ND-19-1272, Unit 4 - Notice of Uncompleted ITAAC 225-days Prior to Initial Fuel Load Item 2.5.04.02.1 (Index Number 557)

From kanterella
(Redirected from ND-19-1272)
Jump to navigation Jump to search
Unit 4 - Notice of Uncompleted ITAAC 225-days Prior to Initial Fuel Load Item 2.5.04.02.1 (Index Number 557)
ML19298B121
Person / Time
Site: Vogtle  Southern Nuclear icon.png
Issue date: 10/25/2019
From: Yox M
Southern Nuclear Operating Co
To:
Document Control Desk, Office of New Reactors
References
ND-19-1272
Download: ML19298B121 (15)
v • d • e


Text

Michael J. Yox 7825 River Road

^Southern Nuclear Regulatory Affairs Director Vogtle 3 & 4 Waynesboro, GA 30830 OCT 2 5 2019 Docket Nos.: 52-025 52-026 ND-19-1272 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 Notice of Uncompleted ITAAC 225-davs Prior to Initial Fuel Load Item 2.5.04.02.1 [Index Number 5571 Ladies and Gentlemen:

Pursuant to 10 CFR 52.99(c)(3), Southern Nuclear Operating Company hereby notifies the NRC that as of October 23, 2019, Vogtle Electric Generating Plant(VEGP) Unit 3 and Unit 4 Uncompleted Inspection, Test, Analysis, and Acceptance Criteria (ITAAC) Item 2.5.04.02.1 [Index Number 557] has not been completed greater than 225-days prior to initial fuel load. The Enclosure describes the plan for completing ITAAC 2.5.04.02.1 [Index Number 557]. 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.

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 Plan for Uncompleted ITAAC 2.5.04.02.1 [Index Number 557]

MJY/JBN/sfr

U.S. Nuclear Regulatory Commission ND-19-1272 Page 2 of 3 To:

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

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

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

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

Mr. G. Chick Mr. M. Page 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. C. Leighty Ms. A. C. Chamberlain Mr. J. C. Haswell Document Services RTYPE: VND.LI.L06 File AR.01.02.06 cc:

Nuclear Regulatory 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. 8. 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 Oqlethorpe Power Corporation Mr. R. B. Brinkman Mr. E. Rasmussen Municipal Electric Authoritv of Georgia Mr. J. E. Fuller Mr. S. M. Jackson

U.S. Nuclear Regulatory Commission ND-19-1272 Page 3 of 3 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-19-1272 Enclosure Page 1 of 12 Southern Nuclear Operating Company ND-19-1272 Enclosure Vogtle Electric Generating Plant(VEGP) Unit 3 and Unit 4 Completion Plan for Uncompleted ITAAC 2.5.04.02.1 [index Number 557]

U.S. Nuclear Regulatory Commission ND-19-1272 Enclosure Page 2 of 12 ITAAC Statement Desion Commitment

2. The DDS provides for the minimum inventory of displays, visual alerts, and fixed position controls, as identified in Table 2.5.4-1. The plant parameters listed with a "Yes" in the "Display" column and visual alerts listed with a "Yes" in the "Alert" column can be retrieved at the RSW.

The controls listed with a "Yes" in the "Control" column are provided at the RSW.

Inspections/Tests/Analvses i) An inspection wiil be performed for retrievability of plant parameters at the RSW.

ii) An inspection and test will be performed to verify that the plant parameters are used to generate visual alerts that identify challenges to critical safety functions.

iii) An operational test of the as-built system will be performed using each RSW control.

Acceptance Criteria i) The piant parameters iisted in Table 2.5.4-1 with a "Yes" in the "Dispiay" column can be retrieved at the RSW.

ii) The plant parameters listed in Table 2.5.4-1 with a "Yes" in the "Alert" column are used to generate visual alerts that identify challenges to critical safety functions. The visuai alerts actuate in accordance with their logic and values.

iii) For each test of a control listed in Table 2.5.4-1 with a "Yes" in the "Control" column, an actuation signal is generated. Tests from the actuation signai to the actuated device(s) are performed as part of the system-reiated inspection, test, analysis and acceptance criteria.

