July 26, 2013, the NRC transmitted a third RAI to DNC. DNC responded to the third RAI in a letter dated August 1, 2013. In an e-mail dated September 4, 2013, the NRC transmitted a fourth RAI to DNC. to this letter contains DNC's response to the fourth RAI.
If you have any questions or require additional information, please contact Wanda Craft at (804) 273-4687.
Sincerely, VICKI L. HULL M
Notary Public Mark D. SartainCommonwealth of Virginia Vice President - Nuclear Engineering and Development C
140542 n
C My Commission Expires May 31, 2014 COMMONWEALTH OF VIRGINIA COUNTY OF HENRICO
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The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Mark D. Sartain, who is Vice President - Nuclear Engineering and Development of Dominion Nuclear Connecticut, Inc. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that company, and that the statements in the document are true to the best of his knowledge and belief.
A//f Acknowledged before me thia day of 2013.
My Commission Expires:
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- 2i ii NOVPublic 71\\
Serial No.13-524 Docket No. 50-336 Page 2 of 2 Commitments made in this letter: None Attachments:
1.
Response to Request for Additional Information Regarding License Amendment Request for Changes to Technical Specifications 3/4.7.11, "Ultimate Heat Sink"
- 2.
FSAR Section 14.8.2 Figures cc:
U.S. Nuclear Regulatory Commission Region I 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 James S. Kim Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 08 C2A 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station Director, Radiation Division Department of Energy and Environmental Protection 79 Elm Street Hartford, CT 06106-5127
Serial No 13-524 Docket No. 50-336 Response to Request for Additional Information Regarding License Amendment Request for Changes to Technical Specifications 3/4.7.11, "Ultimate Heat Sink" Dominion Nuclear Connecticut, Inc.
Millstone Power Station Unit 2
Serial No 13-524 Docket No. 50-336, Page 1 of 5 Response to Request for Additional Information Regarding License Amendment Request for Changes to Technical Specifications 3/4.7.11, "Ultimate Heat Sink" By letter dated May 3, 2013, Dominion Nuclear Connecticut, Inc. (DNC) submitted a license amendment request (LAR) for Millstone Power Station Unit 2 (MPS2).
The proposed amendment would modify Technical Specification (TS) 3/4.7.11, "Ultimate Heat Sink," to increase the current ultimate heat sink (UHS) water temperature limit and change the TS Action to state, "With the ultimate heat sink water temperature greater than 800F, be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />."
In a letter dated June 26, 2013, the Nuclear Regulatory Commission (NRC) provided DNC an opportunity to supplement the LAR identified above. Supplemental information was provided to the NRC in a letter dated June 27, 2013. In a letter dated July 18, 2013, the NRC transmitted a request for additional information (RAI) related to the LAR.
DNC responded to the RAI in a letter dated July 19, 2013. In an e-mail dated July 23, 2013, the NRC transmitted a second RAI to DNC. DNC responded to the second RAI in a letter dated July 30, 2013. In an e-mail dated July 26, 2013, the NRC transmitted a third RAI to DNC.
DNC responded to the third RAI in a letter dated August 1, 2013.
In an e-mail dated September 4, 2013, the NRC transmitted a fourth RAI to DNC. The response to this RAI is as follows.
SCVB-RAI-1 Section 5.10, second paragraph of the LAR mentions about MPS2 LOCA and MSLB containment response analysis using the revised Mass and Energy (M&E) release information and Ultimate Heat Sink (UHS) temperature of 800F.
(a) Describe the basis and reasons for the revision in the M&E release information provided by Westinghouse.
DNC Response (a) Westinghouse identified that non-conservative loss of coolant accident (LOCA)
M&E release data was input to the Westinghouse containment response analysis using the CONTRANS methodology.
Westinghouse determined that the M&E releases generated by the thermal hydraulic response computer code (CEFLASH-4A) during the blowdown phase of the LOCA was not adequately detailed with respect to time step data during the early stages of the transient for use in downstream containment response calculations.
This error resulted in an under prediction of the M&E released to the containment during the blowdown phase of the LOCA. In response to this issue, Dominion had Westinghouse re-analyze the MPS2 LOCA M&E analysis and provide data for use in the NRC-approved Dominion GOTHIC methodology identified in topical report DOM-NAF-3-0.0-P-A.
The objective was to replace the Final Safety Analysis Report (FSAR) Chapter 14
Serial No 13-524 Docket No. 50-336, Page 2 of 5 analysis of record with a corrected LOCA M&E analysis and change the containment response analysis methodology from Westinghouse CONTRANS to Dominion GOTHIC.
