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): Difference between revisions

From kanterella
Jump to navigation Jump to search
VisualWikitext
StriderTol Bot insert
 
StriderTol Bot change
 
Line 16: Line 16:


=Text=
=Text=
{{#Wiki_filter:Michael J. Yox                7825 River Road
{{#Wiki_filter:!ik. Southern Nuclear NOV 1 <\\ 21119 Docket Nos.: 52-025 52-026 Michael J. Yox Regulatory Affairs Director Vogtle 3 & 4 7825 River Road Waynesboro, GA 30830 706-848-6459 tel ND-19-1392 10 CFR 52.99(c)(3)
!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:
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.
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
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.
[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 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).
Line 29: Line 28:


==Enclosure:==
==Enclosure:==
Vogtle Electric Generating Plant(VEGP) Unit 3 and Unit 4 Completion of ITAAC 2.5.02.10 [Index Number 549]
Vogtle Electric Generating Plant (VEGP) Unit 3 and Unit 4 Completion of ITAAC 2.5.02.10 [Index Number 549]
MJY/SBB/sfr
MJY/SBB/sfr


Line 47: Line 46:
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
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 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
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.
: 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.
Inspections/T ests/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.
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.
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
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.
: 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.
WCAP-16361-P, "Westinghouse Setpoint Methodology for Protection Systems -
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.
API 000" (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
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.
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),
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.
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, "API 000 Protection and Safety Monitoring system Preoperational and Component Test Specification" and SV4-PMS-T1-501, "API 000 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).
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.
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)
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"
: 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"
: 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"
: 3. WCAP-16361 -P, February 2011 "Westinghouse Setpoint Metfiodology for Protection Systems-API 000"
: 4. B-ADM-WCO-001,"Operational Readiness Work Management"
: 4. B-ADM-WCO-001, "Operational Readiness Work Management"
: 5. NMP-ES-084-001,"Plant Modification and Configuration Change Processes"
: 5. NMP-ES-084-001, "Plant Modification and Configuration Change Processes"
: 6. VEGP 3&4 UFSAR,
: 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"
* Section 14.2.9.1.12, "Protection and Safety Monitoring System Testing"
: 7. SV3-PMS-T1-501, "API 000 Protection and Safety Monitoring system Preoperational and Component Test Specification"
* Section 15.0.6, "Protection and Safety Monitoring System Setpoints and Time Delays to Trip Assumed in Accident Analyses"
: 8. SV4-PMS-T1-501, "API 000 Protection and Safety Monitoring system Preoperational and Component Test Specification"
: 7. SV3-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"
: 8. SV4-PMS-T1-501,"API000 Protection and Safety Monitoring system Preoperational and Component Test Specification"
: 10. APP-PMS-T2R-007 "API 000 Protection and Safety Monitoring System System-Level Reactor Trip Channel Integration Test Report"
: 9. APP-PMS-T5-001,"Protection and Safety Monitoring System Test Plan"
: 11. APP-PMS-T2R-008 "API 000 Protection and Safety Monitoring System System-Level Engineered Safety Features Channel Integration Test Report"
: 10. APP-PMS-T2R-007"API000 Protection and Safety Monitoring System System-Level Reactor Trip Channel Integration Test Report"
: 12. APP-PMS-T2R-050 "API 000 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
: 11. APP-PMS-T2R-008 "API000 Protection and Safety Monitoring System System-Level Engineered Safety Features Channel Integration Test Report"
: 15. NEI 08-01, "Industry Guideline for the ITAAC Closure Process Under 10 CFR Part 52"
: 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.
U.S. Nuclear Regulatory Commission ND-19-1392 Enclosure Page 5 of 14 Attachment A APP-PMS-T2R-007 "API 000 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.
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
Table 2.5.2-2 RT DivA DivB DIv 0 DlvD Source Range High Neutron Flux Reactor Trip TPS01A-01.1 TPS01B-01.1 TPS01C-01.1 TPS01D-01.1 Intermediate Range High Neutron Flux Reactor Trip TPS01A-02.1 TPS01B-02.1 TPS01C-02.1 TPS01D-02.1 Power Range High Neutron Flux (Low Setpoint) Trip TPS01A-03.1 TPS01B-03.1 TPS01C-03.1 TPS01D-03.1 Power Range High Neutron Flux (High Setpoint) Trip TPS01A-04.1 TPS01B-04.1 TPS01C-04.1 TPS01D-04.1 Power Range High Positive Flux Rate Trip TPS01A-05.1 TPS01B-05.1 TPS01C-05.1 TPS01D-05.1 Reactor Coolant Pump High-2 Bearing Water Temperature Trip RCP1A TPS01A-12.1 TPS01B-12.1 TPS01C-12.1 TPS01D-12.1 Reactor Coolant Pump High-2 Bearing Water Temperature Trip RCP1B TPS01A-12.2 TPS01B-12.2 TPS01C-12.2 TPS01D-12.2 Reactor Coolant Pump High-2 Bearing Water Temperature Trip RCP2A TPS01A-12.3 TPS01B-12.3 TPS01C-12.3 TPS01D-12.3