ITAAC Completion Description Multiple ITAAC are performed to verify that the Data Dispiay and Processing System (DDS) provides for the minimum inventory of displays, visual alerts, and fixed position controls, as identified in COL Appendix C Table 2.5.4-1 (Attachment A), the plant parameters listed with a "Yes" in the "Display" column and visual alerts listed with a "Yes" in the "Alert" column can be retrieved at the Remote Shutdown Workstation (RSW), and the controls listed with a "Yes" in the "Control" column are provided at the RSW.The subject ITAAC performs inspections on the displays in Attachment A to verify the listed plant parameters can be retrieved at the RSW, inspections and testing of the alerts in Attachment A to verify that the listed piant parameters are used to generate visual alerts that identify challenges to critical safety functions, and testing of the controls listed in Attachment A to verify the listed controls generate actuation signals.

The oiant parameters listed in Table 2.5.4-1 with a "Yes" in the "Disolav" column can be retrieved at the RSW.

An inspection is performed to verify the retrievability of the VEGP Unit 3 and Unit 4 plant parameters at the RSW.The inspection for retrievability confirms that the piant parameters

U.S. Nuclear Regulatory Commission ND-19-1272 Enclosure Page 3 of 12 listed in Attachment A with a "Yes" In the "Display" column (Attachment A)can be retrieved at the RSW.

The Inspection Is performed In accordance with 3-DDS-ITPP-520 and 4-DDS-ITPP-520 (Reference 1 and 2)and visually confirms that when each of the plant parameters Identified In Attachment A with a "Yes" In the "Display" column Is recalled at the RSW,the recalled plant parameter appears on a display monitor.

The Inspection results are Included In References 1 and 2 and confirm that the plant parameters listed In Table 2.5.4-1 with a "Yes" In the "Display" column can be retrieved at the RSW.

References 1 and 2 are available for NRG Inspection as part of the Unit 3 and Unit 4 ITAAC 2.5.04.02.1 Completion Packages (References 36 and 37).

The plant parameters listed In Table 2.5.4-1 with a "Yes" In the "Alert" column are used to Generate visual alerts that Identlfv challenges to critical safetv functions. The visual alerts actuate In accordance with their logic and values.

Inspections and testing are performed to verify the retrlevablllty of the VEGP Unit 3 and Unit 4 visual alerts at the RSW.The Inspections and testing confirm that the plant parameters listed In Attachment A with a "Yes" In the "Alert" column are used to generate visual alerts that Identify challenges to critical safety functions(CSF)and actuate In accordance with their logic and values.

This ITAAC Is completed as a combination of:

  • PMS Factory Acceptance Test(FAT)- Functional testing of PMS Inputs, outputs, logic, and function
  • Preoperatlonal Test of communication - Functional testing of the communication between the PMS output and the DDS Input
  • DDS FAT- Functional testing of DDS Inputs, outputs, logic, and function
  • Preoperatlonal Test of the as-bullt RSW - Visual Inspection and test of the visual alerts at the as-bullt RSW The PMS FAT was performed In accordance with PMS Software Program Manual WCAP-16096 (Reference 3), PMS Test Plan APP-PMS-T5-001 (Reference 4)and applicable Codes and Standards described In Vogtle 3 and 4 UFSAR Chapter 7. The DDS FAT was performed In accordance with PLS Test Plan APP-PLS-T5-001 (Reference 5).

The logic that generates the visual alerts Is contained In both the Qualified Data Processing System (QDPS)of the PMS and the Nuclear Applications Programs(NAPs) of the DDS and are tested as follows:

  • Logic and values of visual alerts generated In the QDPS are verified In the PMS FAT
  • Logic and values of visual alerts generated In the NAPs are verified In the DDS FAT

U.S. Nuclear Regulatory Commission ND-19-1272 Enclosure Page 4 of 12 During the RMS FAT, the plant parameters were simulated and adjusted to create applicable alert conditions. RMS outputs were monitored during the test, and it was confirmed that the outputs from the RMS which support the DDS visual alerts were output from the RMS as designed. This testing was performed in accordance with FAT Test Procedures ARR-RMS-T1R-010 and ARR-RMS-T1R-035(References 6 and 14). The results of the RMS testing are documented in the FAT test reports ARR/SV3/SV4-RMS-T2R-010(References 7 through 9) and SV0/SV3/SV4-RMS-T2R-035(References 17 through 19) Test Case TRS35-01. Attachment B provides a listing of test cases used in References 7 through 9.

To provide communication between the RMS and DDS,the Maintenance and Test Panel(MTR) in a given RMS division provides an isolated (optical-to-electrical isolation) pathway from the intra-divisional communication bus to the Advant/Ovation Interface (AOI) Gateway associated with that division. Over the divisional AOI Gateway, the MTR transfers certain real-time data from the division's AF100 bus to the non-safety Real Time Data Network to support control and information system functions performed in non-safety systems, such as the DDS. Testing in 3-RMS-ITRR-521 and 4-RMS-ITRR-521 (References 10 and 11) verifies the AOI gateway by ensuring datapoints on RMS which are output to the DDS match those on the DDS.

