View source for NL-18-0367, Revision 21 to Updated Final Safety Analysis Report, Technical Specification Bases Changes, Technical Requirements Manual Changes, 10 CFR 50.59 Summary Report, and Revised NRC Commitments Report

{{Adams
| number = ML18101A046
| issue date = 03/29/2018
| title = Revision 21 to Updated Final Safety Analysis Report, Technical Specification Bases Changes, Technical Requirements Manual Changes, 10 CFR 50.59 Summary Report, and Revised NRC Commitments Report
| author name = 
| author affiliation = Southern Nuclear Operating Co, Inc
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000424, 05000425
| license number = 
| contact person = 
| case reference number = NL-18-0367
| package number = ML18101A044
| document type = Letter type:NL, Updated Final Safety Analysis Report (UFSAR)
| page count = 25
}}

=Text=
{{#Wiki_filter:~ Southern Nudear                                   Cheryl A. Gayheart Regulatory Affairs Director 40 Inverness Center Parkway Post Office Box l 295 Birmingham, AL 35242 205 992 53 l 6    tel 205 992 760 l     fax cagayhea@southernco.com March 29, 2018 Docket Nos.: 50-424                                                               NL-18-0367 50-425 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Vogtle Electric Generating Plant, Units 1 & 2 Revision 21 to the Updated Final Safety Analysis Report, Technical Specification Bases Changes, Technical Requirements Manual Changes, 10 CFR 50.59 Summary Report. and Revised NRC Commitments Report Ladies and Gentlemen:
In accordance with 10 CFR 50.4(b) and 50.71 (e), Southern Nuclear Operating Company (SNC) hereby submits Revision 21 *to the Vogtle Electric Generating Plant (VEGP) Units 1
* and 2 Updated Final Safety Analysis Report (UFSAR). The revised VEGP Units 1 and 2 UFSAR pages, indicated as Revision 21, reflect changes through February 28, 2018.
The VEGP Units 1 and 2 Technical Specifications, Section 5.5.14, *"Technical Specifications (TS) Bases Control Program," provides for changes to the Bases without prior NRC approval. In addition, TS Section 5.5.14 requires that Bases changes made without prior NRC approval be provided to the NRC on a frequency consistent with 10 CFR 50.71 (e).
Pursuant to TS 5.5.14, SNC hereby submits a complete copy of the VEGP Units 1 and 2 TS Bases. The revised VEGP Units 1 and 2 TS Bases pages, indicated as Revision 52, reflect changes to the TS Bases through February 2018.
In accordance with Regulatory Issue Summary (RIS) 2001-05, "Guidance on Submitting Documents to the NRC by Electronic Information Exchange or on CD-ROM," all of the current pages of the VEGP Units 1 and 2 UFSAR, the VEGP Units 1 and 2 UFSAR reference drawings, the TS Bases, and the Technical Requirements Manual (TRM) are being submitted on CD-ROM in portable document format (PDF). The revised VEGP Units 1 and 2 TRM pages, indicated as Revision 47, reflect changes to the TRM through February 2018.
* In accordance with 10 CFR 50.59(d)(2), SNC hereby submits the 10 CFR 50.59 Report containing a brief description of any changes, tests, or experiments, including a summary of the. safety evaluation of each. This report is based on the same time period as Revision 21 of the UFSAR.
In accordance with NEI 99-04, "Guidelines for Managing NRC Commitment Changes,"
Revision 0, SNC hereby submits a Revised NRC Commitments Report containing the                             At)5_-3
:rEJ/7
                                                                                                               ;v.zl(

U. S. Nuclear Regulatory Commission NL-18-0367 Page 2 original commitment, the revised commitment, and the justification for the change. This report is based on the same time period as Revision 21 of the UFSAR. provides a table of contents with associated file names for the set of three CD-ROMs (Enclosure 2). Enclosure 3 provides the 10 CFR 50.59 Summary Report. provides the Revised NRC Commitments Report.
This letter contains no NRC commitments. If you have any questions, please contact Ken McElroy at 205.992.7369.
I declare under penalty of perjury that the foregoing is true and correct. Executed on the 29th day of March 2018.
Respectfully submitted, Cheryl     yh Regulatory A airs Director CAGffLE/cbg

==Enclosures:==
: 1. CD-ROM Table of Contents
: 2. CD-ROMs (3 discs) containing .Files 001 - 029
: 3. 10 CFR 50.59 Summary Report
: 4. Revised NRC Commitments Report cc: Regional Administrator, Region II (w/o enclosures)
Senior NRR Project Manager - VEGP Units 1 and 2 (w/o enclosures)
Senior Resident Inspector - VEGP Units 1 and 2 (w/o enclosures)
INPO Emergency Management Manager (Enclosure 2, CD ROMs, only)
RType: CVC7000

