ND-19-1392, Unit 4 - Resubmittal Notice of Uncompleted ITAAC 225-days Prior to Initial Fuel Load Item 2.5.02.10 (Index Number 549)

Unit 4 - Resubmittal Notice of Uncompleted ITAAC 225-days Prior to Initial Fuel Load Item 2.5.02.10 (Index Number 549)

ML19319A135
Person / Time
Site:	Vogtle
Issue date:	11/14/2019
From:	Yox M Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
ND-19-1392
Download: ML19319A135 (17)

Text

Michael J. Yox 7825 River Road

!ik. Southern Nuclear Regulatory Affairs Director Waynesboro, GA 30830 Vogtle 3 & 4 706-848-6459 tel NOV 1 <\ 21119 Docket Nos.: 52-025 52-026 ND-19-1392 10 CFR 52.99(c)(3)

U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555-0001 Southern Nuclear Operating Company Vogtle Electric Generating Plant Unit 3 and Unit 4 Resubmittal Notice of Uncompleted ITAAC 225-davs Prior to Initial Fuel Load Item 2.5.02.10 (Index Number 5491 Ladies and Gentlemen:

Pursuant to 10 CFR 52.99(c)(3), Southern Nuclear Operating Company hereby notifies the NRC that as of November 12, 2019, Vogtle Electric Generating Plant(VEGP) Unit 3 and Unit 4 Uncompleted Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) Item 2.5.02.10 [Index Number 549] has not been completed greater than 225-days prior to initial fuel load. The Enclosure describes the plan for completing this ITAAC. Southern Nuclear Operating Company will, at a later date, provide additional notifications for ITAAC that have not been completed 225 days prior to initial fuel load.

Southern Nuclear Operating Company(SNC) previously submitted Notice of Uncompleted ITAAC 225 days Prior to Initial Fuel Load for Item 2.5.02.10 [Index Number 549] ND-19-1311

[ML19301D014] dated October 28, 2019. This resubmittal supersedes ND-19-1311 in its entirety.

This notification is informed by the guidance described in NEI 08-01, Industry Guideline for the ITAAC Closure Process Under 10 CFR Part 52, which was endorsed by the NRC in Regulatory Guide 1.215. In accordance with NEI 08-01, this notification includes ITAAC for which required inspections, tests, or analyses have not been performed or have been only partially completed. All ITAAC will be fully completed and all Section 52.99(c)(1) ITAAC Closure Notifications will be submitted to NRC to support the Commission finding that all acceptance criteria are met prior to plant operation, as required by 10 CFR 52.103(g).

This letter contains no new NRC regulatory commitments.

If there are any questions, please contact Tom Petrak at 706-848-1575.

Respectfully submitted.

Michael J. Yox Regulatory Affairs Director Vogtle 3 & 4

Enclosure:

Vogtle Electric Generating Plant(VEGP) Unit 3 and Unit 4 Completion of ITAAC 2.5.02.10 [Index Number 549]

MJY/SBB/sfr

U.S. Nuclear Regulatory Commission ND-19-1392 Page 2 of 3 To:

Southern Nuclear Operating Company/Georgia Power Company Mr. Peter P. Sena III (w/o enclosures)

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

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

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

Mr. G. Chick Mr. M. Page Mr. P. Martino Mr. M. J. Yox Mr. A. S. Parton Ms. K. A. Roberts Mr. T. G. Petrak Mr. C. T. Defnall Mr. C. E. Morrow Mr. J. L. Hughes Mr. S. Leighty Ms. A. C. Chamberlain Mr. J. C. Haswell Document Services RTYPE: VND.LI.L06 File AR.01.02.06 cc:

Nuclear Reoulatorv Commission Mr. W. Jones (w/o enclosures)

Mr. F. D. Brown Mr. C. P. Patel Mr. G. J. Khouri Ms. S. E. Temple Mr. N. D. Karlovich Mr. A. Lerch Mr. C. J. Even Mr. B. J. Kemker Ms. N. C. Coovert Mr. C. Welch Mr. J. Gaslevic Mr. V. Hall Mr. G. Armstrong Ms. T. Lamb Mr. M. Webb Mr. T. Fredette Mr. C. Weber Mr. S. Smith Mr. C. Santos Oqlethorpe Power Corporation Mr. R. B. Brinkman Mr. E. Rasmussen

U.S. Nuclear Regulatory Commission ND-19-1392 Page 3 of 3 Municipal Electric Authority of Georgia Mr. J. E. Fuller Mr. 8. M. Jackson Dalton Utilities Mr. T. Bundros Westinqhouse Electric Company. LLC Dr. L. Oriani (w/o enclosures)

Mr. D. 0. Durham (w/o enclosures)

Mr. M. M. Corletti Ms. L. G. Iller Mr. Z. 8. Harper Mr. J. L. Coward Other Mr. J. E. Hesler, Bechtel Power Corporation Ms. L. Matis, Tetra Tech NUS, Inc.

