ND-18-0302, Supplement to Request for License Amendment and Exemption Regarding Improvements to Main Control Room (MCR) Post-Accident Radiological Consequences (LAR-17-023S2)

Supplement to Request for License Amendment and Exemption Regarding Improvements to Main Control Room (MCR) Post-Accident Radiological Consequences (LAR-17-023S2)

ML18067A648
Person / Time
Site:	Vogtle
Issue date:	03/08/2018
From:	Whitley B Southern Nuclear Operating Co
To:	Document Control Desk, Office of New Reactors
References
LAR-17-023S2, ND-18-0302
Download: ML18067A648 (10)

Text

B. H. Whitley Southern Nuclear Director Operating Company, Inc.

Regulatory Affairs 42 Inverness Center Parkway Birmingham, AL 35242 Tel 205.992.7079 Fax 205.992.5296 March 8, 2018 Docket Nos.: 52-025 ND-18-0302 52-026 10 CFR 50.90 10 CFR 52.63 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555-0001 Southern Nuclear Operating Company Vogtle Electric Generating Plant Units 3 and 4 Supplement to Request for License Amendment and Exemption Regarding Improvements to Main Control Room (MCR) Post-Accident Radiological Consequences (LAR-17-023S2)

Ladies and Gentlemen:

Pursuant to 10 CFR 52.98(c) and in accordance with 10 CFR 50.90, Southern Nuclear Operating Company (SNC), the licensee for Vogtle Electric Generating Plant (VEGP) Units 3 and 4, requested an amendment to Combined License (COL) Numbers NPF-91 and NPF-92, for VEGP Units 3 and 4, respectively, by SNC letter ND-17-1297, dated August 31, 2017 [ADAMS Accession Number ML17243A352]. License Amendment Request (LAR)17-023, proposes to revise the licensing basis information regarding the nuclear island non-radioactive ventilation system (VBS), the main control room emergency habitability system (VES), and post-accident operator dose analyses, to maintain compliance with 10 CFR Part 50, Appendix A, General Design Criterion (GDC) 19, which requires that main control room (MCR) personnel dose does not exceed 5 rem total effective dose equivalent (TEDE) for the duration of a design basis accident (DBA). Pursuant to the provisions of 10 CFR 52.63(b)(1), an exemption from elements of the design as certified in the 10 CFR Part 52, Appendix D, design certification rule was also requested for the plant-specific DCD Tier 1 material departures.

Enclosures 1 through 6 were provided with the original LAR-17-023. Enclosures 7 through 11 were provided with the supplement to LAR-17-023 (LAR-17-023S1) on February 9, 2018, with SNC letter ND-18-0096 [ADAMS ML18040A488 and ML18040A489], which provided the response to an NRC Staff Request for Additional Information (RAl) that was provided to SNC on January 30, 2018

[ADAMS ML18030B069]. Enclosure 12 to this letter provides a voluntary supplement to the information provided in LAR-17-023S1 by addressing inconsistencies that were identified after ND-18-0096 was submitted. Enclosure 13 contains revised licensing basis document markups to reflect the changes described in Enclosure 12.

U.S. Nuclear Regulatory Commission ND-18-0302 Page 2 of 4 The supplemental information provided in this LAR supplement does not impact the scope, technical content, or conclusions of the Technical Evaluation, Regulatory Evaluation (including the Significant Hazards Consideration Determination), or Environmental Considerations of the original LAR provided in letter ND-17 -1297, Enclosure 1.

This letter contains no regulatory commitments. This letter, including enclosures, has been reviewed and confirmed to not contain security-related information.

SNC now requests NRC Staff review and approval of the license amendment and exemption no later than April 20, 2018, to allow sufficient time to implement licensing basis changes necessary to support operator training.

In accordance with 10 CFR 50.91, SNC is notifying the State of Georgia of this LAR supplement by transmitting a copy of this letter and enclosures to the designated State Official.

Should you have any questions, please contact Mr. Wesley Sparkman at (205) 992-5061.

I declare under penalty of pe~ury that the foregoing is true and correct. Executed on the 8th of March 2018.