NL-18-0367 Vogtle Electric Generating Plant, Units 1 & 2 Revision 21 to the Updated Final Safety Analysis Report, Technical Specification Bases Changes, Technical Requirements Manual Changes, 10 CFR 50.59 Summary Report, and Revised NRC Commitments Report Enclosure 1 CD-ROM Table of Contents to NL-18-0367 CD-ROM Table of Contents I                                       FILENAME SEQ   CONTENT                                           EXTENSION DISC 1 001   VEGP FSAR_CH1, CH2 (PRT 1)                        .pdf Chapter 1 Chapter 2 (PRT 1)
Section 2.1 to 2.3 to Figure 2.3.5-4 002   VEGP FSAR_CH2 (PRT 2) & CH3 (PRT 1)               .pdf Chapter 2 (PRT 2)
Figure 2.3.5-5 to 2.3.5-6 Section 2.4 Table 2.4.1-1 to Table 2.4: 13-1 Figures 2.4.1-1 Section 2.5 Appendix 2A Appendix 28 Chapter 3 (PRT 1)
Section 3.1 to 3.6 Section 3.7 to Figure 3.7.8.2-24 003   VEGP FSAR_CH3 (PRT 2)                             .pdf Chapter 3 (PRT 2)
Figures 3.7.8.2-25 to 3.7.4-1 Section 3.8 to 3.11 Appendix 3A to 3C Appendix 30 to Figure 30-17 004    VEGP FSAR_CH3 (PRT 3)                            .pdf Chapter 3 (PRT 3)
Appendix 30 Figures 30-18 to 30-40 005    VEGP FSAR_CH3 (PRT 4)                            .pdf Chapter 3 (PRT 4)
Appendix 30 Figures 30-41 to 30-65 006    VEGP FSAR~CH3 (PRT 5)                            *.pdf Chapter 3 (Part 5)
Appendix 30 Figure 30-66 to 30-92 007    VEGP FSAR_CH 3 (PRT 6)                           .pdf Chapter 3 (Part 6)
Appendix 30-93 to 30-136 E1-1 to NL-18-0367 CD-ROM Table of Contents I                                         FILENAME SEQ   CONTENT                                                  EXTENSION 008   VEGP FSAR_CH 3 (PRT 7) & CH 4 (PRT 1)                    .pdf Appendix 3D-137 to 3D-148 Appendix 3E Appendix 3F Chapter4 (PRT 1)
Section 4.1 to 4.2 Section 4.3 009   VEGP FSAR_CH 4 PRT 2, CH5, & CH6 (PRT 1)                 .pdf Chapter 4 (PRT 2)
Tables 4.3-1 to 4.3-12 Figures 4.3-1 to 4.3-57 Section 4.4 Section 4.5 Section 4.6 Appendix 4A Chapter 5 Chapter 6 (PRT 1)
Section 6.1             .
Section 6.2 to Figure 6.2.1-15 (SH 14) 010   VEGP FSAR_CH6 (PRT 2)                                    .pdf.
Figure 6.2.1-15 (SH 15) to 6.2-21 (SH 74)
D1SC2 011   VEGP FSAR_CH6 (PRT 3) & CH7 (PRT 1)                      .pdf Figure 6.2.1-22 (SH 1) to 6.2.5-7 Section 6.3 to Section 6.6 Chapter 7 (PRT 1)
Section 7 .1 to 7 .5 Section 7.6 to Figure 7.6.4-1 012   VEGP FSAR_CH7 (PRT 2), CH8 (PRT 1)                       .pdf Chapter 7 (PRT 2)
Figure 7.6.5-1 Section 7.7 to Figure 7.7.2-2 Chapter 8 (PRT 1)
Section 8.1 to 8.2 Section 8.3 to Figure 8.3.1-1 (SH 4)
E1-2 to NL-18-0367 CD-ROM Table of Contents I                                       FILENAME SEQ   CONTENT                                            EXTENSION 013   VEGP FSAR_CH8 (PRT 2)                              .pdf Chapter 8 (PRT 2)
Figure 8.3.1-1 (SHs 5 thru 24) 014   VEGP FSAR_CH 8 (PRT 3), CH 9, CH 10 (PRT 1)        .pdf Figure 8.3.1-1 (SHs 25 thru 34)
Section 8.4 Chapter 9 Section 10. 1 Section 10.2 to Figure 10.2.2-2b 015    VEGP FSAR CH10 (PRT2), CH11, CH12, CH13, CH14, CH .pdf 15(PRT1)
Section 10.3 Section 10.4 Appendix 1OA Chapter 11 Chapter 12 Chapter 13 Chapter 14 Chapter 15 to Figure 15.1.5-10 016    VEGP FSAR CH15 (PRT2), CH16, CH17, CH18, CH19,    .pdf TABLE OF CONTENTS, REV 21 EFFECTIVE PAGE LIST 017    VEGP TECHNICAL REQUIREMENTS MANUAL                .pdf 018    VEGP BASES                                        .pdf DISC 3 019    VEGP FSAR REF DWGS PART 1 (1K5-1305-058-01 thru   .pdf 1X3D-BD-J02D) 020    VEGP FSAR REF DWGS PART 2 (1 X3DG001 thru         .pdf 1X4DB150-1) 021    VEGP FSAR REF DWGS PART 3 (1X4DB151-1 thru        .pdf 1X4DB203) 022    VEGP FSAR REF DWGS PART 4 (1X4DB205-1 thru        .pdf 1X4DE507)
E1-3 to NL-18-0367 CD-ROM Table of Contents I                              FILENAME SEQ   CONTENT                                       EXTENSION 023   VEGP FSAR REF DWGS PART 5 (1X4DE508 thru      .pdf 1X6AA02-00238) 024   VEGP FSAR REF DWGS PART 6 (1X6AA02-00239 thru .pdf 2X40B 170-2) 025   VEGP FSAR REF DWGS PART 7 (2X4DB174-2 thru    .pdf AX1 D11A04-1) 026   VEGP FSAR REF DWGS PART 8 (AX1 D11A04-3 thru  .pdf AX4DB241) 027   VEGP FSAR REF DWGS PART 9 (AX4DB242-1 thru    .pdf AX4DJ8047) 028   VEGP FSAR REF DWGS PART 10 (AX4DJ8048 thru    .pdf AX6DD303) 029   VEGP FSAR REF DWGS PART 11 (AX6DD304 thru     .pdf CX5DT101-58)
VEGP-File Nomenclature                        .doc E1-4

