View source for ND-20-1005, Supplement to Request for License Amendment: Vacuum Relief Valve Technical Specification Changes (LAR-20-005S1)

{{Adams
| number = ML20262H206
| issue date = 09/18/2020
| title = Supplement to Request for License Amendment: Vacuum Relief Valve Technical Specification Changes (LAR-20-005S1)
| author name = Chamberlain A
| author affiliation = Southern Nuclear Operating Co, Inc
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05200025, 05200026
| license number = 
| contact person = 
| case reference number = LAR-20-005S1, ND-20-1005
| document type = Letter type:ND, Response to Request for Additional Information (RAI)
| page count = 17
| project = 
| stage = Supplement
}}

=Text=
{{#Wiki_filter:A Southern Nuclear Southern Nuclear Operating Company, Inc.
3535 Colonnade Parkway Birmingham, AL 35243 Tel 205.992.7079 September 18, 2020 Docket Nos.: 52-025                                                                  ND-20-1005 52-026                                                           10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: 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:
Vacuum Relief Valve Technical Specification Changes (LAR-20-005S1)
Ladies and Gentlemen:
On June 19, 2020, pursuant to 10 CFR 52.98(c) and in accordance with 10 CFR 50.90, Southern Nuclear Operating Company (SNC) submitted License Amendment Request (LAR) LAR-20-005 per ND-20-0668 requesting an amendment to the combined licenses (COLs) for Vogtle Electric Generating Plant (VEGP) Units 3 and 4 (License Numbers NPF-91 and NPF-92, respectively)
(ADAMS Accession Number ML20171A563). On August 18, 2020, the NRC issued a Request for Additional Information (RAI) (ADAMS Accession Number ML20231A905) seeking clarifications related to the requested LAR.
This letter submits a supplement to LAR-20-005 responding to the RAI and including an editorial revision to the proposed COL Appendix A, Technical Specifications (TS) changes.
The LAR proposes changes to COL Appendix A, TS 3.6.3, Containment Isolation Valves, and TS 3.6.9, Vacuum Relief Valves, to exclude the vacuum relief containment isolation valves from TS LCO 3.6.3 and address the containment isolation function, operability, Actions, and Surveillances in TS 3.6.9.
This supplement does not impact the scope or conclusions of the Technical Evaluation, Significant Hazards Consideration Determination, or Environmental Considerations of the original submittal.
Enclosures 1, 2, and 3 were previously submitted to the NRC via ND-20-0668 (LAR-20-005). provides the responses to the RAI. Enclosures 5 and 6 contain revisions to Enclosures 2 through 3, respectively, (shown with Revision Bars).
This letter contains no regulatory commitments. This letter has been reviewed and determined not to contain security-related information.

U.S. Nuclear Regulatory Commission ND-20-1005 Page 2 of 4 SNC requests NRC staff review and approval of this LAR no later than December 30, 2020 to support the initial plant entry into MODE 4, which is the Applicability for the containment isolation function for the vacuum relief valves. Delayed approval of this license amendment could put the plant at increased risk of a TS required shutdown upon discovery of an inoperable vacuum relief valve containment isolation function. SNC expects to implement the proposed amendment within 30 days of approval of the LAR. This requested date is revised from the initial request submitted June 19, 2020 via ND-20-0668.
In accordance with 10 CFR 50.91 , SNC is notifying the State of Georgia by transmitting a copy of this letter and its enclosures to the designated State Official.
Should you have any questions, please contact Ms. Amy Chamberlain at (205) 992-6361 .
I declare under penalty of perjury that the foregoing is true and correct. Executed on the 18th of September 2020.
Respectfully submitted, C. Chamberlain Licensing Manger, Regulatory Affairs Southern Nuclear Operating Company Enclosures       1) through 3) previously provided in original submittal of LAR-20-005 dated June 19, 2020 (ML20171A563)
: 4) Vogtle Electric Generating Plant (VEGP) Units 3 and 4 - Responses to Requests for Additional Information (LAR-20-005S1)
: 5) Vogtle Electric Generating Plant (VEGP) Units 3 and 4 - Revised Proposed Changes to Licensing Basis Documents (LAR-20-005S 1)
: 6) Vogtle Electric Generating Plant (VEGP) Units 3 and 4 - Revised Conforming Changes to the Technical Specifications Bases (For Information Only)
(LAR-20-005S 1)