Dr. W. R. Jacobs, Jr., Ph.D., CDS Associates, inc.

Mr. 8. Roetger, Georgia Public Service Commission Ms. 8. W. Kernizan, Georgia Public Service Commission Mr. K. C. Greene, Troutman Sanders Mr. 8. Blanton, Baich Bingham

U.S. Nuclear Regulatory Commission ND-19-1392 Enclosure Page 1 of 14 Southern Nuclear Operating Company ND-19-1392 Enclosure Vogtle Electric Generating Plant(VEGP) Unit 3 and Unit 4 Completion Plan for Uncompleted ITAAC 2.5.02.10[Index Number 549]

U.S. Nuclear Regulatory Commission ND-19-1392 Enclosure Page 2 of 14 ITAAC Statement Design Commitment

10. Setpolnts are determined using a methodology which accounts for loop Inaccuracies, response testing, and maintenance or replacement of instrumentation.

Inspections/Tests/Analvses Inspection will be performed for a document that describes the methodology and input parameters used to determine the RMS setpoints.

Acceptance Criteria A report exists and concludes that the RMS setpoints are determined using a methodology which accounts for loop inaccuracies, response testing, and maintenance or replacement of instrumentation.

ITAAC Determination Basis This ITAAC requires that inspection be performed to verify that the Protection and Safety Monitoring System (RMS)setpoints are determined using a methodology which accounts for loop inaccuracies, response testing, and maintenance or replacement of instrumentation.

Results of the inspections noted below are documented in 2.5.02.10-U3-SumRep-Rev X,"Unit 3 Protection and Safety Monitoring System Setpoint Methodology Summary Report"(Reference

1) and 2.5.02.10-U4-SumRep-Rev X,"Unit 4 Protection and Safety Monitoring System Setpoint Methodology Summary Report"(Reference 2)for Units 3 and 4, respectively.

WCAP-16361-P,"Westinghouse Setpoint Methodology for Protection Systems- API000" (Reference 3), was inspected and identifies the methodology used to determine the overall instrument uncertainty (i.e. loop inaccuracy)for a Reactor Trip System (RTS) and Engineered Safeguards Features Actuation System (ESFAS)function. Reference 3 provides specific instructions for calculating instrument and loop uncertainty setpoints consistent with ANSI/ISA-67.04.01-2000 and Regulatory Guide 1.105, Revision 3. An inspection of the RTS/ESFAS function setpoint and uncertainty calculations was performed and confirmed calculations employ the WCAP-16361-P methodology to determine loop inaccuracies, as documented in references 1 and 2.

Section 5.5.14 of the AP1000 Technical Specifications (TS) requires the nominal trip setpoint, As-Found Tolerance (AFT), and As-Left Tolerance (ALT)for each TS-required automatic protection instrumentation function be calculated in conformance with WCAP-16361-P. These requirements are used to determine if maintenance or replacement of instrumentation is needed. If maintenance or replacement is required, the "Operational Readiness Work Management' procedure (Reference 4), and "Plant Modification and Configuration Change Processes"(Reference 5), accounts for any impacts on instrumentation or issued calculations.

An inspection of the RTS/ESFAS function setpoint and uncertainty calculations was performed

U.S. Nuclear Regulatory Commission ND-19-1392 Enclosure Page 3 of 14 and confirmed calculations employed the WCAP-16361-P methodology for AFT and ALT, as documented in references 1 and 2.