NL-18-0367 Vogtle Electric Generating Plant, Units 1 & 2 Revision 21 to the Updated Final Safety Analysis Report, Technical Specification Bases Changes, Technical Requirements Manual Changes, 10 CFR 50.59 Summary Report, and Revised NRC Commitments Report Enclosure 2 CD-ROMs (3 discs) containing Files 001 - 029

NL-18-0367 Vogtle Electric Generating Plant, Units 1 & 2 Revision 21 to the Updated Final Safety Analysis Report, Technical Specification Bases Changes, Technical Requirements Manual Changes, 10 CFR 50.59 Summary Report, and Revised NRC Commitments Report Enclosure 3 10 CFR 50.59 Summary Report to NL-18-0367 10 CFR 50.59 Summary Report 10 CFR 50.59 Summary Report Activity: DCP SNC656039

==Title:==
Unit 1 IPC Cyber Security Upgrades 10 CFR 50.59 Evaluation Summary:
DCP SNC656039 provides an upgraded IPC system for Unit 1. The upgrade will provide new hardware and software to meet current cyber security requirements while maintaining all existing functions of the IPC. The IPC provides calculations, alarms, and trending to the operators, reducing their workload; however, it is not a Technical Specification (TS} instrument.
For all values displayed by the IPC, there exist alternate indications or means of calculating values. The functions provided to the operators and to other plant systems, and the way in which the operators interact with the upgraded IPC system is the same as in the existing system. Operators will not be adversely impacted by the upgraded IPC.
This is a digital-for-digital upgrade, and the guidance of NEI 01-01 has been utilized to evaluate this upgrade in Section C. The proposed activity cannot result in increased frequency, nor in increased consequences of an accident evaluated in the Updated FSAR. The IPC is not an accident initiator and does not have any mitigating control functions. The upgraded IPC will not adversely impact the environment in which it is installed. The FSAR described functions of the IPC remain unchanged by this modification. The proposed activity cannot result in increased occurrences or consequences of an SSC malfunction.
The possibility of adverse impacts from software changes are mitigated by validation of the system software by the factory acceptance test and the site acceptance test, consistent with plant procedure NMP-GM-007-002. Based on this 10CFR50.59 evaluation, the proposed activity may be implemented without prior NRC approval.
E3-1.
to NL-18-0367 10 CFR 50.59 Report Activity: DCP SNC656040

==Title:==
Unit 2 IPC Cyber Security Upgrades 10 CFR 50.59 Evaluation Summary:
DCP SNC656040 provides an upgraded IPC system for Unit 2. The upgrade will provide new hardware and software to meet current cyber security requirements while maintaining all existing functions of the IPC. The IPC provides calculations, alarms, and trending to the operators, reducing their workload; however, it is not a Technical Specification (TS) instrument.
For all values displayed by the IPC, there exist alternate indications or means of calculating values. The functions provided to the operators and to other plant systems, and the way in which the operators interact with the upgraded IPC system is the same as in the existing system. Operators will not be adversely impacted by the upgraded IPC.
This is a digital-for-digital upgrade, and the guidance of NEI 01-01 has been utilized to evaluate this upgrade in Section C. The proposed activity cannot result in increased frequency, nor in increased consequences of an accident evaluated in the Updated FSAR. The IPC is not an accident initiator and does not have any mitigating control functions. The upgraded IPC will not adversely impact the environment in which it is installed. The FSAR described functions of the IPC remain unchanged by this modification. The proposed activity cannot result in increased occurrences or consequences of an SSC malfunction.
The possibility of adverse impacts from software changes are mitigated by validation of the system software by the factory acceptance test and the site acceptance test, consistent with plant procedure NMP-GM-007-002. Based on this 10CFR50.59 evaluation, the proposed activity may be implemented without prior NRC approval.
E3-2 to NL-18-0367 10 CFR 50.59 Report Activity: SNC799988