U.S. Nuclear Regulatory Commission ND-20-1005 Page 3 of 4 cc:
Southern Nuclear Operating Company / Georgia Power Company Mr. S. E. Kuczynski (w/o enclosures)
Mr. P. P. Sena III (w/o enclosures)
Mr. M. D. Meier (w/o enclosures)
Mr. G. Chick Mr. S. Stimac Mr. P. Martino Mr. D. L. McKinney (w/o enclosures)
Mr. T. W. Yelverton (w/o enclosures)
Mr. B. H. Whitley Ms. C. A. Gayheart Ms. M. Ronnlund Mr. D. L. Fulton Mr. M. J. Yox Mr. C. T. Defnall Mr. J. Tupik Ms. S. Agee Mr. M. Humphrey Ms. A. C. Chamberlain Mr. S. Leighty Mr. N. Kellenberger Mr. E. Riffle Ms. K. Roberts Mr. J. Haswell Mr. D. T. Blythe Mr. K. Warren Mr. A. S. Parton Document Services RTYPE: VND.LI.L00 File AR.01.02.06 Nuclear Regulatory Commission Mr. M. King (w/o enclosures)
Ms. M. Bailey w/o enclosures)
Mr. C. Patel Mr. C. Santos Mr. B. Kemker Mr. J. Eargle Mr. G. Khouri Mr. C. J. Even Mr. A. Lerch Mr. S. Walker Mr. N.D. Karlovich Ms. N. C. Coovert Mr. C. Welch Mr. J. Gaslevic Mr. V. Hall Ms. K. P. Carrington Mr. M. Webb

U.S. Nuclear Regulatory Commission ND-20-1005 Page 4 of 4 State of Georgia Mr. R. Dunn Oglethorpe Power Corporation Mr. M. W. Price 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. T. Rubenstein (w/o enclosures)
Mr. M. Corletti 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 Mr. R.L. Trokey, Georgia Public Service Commission Mr. K. C. Greene, Troutman Sanders Mr. S. Blanton, Balch Bingham

Southern Nuclear Operating Company ND-20-1005 Enclosure 4 Vogtle Electric Generating Plant (VEGP) Units 3 and 4 Responses to Requests for Additional Information (LAR-20-005S1)
(Enclosure 1 consists of 9 pages, including this cover page.)

ND-20-1005 Responses to Requests for Additional Information (LAR-20-005S1)
On August 18, 2020, the NRC issued a Request for Additional Information (RAI) (ADAMS Accession Number ML20231A905) seeking clarifications related to the requested License Amendment Request (LAR) 20-005 per ND-20-0668 (ML20171A563).
Additionally, based on discussions with the NRC Staff on August 6, 2020 (ML20212M116) regarding Surveillance Requirement (SR) 3.6.9.3 (proposed renumbering), Enclosure 5 provides an editorial revision to the proposed COL Appendix A, Technical Specifications (TS) changes provided in the original LAR. The original requested SR change required (emphasis added)
Verify each vacuum relief valve ... is being revised to provide additional clarity. The SR revision show in Enclosure 5 now requires (emphasis added) Verify each vacuum relief check valve and each vacuum relief isolation valve ... . This is an editorial enhancement to clearly articulate that each vacuum relief valve includes both the vacuum relief check valves and the vacuum relief isolation valves, which covers the entire scope of valves required to be operable by Limiting Condition for Operation (LCO) 3.6.9, Vacuum Relief Valves. The original request provided the technical justification for the change flow path to valve as a clarification. The revised change improves the clarity by detailing that the current flow path consists of both vacuum relief check valves and vacuum relief isolation valves.
Conforming TS Bases changes consistent with this SR 3.6.9.3 revision will be incorporated, following NRC approval of the license amendment request, in accordance with TS 5.5.6, Technical Specification Bases Control Program. The markups showing these revisions are provided in Enclosure 6 for information only.
The August 18, 2020 NRC RAI and Southern Nuclear Operating Company (SNC) responses follow.
NRC RAI:
The LAR does not appear to consider any insights from low power and shutdown operations. The LAR does not include a summary or description of the risk insights considered for the proposed change. To demonstrate that risk insights have been considered and as applicable, included, in the proposed change, including the determination of adequate defense-in-depth and safety margin, provide the following:
NRC RAI:
: a. Describe the dominant risk scenarios affected by the proposed change.
SNC Response:
The dominant risk scenarios detailed below are defined as the impacted cutsets that differ from the baseline results when the outboard motor operated containment vacuum relief isolation valves VFS-PL-800A and B or inboard containment vacuum relief isolation check valves VFS-PL-803A and B are being controlled open, reflecting the proposed change to allow inoperability for closure for 7 days. Note that for this risk-assessment no credit was taken to manually close the vacuum relief valves on a containment isolation signal.
Page 2 of 9