The methodology utilized for response time determination is per UFSAR Chapter 15 criteria (Reference 6). Response testing is conducted within the factory acceptance and preoperational test program per UFSAR Chapter 14(Reference 6). UFSAR section 15.0.6 (Reference 6),

discusses the PMS time delay methodology that is assumed in the accident analysis for RTS and equipment actuated by ESFAS functions. An inspection of SV3-PMS-T1-501,"API000 Protection and Safety Monitoring system Preoperational and Component Test Specification" and SV4-PMS-T1-501,"API000 Protection and Safety Monitoring system Preoperational and Component Test Specification"(References 7 and 8)and APP-PMS-T5-001,"Protection and Safety Monitoring System Test Plan"(Reference 9), was performed to confirm that the PMS preoperational test program includes the response time testing as required. Attachment A provides a cross-reference of each applicable PMS function to its test case.

References 1 through 9 are available for NRC inspection as part of Unit 3 and Unit 4ITAAC 2.5.02.10 Completion Packages (Reference 13 and 14, respectively).

ITAAC Finding Review In accordance with plant procedures for ITAAC completion. Southern Nuclear Operating Company(SNC) performed a review of all ITAAC findings pertaining to the subject ITAAC and associated corrective actions. This review found that there are no relevant ITAAC findings associated with this ITAAC.

U.S. Nuclear Regulatory Commission ND-19-1392 Enclosure Page 4 of 14 References favailable for NRG Inspection)

1. 2.5.02.10-U3-SumRep-Rev X "Unit 3 Protection and Safety Monitoring System Setpoint Metfiodology Summary Report"

2. 2.5.02.10-U4-SumRep-Rev X "Unit 4 Protection and Safety Monitoring System Setpoint Metfiodology Summary Report"

3. WCAP-16361-P, February 2011 "Westinghouse Setpoint Metfiodology for Protection Systems-API000"

4. B-ADM-WCO-001,"Operational Readiness Work Management"

5. NMP-ES-084-001,"Plant Modification and Configuration Change Processes"

6. VEGP 3&4 UFSAR,

• Section 14.2.9.1.12, "Protection and Safety Monitoring System Testing"

• Section 15.0.6, "Protection and Safety Monitoring System Setpoints and Time Delays to Trip Assumed in Accident Analyses"

7. SV3-PMS-T1-501,"API000 Protection and Safety Monitoring system Preoperational and Component Test Specification"

8. SV4-PMS-T1-501,"API000 Protection and Safety Monitoring system Preoperational and Component Test Specification"

9. APP-PMS-T5-001,"Protection and Safety Monitoring System Test Plan"

10. APP-PMS-T2R-007"API000 Protection and Safety Monitoring System System-Level Reactor Trip Channel Integration Test Report"

11. APP-PMS-T2R-008 "API000 Protection and Safety Monitoring System System-Level Engineered Safety Features Channel Integration Test Report"

12. APP-PMS-T2R-050 "API000 Protection and Safety Monitoring System Fuel Load Regression Test Report" 13.2.5.02.10-U3-CP-Rev 0, ITAAC Completion Package 14.2.5.02.10-U4-CP-Rev 0, ITAAC Completion Package

15. NEI 08-01,"Industry Guideline for the ITAAC Closure Process Under 10 CFR Part 52"

U.S. Nuclear Regulatory Commission ND-19-1392 Enclosure Page 5 of 14 Attachment A APP-PMS-T2R-007"API000 Protection and Safety Monitoring System System-Level Reactor Trip Channel Integration Test Report"(Reference 10), documents each reactor trip bistable.

These tests verify the change of state of the Bistable trip that initiates the Reactor Trip logic.

Table 2.5.2-2 RT DivA DivB DIv 0 DlvD Source Range High TPS01A-01.1 TPS01B-01.1 TPS01C-01.1 TPS01D-01.1 Neutron Flux Reactor Trip Intermediate Range TPS01A-02.1 TPS01B-02.1 TPS01C-02.1 TPS01D-02.1 High Neutron Flux Reactor Trip Power Range High TPS01A-03.1 TPS01B-03.1 TPS01C-03.1 TPS01D-03.1 Neutron Flux (Low Setpoint) Trip Power Range High TPS01A-04.1 TPS01B-04.1 TPS01C-04.1 TPS01D-04.1 Neutron Flux (High Setpoint) Trip Power Range High TPS01A-05.1 TPS01B-05.1 TPS01C-05.1 TPS01D-05.1 Positive Flux Rate Trip Reactor Coolant Pump TPS01A-12.1 TPS01B-12.1 TPS01C-12.1 TPS01D-12.1 High-2 Bearing Water Temperature Trip RCP1A Reactor Coolant Pump TPS01A-12.2 TPS01B-12.2 TPS01C-12.2 TPS01D-12.2 High-2 Bearing Water Temperature Trip RCP1B Reactor Coolant Pump TPS01A-12.3 TPS01B-12.3 TPS01C-12.3 TPS01D-12.3 High-2 Bearing Water Temperature Trip RCP2A