==Title:==
RWST Switchover Change for GSl-191 10 CFR 50.59 Evaluation Summary:
In an effort to improve the performance of the containment sump strainers following a post-LOCA response, it is desirable to increase the amount of transferred inventory of RWST water to containment for certain scenarios which do not currently reach the RWST empty level alarm.
The proposed change, a delay in ECCS switchover from injection to recirculation, allows a greater amount of RWST inventory to be transferred to containment for the previously mentioned scenarios. However, the proposed change results in reduced allowable response times (and fewer required operator actions) between low-low and empty levels for operator action in order to proceed from ECCS injection mode to ECCS cold leg recirculation mode.
MC-V-16-0062 determines new required operator action response times to avoid potential air ingestion into RHR and Containment Spray suction lines. Some of these new operator response times are shorter than the previous required times. NMP-OS-014-003, VNP Time Critical Operator Action Program demonstrates that operator actions can be completed within the required times. Furthermore, the ECCS and Containment Spray systems are designed to meet single failure criteria. Assuming a bounding single failure where one of the RWST to RHR suction valves (HV-8812 A or B) or CSS (HV-9017A or B) fails to close, air ingestion into the RHR pumps could potentially cause a single train failure. The failure of a given train of the CSS or RHR system is already considered in a design basis response and thus is no different than that already considered by the Updated FSAR.
Therefore, the proposed activity does not:
* Result in more than a minimal increase in the frequency of occurrence or consequences of an accident previously evaluated in the UFSAR;                          ,
* Result in more than a minimal increase in the likelihood of occurrence or consequences of a malfunction of an SSC important to safety previously evaluated in the UFSAR;
* Create the possibility for an accident of a different type or a malfunction of an SSC important to safety with a different result than any previously evaluated in the UFSAR;
* Have an impact on the integrity of the fuel cladding, reactor coolant pressure boundary, or containment.
E3-3 to NL-18-0367 10 CFR 50.59 Report Activity: SNC887864

==Title:==
Unit 1 Alternate Means of NSCW Transfer 10 CFR 50.59 Evaluation Summary:
The proposed activity involves design and guidance to provide a method for transferring the 18 NSCW Tower inventory to the 1A NSCW Tower as required in Technical Specification 3.7.9.D.
This includes guidance to route a temporary hose at a preexisting blanked pipe connection to be used in the event of NSCW transfer pump inoperability and after a design basis accident. The 1A NSCW Tower would be filled through this connection and hose. The activity does not increase the likelihood of an accident or the consequences of an accident. The guidance given is used as an additional option during NSCW transfer pump inoperability. The combination of the 1A and 18 NSCW basins will remain the credited 30-day NSCW supply.
Activity: SNCB 12963

==Title:==
Unit 2 Alternate Means of NSCW Transfer 10 CFR 50.59 Evaluation Summary:
The proposed activity involves design and guidance to provide a method for transferring the 2A NSCW Tower inventory to the 28 NSCW Tower as required in Technical Specification 3.7.9.D.
This includes guidance to route a temporary hose at a preexisting blanked pipe connection to be used in the event of NSCW transfer pump inoperability and after a-design basis accident. The 28 NSCW Tower would be filled through this connection and hose. The activity does not increase the likelihood of an accident or the consequences of an accident. The guidance given is used as an additional option during NSCW transfer pump inoperability. The combination of the 2A and 28 NSCW basins will remain the credited 30-day NSCW supply.
E3-4 to NL-18-0367 10 CFR 50.59 Report Activity: Technical Evaluation #214939 (LDCR 2012045)

==Title:==
Loss of Normal Feedwater (LONF) and Loss of Nonemergency AC (LOAC) Reanalysis to Address NSAL-11-01 10 CFR 50.59 Evaluation Summary:
The proposed change consists of the re-analyses of the LONF and LOAC design basis accidents and the AFW System reliability concerns. The re-analyses used a dual analysis approach for the design basis accidents and the AFW System reliability concerns. Previously, a single bounding analysis was performed that combined the conservative Chapter 15 safety analysis assumptions and the reduced AFW flow consistent with a single MDAFP. This resulted in an analysis that was overly conservative. Utilizing the dual analysis approach, with both analyses assuming the failure of the TDAFP as the limiting single failure, allows both analyses to be addressed separately, while continuing to show that the conservative acceptance criterion used by Westinghouse for this event (preventing pressurizer filling) is met for both scenarios.
By demonstrating that acceptable results are achieved in this separate analysis crediting a single MDAFP, the Chapter 15 analyses can be performed assuming the operation of both available MDAFW pumps. Therefore, the use of the dual analysis method can be considered to be an alternate method of evaluation from that described in the Updated FSAR.
Although the dual analysis approach may represent the use of an alternate methodology, the dual analysis approach was approved by the NRC in the license amendment granted for the Callaway Replacement Steam Generator (RSG) project (Reference 8). As such, the use of the dual analysis approach for this specific purpose has been previously reviewed and approved by the NRC. Therefore, the re-analyses performed for Vogtle Units 1 and 2 to address the AFW reliability concerns and the design basis accidents was found to be acceptable, and is not considered to be a departure from the methodology described in the Updated FSAR.
Activity: Technical Evaluation #934802 (LDCR 2015036)