ND-20-1005 Responses to Requests for Additional Information (LAR-20-005S1)
The dominant impacted Core Damage Frequency (CDF) scenarios are due to loss of long-term containment cooling sequences. The following is a summary of the sequence for internal events.
The leading initiating events: loss of offsite power (LOOP) or general transient events with failure of long-term Startup Feedwater (SFW) operation due to Condensate Storage Tank (CST) make-up failure. The initiating event and loss of secondary side heat removal is an independent failure from containment vacuum relief isolation failure.
Containment isolation failure in vacuum relief penetration due to independent failure of the unaffected redundant isolation valve(s).
Independent failures of Passive Containment Cooling (PCS) actuation on a Diverse Actuation System (DAS) signal and operator failure to manually actuate PCS. Note that Protection and Safety Monitoring System (PMS) automatic actuation of PCS is assumed not to occur. PMS automatic PCS actuation setpoint is based on high containment pressure. If containment isolation for the vacuum relief flow path fails, the PMS actuation on Containment pressure high setpoint may not be reached.
The following is a summary of dominant impacted Large Early Release Frequency (LERF) scenarios for internal events.
Top sequences include general transients and smaller Loss of Coolant Accident (LOCA) events coupled with independent failures leading to core damage.
Containment isolation failure in vacuum relief penetration due to independent failure of the unaffected redundant isolation valve(s).
The following is a summary of the impacted dominant sequence for other hazards. The top impacted CDF sequences for internal fire and internal flood is similar to the internal event top CDF sequences with fire or flood induced initiating events. Seismic CDF results showed no sensitivity to VFS-PL-800A and B or VFS-PL-803A and B being controlled open.
Internal Fire LERF o The leading fire LERF sequence is a fire induced spurious Automatic Depressurization System (ADS) stage 3 with failure of Core Makeup Tank (CMT) injection (due to failure of Reactor Coolant Pump (RCP) to trip) leading to core damage.
Internal Flooding LERF o Top sequences include flooding induced general transient events coupled with independent failures leading to core damage.
Seismic LERF o Top sequences include seismic induced direct core damage scenarios.
NRC RAI:
: b. For the dominant risk scenarios identified in item (a), discuss risk management actions that will be used to manage the risk from the scenarios or provide justification that such actions are unnecessary. Such actions can include existing or enhanced procedures (e.g., checking that the unaffected containment isolation valves (CIVs) are in their proper position prior to performance of corrective maintenance or repair on the affected CIV(s)).
Page 3 of 9