During the DDS FAT, inputs to the DDS were simulated and adjusted to create applicable alert conditions and it was confirmed that the logic and functionality of the DDS supports the visual alerts. This testing was performed in accordance with FAT Test Procedures ARR-DDS-T1R-001 and ARR-RCS-T1R-100(References 12 and 13). The results of the testing are documented in the FAT test reports ARR-DDS-T1 R-001 and ARR-RCS-T2R-100(References 15 and 16). provides a listing of test cases used in these tests.

Testing is performed in accordance with 3-DDS-ITRR-520 and 4-DDS-ITRR-520 (References 1 and 2)to verify that when the applicable DDS input is simulated, each plant parameter listed in Attachment A with a "Yes" in the "Alert" column is used to generate visual alerts that identify challenges to CSF at the RSW.Testing in References 1 and 2 forces the applicable DDS input parameter from an engineering workstation and visually confirms that when each of the plant parameters identified in Attachment A with a "Yes" in the "Alert" column is used to generate visual alerts that identify challenges to CSF,the summoned plant visual alert appears on a display monitor at the RSW.

The completed Unit 3 and Unit 4 factory test results (References 7 through 9 and 15 through 19) and preoperational test results (References 1,2, 10, and 11) confirm that the plant parameters listed in Table 2.5.4-1 with a "Yes" in the "Alert" column are used to generate visual alerts that identify challenges to critical safety functions. The visual alerts actuate in accordance with their logic and values.

References 1,2,7 through 11, and 15 through 19 are available for NRG inspection as part of the Unit 3 and Unit 4 ITAAG 2.5.04.02.i Completion Packages (References 36 and 37).

For each test of a control listed in Table 2.5.4-1 with a "Yes" in the "Control" column, an actuation signal is Generated. Tests from the actuation sional to the actuated devicefsl are performed as part of the svstem-related inspection, test, analvsis and acceptance criteria.

An operational test of the as-built system is performed using each RSW control. The test confirms that for each test of a control listed in Table 2.5.4-1 (Attachment A) with a "Yes" in the "Control" column, an actuation signal is generated. Tests from the actuation signal to the

U.S. Nuclear Regulatory Commission ND-19-1272 Enclosure Page 5 of 12 actuated device(s) are performed as part of tfie system-related inspection, test, analysis and acceptance criteria.

This ITAAC is completed as a combination of:

  • Factory Acceptance Test- Functional testing of the RMS control circuit
  • Site software installation and regression test - Hardware and software integration verification and testing of post system delivery changes
  • Component test- testing of the remote shutdown room switches, including their interface with RMS and full testing of the hydrogen igniter soft controls at the RSW The Factory Acceptance Testing (FAT)follows the guidance of NEI 08-01 Section 9.4 (Reference 38)for the as-built tests to be performed at other than the final installed location.

The FAT was performed in accordance with RMS Software Program Manual WCAR-16096 (Reference 3), RMS Test Plan ARR-RMS-T5-001 (Reference 4), and applicable Codes and Standards described in Vogtle 3 and 4 UFSAR Chapter 7.

The FAT included testing of RMS inputs and outputs, logic, and functionality. During this test, the manual inputs to the RMS were simulated and it was confirmed that the actuation signals were generated for the minimum inventory of controls at the RSW identified in Attachment A.

This testing was performed in accordance with the RMS FAT procedures ARR-RMS-T1R-007 and ARR-RMS-T1R-008(References 20 and 21). The results of the tests are documented in the FAT test reports SV0/SV3/SV4-RMS-T2R-007 and SV0/SV3/SV4-RMS-T2R-008 (References 22 through 27). Attachment C provides a listing of test cases used in these tests.

Additional hardware and software installation and associated inspections and testing are performed on-site to verify that the cabinets are intact and functional in accordance with Units 3 and 4 for applicable Field Change Notifications(FCNs) API000 Vogtle Unit 3 RMS Initial Software Installation - Software Release 8.7.0.1 and API000 Vogtle Unit 4 RMS Initial Software Installation - Software Release 8.7.0.1 (References 28 and 29). References 28 and 29 include steps that confirm and document successful software load and further confirm the physical properties of the as-built RMS. A regression analysis (i.e., change evaluation) is performed post-delivery and installation for hardware changes (References 33 and 34) and software changes (Reference 35)to determine if additional testing is needed for the as-built system.