==Title:==
1-E Battery Charger Load Sharing Licensing Documents Modification 10 CFR 50.59 Evaluation Summary:
This change will change the battery chargers on the 1-E buses so that only one charger per battery has to be in operation at a time. This change will not impact any design functions as the batteries will still be able to provide the required voltage to the bus even if the battery charger is disconnected for a period of time before the.second charger can be brought online. This change will not create any new accidents, impact the consequences of an accident, increase the likelihood of an accident, or impact any of the fission product barriers in any way.
E3-5 to NL-18-0367 10 CFR 50.59 Report Activity: GP-19371 (LDCR 2016012)

==Title:==
Revision to WRB-2 DNB Correlation in UFSAR Section 4.4 10 CFR 50.59 Evaluation Summary:
Westinghouse NSAL-14-5 (June 17, 2014) reported a potential non-conservatism in Departure from Nucleate Boiling (DNB) correlations used in the safety analyses for Westinghouse plants.
This was based on test data for a fuel design and DNB correlation that are not used in SNC plants. However, to address the NSAL, Westinghouse has chosen to conservatively apply a penalty to other fuel designs and DNB correlations under certain conditions. This results in a change to the DNB method of evaluation as described in UFSAR Section 4.4. The standards of 10 CFR 50.59 were applied to the change in the method of evaluation to demonstrate that prior NRG approval to make the change is not required.
Activity: Caution Tagout 1-CA-16-1817-00189{001)

==Title:==
10 CFR 50.59 Evaluation for Caution Tagout 1-CA-16-1817-00189(001) 10 CFR 50.59 Evaluation Summary:
Caution Tagout 1-CA-16-1817-00189(001) maintains Heat Trace Panel (HTP) SA, 11817U3005A, in a de-energized configuration due to the panel and panel fan supply breakers being found tripped. The heat trace loads of HTP SA include instrumentation and lines for the RWST, RMWST, and RMWST Degasification.
The design functions of the RMWST and Degasification primarily support the eves function in maintaining design RCS inventory at power. Provisions are made in the UFSAR and plant design for continued charging pump suction provision upon a loss of RMWST makeup supply.
The other makeup water functions of the tank are in an emergency backup mode, with other sources of makeup available to the users.
Since the application of RWST level indication is solely for support of ECCS post-accident function, the tagout would not increase the frequency of occurrence of an accident. Technical Specification requirements on RWST volume verification provide protection against another accident type, forcing the plant to Mode 4 if unable to verify RWST inventory.
Freeze protection measures implemented under site procedure 11877-1, "Cold Weather Checklist," would reduce the likelihood of malfunction of the untraced instrumentation or lines for the RWST and RMWST, or of any change or increase in the consequences or results of a malfunction, including fuel cladding damage. The tagout is determined to be evaluated out, and not require NRG approval for its continued implementation.
E3-6 to NL-18-0367 10 CFR 50.59 Report Activity: Caution Tagout 1-CA-16-1817-00189(002)

==Title:==
10 CFR 50.59 Evaluation for Caution Tagout 1-CA-16-1817-00189(002) 10 CFR 50.59 Evaluation Summary:
Caution Tagout 1-CA-16-1817-00189(002) maintains Heat Trace Panel (HTP) SA, 11817U3005A, in a de-energized configuration due to the panel and panel fan supply breakers being found tripped. The heat trace loads of HTP SA include instrumentation and lines for the RWST, RMWST, and RMWST Degasification.
The design functions of the RMWST and Degasification primarily support the CVCS function in maintaining design RCS inventory at power. Provisions are made in the UFSAR and plant design for continued charging pump suction provision upon a loss of RMWST makeup supply.
The other makeup water functions of the tank are in an emergency backup mode, with other sources of makeup available to the users.
Since the application of RWST level indication is solely for support of ECCS post-accident function, the tagout would not increase the frequency of occurrence of an accident. Technical Specification requirements on RWST volume verification provide protection against another accident type, forcing the plant to Mode 4 if unable to verify RWST inventory.
Freeze protection measures implemented under site procedure 11877-1, "Cold Weather Checklist," would reduce the likelihood of malfunction of the untraced instrumentation or lines for the RWST and RMWST, or of any change or increase in the consequences or results of a malfunction, including fuel cladding damage. The tagout is determined to be evaluated out, and not require NRC approval for its continued implementation.
E3-7 to NL-18-0367 10 CFR 50.59 Report Activity: Caution Tagout 2-CA-17-1208-00114