ND-20-1005 Responses to Requests for Additional Information (LAR-20-005S1)
SNC Response:
Planned maintenance is screened for impacts related to nuclear safety. Examples include consideration of maintenance that results in entry into Technical Specifications Required Action Statements, inability to control a critical safety function (e.g., containment integrity), and inability to perform an Emergency Operating Procedure. Planned maintenance also bundles work to minimize plant risk and to minimize out of service time. Operations is the final authority for determining if work is done on-line or in an outage.
Operational Risk Awareness procedures also screen emergent, as well as planned maintenance.
These activities are classified as either medium or high risk. The associated risk management plans identify work activities that pose risk to personnel, plant equipment, or the environment are clearly identified, and an appropriate mitigation plan(s) developed to minimize or eliminate the likelihood of an unacceptable event.
The dominant risk scenarios identified in item (a) show that containment isolation failure of the vacuum relief flow path as the dominant failure driving the top cutsets. When operating with inoperable-for-closing vacuum relief check or vacuum relief isolation valve(s), the Operational Risk Awareness risk management actions would:
(1)      Suspend work currently in progress on vacuum relief check or vacuum relief isolation valve(s) (other than repairs necessary to restore operability);
(2)      Reschedule the start of pending work on any component associated with the containment vacuum relief penetration; (3)      Initiate restoration of out-of-service equipment associated with the vacuum relief penetration; and (4)      Maintain isolation of the vacuum relief penetration by maintaining the unaffected vacuum relief check or vacuum relief isolation valves closed; which avoids entry into proposed TS 3.6.9 Action D.
NRC RAI:
: c. For operation in Mode 4 with decay heat removal provided by the decay heat removal system (RNS), discuss any risk insights that were considered in the evaluation of the risk impact of the proposed change. The discussion should explain how the RNS, which is not safety related or included in the Technical Requirements Manual for Mode 4, is treated as part of the risk management process (e.g., existing TSs, regulatory treatment of non-safety systems (RTNSS), maintenance).
SNC Response:
The risk impact estimates performed for Mode 1 are conservative in comparison to Mode 4 operation with respect to the time until the containment isolation function is important, the time until core damage, and the time to pressurizing containment leading to release are all longer at Mode 4 when compared to at-power core damage scenarios. Therefore, there are no unique adverse risk impacts from the proposed change for operation in Mode 4 with decay heat removal provided by the Normal Residual Heat Removal System (RNS).
Page 4 of 9

ND-20-1005 Responses to Requests for Additional Information (LAR-20-005S1)
For Mode 4 with RNS in service, the plant has the same safety-related decay heat removal defenses as Mode 1, as well as the non-safety related steam generators (SGs) providing active decay heat removal. Core Makeup Tanks (CMTs), Passive Residual Heat Removal Heat Exchanger (PRHR HX), and In-Containment Refueling Water Storage Tank (IRWST) injection/
passive recirculation cooling methods continue to be available during Mode 4 whether or not RNS is providing decay heat removal.
In the event that RNS is lost during Mode 4 operation with RNS in service, the operators utilize the Loss of Normal Residual Heat Removal procedure to address the cause with the intent of restoring RNS for decay heat removal. If RNS cannot be restored, operators establish cooling using a secondary heat sink consisting of at least one intact SG, a means to add feedwater (main or startup feedwater pumps), and a means to dump steam in a controlled manner. If a secondary heat sink cannot be established, then PRHR HX cooling is initiated. If no means of RCS cooling can be established, then passive feed and bleed is established by actuation of Safeguards and transitioning to the Emergency Operation Procedures. Unless a heat sink is established that directs reactor decay heat outside containment using the RNS system or a SG, the PRHR HX is used to transfer heat to the IRWST. Heat is then transferred to the containment atmosphere by allowing the IRWST to heat up and boil. The IRWST can be cooled by the spent fuel pool cooling system if available. Potential actions to cool the containment atmosphere include operation of the containment fan coolers.
Also, as discussed in UFSAR Section 17.4 risk significant non-safety systems, structures, and components (SSCs) (including RNS as shown in Table 17.4-1) are included in the Operational Phase Reliability Assurance Activities (OPRAAS). These reliability assurance and investment protection programs include:
Maintenance Rule Program Quality Assurance Program Inservice Testing Program Inservice Inspection Program Investment Protection Short Term Availability Controls Site Maintenance Program There is no change to the RTNSS evaluation as a result of the risk impact estimates for the proposed changes to TS 3.6.3 and TS 3.6.9. The risk impact estimates for Mode 1 are conservative in comparison to Mode 4 operation with respect to the containment isolation function, the time until core damage, and time to pressurizing containment leading to release, which are all longer when compared to at-power core damage scenarios.
Safety-related normal decay heat removal is provided by the Passive Core Cooling System (PXS).
The requirements for Mode 4 operations of the PXS are provided in TS Section 3.5, Passive Core Cooling System. Specifically, TS 3.5.3, Core Makeup Tanks (CMTs) - Shutdown, RCS Intact, and TS 3.5.5, PRHR HX - Shutdown, RCS Intact, address the PXS component operability required when in Mode 4 with the RCS cooling provided by the RNS to provide the availability of safety-related decay heat removal in the event RNS is lost. TS 3.5.6, In-Containment Refueling Water Storage Tank (IRWST) - Operating, addresses the operability requirements for the IRWST in Mode 4.
Page 5 of 9