Component testing of the dedicated RSW controls identified in Attachment A is performed in accordance with component test packages SNCXXXXXX (Unit 3) and SNCYYYYYY (Unit 4)

(References 30 and 31). These component test packages utilize B-GEN-ITRCI-006,(Reference

32) to test the RSW manual controls. Selected RSW manual controls are actuated and confirmed at the RMS input, by visually inspecting the digital input light emitting diodes. The completed Unit 3 and Unit 4 component test packages confirm that select RSW manual control actuations are received at the RMS.

For the containment hydrogen igniters, testing is performed in accordance with 3-DDS-ITRR-520 and 4-DDS-ITRR-520 (References 1 and 2). Testing in References 1 and 2 verifies the Hydrogen Control System (VLS) is available and capable of energizing the hydrogen igniters, then the containment hydrogen igniters are energized using soft controls from the RSW. Local

U.S. Nuclear Regulatory Commission ND-19-1272 Enclosure Page 6 of 12 voltage verification at the igniter control relays verifies the hydrogen igniter soft controls generate an actuation signal and is documented in the test.

The completed Unit 3 and Unit 4 FAT results (References 22 through 27), PONs(References 28 and 29), regression test results (References 33 through 35), completed component test packages(References 30 and 31), and preoperational test results (References 1 and 2) confirm that for each test of a control listed in Table 2.5.4-1 with a "Yes" in the "Control" column, an actuation signal is generated.

References 1, 2, 22 through 27, 30, and 31 are available for NRC inspection as part of the Unit 3 and Unit 4 ITAAC 2.5.04.02.i Completion Packages(References 36 and 37).

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

References(available for NRC inspection)

1. 3-DDS-ITPP-520,"Data and Display Processing System Remote Shutdown Room Preoperational Test Procedure"
2. 4-DDS-ITPP-520,"Data and Display Processing System Remote Shutdown Room Preoperational Test Procedure"
3. WCAP-16096 "Software Program Manual for Common Q Systems" Revision 4A
4. APP-PMS-T5-001,"API000 Protection and Safety Monitoring System Test Plan"
5. APP-PLS-T5-001,"API000 Plant Control System/Data Display and Processing System Test Plan"
6. APP-PMS-T1P-010,"API000 Protection and Safety Monitoring System Qualified Data Processing System Channel Integration Test Procedure"
7. APP-PMS-T2R-010 "API000 Protection and Safety Monitoring System Qualified Data Processing System Channel Integration Test Report"
8. SV3-PMS-T2R-010 "Vogtle Unit 3 API000 Protection and Safety Monitoring System Qualified Data Processing System Channel Integration Test Report"
9. SV4-PMS-T2R-010 "Vogtle Unit 4 API000 Protection and Safety Monitoring System Qualified Data Processing System Channel Integration Test Report"
10. 3-PMS-ITPP-521,"Protection and Safety Monitoring System Logic Test Preoperational Test Procedure" 11.4-PMS-ITPP-521,"Protection and Safety Monitoring System Logic Test Preoperational Test Procedure"
12. APP-DDS-T1P-001,"API000 Data Display and Processing System Application Programs Test Procedure"
13. APP-PCS-T1P-100,"API000 Plant Control System Passive Containment Cooling System (PCS) Software Test Procedure"
14. APP-PMS-T1P-035,"API000 Protection and Safety Monitoring System Display Calibration Data Test Procedure"
15. APP-DDS-T1R-001,"API000 Data Display and Processing System Application Programs Test Report"

U.S. Nuclear Regulatory Commission ND-19-1272 Enclosure Page 7 of 12

16. APP-PCS-T2R-100,"API000 Plant Control System Passive Containment Cooling System (PCS) Software Test Report"
17. SV0-PMS-T2R-035,"API000 Protection and Safety Monitoring System - Reactor Trip Channel Integration Test Report"
18. SV3-PMS-T2R-035,"Vogtle Unit 3 API000 Protection and Safety Monitoring System -

Reactor Trip Channel Integration Test Report"

19. SV4-PMS-T2R-035,"Vogtle Unit 4 API000 Protection and Safety Monitoring System -

Reactor Trip Channel Integration Test Report"