==Title:==
10 CFR 50.59 Evaluation for Caution Tagout 2-CA-17-1208-00114 10 CFR 50.59 Evaluation Summary:
Caution Tagout 2-CA-17-1208-00114 provides information to aid operators in maintaining appropriate valve positions to bypass the Unit 2 Seal Water Return Backflushable Filter (SWRBF). The filter is being bypassed due to excessive leakage from the filter drain valve, 2HV41329C, when the filter is in service.
The SWRBF passes the combined RCP seal water return and excess letdown flows to remove particulates prior to return to the suction of the Normal Charging Pump. The RCP seal water loop and the SWRBF are both within the Chemical and Volume Control System (CVCS), and the large majority of equipment malfunction and accident response adverse effect possibilities are associated with normal and emergency (ECCS) CVCS component function.
Analysis of the applicable 10 CFR 50.59 evaluation questions has determined that the SWRBF bypass maintained by the tagout does not cause more than minimal increase in the frequency or consequences of UFSAR analyzed accident conditions, nor does it create the possibility for an unanalyzed accident condition of consequence. It has also been determined that the tagout does not increase the likelihood or consequence or component malfunctions. Finally, no adverse impact was identified on fuel cladding or RCS integrity.
Acti~ity: Danger Tagout 2-DT-17-1208-00220(002)

==Title:==
10 CFR 50.59 Evaluation for Danger Tagout 2-DT-17-1208-00220(002) 10 CFR 50.59 Evaluation Summary:
Caution Tagout 2-CA-17-1208-00114 provides information to aid operators in maintaining appropriate valve positions to bypass the Unit 2 Seal Water Return Backflushable Filter (SWRBF). The filter is being bypassed due to excessive leakage from the filter drain valve, 2HV41329C, when the filter is in service.
The SWRBF passes the combined RCP seal water return and excess letdown flows to remove particulates prior to return to the suction of the Normal Charging Pump. The RCP seal water loop and the SWRBF are both within the Chemical and Volume Control System (CVCS), and the large majority .of equipment malfunction and accident response adverse effect possibilities are associated with normal and emergency (ECCS) CVCS component function.
Analysis of the applicable 10 CFR 50.59 e,valuation questions has determined that the SWRBF bypass maintained by the tagout does not cause more than minimal increase in the frequency or consequences of UFSAR analyzed accident conditions, nor does it create the possibility for an unanalyzed accident condition of consequence. It has also been determined that the tagout does not increase the likelihood or consequence or component malfunctions. Finally, no adverse impact was identified on fuel cladding or RCS integrity.
E3-8 to NL-18-0367 10 CFR 50.59 Report Activity: 1-DT-12-1202-00404

==Title:==
Safety Screening for Tagout 1-DT-12-1202-00404 10 CFR 50.59 Evaluation Summary:
The proposed activity involves the evaluation of the isolation of the Unit 1 and 2 NSCW corrosion monitoring system. The isolation valves for the corrosion monitoring system at the interface of the safety and non-safety related piping were closed due to concerns of loss of inventory in the NSCW System due to a line break on the non-safety side of the NSCW piping for this system. Calculation X4C1202V71 determined that the loss of inventory due to a line break on the non-safety lines installed under DCP 91-V1 N0045 would exceed the allowable loss of inventory of the UHS during accident conditions. The existing chemical control/procedures and heat exchanger periodic inspection and cleaning associated with the U1 and U2 NSCW System ensure that the corrosion is mitigated and long-term reliability is ensured. This also ensures the structural integrity of the NSCW System and its components are maintained. In addition, the NSCW System is comprised largely of stainless steel components which are inherently resistant to corrosion.
E3-9 to NL-18-0367 10 CFR 50.59 Report Activity: DCP SNC526823