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


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 7 of 14 Table 2.5.2-2 RT DivA DivB DIvC DlvD High-3 Steam Generator Water Level Trip SG2 TPS01A-18.1 TPS01B-18.1 TPS01C-18.1 TPS01D-18.1


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.
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-007 "API 000 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-008 "API 000 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)
APP-PMS-T2R-050 "API 000 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.
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 begin with TPS01 are located in APP-PMS-T2R-007.
Test Cases that being with TPS02 are located in APP-PMS-T2R-008.
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.
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
Table 2.5.2-3 ESP DIv A DivB DIv C DlvD Safeguards Actuation on Pressurizer Pressure Low-3 TPS02A-01.1 TPS02B-01.1 TPS02C-01.1 TPS02D-01.1 Safeguards Actuation on RCS Cold Leg Low-2 TPS02A-01.3 TPS02A-01.6 TPS02B-01.3 TPS02B-01.6 TPS02C-01.3 TPS02C-01.6 TPS02D-01.3 TPS02D-01.6 Safeguards Actuation on SG 1 Steam Line Pressure Low-2 TPS02A-01.4 TPS02B-01.4 TPS02C-01.4 TPS02D-01.4 Safeguards Actuation on SG 2 Steam Line Pressure Low-2 TPS02A-01.5 TPS02B-01.5 TPS02C-01.5 TPS02D-01.5 Safeguards Actuation on Containment Pressure High-2 TPS02A-01.7 TPS02B-01.7 TPS02C-01.7 TPS02D-01.7 ADS Actuation stage 1, 2, 3 on CMT Level Low-3 TPS02A-05.3 TPS02B-05.3 TPS02C-05.3 TPS02D-05.3 ADS Actuation Stage 4 on RCS Hot Leg 1 & 2 Level Low-4 N/A TPS02B-05.5 TPS02C-05.5 N/A ADS Actuation Stage 4 on Core Makeup A Tank MR Lower Level Low-6 TPS02A-05.6 TPS02B-05.6 TPS02C-05.6 TPS02D-05.6 ADS Actuation Stage 4 on Core Makeup B Tank NR Lower Level Low-6 TPS02A-05.6 TPS02B-05.6 TPS02C-05.6 TPS02D-05.6


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


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*
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 on SG 1 and SG 2 WR Level Low -2 Plus Thot High TPS02A-04.3 TPS02S-04.3 TPS02C-04.3 TPS02D-04.3 Turbine Trip Reactor Trip and Main Feedwater Isolation on SGI NR Level High-3 TPS02A-09.2 TPS02S-09.2 TPS02C-09.2 TPS02D-09.2 Turbine Trip Reactor Trip and Main Feedwater Isolation on SG 2 NR Level High-3 TPS02A-09.2 TPS02S-09.2 TPS02C-09.2 TPS02D-09.2 Steam Line Isolation on SG 1 Pressure Low -2 TPS02A-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
TPS02S-01.4 TPS02C-01.4*
TPS02D-01.4*
Steam Line Isolation on SG 2 Pressure Low -2 TPS02A-01.5*
TPS02S-01.5 TPS02C-01.5*
TPS02D-01.5 Steam Line Isolation on Steam Line High Negative Rate SG 1 TPS02A-11*
TPS02S-11.2 TPS02C-11*
TPS02D-11.2 Steam Line Isolation on Steam Line High Negative Rate SG 2 TPS02A-11" TPS02S-11.2 TPS02C-11*
TPS02D-11.2 Steam Line Isolation on Containment Pressure High-2 TPS02A-01.7*
TPS02S-11. 4 TPS02C-01.r TPS02D-11.4 Steam Line Isolation on T
COLD Low-2 TPS02A-01.3*
TPS02A-01.6*
TPS02S-01.3 TPS02S-01.6 TPS02C-01.3*
TPS02C-01.6" TPS02D-01.3 TPS02D-01.6 SG 1 Relief Isolation on SG 1 Steam line Press Low-2 TPS02A-11*
TPS02S-25.2 TPS02C-11" TPS02D-25.2 SG 2 Relief Isolation on SG 2 Steam Line Press Low-2 TPS02A-ir TPS02S-25.2 TPS02C-11*
TPS02D-25.2 SG Slowdown Isolation on SG 1 NR Level Low 2 TPS01A-15.1 TPS02S-12.1 TPS01C-15.1 TPS02D-12.1 SG Slowdown Isolation on SG 2 NR Level Low 2 TPS01A-16.1 TPS02S-12.1 TPS01C-16.1 TPS02D-12.1