20. APP-PMS-T1P-007,"API000 Protection and Safety Monitoring System - Reactor Trip Channel Integration Test Procedure"
21. APP-PMS-T1P-008,"API000 Protection and Safety Monitoring System - System-Level Engineered Safety Features Channel Integration Test Procedure"
22. SV0-PMS-T2R-007,"API000 Protection and Safety Monitoring System - Reactor Trip Channel Integration Test Report"
23. SV3-PMS-T2R-007,"Vogtle Unit 3 API000 Protection and Safety Monitoring System -

Reactor Trip Channel Integration Test Report"

24. SV4-PMS-T2R-007,"Vogtle Unit 4 API000 Protection and Safety Monitoring System -

Reactor Trip Channel Integration Test Report"

25. SV0-PMS-T2R-008,"API000 Protection and Safety Monitoring System System-Level Engineered Safety Features Channel Integration Test Reporf
26. SV3-PMS-T2R-008,"Vogtle Unit 3 API000 Protection and Safety Monitoring System System-Level Engineered Safety Features Channel Integration Test Report"
27. SV4-PMS-T2R-008,"Vogtle Unit 4 API000 Protection and Safety Monitoring System System-Level Engineered Safety Features Channel Integration Test Report"
28. SV3-GW-GCW-300, Field Change Notice "API000 Vogtle Unit 3 PMS Initial Software Installation - Software Release 8.7.0.1"(WO SCNXXXXXX)
29. SV4-GW-GCW-XXX, Field Change Notice "API000 Vogtle Unit 4 PMS Initial Software Installation - Software Release 8.7.0.1"(WO SCNYYYYYY)
30. SNCXXXXXX
31. SNCYYYYYY
32. B-GEN-ITPCI-006,"Main Control Room & Remote Shutdown Room"
33. GIC-AP1000-HEDS-19-001, Rev. 0"Regression Testing Analysis for Vogtle Unit 3 Protection and Safety Monitoring System (PMS) Baseline 8.2 to 8.4 Hardware Modifications Performed at Site"
34. GIC-AP1000-HEDS- YY-XXX, Rev.0"Regression Testing Analysis for Vogtle Unit 4 Protection and Safety Monitoring System (PMS) Baseline X.X to Y.Y Hardware Modifications Performed at Site"(YY-XXX is the Year-Letter #)
35. SV0-PMS-T2R-050,"API000 Protection and Safety Monitoring System Channel Integration Test Integrated System Validation Test Report" 36.2.5.04.02.i-U3-CP-Rev 0, ITAAC Completion Package 37.2.5.04.02.i-U4-CP-Rev 0, ITAAC Completion Package
38. NEI 08-01,"Industry Guideline for the ITAAC Closure Process Under 10 CFR Part 52"

U.S. Nuclear Regulatory Commission ND-19-1272 Enclosure Page 8 of 12 Attachment A*

Minimum inventory of Controis, Dispiays, and Alerts at the RSW Description* Control* Display* Alert<^)*

Neutron Flux -

Yes Yes Neutron Flux Doubling -

No Yes Startup Rate -

Yes Yes Reactor Coolant System (RCS) Pressure -

Yes Yes Wide-range Hot Leg Temperature -

Yes No Wide-range Cold Leg Temperature -

Yes Yes RCS Cooldown Rate Compared to the Limit Based on RCS Pressure -

Yes Yes Wide-range Cold Leg Temperature Compared to the Limit Based on RCS -

Yes Yes Pressure Change of RCS Temperature by more than 5°F in the last 10 minutes -

No Yes Containment Water Level -

Yes Yes Containment Pressure -

Yes Yes Pressurizer Water Level -

Yes Yes Pressurizer Water Level Trend -

Yes No Pressurizer Reference Leg Temperature -

Yes No Reactor Vessel-Hot Leg Water Level -

Yes Yes Pressurizer Pressure -

Yes No Core Exit Temperature -

Yes Yes RCS Subcooling -

Yes Yes RCS Cold Overpressure Limit -

Yes Yes In-containment Refueling Water Storage Tank (IRWST)Water Level -

Yes Yes Passive Residual Heat Removal(PRHR) Flow -

Yes Yes PRHR HX Outlet Temperature -

Yes Yes PRHR HX Inlet Isolation and Control Valve Status -

Yes Yes Passive Containment Cooling System (PCS) Storage Tank Water Level -

Yes No PCS Cooling Flow -

Yes No IRWST to Normal Residual Heat Removal System (RNS) Suction Valve -

Yes Yes Status Remotely Operated Containment Isolation Valve Status -

Yes No Containment Area High-range Radiation Level -

Yes Yes Containment Pressure (Extended Range) -

Yes No Core Makeup Tank(CMT) Level -

Yes No Manual Reactor Trip (also initiates turbine trip) Yes - -

Manual Safeguards Actuation Yes - -

Manual CMT Actuation Yes - -

Manual Automatic Depressurization System (ADS)Stages 1, 2, and 3 Yes - -

Actuation Manual ADS Stage 4 Actuation Yes - -

Manual PRHR Actuation Yes - -

Manual Containment Cooling Actuation Yes - -

Manual IRWST Injection Actuation Yes - -

U.S. Nuclear Regulatory Commission ND-19-1272 Enclosure Page 9 of 12 Attachment A*