==Title:==
Unit 2 Mechanical Stress Improvement Process (MSIP) Evaluation and Analysis 10 CFR 50.59 Evaluation Summary:
The proposed activity includes the application of the MSIP to the reactor vessel outlet nozzles.
The new stress profile established by the MSIP (compressive instead of tensile) mitigates      sec in the affected piping weldments. As such, the reactor coolant pressure boundary is impacted by the activity. However, the application of the MSIP to the reactor vessel outlet nozzles mitigates  sec   in the piping and improves the ability of the affected RCS piping to continue to meet the applicable design basis requirements. As such, the ability of the RCS to meet the design function of limiting leakage and activity release (as described in Updated FSAR Section 5.1.1, "Design Basis") is enhanced by the proposed activity.
     ~ The proposed activity does not affect any existing accident initiators and will not affect the frequency of occurrence of any accident previously evaluated in the Updated FSAR.
Therefore, the frequency of occurrence of an accident previously evaluated in the Updated FSAR is not increased.
* The proposed activity does not create any additional malfunctions of SSCs important to safety. The SSCs important to safety will remain capable of performing their safety function. Since the SSCs will continue to operate within design and analysis limits, no new failure modes are created. Therefore, there is no increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the Updated FSAR.
* The proposed activity does not adversely impact the assumptions of the safety analysis or the RCS design basis requirements. Therefore, the proposed activity does not result in an i~crease in the consequences of an accident previously evaluated in the Updated FSAR.
* The SSCs important to safety will continue to operate within the assumed design and analysis limits. Therefore, the consequences of a malfunction of an SSC previously analyzed are not impacted. Thus, the proposed activity does not result in an increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the Updated FSAR.
* The proposed activity does not create any new accident scenarios. The proposed activity does not alter or place any SSC in a configuration outside design or analysis limits. The SSCs important to safety will continue to perform their safety function as assumed in the safety analyses in the same manner as before. Therefore, the proposed activity does not create the possibility for an accident of a different type than any previously evaluated in the Updated FSAR.
* No protection or control system hardware is modified or altered; therefore, no new single failures or failure modes will be introduced by the change. In addition, the ability of the SSCs important to safety to perform their safety function as assumed in the safety analyses is not affected by this activity. Therefore, the proposed activity does not create the possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the Updated FSAR.
* The proposed activity does not affect the operating limits or characteristics of the fuel, or modify the containment boundary in any way. The DNB events are not impacted. The proposed activity (the application of the MSIP) does modify the RCS piping which is part of the reactor coolant pressure boundary. Therefore, the proposed change does have an impact on the reactor coolant pressure boundary.
E3-10 to NL-18-0367 10 CFR 50.59 Report
* The ability of the RCS to meet the design function of limiting leakage and activity release (as described in Updated FSAR Section 5.1.1, "Design Basis") is enhanced by the proposed activity. The proposed activity does not impact any design basis limit for a fission product barrier. Therefore, the proposed activity does not result in a design basis limit for a fission product barrier as described in the Updated FSAR being exceeded or altered.
* Performing the fatigue analyses using the WESTEMS' computer program, that considered the application of the MSIP to the Vogtle Unit 2 Loops 1, 2, 3, and 4 reactor vessel outlet nozzles does not result in a departure from a method of evaluation described in the Updated FSAR used in establishing the design bases or in the safety analyses. Fatigue analysis performed with WESTEMS' in accordance with ASME Ill, Subsection NB-3650 (Subsection NB-3600) are not specific to the AP1000 design, and are consistent with the requirements of the ASME Code for the design of piping components, and therefore can be used to evaluate ASME Class 1 piping components for any nuclear power plant to demonstrate compliance with the ASME Code.
E3-11 to NL-18-0367 10 CFR 50.59 Report Activity: NUCDEV Technical Evaluation #794354