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


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


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*
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 RCS Tavg Low-2 TPS02A-07.2" TPS02B-07.2 TPS02C-07.2*
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
TPS02D-07.2 Main Control Room Isolation, Air Supply Initiation, Electrical Load De-energization on MCR Radiation Particulate High-2 N/A TPS02B-18.2 TPS02C-18.2 N/A Main Control Room Isolation, Air Supply Initiation, Electrical Load De-energization on MCR Radiation Iodine High-2 N/A TPS02B-18.2 TPS02C-18.2 N/A Main Control Room Isolation, Air Supply Initiation, Electrical Load De-energization on VES MCR Differential Pressure Low N/A TPS02B-18.4 TPS02C-18.4 N/A Auxiliary Spray Isolation on Pressurizer Water Level Low-1 TPS02A-19*
TPS02B-19*
TPS02C-19 TPS02D-19" Letdown Purification Line Isolation on Pressurizer Water Level Low-1 TPS02A-19*
TPS02B-19*
TPS02C-19 TPS02D-19*
Zinc and Hydrogen injection containment Isolation on Pressurizer Water Level Low-1 TPS02A-19 TPS02B-19*
TPS02C-19*
TPS02D-19 Containment Air Filtration System Isolation on Containment Radiation High-1 TPS02A-20 TPS02B-20' TPS02C-20 TPS02D-20


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*
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 Removal Isolation on Containment Radiation High-2 TPS02A-16.2 TPS02B-21.2 TPS02C-20*
Relief on Containment Pressure Low-2
TPS02D-16.2 Refueling Cavity and Spent Fuel Pool Cooling System SFS Isolation on Spent Fuel Pool Level Low-2 TPS02A-22 TPS02B-22 TPS02C-22 N/A Refueling Cavity and Spent Fuel Pool Cooling System SFS Isolation on IRWSTWR Level Low N/A N/A TPS02C-22.2 TPS02D-22.2 IRWST Containment Recirculation on IRWST NR Lower Level Low-3 TPS02A-10.2 TPS02B-10.2 TPS02C-10.2 TPS02D-10.2 CVS Letdown Isolation on RCS Hot Leg Level Low-2 N/A TPS02B-23 TPS02C-23 N/A Pressurizer Heater Block Isolated signal to non-safety equipment on Pressurizer Water level High-3 TPS02A-08.2 TPS02B-08.2 TPS02C-08.2 TPS02D-08.2 Pressurizer Heater Breaker Trip on Pressurizer Water level High-3 TPS02A-08.2 TPS02B-08.2*
  *- Bistable operation only. No equipment operation included in the division with the listed actuation.}}
TPS02C-08.2*
TPS02D-08.2" Containment Vacuum Relief on Containment Pressure Low-2 TPS02A-27 TPS02B-11.4*
TPS02C-27 TPS02D-11.4*
*- Bistable operation only. No equipment operation included in the division with the listed actuation.}}

Latest revision as of 20:11, 1 January 2025

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)
v • d • e


Text

!ik. Southern Nuclear NOV 1 <\\ 21119 Docket Nos.: 52-025 52-026 Michael J. Yox Regulatory Affairs Director Vogtle 3 & 4 7825 River Road Waynesboro, GA 30830 706-848-6459 tel 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/T ests/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 -