Respectfully submitted, f

Brian H. Whitley Director, Regulatory Affairs Southern Nuclear Operating Company

Enclosures:

1 - 6) (previously submitted with the original LAR, LAR-17-023, in SNC letter ND-17-1297) 7- 11) (previously submitted with LAR-17-023S1 in SNC letter ND-18-0096)

12) Vogtle Electric Generating Plant (VEGP) Units 3 and 4- Voluntary Supplemental Information Regarding the LAR-17-023 Review (LAR-17 -023S2)

13) Vogtle Electric Generating Plant (VEGP) Units 3 and 4- Revised Proposed Changes to the Licensing Basis Documents (LAR-17-023S2)

U.S. Nuclear Regulatory Commission ND-18-0302 Page 3 of 4 cc:

Southern Nuclear Operating Company / Georgia Power Company Mr. S. E. Kuczynski (w/o enclosures)

Mr. M. D. Rauckhorst Mr. D. G. Bost (w/o enclosures)

Mr. M. D. Meier (w/o enclosures)

Mr. D. H. Jones (w/o enclosures)

Mr. D. L. McKinney (w/o enclosures)

Mr. T. W. Yelverton (w/o enclosures)

Mr. B. H. Whitley Mr. J. J. Hutto Mr. C. R. Pierce Ms. A. G. Aughtman Mr. D. L. Fulton Mr. M. J. Yox Mr. J. Tupik Mr. W. A. Sparkman Ms. A. C. Chamberlain Ms. A. L. Pugh Mr. J. D. Williams Mr. F. J. Redwanz Document Services RTYPE: VND.LI.L00 File AR.01.02.06 Nuclear Regulatory Commission Mr. W. Jones (w/o enclosures)

Ms. J. Dixon-Herrity Mr. C. Patel Ms. J. M. Heisserer Mr. B. Kemker Mr. G. Khouri Ms. S. Temple Mr. F. Brown Mr. T.E. Chandler Ms. P. Braxton Mr. T. Brimfield Mr. C. J. Even Mr. A. Lerch State of Georgia Mr. R. Dunn

U.S. Nuclear Regulatory Commission ND-18-0302 Page 4 of 4 Oglethorpe Power Corporation Mr. M. W. Price Mr. K. T. Haynes Ms. A. Whaley Municipal Electric Authority of Georgia Mr. J. E. Fuller Mr. S. M. Jackson Dalton Utilities Mr. T. Bundros Westinghouse Electric Company, LLC Mr. L. Oriani (w/o enclosures)

Mr. G. Koucheravy (w/o enclosures)

Mr. M. Corletti Mr. M. L. Clyde Ms. L. Iller Mr. D. Hawkins Mr. J. Coward Other Mr. S. W. Kline, Bechtel Power Corporation Ms. L. A. Matis, Tetra Tech NUS, Inc.

Dr. W. R. Jacobs, Jr., Ph.D., GDS Associates, Inc.

Mr. S. Roetger, Georgia Public Service Commission Ms. S. W. Kernizan, Georgia Public Service Commission Mr. K. C. Greene, Troutman Sanders Mr. S. Blanton, Balch Bingham Mr. R. Grumbir, APOG NDDocumentinBox@duke-energy.com, Duke Energy Mr. S. Franzone, Florida Power & Light

Southern Nuclear Operating Company ND-18-0302 Enclosure 12 Vogtle Electric Generating Plant (VEGP) Units 3 and 4 Voluntary Supplemental Information Regarding the LAR-17-023 Review (LAR-17-023S2)

(This Enclosure consists of three pages, including this cover page.)

ND-18-0302 2 Voluntary Supplemental Information Regarding the LAR-17-023 Review (LAR-17-023S2)

The following supplements the information provided by Southern Nuclear Operating Company (SNC) in letter ND-18-0096, dated February 9, 2018 [ADAMS Accession Numbers

[ML18040A488 and ML18040A489], which provided responses to questions raised by the Staff during the review of SNC License Amendment Request (LAR)17-023.