U.S. Nuclear Regulatory Commission ND-19-1392 Enclosure Page 6 of 14 Table 2.5.2-2 RT DIv A DivB DivO DlvD Reactor Coolant Pump TPS01A-12.4 TPS01B-12.4 TPS01C-12.4 TPS01D-12.4 High-2 Bearing Water Temperature Trip RCP2B Overtemperature Delta- TPS01A-07.1 TPS01B-07.1 TPS01C-07.1 TPS01D-07.1 TTrip Overtemperature Delta- TPS01A-07.2 TPS01B-07.2 TPS01C-07.2 TPS01D-07.2 TTrip Overpower Delta-T Trip TPS01A-08.1 TPS01B-08.1 TPS01C-08.1 TPS01D-08.1 Pressurizer Low-2 TPS01A-09.1 TPS01B-09.1 TPS01C-09.1 TPS01D-09.1 Pressure Trip Pressurizer High-2 TPS01A-13.1 TPS01B-13.1 TPS01C-13.1 TPS01D-13.1 Pressure Trip Pressurizer High-3 TPS01A-14.1 TPS01B-14.1 TPS01C-14.1 TPS01D-14.1 Water Level Trip Low-2 Reactor Coolant TPS01A-10.1 TPS01B-10.1 TPS01C-10.1 TPS01D-10.1 Flow Trip ML 1 Low-2 Reactor Coolant TPS01A-10.2 TPS01B-10.2 TPS01C-10.2 TPS01D-10.2 Flow Trip ML 2 Low-2 Reactor Coolant TPS01A-11.1 TPS01B-11.1 TPS01C-11.1 TPS01D-11.1 Pump Speed Trip Low-2 Steam Generator TPS01A-15.1 TPS01B-15.1 TPS01C-15.1 TPS01D-15.1 Narrow Range Water Level Trip SGI Low-2 Steam Generator TPS01A-16.1 TPS01B-16.1 TPS01C-16.1 TPS01D-16.1 Narrow Range Water Level Trip SG2 High-3 Steam TPS01A-17.1 TPS01B-17.1 TPS01C-17.1 TPS01D-17.1 Generator Water Level Trip SGI

U.S. Nuclear Regulatory Commission ND-19-1392 Enclosure Page 7 of 14 Table 2.5.2-2 RT DivA DivB DIvC DlvD High-3 Steam TPS01A-18.1 TPS01B-18.1 TPS01C-18.1 TPS01D-18.1 Generator Water Level Trip SG2

U.S. Nuclear Regulatory Commission ND-19-1392 Enclosure Page 8 of 14 The following test reports document the logic for engineered safety feature trip bistables.

• APP-PMS-T2R-007"API000 Protection and Safety Monitoring System System-Level Reactor Trip Channel Integration Test Report"(Reference 10)

• APP-PMS-T2R-008 "API000 Protection and Safety Monitoring System System-Level Engineered Safety Features Channel Integration Test Report"(Reference 11)

• APP-PMS-T2R-050 "API000 Protection and Safety Monitoring System Fuel Load Regression Test Report"(Reference 12)

These tests verify the change of state of the Bistable Trip computer points.

Test cases that begin with TPS01 are located in APP-PMS-T2R-007.

Test Cases that being with TPS02 are located in APP-PMS-T2R-008.

APP-PMS-T2R-050 is the regression testing that contains each applicable test case.