API 000" (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, "API 000 Protection and Safety Monitoring system Preoperational and Component Test Specification" and SV4-PMS-T1-501, "API 000 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-API 000"
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, "API 000 Protection and Safety Monitoring system Preoperational and Component Test Specification"
8. SV4-PMS-T1-501, "API 000 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 "API 000 Protection and Safety Monitoring System System-Level Reactor Trip Channel Integration Test Report"
11. APP-PMS-T2R-008 "API 000 Protection and Safety Monitoring System System-Level Engineered Safety Features Channel Integration Test Report"
12. APP-PMS-T2R-050 "API 000 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 "API 000 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 Neutron Flux Reactor Trip TPS01A-01.1 TPS01B-01.1 TPS01C-01.1 TPS01D-01.1 Intermediate Range High Neutron Flux Reactor Trip TPS01A-02.1 TPS01B-02.1 TPS01C-02.1 TPS01D-02.1 Power Range High Neutron Flux (Low Setpoint) Trip TPS01A-03.1 TPS01B-03.1 TPS01C-03.1 TPS01D-03.1 Power Range High Neutron Flux (High Setpoint) Trip TPS01A-04.1 TPS01B-04.1 TPS01C-04.1 TPS01D-04.1 Power Range High Positive Flux Rate Trip TPS01A-05.1 TPS01B-05.1 TPS01C-05.1 TPS01D-05.1 Reactor Coolant Pump High-2 Bearing Water Temperature Trip RCP1A TPS01A-12.1 TPS01B-12.1 TPS01C-12.1 TPS01D-12.1 Reactor Coolant Pump High-2 Bearing Water Temperature Trip RCP1B TPS01A-12.2 TPS01B-12.2 TPS01C-12.2 TPS01D-12.2 Reactor Coolant Pump High-2 Bearing Water Temperature Trip RCP2A TPS01A-12.3 TPS01B-12.3 TPS01C-12.3 TPS01D-12.3

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 High-2 Bearing Water Temperature Trip RCP2B TPS01A-12.4 TPS01B-12.4 TPS01C-12.4 TPS01D-12.4 Overtemperature Delta-TTrip TPS01A-07.1 TPS01B-07.1 TPS01C-07.1 TPS01D-07.1 Overtemperature Delta-TTrip TPS01A-07.2 TPS01B-07.2 TPS01C-07.2 TPS01D-07.2 Overpower Delta-T Trip TPS01A-08.1 TPS01B-08.1 TPS01C-08.1 TPS01D-08.1 Pressurizer Low-2 Pressure Trip TPS01A-09.1 TPS01B-09.1 TPS01C-09.1 TPS01D-09.1 Pressurizer High-2 Pressure Trip TPS01A-13.1 TPS01B-13.1 TPS01C-13.1 TPS01D-13.1 Pressurizer High-3 Water Level Trip TPS01A-14.1 TPS01B-14.1 TPS01C-14.1 TPS01D-14.1 Low-2 Reactor Coolant Flow Trip ML 1 TPS01A-10.1 TPS01B-10.1 TPS01C-10.1 TPS01D-10.1 Low-2 Reactor Coolant Flow Trip ML 2 TPS01A-10.2 TPS01B-10.2 TPS01C-10.2 TPS01D-10.2 Low-2 Reactor Coolant Pump Speed Trip TPS01A-11.1 TPS01B-11.1 TPS01C-11.1 TPS01D-11.1 Low-2 Steam Generator Narrow Range Water Level Trip SGI TPS01A-15.1 TPS01B-15.1 TPS01C-15.1 TPS01D-15.1 Low-2 Steam Generator Narrow Range Water Level Trip SG2 TPS01A-16.1 TPS01B-16.1 TPS01C-16.1 TPS01D-16.1 High-3 Steam Generator Water Level Trip SGI TPS01A-17.1 TPS01B-17.1 TPS01C-17.1 TPS01D-17.1

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 Generator Water Level Trip SG2 TPS01A-18.1 TPS01B-18.1 TPS01C-18.1 TPS01D-18.1