==Title:==
10 CFR 50.59 Screening/Evaluation to Change NRC Commitment SNC12175 10 CFR 50.59 Evaluation Summary:
This 50.59 evaluation is being performed to change NRC commitment SNC12175 which states the following:
ITS NRC Commitment: GPC will control future changes to this information via 10 CFR 50.59 - Original TS 4.8.1.1.2.F - Each diesel generator shall be demonstrated operable: At least once per 92 days and from new fuel oil prior to addition to the storage tank to obtain a sample and verify that the neutralization number is less than 0.2 and the mercaptan is less than 0.01%. #, ##-#
mercaptan content shall not be required to be verified within specification for new fuel prior to its addition, for up to 15,000 gallons of fuel added to the tank, if the last tank sample had a mercaptan content of less than 0.007%. All subsequent new fuel addition will require mercaptan content verification prior to its addition until the tank contents are verified to be less than .007%.
The intent of the change is to remove the testing requirements of this commitment to determine mercaptan content and neutralization number. A technical justification shows that based on switching to ultra-low sulfur diesel fuel oil and reviewing historical data that mercaptan and neutralization number testing is unneeded. Any degradation of the zinc coating would be caught by tests required by the Updated FSAR and Technical Specifications. Vogtle also has seen no issues with degradation during the once every 10-year inspection of the DFOST.
Particulates and the presence of water are checked once every 31 days. The diesel fuel oil duplex filters differential pressure is checked once every 28 days. These tests will be able to detect degradation of the inorganic zinc coating. Removing the testing requirements for mercaptan and neutralization number will not more than minimally increase the chance of an emergency diesel generator failure ..
The changes evaluated by this 50.59 do not affect or minimally increase the chance of occurrence of an accident.
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NL-18-0367 Vogtle Electric Generating Plant, Units 1 & 2 Revision 21 to the Updated Final Safety Analysis Report, Technical Specification Bases Changes, Technical Requirements Manual Changes, 10 CFR 50.59 Summary Report, and Revised NRC Commitments Report Enclosure 4 Revised NRC Commitments Report to NL-18-0367 Revised NRC Commitments Report Original Commitment: Perform the Maintenance Rule Periodic Assessment (18 month)
SNC16544 (Legacy Commitment Number CO 37180)
Perform the Maintenance Rule periodic assessment. Eighteen (18) month target date for, completion of assessment per this open item. 10 CFR 50.65(a)(3): "Performance and condition monitoring activities and associated goals and preventive maintenance activities shall be evaluated at least every refueling cycle provided the interval between evaluations does not exceed 24 months. The evaluations shall take into account, where practical, industry-wide operating experience. Adjustments shall be made where necessary to ensure that the objective of preventing failures of structures, systems, and components through maintenance is appropriately balanced against the objective of minimizing unavailability of structures, systems, and components due to monitoring or preventive maintenance."
Revised Commitment: Procedures NMP-ES-027, "Maintenance Rule Program" and NMP-ES-027-001, "Maintenance Rule Implementation", have been revised to incorporate the Living (a)(3) process that will satisfy the requirements of 10 CFR 50.65, "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," paragraph (a)(3) by leveraging the current state of the art processes for Performance Monitoring, Program and System Health Reports. For SNC to credit on-going processes to meet the requirements of paragraph (a)(3) periodic assessment, NMP-ES-027 Attachment 2 was developed to map specifically how the requirements are met within the fleet-specific program. This attachment will form the basis for the written guidance on how the site is meeting the requirements of paragraph (a)(3).
Justification for Change: Use of routine periodic monitoring satisfies the guidance in NUMARC 93-01 using the processes evolved since origination. Table "SNC Living (a)(3)
Assessment Process Controls" in NMP-ES-027 relates specific relevant NUMARC 93-01 sections to the current Maintenance Rule process elements and will take the place of discrete periodic assessments. Support for a living (a)(3) process is found in NRC Inspection Manual 71111.12, Block 22, and NUREG-1526, Section 2.6. Both discuss the continuous balancing of unavailability and reliability as an integral aspect of monitoring against performance criteria and that the requirements for the evaluation can be satisfied through on-going assessments. For SNC to credit on-going processes to meet the requirements of paragraph (a)(3) periodic assessment, the above-mentioned table was developed to specifically map how the requirements are met within the Fleet-specific program. This table will form the basis for the written guidance on how the site is meeting the requirements of paragraph (a)(3).
E4-1 to NL-18-0367 Revised NRC Commitments Report Original Commitment: Endorsed by Generic Letter 83-22 SNC10405 (Legacy Commitment Number 1985307152)
... that have been endor~ed by Generic Letter 83-22. These guidelines include instructions for coping with ATWS. The staff concludes that the applicant's commitment to implement procedures based on these guidelines is acceptable on an interim basis for full power operation.
Future modifications will be needed to implement the ATWS rule; the staff will determine the required schedule for implementing such modifications. Further, the staff has identified Generic Letter 83-28 as an open item and will report on its resolution in a supplement to the SER.
Revised Commitment: No change in commitment description; status change only. See discussion in Justification for Change.
Justification for Change: This commitment is currently CT-CLOSED in Maximo. The CT-CLOSED status was based on previous procedure guidance; however, per NMP-GM-019 the status would be CT-IMPL. This discussion documents justification of changing the status of this commitment to CT-CLOSED, as defined in the current version of NMP-GM-019, and to ensure proper notification to the NRC.
The commitment addresses both GL 83-22 and GL 83-28. The commitment associated with GL 83-22 was interim and has long since been settled. Moreover, it does not apply to procedure 08-002. There is very little in this commitment concerning GL 83-28 beyond noting the fact that the NRC Staff identified it as an open item and will report on its resolution in a supplement to a SER. Resolution occurred later and resulted in multiple commitments. This Commitment Tracking item does not establish or document a commitment to GL 83-028 (or GL 90-03), and therefore, it is recommended that this commitment be removed from DS-002 references and the commitment status be changed to CT-CLOSED.
E4-2 to NL-18-0367 Revised NRC Commitments Report Original Commitment: Track the Corporate Action to Ensure a Contact Program for Vendors SNC15026 (Legacy Commitment Number 1990320117)
Track the Corporate action to ensure a contact program for vendors of key safety-related components outside of the NSSS scope of supply, such as: batteries, battery chargers, inverters, pumps, valve operators, and switchgear.
Revised Commitment: No change in commitment description; status change only. See discussion in Justification for Change.
Justification for Change: This commitment documented the response of Georgia Power Company to the NRC regarding implementation of GL 90-03(b). Benchmarking has shown the industry has weighed the cost and the time required to complete the vendor contact effort balanced against the risks to safe operation and the goals of equipment reliability. Many operators have eliminated or are in the process of eliminating the non-NSSS portion of their programs. Elimination of undue administrative burden aligns with the goals of Delivering the Nuclear Promise and reducing regulatory risk. Furthermore, there are multiple other programs currently in place in the SNC fleet that achieve the same goals as the original intent of this aspect of the generic letter.
The technological advancements in equipment performance information exchanges and other OE sharing meet the intent of GL 83-28 and GL 90-03 and have obviated the need for direct periodic contact with vendors. This proposed change will not affect SNC contact with fleet NSSS vendors as defined in GL 90-03(a). The requirement to have direct contact with our key safety-related vendors was made in a time when computers, email and other regulatory and industry processes were not well integrated into every-day business. Therefore, sunsetting (status of CT-CLOSED) of this commitment is justified.
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