ML13275A240
| ML13275A240 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 09/19/2013 |
| From: | Stoddard D Dominion, Dominion Nuclear Connecticut |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 13-474 | |
| Download: ML13275A240 (6) | |
Text
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-- Air Dominion Nuclear Connecticut, Inc.
5000 Dominion Boulevard, Glen Allen, VA 23060 JFDominioW Web Address: www.dom.com September 19, 2013 U.S. Nuclear Regulatory Commission Serial No.13-474 Attention: Document Control Desk NSSL/MLC RO Washington, DC 20555 Docket No. 50-423 License No. NPF-49 DOMINION NUCLEAR CONNECTICUT, INC.
MILLSTONE POWER STATION UNIT 3 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE AMENDMENT REQUEST TO REVISE TECHNICAL SPECIFICATION 6.8.4.F FOR PEAK CALCULATED CONTAINMENT INTERNAL PRESSURE By letter dated April 25, 2013, Dominion Nuclear Connecticut, Inc. (DNC) submitted a license amendment request (LAR) for Millstone Power Station Unit 3 (MPS3). The proposed amendment would revise the peak calculated containment internal pressure for the design basis loss of coolant accident described in Technical Specification (TS) 6.8.4.f, "Containment Leakage Rate Testing Program." The peak calculated containment internal pressure, Pa, would increase from 41.4 psig to 41.9 psig. In a letter dated August 8, 2013, the Nuclear Regulatory Commission (NRC) transmitted a request for additional information (RAI) to DNC related to the LAR. DNC agreed to respond to the RAI by September 23, 2013.
The attachment to this letter provides DNC's response to the NRC's RAI.
If you have any questions regarding this submittal, please contact Wanda Craft at (804) 273-4687.
Sincerely, Q)1%1 Daniel G. Stoddard Senior Vice President - Nuclear Operations COMMONWEALTH OF VIRGINIA )
)
COUNTY OF HENRICO The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Daniel G. Stoddard, who is Senior Vice President - Nuclear Operations 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.
Acknowledged before me this J 9-A day of 5, 2013.
My Commission Expires: IA 3' (G D NotayPbi Notary Public
=SLY Commonwealth of Virginia
- CRAIG Reg. # 7518653 My Commission Expires December 31, 201..k AD,0 0'4'L
Serial No.13-474 Docket No. 50-423 Page 2 of 2 Commitments made in this letter: None
Attachment:
Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 6.8.4.f for Peak Calculated Containment Internal Pressure cc: U.S. Nuclear Regulatory Commission Region I 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 J. 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.13-474 Docket No. 50-423 ATTACHMENT RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE AMENDMENT REQUEST TO REVISE TECHNICAL SPECIFICATION 6.8.4.F FOR PEAK CALCULATED CONTAINMENT INTERNAL PRESSURE DOMINION NUCLEAR CONNECTICUT, INC.
MILLSTONE POWER STATION UNIT 3
Serial No.13-474 Docket No. 50-423 Attachment 1, Page 1 of 3, By letter dated April 25, 2013, Dominion Nuclear Connecticut, Inc. (DNC) submitted a license amendment request (LAR) for Millstone Power Station Unit 3 (MPS3). The proposed amendment would revise the peak calculated containment internal pressure for the design basis loss of coolant accident (LOCA) described in Technical Specification (TS) 6.8.4.f, "Containment Leakage Rate Testing Program." The peak calculated containment internal pressure, Pa, would increase from 41.4 psig to 41.9 psig. In a letter dated August 8, 2013, the Nuclear Regulatory Commission (NRC) transmitted a request for additional information (RAI) to DNC related to the LAR. This attachment provides DNC's response to the NRC's RAI.
Question 1 Please verify that no change is needed in Technical Specification 3.6.1.4, "Containment Pressure," in order to maintain the calculated peak containment internal pressure (Pa) below the design limit. Please indicate the containment operatingpressure used as the initial condition for the Design-Basis Accident Loss-of-Coolant Accident analysis performed.
DNC Response When Dominion first applied the approved DOM-NAF-3-0.0-P-A GOTHIC containment analysis methodology to MPS3, parametric studies were performed spanning the steady state containment pressure requirements of TS 3.6.1.4 (10.6 psia to 14.0 psia). The most adverse initial pressure for the peak containment pressure cases was an analytical value of 14.2 psia.
The value of 14.2 psia was used as the initial condition in the reanalysis of the peak containment pressure for the design basis LOCA with the corrected Westinghouse mass and energy (M&E) data. No change was needed for TS 3.6.1.4 in order to maintain Pa below the design limit.
Question 2 Please verify that this issue is not related to the recent EPITOME computer modeling errors discoveredby Westinghouse.