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 "API 000 Protection and Safety Monitoring System System-Level Reactor Trip Channel Integration Test Report" (Reference 10)

APP-PMS-T2R-008 "API 000 Protection and Safety Monitoring System System-Level Engineered Safety Features Channel Integration Test Report" (Reference 11)

APP-PMS-T2R-050 "API 000 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 Pressurizer Pressure Low-3 TPS02A-01.1 TPS02B-01.1 TPS02C-01.1 TPS02D-01.1 Safeguards Actuation on RCS Cold Leg Low-2 TPS02A-01.3 TPS02A-01.6 TPS02B-01.3 TPS02B-01.6 TPS02C-01.3 TPS02C-01.6 TPS02D-01.3 TPS02D-01.6 Safeguards Actuation on SG 1 Steam Line Pressure Low-2 TPS02A-01.4 TPS02B-01.4 TPS02C-01.4 TPS02D-01.4 Safeguards Actuation on SG 2 Steam Line Pressure Low-2 TPS02A-01.5 TPS02B-01.5 TPS02C-01.5 TPS02D-01.5 Safeguards Actuation on Containment Pressure High-2 TPS02A-01.7 TPS02B-01.7 TPS02C-01.7 TPS02D-01.7 ADS Actuation stage 1, 2, 3 on CMT Level Low-3 TPS02A-05.3 TPS02B-05.3 TPS02C-05.3 TPS02D-05.3 ADS Actuation Stage 4 on RCS Hot Leg 1 & 2 Level Low-4 N/A TPS02B-05.5 TPS02C-05.5 N/A ADS Actuation Stage 4 on Core Makeup A Tank MR Lower Level Low-6 TPS02A-05.6 TPS02B-05.6 TPS02C-05.6 TPS02D-05.6 ADS Actuation Stage 4 on Core Makeup B Tank NR Lower Level Low-6 TPS02A-05.6 TPS02B-05.6 TPS02C-05.6 TPS02D-05.6

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 on RCS Wide Range Pressure Low TPS02A-05.6 TPS02B-05.6 TPS02C-05.6 TPS02D-05.6 Main Feedwater Isolation on SG 1 NR Level High-3 TPS02A-28*

TPS02B-07.1 TPS02C-09.2 TPS02D-07.1 Main Feedwater Isolation on SG 2 NR Level High-3 TPS02A-09.2*

TPS02B-09.2 TPS02C-09.2*

TPS02D-09.2 Main Feedwater Isolation on RCS Tavg Low-2 TPS02A-07.2*

TPS02B-07.2 TPS02C-07.2*

TPS02D-07.2 Main Feedwater Isolation Control Valves on RCS Tavg Low-1 TPS02A-07.2*

TPS02B-07.2*

TPS02C-07.2*

TPS02D-07.3 Main Feedwater Isolation Control Valves on SG NR WL High-3 TPS02A-09.2*

TPS02B-09.2*

TPS02C-09.2*

TPS02D-09.2 Reactor Coolant Pump Trip & CCS Containment Isolation on RCP 1 -A Bearing Water Temperature High-2 TPS02A-06 TPS02B-06 TPS02C-06 TPS02D-06 Reactor Coolant Pump Trip & CCS Containment Isolation on RCP 1-B Bearing Water Temperature High-2 TPS01A-12.2 TPS01B-12.2 TPS01C-12.2 TPS01D-12.2 Reactor Coolant Pump Trip & CCS Containment Isolation on RCP 2-A Bearing Water Temperature High-2 TPS01A-12.3 TPS01B-12.3 TPS01C-12.3 TPS01D-12.3 Reactor Coolant Pump Trip & CCS Containment Isolation on RCP 2-B Bearing Water Temperature High-2 CCS Containment Isolation TPS01A-12.4 TPS01B-12.4 TPS01C-12.4 TPS01D-12.4 CMT Actuation on Pzr Level Low-2 TPS02A-04.2 TPS02B-04.2 TPS02C-04.2 TPS02D-04.2

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 on SG 1 and SG 2 WR Level Low -2 Plus Thot High TPS02A-04.3 TPS02S-04.3 TPS02C-04.3 TPS02D-04.3 Turbine Trip Reactor Trip and Main Feedwater Isolation on SGI NR Level High-3 TPS02A-09.2 TPS02S-09.2 TPS02C-09.2 TPS02D-09.2 Turbine Trip Reactor Trip and Main Feedwater Isolation on SG 2 NR Level High-3 TPS02A-09.2 TPS02S-09.2 TPS02C-09.2 TPS02D-09.2 Steam Line Isolation on SG 1 Pressure Low -2 TPS02A-01.4*