Supplemental information regarding RAI No. 06.04-1 response:

SNCs response to RAI No. 06.04-1 provided an assessment of the iodine decontamination factor applicable to a pool water depth of 18.6 feet above fuel damaged in the design basis fuel handling accident (FHA), including the basis for the method used to determine the overall effective iodine decontamination factor. However, SNCs response indicates that the pool water depth of a dropped fuel assembly lying horizontally on top of the spent fuel racks would be greater than 18.6 feet; therefore, SNC is revising the proposed UFSAR change to more closely align with the available water depth (i.e., 21.5 feet, providing some margin), rather than the calculated minimum water depth (i.e., 18.6 feet). Accordingly, the proposed change to UFSAR subsection 15.7.4.2 is revised to include the following paragraph:

In the case of a single bundle dropped and lying horizontally on top of the spent fuel racks, there may be less than 23 feet of water above the top of the fuel bundle and the surface of the water, indicated by the width of the bundle. The fuel handling accident analysis bounds the case of a single bundle lying horizontally on top of the spent fuel racks by demonstrating that the overall decontamination factor of 200 is valid for pool depths of 21.5 feet.

The revised proposed licensing basis change is provided in Enclosure 13 of this letter.

Supplemental information regarding RAI No. 06.04-3 response:

SNCs response to RAI No. 06.04-3 provided information used in a shielding and radiation protection assessment and calculation regarding a previously unidentified post-accident mission travel path. This travel path is associated with the task of retrieving the main control room ancillary fans in postulated post-accident conditions, as depicted in the markup of Figure 12.3-2, Sheet 7, in LAR 17-023, Enclosure 4. The response included a detailed list, Table 3-1, of the most critical inputs and assumptions used in the assessment. Table 3-1 identified a value of 35 Rem/hr for the input parameter identified as the Highest Inhalation Dose Rate in Annex Building (APF=1). It was subsequently identified that this inhalation dose rate corresponds to an Assigned Protection Factor (APF) of 50, not 1, as identified in the table. Most post-accident vital area actions were determined to require some level of respiratory protection to meet the 5 Rem limit; therefore, the dose rate corresponding to the APF of 50, as addressed in the sensitivity analysis discussed in the response to RAI No. 06.04-3, was used.

Supplemental information regarding the Correction to the Markup Convention for UFSAR Table 2.3-201:

In letter ND-18-0096, Enclosure 7, SNC provided a correction to the markup convention that was used for the proposed changes to the plant-specific atmospheric dispersion factors, also referred to as /Q values, in UFSAR Section 2.3, Meteorology, Table 2.3-201, ARCON96 /Q Values at the Control Room HVAC Intake. The correct convention was used in the revised marked up Table 2.3-201 provided in Enclosure 8 of ND-18-0096.

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ND-18-0302 2 Voluntary Supplemental Information Regarding the LAR-17-023 Review (LAR-17-023S2)

However, it was subsequently identified that two of the proposed /Q values included editorial errors. Specifically, for the Radwaste Building Truck Staging Area Door, the proposed /Q value for 1 to 4 days should have been 2.98E-04, not 2.98E-05 as presented, and the proposed

/Q value for 4 to 30 days should have been 2.09E-04, not 2.09E-05 as presented (emphasis added). All other values in this table are consistent with the design basis calculation and the markup of UFSAR Table 2.0-202 (Sheet 1 of 2).

To correct this editorial error, the marked-up version of UFSAR Table 2.3-201 provided in ND-18-0096 is replaced in its entirety with a revised marked-up version in Enclosure 13 of this letter.

Supplemental Information regarding a Radiation Dose Analyses Assumption During the conduct of an audit of Westinghouse radiation dose analyses supporting LAR-17-023, it was identified that the analyses assumed the use of personnel respiratory protection equipment to maintain radiation exposure within personnel exposure limits; however, this assumption is not addressed in the UFSAR. To avoid ambiguity regarding this assumption, a change is proposed to revise UFSAR subsection 12.4.1.8 to identify this assumption. The marked-up version of UFSAR subsection 12.4.1.8 is provided in Enclosure 13 of this letter.

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Southern Nuclear Operating Company ND-18-0302 Enclosure 13 Vogtle Electric Generating Plant (VEGP) Units 3 and 4 Revised Proposed Changes to the Licensing Basis Documents (LAR-17-023S2)

Note:

Added text is shown as bold Blue Underline Deletions are shown as Red Strike-out New changes made in this supplement are shown as bold Black Underline, with a change bar in the right-hand margin adjacent to the changed text.

(This Enclosure consists of three pages, including this cover page.)