ND-20-1005 Responses to Requests for Additional Information (LAR-20-005S1)
NRC RAI:
: d. Based on the dominant risk scenarios and design basis accidents, identify the initiators that result in the earliest Containment Radioactivity High and/or Containment Isolation signal. For the identified initiators, if the unaffected CIV(s) fail to remain closed while the affected CIV(s) are in maintenance or repair, discuss the cues, available time, and procedures for operators to take action to isolate the containment.
SNC Response:
Probabilistic View:
The VFS isolation function supports core damage mitigation for loss of long-term containment cooling events within the PRA. For these long-term containment cooling events longer time windows are available for the operator to recover. Core damage due to loss of containment inventory during the loss of longer-term containment cooling events is not anticipated until after 24 hours into the event.
The VFS isolation supports LERF mitigation for the containment isolation function within the PRA.
The dominant internal events and internal flooding LERF sequences include general transients and smaller LOCA events. The bounding operator time window for containment isolation for small LOCA and general transient events is based on an upper bound 4 inch diameter small LOCA, with successful reactor trip, and no other means of mitigation credited. Available time from the start of the initiator until core damage is expected is 1289 seconds (or 21 minutes). The dominant internal Fire LERF sequences included fire induced spurious ADS Stage 1, 2, and 3 events. The dominant seismic sequences included seismic induced direct to core damage scenarios. The Vogtle PRA model does not credit manual containment isolation recovery to prevent LERF for larger LOCAs as assumed during the fire spurious ADS Stage 1, 2, and 3 scenario and the seismic direct core damage scenario.
Containment Isolation status monitoring is contained within the Emergency Operating Procedure (EOP) network. The cue to the operator to recover containment isolation failure is received early in the scenario on a safeguards actuation signal. This can occur at the start of the event like for the LOCA examples or at the point when PRHR HX is actuated during transient core damage scenarios. If proper isolation has not occurred, then manual Containment Isolation is performed from the Main Control Room (MCR). If containment penetrations are still not isolated, then operators are dispatched to perform actions locally to isolate Containment. Containment isolation status is also visually represented on the Critical Safety Function Status Trees to the operators in the MCR. Additionally, the Alarm Presentation System (APS) notifies the Operators both visually and audibly of a Containment Isolation actuation failure, which the operator addresses with an Alarm Response Procedure (ARP).
The position of the vacuum relief motor operated valves is a Post-Accident Monitored variable, is indicated in the MCR, and is provided as input to the plant computer. Position indication is based on actual valve position rather than demanded position.
Deterministic view:
The large-break LOCA is the limiting accident with respect to radiological consequences o Consistent with Position 3.3 of Reg. Guide 1.183 regarding leak-before-break plants, the gap release is assumed to start at 10 minutes. The initial coolant activity is released to containment and the containment is isolated per the onset of activity Page 6 of 9