Table 2.5.2-3 ESP DIv A DivB DIv C DlvD Safeguards Actuation on TPS02A-01.1 TPS02B-01.1 TPS02C-01.1 TPS02D-01.1 Pressurizer Pressure Low-3 Safeguards Actuation on TPS02A-01.3 TPS02B-01.3 TPS02C-01.3 TPS02D-01.3 RCS Cold Leg Low-2 TPS02A-01.6 TPS02B-01.6 TPS02C-01.6 TPS02D-01.6 Safeguards Actuation on TPS02A-01.4 TPS02B-01.4 TPS02C-01.4 TPS02D-01.4 SG 1 Steam Line Pressure Low-2 Safeguards Actuation on TPS02A-01.5 TPS02B-01.5 TPS02C-01.5 TPS02D-01.5 SG 2 Steam Line Pressure Low-2 Safeguards Actuation on TPS02A-01.7 TPS02B-01.7 TPS02C-01.7 TPS02D-01.7 Containment Pressure High-2 ADS Actuation stage 1, TPS02A-05.3 TPS02B-05.3 TPS02C-05.3 TPS02D-05.3 2, 3 on CMT Level Low-3 ADS Actuation Stage 4 N/A TPS02B-05.5 TPS02C-05.5 N/A on RCS Hot Leg 1 & 2 Level Low-4 ADS Actuation Stage 4 TPS02A-05.6 TPS02B-05.6 TPS02C-05.6 TPS02D-05.6 on Core Makeup A Tank MR Lower Level Low-6 ADS Actuation Stage 4 TPS02A-05.6 TPS02B-05.6 TPS02C-05.6 TPS02D-05.6 on Core Makeup B Tank NR Lower Level Low-6

U.S. Nuclear Regulatory Commission ND-19-1392 Enclosure Page 9 of 14 Table 2.5.2-3 ESF DivA DivB DivC DivD ADS Actuation Stage 4 TPS02A-05.6 TPS02B-05.6 TPS02C-05.6 TPS02D-05.6 on RCS Wide Range Pressure Low Main Feedwater Isolation TPS02A-28* TPS02B-07.1 TPS02C-09.2 TPS02D-07.1 on SG 1 NR Level High-3 Main Feedwater Isolation TPS02A-09.2* TPS02B-09.2 TPS02C-09.2* TPS02D-09.2 on SG 2 NR Level High-3 Main Feedwater Isolation TPS02A-07.2* TPS02B-07.2 TPS02C-07.2* TPS02D-07.2 on RCS Tavg Low-2 Main Feedwater Isolation TPS02A-07.2* TPS02B-07.2* TPS02C-07.2* TPS02D-07.3 Control Valves on RCS Tavg Low-1 Main Feedwater Isolation TPS02A-09.2* TPS02B-09.2* TPS02C-09.2* TPS02D-09.2 Control Valves on SG NR WL High-3 Reactor Coolant Pump TPS02A-06 TPS02B-06 TPS02C-06 TPS02D-06 Trip & CCS Containment Isolation on RCP 1 -A Bearing Water Temperature High-2 Reactor Coolant Pump TPS01A-12.2 TPS01B-12.2 TPS01C-12.2 TPS01D-12.2 Trip & CCS Containment Isolation on RCP 1-B Bearing Water Temperature High-2 Reactor Coolant Pump TPS01A-12.3 TPS01B-12.3 TPS01C-12.3 TPS01D-12.3 Trip & CCS Containment Isolation on RCP 2-A Bearing Water Temperature High-2 Reactor Coolant Pump TPS01A-12.4 TPS01B-12.4 TPS01C-12.4 TPS01D-12.4 Trip & CCS Containment Isolation on RCP 2-B Bearing Water Temperature High-2 CCS Containment Isolation CMT Actuation on Pzr TPS02A-04.2 TPS02B-04.2 TPS02C-04.2 TPS02D-04.2 Level Low-2

U.S. Nuclear Regulatory Commission ND-19-1392 Enclosure Page 10 of 14 Table 2.5.2-3 ESF DivA DivB DivC DivD GMT Actuation TPS02A-04.3 TPS02S-04.3 TPS02C-04.3 TPS02D-04.3 on SG 1 and SG 2 WR Level Low -2 Plus Thot High Turbine Trip TPS02A-09.2 TPS02S-09.2 TPS02C-09.2 TPS02D-09.2 Reactor Trip and Main Feedwater Isolation on SGI NR Level High-3 Turbine Trip TPS02A-09.2 TPS02S-09.2 TPS02C-09.2 TPS02D-09.2 Reactor Trip and Main Feedwater Isolation on SG 2 NR Level High-3 Steam Line Isolation TPS02A-01.4* TPS02S-01.4 TPS02C-01.4* TPS02D-01.4*