ND-18-0302 Enclosure 13 Revised Proposed Changes to the Licensing Basis Documents (LAR-17-023S2)

NOTE:

The following change replaces the proposed change to UFSAR Table 2.3-201 that was provided in SNC letter ND-18-0096, Enclosure 8, page 6 of 6, in its entirety. A change bar is provided in the right-hand margin adjacent to the text changed by this supplement.

Revise UFSAR Tier 2 Table 2.3-201, ARCON96 /Q Values at the Control Room HVAC Intake, as follows:

0-2 2-8 8 - 24 1-4 4 - 30 Release Point hours hours hours days days 2.02E-03 1.58E-03 6.37E-04 5.12E-04 3.82E-04 Plant Vent 2.27E-03 1.86E-03 7.36E-04 5.99E-04 4.31E-04 1.68E-03 1.29E-03 5.47E-04 4.55E-04 3.34E-04 PCS Air Diffuser 1.71E-03 1.32E-03 5.56E-04 4.63E-04 3.43E-04 Fuel Auxiliary Building Fuel 1.54E-03 1.11E-03 4.42E-04 3.57E-04 2.59E-04 Handling Area Blowout Panel 1.57E-03 1.15E-03 4.62E-04 3.72E-04 2.68E-04 Fuel Radwaste Building Rail Bay 1.15E-03 8.29E-04 3.35E-04 2.62E-04 1.86E-04 Truck Staging Area Door 1.30E-03 9.36E-04 3.78E-04 2.98E-04 2.09E-04 1.48E-02 1.20E-02 5.41E-03 3.93E-03 3.26E-03 Steam Line Break 1.87E-02 1.51E-02 6.79E-03 4.94E-03 4.14E-03 1.31E-02 1.02E-02 4.62E-03 3.29E-03 2.77E-03 PORV & Safety Valves 1.77E-02 1.41E-02 6.25E-03 4.61E-03 3.87E-03 6.23E-04 4.57E-04 2.05E-04 1.49E-04 1.12E-04 Condenser Air Removal Stack 6.60E-04 4.83E-04 2.17E-04 1.57E-04 1.17E-04 Containment Shell 3.20E-03 1.82E-03 8.27E-04 7.22E-04 5.70E-04 (As Diffuse Area Source) 2.93E-03 1.75E-03 7.78E-04 6.81E-04 5.30E-04 Page 2 of 3

ND-18-0302 Enclosure 13 Revised Proposed Changes to the Licensing Basis Documents (LAR-17-023S2)

Revise UFSAR Tier 2 Subsection 12.4.1.8, Post-Accident Actions, by adding a new sentence to identify the assumption of the use of respiratory protection equipment, as shown below.

12.4.1.8 Post-Accident Actions Requirements of 10 CFR 52.79(b) The analyses that confirm that the individual personnel exposure limits following an accident are not exceeded reflect the time-dependency of the area dose rates and the required post-accident access times. The analyses include the assumption that the appropriate respiratory protection equipment is used to maintain radiation exposure within the exposure limits. The areas that require post-accident accessibility are:

NOTE:

The following change replaces the proposed change to UFSAR Subsection 15.7.4.2 that was originally provided in SNC letter ND-17-1297, Enclosure 3, page 60 of 64, in its entirety.

A change bar is provided in the right-hand margin adjacent to the text changed by this supplement.

Revise UFSAR Tier 2 Subsection 15.7.4.2, Release Pathways, by adding a new paragraph after the first paragraph, to describe the column of water above a dropped fuel bundle, as shown below.

15.7.4.2 Release Pathways The spent fuel handling operations take place underwater. Thus, activity releases are first scrubbed by the column of water 23 feet in depth. This has no effect on the releases of noble gases or organic iodine but there is a significant removal of elemental iodine.

Consistent with the guidance in Regulatory Guide 1.183, the overall pool scrubbing decontamination factor for iodine is assumed to be 200.

In the case of a single bundle dropped and lying horizontally on top of the spent fuel racks, there may be less than 23 feet of water above the top of the fuel bundle and the surface of the water, indicated by the width of the bundle. The fuel handling accident analysis bounds the case of a single bundle lying horizontally on top of the spent fuel racks by demonstrating that the overall decontamination factor of 200 is valid for pool depths of 21.5 feet.

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