ND-20-1005 Responses to Requests for Additional Information (LAR-20-005S1) release from the gap. In the event of the vacuum relief penetration were not isolated, the EOP procedures described above would be followed.
NRC RAI:
: e. Discuss whether the performance of the vacuum relief isolation valves and vacuum relief check valves will be monitored by the maintenance rule; or discuss how existing or planned performance monitoring strategies will ensure that the conclusions drawn from the risk insights remain valid.
SNC Response:
The vacuum relief isolation valves and vacuum relief check valves are monitored within the scope of the maintenance rule program as Maintenance Rule Functions VFS-001 and CNS-001.
NRC RAI:
: f. Describe potential impacts to defense-in-depth aspects of the plants design and operation and the adequacy of safety margins following the proposed change.
SNC Response:
There is no change to the safety-related functions (containment vacuum relief and containment isolation) due to the proposed TS 3.6.3 and TS 3.6.9 changes. The vacuum relief flow path includes two inboard and two outboard automatic isolation valves to provide redundant valve sets. The vacuum relief penetration provides inboard and outboard automatic isolation valves in accordance with General Design Criteria (GDC) 56 of 10 CFR 50, Appendix A.
There are no defense-in-depth functions performed by the Containment Air Filtration System (VFS).
There is no change to the input parameters for the containment vacuum relief analysis presented in UFSAR Subsection 6.2.1.1.4, External Pressure Analysis. Therefore, there is no impact to the safety margins with respect to the containment negative pressure event.
There is no change to the input parameters or assumptions for the safety analyses and radiological consequence analyses in UFSAR Chapter 15, and therefore, there is no reduction in safety margins for these analyses.
NRC RAI:
: g. Provide the basis for screening the impact of other external hazards and low power and shutdown on the proposed change.
SNC Response:
External Hazards o The identified at-power external hazards (excluding Seismic) were screened out based on the conclusions of the at-power External Hazards Assessment. Qualitative and quantitative screening of external hazards was performed consistent with the requirements of the ASME/ANS PRA Standard (ASME/ANS RA-Sa-2009) for this analysis. These assessments indicate external hazards are individually screened with CDFs between 1.E-09 per year to approximately 1E-14 per year. During the 7-day Completion Time proposed in the TS change the associated core damage probability, Page 7 of 9

ND-20-1005 Responses to Requests for Additional Information (LAR-20-005S1)
CDP, would be approximately in the range of 2E-11 to 2E-16 per year. Given the low probability of external hazards (other than Seismic) they may be screened from the impact of VFS-PL-V800A and B or VFS-PL-V803A and B being maintained open for a 7-day period. These contributions are considered insignificant, particularly when compared to the impact on Internal Events, Internal Flooding, Internal Fire and Seismic.
Low Power and Shutdown o For Modes 1, 2, and 3 the insights from at-power PRA are applicable. Modes 2 and 3 are considered similar to the at-power PRA model. The at-power model would conservatively bound the Mode 3 configuration (no ATWS and potentially lower LOCA frequencies). Initiating events during Mode 2 and 3 also provide more time for operator responses.
o For Mode 4 operation, the insights from the at-power PRA model are conservative with respect to the containment isolation function, the time until core damage, and the time to pressurize containment leading to release, which are all are longer when compared to at-power core damage scenarios.
o For Mode 4 operation with RNS in service, the plant decay heat removal defenses are not reduced. For example, SGs, PRHR HX and IRWST injection/ passive recirculation cooling methods are available.
o Therefore, it is acceptable to use the at-power PRA model in lieu of developing quantitative impacts for low power and shutdown operations.
NRC RAI:
: h. Discuss if and how key assumptions and sources of uncertainty in the inputs for the risk insights impact the responses to the above items. If key assumptions and sources of uncertainty impact the responses, explain how they will be dispositioned for the proposed change.
SNC Response:
No new sources of uncertainty were identified during the development of this sensitivity analysis.
The following table includes sources of model uncertainty identified related to the containment vacuum relief isolation function from the baseline at power model. Note that no Internal Flooding, Internal Fire, and Seismic specific related sources of model uncertainty pertinent to this application were identified. The following table includes the related sources of model uncertainty and a generic baseline characterization. The final column provides a characterization for this application.
The impact of these uncertainties is assessed to be enveloped within the RG 1.177 small quantitative impact threshold (Incremental Conditional Core Damage Probability (ICCDP) of less than 1.0x10-6 and Incremental Conditional Large Early Release Probability (ICLERP) of less than 1.0x10-7) with sufficient margin. These uncertainties are not expected to impact the conclusions that a Completion Time of 7 days is acceptable.
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ND-20-1005 Responses to Requests for Additional Information (LAR-20-005S1)
Generic/
Plant Specific Assumptions Made/                 Baseline           LAR-234 Risk Assessment -
Approach Uncertainty                 Impact on                Characterization Taken Model Small internal     Small internal leak size has been     Conservative Conservative approach.
leak failure       conservatively binned into the        approach modes have         same Level 2 category (LERF,                       This assumption has a conservative impact been               Small Early Release Frequency                      on the loss of VFS containment isolation conservatively     (SERF), or Late releases) as the                   function. Small leakage failure modes for binned into large  large leak and the penetration                     VFS-PL-V800A/B and VFS-PL-V803A/B release groups     itself.                                            are included in the model. A small leak for level 2                                                           <50 gpm may not generate a large release.
Manual             Only automatic containment            Conservative This assumption is conservative with containment        isolation is credited for the low     approach     respect to baseline LERF values.
isolation is not   RCS pressure scenarios due the credited for low   potential of core damage                           Since no credit was taken to recover and RCS pressure       occurring early in the event before                manually close VFS-PL-V800A and B or core damage        operator intervention is                           VFS-PL-V803A and B, the change in LERF scenarios to       reasonable. Low pressure core                      from the baseline to the sensitivity cases prevent LERF       damage sequences are                               may increase with more model refinement determined based on                                on the baseline model.
depressurization that results from the initiating event (e.g., LLOCA                  With an estimated ICLERP of 6.96E-09 for and MLOCA) and the status of                       the VFS valves to be control open for 7 ADS. Sufficient time may be                        days sufficient margin is available to 1E-7 available for smaller LOCAs and                    limit to envelope this uncertainty.
transients with successful ADS to credit manual containment isolation, however it is conservatively not credited in the fault trees.
Containment        Lower bound containment               Conservative This assumption has a conservative impact isolation failures isolation failure size that permits   approach     on the loss of long-term cooling CDF are assumed to     containment to pressurize with                     results.
fail PCS PMS       lower bound energy release to actuation on       containment to actuate PCS water                   Failure of the VFS pathway due to valve high               is unknown.                                        leakage may not fail automatic PMS containment                                                           actuation (high containment pressure pressure                                                              setpoint may be reached). This assumption could also reduce the baseline CDF if refined.
With an estimated ICCDP of 6.17E-09 for the VFS valves to be control open for 7 days sufficient margin is available to 1E-06 limit to envelope this uncertainty.
Page 9 of 9