TPS02S-01.4 TPS02C-01.4*

TPS02D-01.4*

Steam Line Isolation on SG 2 Pressure Low -2 TPS02A-01.5*

TPS02S-01.5 TPS02C-01.5*

TPS02D-01.5 Steam Line Isolation on Steam Line High Negative Rate SG 1 TPS02A-11*

TPS02S-11.2 TPS02C-11*

TPS02D-11.2 Steam Line Isolation on Steam Line High Negative Rate SG 2 TPS02A-11" TPS02S-11.2 TPS02C-11*

TPS02D-11.2 Steam Line Isolation on Containment Pressure High-2 TPS02A-01.7*

TPS02S-11. 4 TPS02C-01.r TPS02D-11.4 Steam Line Isolation on T

COLD Low-2 TPS02A-01.3*

TPS02A-01.6*

TPS02S-01.3 TPS02S-01.6 TPS02C-01.3*

TPS02C-01.6" TPS02D-01.3 TPS02D-01.6 SG 1 Relief Isolation on SG 1 Steam line Press Low-2 TPS02A-11*

TPS02S-25.2 TPS02C-11" TPS02D-25.2 SG 2 Relief Isolation on SG 2 Steam Line Press Low-2 TPS02A-ir TPS02S-25.2 TPS02C-11*

TPS02D-25.2 SG Slowdown Isolation on SG 1 NR Level Low 2 TPS01A-15.1 TPS02S-12.1 TPS01C-15.1 TPS02D-12.1 SG Slowdown Isolation on SG 2 NR Level Low 2 TPS01A-16.1 TPS02S-12.1 TPS01C-16.1 TPS02D-12.1

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 on Pressurizer Water Level High-3 TPS02A-08.2*

TPS02B-08.2 TPS02C-08.2' TPS02D-08.2 SG Slowdown Isolation on Startup Feedwater Flow Low-2 N/A TPS02B-08.3 N/A TPS02D-08.3 SG Slowdown Isolation on Pressurizer Water Level High-3 TPS02A-08.2*

TPS02B-08.2 TPS02C-08.2*

TPS02D-08.2 Passive Containment Cooling Actuation on Containment Pressure High-2 TPS02A-13 TPS02B-13 TPS02C-13 TPS02D-01.r Startup Feedwater Isolation on SG 1 NR Level High-3 TPS02A-28" TPS02B-07.1 TPS02C-09.2 TPS02D-07.1 Startup Feedwater Isolation on SG 2 NR Level High-3 TPS02A-09.2' TPS02B-09.2 TPS02C-09.2*

TPS02D-09.2 Startup Feedwater Isolation on RCS Tcold Low-2 TPS02A-01.3*

TPS02A-01.6" TPS02B-01.3 TPS02B-14.2 TPS02C-01.3*

TPS02C-01.6*

TPS02D-01.3 TPS02D-14.2 Startup Feedwater Isolation on SG 1 NR Level High-1 with P-4 TPS02A-09.2" TPS02B-14.3 TPS02C-09.2' TPS02D-14.3 Startup Feedwater Isolation on SG 2 NR Level High-1 with P-4 TPS02A-09.2' TPS02B-14.3 TPS02C-09.2*

TPS02D-14.3 PRHR Heat Exchanger Alignment on Pressurizer Water Level High-3 TPS02A-08.2 TPS02B-08.2 TPS02C-08.2 TPS02D-08.2 PRHR Heat Exchanger Alignment on SG 1 Start Up Flow Low -2 and NR Level Low-2 N/A TPS02B-08.3 N/A TPS02D-08.3 PRHR Heat Exchanger Alignment on SG 2 Start Up Flow Low -2 and NR Level Low-2 N/A TPS02B-08.3 N/A TPS02D-08.3

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 Alignment on SG 1 Wide Range Level Low-2 TPS02A-08.4 TPS02B-08.4 TPS02C-08.4*