DNC Response MPS3 calculates containment response using Dominion's NRC-approved DOM-NAF 0.0-P-A GOTHIC methodology. As part of that methodology, the vendor (Westinghouse) provides M&E inputs for the blowdown, refill and reflood phases of the LOCA transient. For the post-reflood phase, the Dominion GOTHIC model contains an
Serial No.13-474 Docket No. 50-423 Attachment 1, Page 2 of 3 internal reactor coolant system (RCS) model that is used to calculate the M&E releases to the GOTHIC containment model.
Westinghouse Nuclear Safety Advisory Letter (NSAL)-1 1-5 identified six issues that could potentially impact the MPS3 large break LOCA M&E calculations. The six issues, which include generic errors, are as follows:
- 1. The reactor vessel modeling did not include all the appropriate vessel metal mass available from the component drawings.
- 2. The reactor vessel modeling did not include all the appropriate vessel metal mass in the reactor vessel barrel/baffle region.
- 3. The reactor coolant pump (RCP) homologous curve input incorrectly included an absolute zero point coordinate.
- 4. The RCP homologous curve input incorrectly contained a sign error in a coordinate value.
- 5. The LOCA M&E release analysis initializes at a non-conservative (low) steam generator (SG) secondary pressure condition.
- 6. An error was found in the EPITOME computer code that is used to determine the M&E release rate during the long-term (i.e., post-reflood) SG depressurization phase of the LOCA transient.
Of these six identified issues, three were determined by Westinghouse in Table 1 of NSAL-1 1-5 to affect the MPS3 large break LOCA containment M&E release analysis, specifically, Issues 1, 2 and 5 (above) regarding the reactor vessel metal mass error, vessel barrel/baffle metal mass modeling, and the SG secondary pressure. Issues 1, 2, and 5 were addressed in revised M&E inputs that were used in the GOTHIC-based containment reanalysis for MPS3 that generated the requested change in Pa.
The EPITOME error listed as Issue 6 in NSAL-11-5 impacts the post-reflood M&E releases calculated by Westinghouse. MPS3 does not use Westinghouse M&E data during the post-reflood phase, relying instead on the RCS model embedded in the DOM-NAF-3-0.0-P-A GOTHIC model. Therefore, MPS3 was not affected by the EPITOME computer code error.
Question 3 Please state if the Appendix J Type B and C test procedures do not require revision upon approval of this proposed LAR. The American National Standards Institute (ANSI) 56.8-1994, Section 3.3.2 requires that Type B and C testing be performed at a pressure not less than Pa (except for airlock door seals, which may have a lower pressure
Serial No.13-474 Docket No. 50-423 Attachment 1, Page 3 of 3 specified) and not more than 1.1 times Pa when a higher differential pressure results in increased sealing. Please discuss the site procedures for Type B and C testing, and provide a discussion on the requirement that the testing be performed within a range of pressures that, with the revised Pa, will continue to be within the range of pressures requiredby ANSI 56.8-1994.
DNC Response The MPS3 Appendix J test procedures for Type B penetrations and Type C penetrations will require changes to identify the higher Pa value consistent with the proposed change to TS 6.8.4.f, "Containment Leakage Rate Testing Program."
The American National Standards Institute (ANSI) 56.8-1994, Section 3.3.2, requires that Type B and Type C testing be performed at a pressure not less than Pa and not more than 1.1 times Pa when a higher differential pressure results in increased sealing.
At MPS3, leakage rate testing is conducted using the "makeup flowrate" method as described in Section 6.4.2 of ANSI 56.8-1994 which states: "the test volume shall be pressurized and maintained to at least Pa, using a pressure regulator to maintain pressure". The makeup flowrate method is used to detect and measure local leakage across pressure containing boundaries based on a constant pressure process, while measuring discharge flow. In the makeup flow rate mode, the flow rate of the leak (discharge flow) can be read directly from the Volumetric Leak Rate Monitors (VLRM) digital flow rate readout. Both procedures for Type B and Type C local leakage rate testing identify the test pressure, Pa, as specified in TS 6.8.4.f and a maximum "not to exceed" design pressure, Pd (45 psig). Both specified pressures bound the range of pressures required by ANSI 56.8-1994.
Question 4 Please describe the steps that will be taken following approval of this amendment with regardto Appendix J Type A testing procedures.
DNC Response The Type A integrated leakage rate test procedure used during the most recent Type A testing at MPS3 is a vendor-controlled document that was approved by the Millstone technical subject matter experts and the facility safety review committee. This procedure is identified as a Category 1 (could affect containment isolation), infrequently conducted or complex evolution. As required by the procedure, the technical content will be validated prior to each use. The Type A test pressure is identified in the "Definitions" Section 2.2.9 of the vendor procedure and, as required, will be validated for accuracy prior to the next scheduled test, currently planned for 2025. The value of Pa identified in the Type A test procedure is consistent with the Pa value identified in TS 6.8.4.f.