As identified in Section 5.10 of the May 3, 2013 LAR, Dominion has implemented the revised LOCA containment analyses using the NRC-approved Dominion GOTHIC methodology in the current MPS2 Licensing Basis. The analyses conservatively assumed a UHS temperature of 800F.
For the LOCA containment analysis, Westinghouse provided revised M&E release data during the LOCA blowdown and reflood phases (using CEFLASH-4A and FLOOD3 methods), and Dominion used the Dominion GOTHIC methodology to determine M&E releases during the post-reflood and long-term phases of the LOCA.
The revised M&E release data corrected the error in the Westinghouse blowdown M&E releases discussed above.
To ensure the limiting LOCA peak containment pressure-temperature cases were identified using the Dominion GOTHIC methodology, Westinghouse generated new blowdown and reflood M&E release data for double ended guillotine and slot LOCAs and smaller LOCAs in the reactor coolant system cold leg (in both the reactor coolant pump suction and discharge legs) and hot legs.
In concert with the LOCA containment reanalysis project that corrected the vendor M&E errors and implemented the Dominion GOTHIC methodology, MPS2 also revised the FSAR Chapter 14 containment analyses for the main steam line break (MSLB) event using the Dominion GOTHIC analysis methodology. To ensure the limiting MSLB peak containment pressure-temperature cases were identified using the Dominion GOTHIC methodology, Westinghouse provided revised M&E release data using the SGN-3 methodology, as described in the MPS2 FSAR. M&E release data was provided for MSLB events initiated from 0%, 25%, 50%, 75%, and 102%
power for various feedwater system and containment heat removal single failures.
Cases with and without the concurrent loss of offsite power were evaluated. These analyses assumed a UHS temperature of 80'F for conservatism and were implemented under the provisions of 10 CFR 50.59 as part of the containment analysis upgrade project.
(b) Please provide the results of the revised containment peak pressure licensing basis analysis. Please describe the impact that an increase in the UHS temperature from the current TS limit to 80OF has on the containment peak pressure, containment peak gas temperature for EQ, and peak containment liner temperature.
DNC Response (b) The results of the current containment peak pressure licensing basis analysis are described in MPS2 FSAR Section 14.8.2, which was provided in Attachment 12 to DNC's RAI response dated July 19, 2013. The FSAR figures associated with this attachment were inadvertently omitted. These figures are provided in Attachment 2 of this submittal.
Serial No 13-524 Docket No. 50-336, Page 3 of 5 The current licensing basis LOCA containment analysis, which uses the Dominion GOTHIC methodology and assumes a UHS temperature of 800F, predicts a maximum containment pressure of 52.5 psig (compared to containment design limit of 54 psig) and a maximum containment gas temperature of 279.21F. Increasing the UHS temperature from the current 771F TS limit to the proposed limit of 80OF increases the predicted containment maximum pressure by 0.03 psi. This pressure increase has insignificant impact on the predicted peak gas containment temperature for equipment qualification (EQ) following a LOCA.
Because the maximum containment gas temperature for EQ following a LOCA is less than the containment structure and liner design temperature of 2891F, a specific calculation for the peak containment liner temperature is not required.
The current licensing basis MSLB containment analyses, which use the Dominion GOTHIC methodology and assume a UHS temperature of 801F, predict a maximum containment pressure of 53.8 psig (compared to containment design limit of 54 psig) and a maximum containment gas temperature of 360.91F.
Increasing the UHS temperature from the current 77 0F TS limit to the proposed limit of 80°F increases the predicted containment maximum pressure of the MSLB by less than 0.1 psi and the predicted peak containment gas temperature by less than 0.1 0F.
The containment liner thermal response was determined using the method described in Section 3.3.3 of the NRC-approved Dominion GOTHIC methodology identified in topical report DOM-NAF-3-0.0-P-A.
The maximum predicted containment liner temperature following a MSLB, assuming an 800F UHS temperature, was 259.71F, which is less than the containment structure and liner design temperature of 2890F.
Increasing the UHS temperature from the current 770F TS limit to the proposed limit of 80OF increases the containment liner temperature by less than 0.10F.
In conclusion, the FSAR Chapter 14 containment analyses for LOCA and MSLB assume 80°F UHS temperature, use the NRC-approved Dominion GOTHIC methodology in DOM-NAF-3-0.0-P-A, and demonstrate compliance with the containment design limits.