TPS02D-08.4 PRHR Heat Exchanger Alignment on SG 2 Wide Range Level Low-2 TPS02A-08.4 TPS02B-08.4 TPS02C-08.4*

TPS02D-08.4 Block of Boron Dilution on Flux Doubling TPS02A-15.1 TPS02B-15" TPS02C-15.1 TPS02D-15.r CVS Makeup Isolation on Flux Doubling TPS02A-15.1 TPS02B-15*

TPS02C-15.r TPS02D-15.1 Chemical and Volume Control System CVS Makeup Line Isolation on Pressurizer Water Level High-1 TPS02A-16.1 TPS02B-16 TPS02C-16.1 TPS02D-16.1 Chemical and Volume Control System CVS Makeup Line Isolation on Pressurizer Water Level High-2 TPS02A-16.2 TPS02B-16*

TPS02C-16.1*

TPS02D-16.2 Chemical and Volume Control System CVS Makeup Line Isolation on Containment Radioactivity High-2 TPS02A-16.2 TPS02B-21.2*

TPS02C-20*

TPS02D-16.2 Chemical and Volume Control System CVS Makeup Line Isolation on SG NR Level High - 3 TPS02A-09.2 TPS02B-09.2*

TPS02C-09.2" TPS02D-09.2 Chemical and Volume Control System CVS Makeup Line Isolation on SG Level High w/ P-4 TPS02A-16.3 TPS02B-14.3*

TPS02C-09.2" TPS02D-14.3

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 RCS Tavg Low-2 TPS02A-07.2" TPS02B-07.2 TPS02C-07.2*

TPS02D-07.2 Main Control Room Isolation, Air Supply Initiation, Electrical Load De-energization on MCR Radiation Particulate High-2 N/A TPS02B-18.2 TPS02C-18.2 N/A Main Control Room Isolation, Air Supply Initiation, Electrical Load De-energization on MCR Radiation Iodine High-2 N/A TPS02B-18.2 TPS02C-18.2 N/A Main Control Room Isolation, Air Supply Initiation, Electrical Load De-energization on VES MCR Differential Pressure Low N/A TPS02B-18.4 TPS02C-18.4 N/A Auxiliary Spray Isolation on Pressurizer Water Level Low-1 TPS02A-19*

TPS02B-19*

TPS02C-19 TPS02D-19" Letdown Purification Line Isolation on Pressurizer Water Level Low-1 TPS02A-19*

TPS02B-19*

TPS02C-19 TPS02D-19*

Zinc and Hydrogen injection containment Isolation on Pressurizer Water Level Low-1 TPS02A-19 TPS02B-19*

TPS02C-19*

TPS02D-19 Containment Air Filtration System Isolation on Containment Radiation High-1 TPS02A-20 TPS02B-20' TPS02C-20 TPS02D-20

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 Removal Isolation on Containment Radiation High-2 TPS02A-16.2 TPS02B-21.2 TPS02C-20*

TPS02D-16.2 Refueling Cavity and Spent Fuel Pool Cooling System SFS Isolation on Spent Fuel Pool Level Low-2 TPS02A-22 TPS02B-22 TPS02C-22 N/A Refueling Cavity and Spent Fuel Pool Cooling System SFS Isolation on IRWSTWR Level Low N/A N/A TPS02C-22.2 TPS02D-22.2 IRWST Containment Recirculation on IRWST NR Lower Level Low-3 TPS02A-10.2 TPS02B-10.2 TPS02C-10.2 TPS02D-10.2 CVS Letdown Isolation on RCS Hot Leg Level Low-2 N/A TPS02B-23 TPS02C-23 N/A Pressurizer Heater Block Isolated signal to non-safety equipment on Pressurizer Water level High-3 TPS02A-08.2 TPS02B-08.2 TPS02C-08.2 TPS02D-08.2 Pressurizer Heater Breaker Trip on Pressurizer Water level High-3 TPS02A-08.2 TPS02B-08.2*

TPS02C-08.2*

TPS02D-08.2" Containment Vacuum Relief on Containment Pressure Low-2 TPS02A-27 TPS02B-11.4*

TPS02C-27 TPS02D-11.4*

  • - Bistable operation only. No equipment operation included in the division with the listed actuation.
Retrieved from "https://kanterella.com/w/index.php?title=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)&oldid=4787118"