on SG 1 Pressure Low -2 Steam Line Isolation on TPS02A-01.5* TPS02S-01.5 TPS02C-01.5* TPS02D-01.5 SG 2 Pressure Low -2 Steam Line Isolation on TPS02A-11* TPS02S-11.2 TPS02C-11* TPS02D-11.2 Steam Line High Negative Rate SG 1 Steam Line Isolation on TPS02A-11" TPS02S-11.2 TPS02C-11* TPS02D-11.2 Steam Line High Negative Rate SG 2 Steam Line Isolation on TPS02A-01.7* TPS02S-11. 4 TPS02C-01.r TPS02D-11.4 Containment Pressure High-2 Steam Line Isolation on TPS02A-01.3* TPS02S-01.3 TPS02C-01.3* TPS02D-01.3 TCOLD Low-2 TPS02A-01.6* TPS02S-01.6 TPS02C-01.6" TPS02D-01.6 SG 1 Relief Isolation on TPS02A-11* TPS02S-25.2 TPS02C-11" TPS02D-25.2 SG 1 Steam line Press Low-2 SG 2 Relief Isolation on TPS02A-ir TPS02S-25.2 TPS02C-11* TPS02D-25.2 SG 2 Steam Line Press Low-2 SG Slowdown Isolation TPS01A-15.1 TPS02S-12.1 TPS01C-15.1 TPS02D-12.1 on SG 1 NR Level Low 2 SG Slowdown Isolation TPS01A-16.1 TPS02S-12.1 TPS01C-16.1 TPS02D-12.1 on SG 2 NR Level Low 2

U.S. Nuclear Regulatory Commission ND-19-1392 Enclosure Page 11 of 14 Table 2.5.2-3 ESF DivA Div B DivO Div D SG Slowdown Isolation TPS02A-08.2* TPS02B-08.2 TPS02C-08.2' TPS02D-08.2 on Pressurizer Water Level High-3 SG Slowdown Isolation N/A TPS02B-08.3 N/A TPS02D-08.3 on Startup Feedwater Flow Low-2 SG Slowdown Isolation TPS02A-08.2* TPS02B-08.2 TPS02C-08.2* TPS02D-08.2 on Pressurizer Water Level High-3 Passive Containment TPS02A-13 TPS02B-13 TPS02C-13 TPS02D-01.r Cooling Actuation on Containment Pressure High-2 Startup Feedwater TPS02A-28" TPS02B-07.1 TPS02C-09.2 TPS02D-07.1 Isolation on SG 1 NR Level High-3 Startup Feedwater TPS02A-09.2' TPS02B-09.2 TPS02C-09.2* TPS02D-09.2 Isolation on SG 2 NR Level High-3 Startup Feedwater TPS02A-01.3* TPS02B-01.3 TPS02C-01.3* TPS02D-01.3 Isolation on RCS Tcold TPS02A-01.6" TPS02B-14.2 TPS02C-01.6* TPS02D-14.2 Low-2 Startup Feedwater TPS02A-09.2" TPS02B-14.3 TPS02C-09.2' TPS02D-14.3 Isolation on SG 1 NR Level High-1 with P-4 Startup Feedwater TPS02A-09.2' TPS02B-14.3 TPS02C-09.2* TPS02D-14.3 Isolation on SG 2 NR Level High-1 with P-4 PRHR Heat Exchanger TPS02A-08.2 TPS02B-08.2 TPS02C-08.2 TPS02D-08.2 Alignment on Pressurizer Water Level High-3 PRHR Heat Exchanger N/A TPS02B-08.3 N/A TPS02D-08.3 Alignment on SG 1 Start Up Flow Low -2 and NR Level Low-2 PRHR Heat Exchanger N/A TPS02B-08.3 N/A TPS02D-08.3 Alignment on SG 2 Start Up Flow Low -2 and NR Level Low-2