SCVB-RAI-2 Please describe the impact that an increase in the UHS temperature from the current TS limit to 80OF has on the maximum sump water temperature and available Net Positive Suction Head (NPSH) of the ECCS and containment spray pumps post-LOCA following the Sump Recirculation Actuation Signal (SRAS), and describe how the NPSH analysis complies with Safety Guide I (Regulatory Guide 1.1) as described in FSAR Sections 6.2.3.1 and 6.4.4.1.
Serial No 13-524 Docket No. 50-336, Page 4 of 5 DNC Response The current licensing basis LOCA analyses, which use the Dominion GOTHIC methodology and an UHS of 80 0F, predict a maximum containment sump water temperature of 233.51F following SRAS. The increase in the UHS temperature from 771F to 801F increased the maximum containment sump water temperature following SRAS by 1.20F.
The minimum available NPSH analysis for the emergency core cooling system (ECCS) and containment spray pumps is conservatively calculated during the sump recirculation mode in accordance with Safety Guide 1 assuming a sump water temperature and pressure of 212OF and 14.7 psia, as noted in FSAR Sections 6.2.3.1 and 6.4.4.1.
To determine the minimum available NPSH for the ECCS and containment spray pumps, the minimum containment sump level at the time of SRAS is determined by conservatively assuming that sump water greater than 212OF boils off, converting a small amount of the sump volume to vapor.
(This available NPSH determination does not credit the containment to maintain pressure greater than 14.7 psia).
This decreases the containment sump level and the calculated NPSH available for the ECCS and containment spray pumps. The available NPSH, conservatively calculated in accordance with Safety Guide 1, exceeds the required NPSH of the ECCS and containment spray pumps following SRAS.
SCVB-RAI-3 NUREG-0800, Standard Review Plan (SRP) 6.2.1.5 describes the minimum containment pressure analysis for emergency core cooling system (ECCS) performance capability.
Regulatory Guide (RG) 1.157, Section 3.12.1 provides guidance for calculating the containment pressure response used for evaluating cooling effectiveness during the post-blowdown phase of a LOCA.
The RG states that the containment pressure should be calculated by including the effects of containment heat sinks and operation of all pressure-reducing equipment assumed to be available. Using the above guidance please describe the impact of the changes in M&E input on the minimum containment pressure analyses for ECCS performance.
DNC Response There were no changes to the M&E input to the minimum containment pressure analyses for ECCS performance. The M&E input to the 10 CFR 50.46 ECCS performance analysis is developed by AREVA, the current MPS2 fuel supplier, using the 10 CFR 50.46 large break LOCA analysis methodology for MPS2. The AREVA methodology was not affected by the Westinghouse LOCA M&E release error for use in containment integrity analyses.
The AREVA calculated minimum containment pressure following a LOCA includes the effects of containment heat sinks and the operation of the available containment pressure-reducing equipment, including the Containment Air Recirculation (CAR) cooling units and
Serial No 13-524 Docket No. 50-336, Page 5 of 5 the containment spray system. As identified in Section 5.10 of the May 3, 2013 LAR, to minimize the containment pressure response following a large break LOCA, the 10 CFR 50.46 ECCS performance analysis maximizes the containment heat removal by the CAR cooling units by using a Reactor Building Closed Cooling Water (RBCCW) water temperature of 350F. This RBCCW water temperature is consistent with a minimum UHS temperature that may occur in the winter months. As such, the proposed LAR requesting an increase in the UHS temperature to 801F does not impact this analysis.
SCVB-RAI-4 Please confirm that in the 10 CFR 50.59 changes implemented, the requirements of General Design Criterion (GDC) 16, 38, and 50 of 10 CFR 50 Appendix A are met.
DNC Response Following the incorporation of the MSLB and LOCA containment analysis using the Dominion GOTHIC methodology and an 80°F UHS temperature into the current licensing basis under the provisions of 10 CFR 50.59, MPS2 continues to comply with the requirements of General Design Criterion (GDC) 16, 38, and 50 of 10 CFR 50 Appendix A as described in Appendix 1A to Chapter 1 of the MPS2 FSAR.
Serial No 13-524 Docket No. 50-336 FSAR Section 14.8.2 Figures Dominion Nuclear Connecticut, Inc.