Southern Nuclear Operating Company ND-20-1005 Enclosure 5 Vogtle Electric Generating Plant (VEGP) Units 3 and 4 Revised Proposed Changes to Licensing Basis Documents (LAR-20-005S1)
Insertions Denoted by Blue Underline and Deletions by Red Strikethrough Omitted text is identified by three asterisks ( * * * )
Revisions from LAR-20-005 annotated with Revision Bars (Enclosure 5 consists of two pages, including this cover page.)

ND-20-1005 Revised Proposed Changes to Licensing Basis Documents (LAR-20-005S1)
Technical Specification 3.6.9, Vacuum Relief Valves:
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                 FREQUENCY SR 3.6.9.32      Verify each vacuum relief flow path check valve and In accordance each vacuum relief isolation valve is OPERABLE in   with the Inservice accordance with the Inservice Testing Program.      Testing Program Page 2 of 2

Southern Nuclear Operating Company ND-20-1005 Enclosure 6 Vogtle Electric Generating Plant (VEGP) Units 3 and 4 Revised Conforming Changes to the Technical Specifications Bases (For Information Only)
(LAR-20-005S1)
Insertions Denoted by Blue Underline and Deletions by Red Strikethrough Omitted text is identified by three asterisks ( * * * )
Revisions from LAR-20-005 annotated with Revision Bars (Enclosure 6 consists of two pages, including this cover page.)

ND-20-1005 Revised Conforming Changes to the Technical Specifications Bases (For Information Only)
(LAR-20-005S1)
Technical Specifications Bases B 3.6.9, Vacuum Relief Valves SURVEILLANCE        * *
* REQUIREMENTS SR 3.6.9.32 Each vacuum relief check valve and each vacuum relief isolation valve is tested to be OPERABLE for opening and OPERABLE for closing. This SR cites the Inservice Testing Program, which establishes the requirement that inservice testing of the ASME Code Class 1, 2, and 3 valves shall be performed in accordance with the ASME OM Code (Ref. 2). Therefore, SR Frequency is governed by the Inservice Testing Program.
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