U.S. Nuclear Regulatory Commission ND-19-1392 Enclosure Page 12 of 14 Table 2.5.2-3 ESF Div A DivB DivC Div D PRHR Heat Exchanger TPS02A-08.4 TPS02B-08.4 TPS02C-08.4* TPS02D-08.4 Alignment on SG 1 Wide Range Level Low-2 PRHR Heat Exchanger TPS02A-08.4 TPS02B-08.4 TPS02C-08.4* TPS02D-08.4 Alignment on SG 2 Wide Range Level Low-2 Block of Boron Dilution TPS02A-15.1 TPS02B-15" TPS02C-15.1 TPS02D-15.r on Flux Doubling CVS Makeup Isolation TPS02A-15.1 TPS02B-15* TPS02C-15.r TPS02D-15.1 on Flux Doubling Chemical and Volume TPS02A-16.1 TPS02B-16 TPS02C-16.1 TPS02D-16.1 Control System CVS Makeup Line Isolation on Pressurizer Water Level High-1 Chemical and Volume TPS02A-16.2 TPS02B-16* TPS02C-16.1* TPS02D-16.2 Control System CVS Makeup Line Isolation on Pressurizer Water Level High-2 Chemical and Volume TPS02A-16.2 TPS02B-21.2* TPS02C-20* TPS02D-16.2 Control System CVS Makeup Line Isolation on Containment Radioactivity High-2 Chemical and Volume TPS02A-09.2 TPS02B-09.2* TPS02C-09.2" TPS02D-09.2 Control System CVS Makeup Line Isolation on SG NR Level High - 3 Chemical and Volume TPS02A-16.3 TPS02B-14.3* TPS02C-09.2" TPS02D-14.3 Control System CVS Makeup Line Isolation on SG Level High w/ P-4

U.S. Nuclear Regulatory Commission ND-19-1392 Enclosure Page 13 of 14 Table 2.5.2-3 ESF DIv A DivB DIvC DivD Steam Dump Block on TPS02A-07.2" TPS02B-07.2 TPS02C-07.2* TPS02D-07.2 RCS Tavg Low-2 Main Control Room N/A TPS02B-18.2 TPS02C-18.2 N/A Isolation, Air Supply Initiation, Electrical Load De-energization on MCR Radiation Particulate High-2 Main Control Room N/A TPS02B-18.2 TPS02C-18.2 N/A Isolation, Air Supply Initiation, Electrical Load De-energization on MCR Radiation Iodine High-2 Main Control Room N/A TPS02B-18.4 TPS02C-18.4 N/A Isolation, Air Supply Initiation, Electrical Load De-energization on VES MCR Differential Pressure Low Auxiliary Spray Isolation TPS02A-19* TPS02B-19* TPS02C-19 TPS02D-19" on Pressurizer Water Level Low-1 Letdown Purification Line TPS02A-19* TPS02B-19* TPS02C-19 TPS02D-19*

Isolation on Pressurizer Water Level Low-1 Zinc and Hydrogen TPS02A-19 TPS02B-19* TPS02C-19* TPS02D-19 injection containment Isolation on Pressurizer Water Level Low-1 Containment Air TPS02A-20 TPS02B-20' TPS02C-20 TPS02D-20 Filtration System Isolation on Containment Radiation High-1

U.S. Nuclear Regulatory Commission ND-19-1392 Enclosure Page 14 of 14 Table 2.5.2-3 ESF Div A DivB Div C DivD Normal Residual Heat TPS02A-16.2 TPS02B-21.2 TPS02C-20* TPS02D-16.2 Removal Isolation on Containment Radiation High-2 Refueling Cavity and TPS02A-22 TPS02B-22 TPS02C-22 N/A Spent Fuel Pool Cooling System SFS Isolation on Spent Fuel Pool Level Low-2 Refueling Cavity and N/A N/A TPS02C-22.2 TPS02D-22.2 Spent Fuel Pool Cooling System SFS Isolation on IRWSTWR Level Low IRWST Containment TPS02A-10.2 TPS02B-10.2 TPS02C-10.2 TPS02D-10.2 Recirculation on IRWST NR Lower Level Low-3 CVS Letdown Isolation N/A TPS02B-23 TPS02C-23 N/A on RCS Hot Leg Level Low-2 Pressurizer Heater Block TPS02A-08.2 TPS02B-08.2 TPS02C-08.2 TPS02D-08.2 Isolated signal to non-safety equipment on Pressurizer Water level High-3 Pressurizer Heater TPS02A-08.2 TPS02B-08.2* TPS02C-08.2* TPS02D-08.2" Breaker Trip on Pressurizer Water level High-3 Containment Vacuum TPS02A-27 TPS02B-11.4* TPS02C-27 TPS02D-11.4*

Relief on Containment Pressure Low-2

• - Bistable operation only. No equipment operation included in the division with the listed actuation.