Millstone Power Station Unit 2
Serial No 13-524 Docket No. 50-336, Page 1 of 9 MPS-2 FSAR FIGURE 14.8.2-1 MAIN STEAM LINE BREAK ANALYSIS - 102% POWER WITH LOSS OF OFFSITE POWER AND FAILURE OF VITAL BUS CABINET VA-10 OR VA CONTAINMENT PRESSURE VS. TIME 13-2 60 50 40
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0 100 200 300 400 500 Time, seconds 600 700 800 900 1000 Rev. 31.1
Serial No 13-524 Docket No. 50-336, Page 2 of 9 NfPS-2 FSAR FIGURE 14.8.2-2 MAIN STEAM LINE BREAK ANALYSIS - 102 % POWER WITH LOSS OF OFFSITE POWER AND FAILURE OF VITAL BUS CABINET VA-10 OR VA CONTAINMENT TEMPERATURE VS. TIME 13-2 350 300
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Serial No 13-524 Docket No. 50-336, Page 3 of 9 MIPS-2 FSAR FIGURE 14.8.2-3 MAIN STEAM LINE BREAK ANALYSIS - 102 % POWER WrITH LOSS OF OFFSITE POWER AND FAILURE OF VITAL BUS CABINET VA-10 OR VA-20 -MASS FLOW RATE VS. TIME 13-2 7000 70 0 0 6 0 0 0 000
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- 1. 3000 2 0 0 0 low from Break iI 1000 I FW Flashing Flow 0
100 200 300 400 500 600 Time, seconds Rev. 31.1
Serial No 13-524 Docket No. 50-336, Page 4 of 9 MPS-2 FSAR FIGURE 14.8.2-4 MAIN STEAM LINE BREAK ANALYSIS - 102 % POWER WITH LOSS OF OFFSITE POWER AND FAILURE OF VITAL BUS CABINET VA-10 OR VA ENERGY RELEASE RATE VS. TIME i13-2 8.OE+06 7.OE-'O6 6.OE+06 5.OE+06 4.OE+O6 3.OE+O6 2.OE+O6 1.OE+O6 O.OE+OO 600 0
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Serial No 13-524 Docket No. 50-336, Page 5 of 9 MPS-2 FSAR FIGURE 14.8.2'-5 MAIN STFEAM LINE BREAK ANALYSIS - 102 % POWER WITH LOSS OF OFFSITE POVER AND FA.ILURE OF VITAL BUS CABINET VA-10 OR VA INTEGRATED MASS FLOW VS. TIME 13-2 350 300 250
= 200 150 100 50 0
0 100 200 300 400 Time, seconds 500 600 Rev. 31.1
Serial No 13-524 Docket No. 50-336, Page 6 of 9 MPS-2 FSAR FIGURE 14.8.2-6 MAIN STEAM LINE BREAK ANALYSIS - 10' % POWER WITH LOSS OF OFFS1TE POWVER AND FAILURE OF VITAL BUS CABINET VA-10 OR VA-20 -INTEGRATED ENERGY RELEASE VS. TIME 1 3-2 400 350 300 S250 200 150 100 50 0
0 100 200 300 Time, seconds 400 500 600 Rev. 31. 1
Serial No 13-524 Docket No. 50-336, Page 7 of 9 MPS-2 FSAR FIGURE 14.8.2-7 MAIN STEAM LINE BREAK ANALYSIS - 102 % POWER WITH LOSS OF OFFSITE POWER AND)
FAILURE OF VITAL BUS CABINET VA-10 OR VA AFFECTED STEAM GENERATOR PRESSURE VS. TIME 13-2
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Serial No 13-524 Docket No. 50-336, Page 8 of 9 MPS-2 FSAR FIGURE 14.8&2-8 MAIN STEAM LINE BREAK ANALYSIS - 102 % POWER WITH LOSS OF OFFSITE POWER AND FAILURE OF VITAL BUS CABINET VA-10 OR VA UNA2FFECTED STEAM GENERATOR PRESSURE VS, TIME 13-2 1100 1000 900 800 700 600 500 400 300 2 0 0.--
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Serial No 13-524 Docket No. 50-336, Page 9 of 9 MPS-2 FSAR FIGURE 14.8.2-9 MAIN STEAM LINE BREAK ANALYSIS - 102 % POWER WITH LOSS OF OFFSITE POWVER AND FAILURE OF VITAL BUS CABINET VA-I0 OR VA AFFECTED STEAM GENERATOR LIOUID MASS VS. TIME 13-2 160000 140000 120000 100000 80000 60000 40000 20000 4
i 0
100 200 300 400 500 600 700 lime, seconds 800 900 1000 Rev. 31.1