Minimum Inventory of Controls, Displays, and Alerts at the RSW Description* Control* Display* Alert'^)*

Manual Containment Recirculation Actuation Yes - -

Manual Containment Isolation Yes - -

Manual Main Steam Line Isolation Yes - -

Manual Feedwater Isolation Yes - -

Manual Containment Hydrogen Igniter (Nonsafety-related)'^* Yes - -

Manual Containment Vacuum Relief Yes - -

Note: Dash (-) indicates not applicable.

1. These parameters are used to generate visual alerts that identify challenges to the critical safety functions. For the RSW,the visual alerts are embedded in the nonsafety-related displays as visual signals.
2. Containment hydrogen igniter control is provided as a "soft" control.

Excerpt from COL Appendix C Table 2.5.4-1

U.S. Nuclear Regulatory Commission ND-19-1272 Enclosure Page 10 of 12 Attachment B*

Minimum inventory of Alerts at the RSW Description* APP/SV3/SV4- DOS Test Document if Test Table PMS-T2R-010 required Test Case(s)

Neutron Flux TPS04-01 APP-DDS-T1R-001/BAP DDS-AP-BAP QT2 Neutron Flux Doubling TPS04-27 APP-DDS-T1R-001/BAP DDS-AP-BAP QT3 Startup Rate TPS04-01 APP-DDS-T1R-001/BAP/SPD DDS-AP-BAP QT Reactor Coolant System (RCS) TPS04-04 APP-DDS-T1R-001/BAP/SSF DDS-AP-SSF_QT Pressure Wide-range Cold Leg Temperature TPS04-09 APP-DDS-T1R-001/BAP/SPD DDS-AP-SPD QT RCS Cooldown Rate Compared to TPS04-34 Generated in the QDPS N/A the Limit Based on RCS Pressure Wide-range Cold Leg Temperature TPS04-34 Generated in the QDPS N/A Compared to the Limit Based on RCS Pressure Change of RCS Temperature by TPS04-27 Generated in the QDPS N/A more than 5°F in the last 10 minutes Containment Water Level TPS04-26 APP-DDS-T1R-001/BAP DDS-AP-BAP QT Containment Pressure TPS04-11 APP-PCS-T2R-100 PCS PT005 VLA MDFO Pressurizer Water Level TPS04-12 APP-DDS-T1R-001/SPD DDS-AP-SPD_QT Reactor Vessel-Hot Leg Water TPS04-19 Generated in the QDPS N/A Level Core Exit Temperature TPS04-02 APP-DDS-T1R-001/BAP DDS-AP-SPD_QT TPS04-03.1 TPS04-03.2 TPS04-03.3 TPS04-03.4 RCS Subcooling TPS04-04 APP-DDS-T1R-001/BAP/SPD DDS-AP-SPD_QT RCS Cold Overpressure Limit TPS04-34 Generated in the QDPS N/A In-containment Refueling Water TPS04-16 Generated in the QDPS N/A Storage Tank(IRWST) Water Level Passive Residual Heat Removal TPS04-17 APP-DDS-T1R-001/BAP/SPD DDS-AP-SPD_QT (PRHR) Flow PRHR HX Outlet Temperature TPS04-17 Generated in the QDPS N/A PRHR HX Inlet Isolation and Control TPS04-23 Generated in the QDPS N/A Valve Status IRWST to Normal Residual Heat TPS04-19 Generated in the QDPS N/A Removal System (RNS)Suction Valve Status Containment Area High-range TPS04-18 Generated in the QDPS N/A Radiation Level Excerpt from COL Appendix C Table 2.5.4-1

U.S. Nuclear Regulatory Commission ND-19-1272 Enclosure Page 11 of 12 Attachment 0*

Minimum Inventory of Controls at the RSW Description* Control Tag Test Report Test Case Number Manual Reactor Trip (also initiates DDS-HS325 SV3/SV4-PMS-T2R-007 TPS01A-22.1 turbine trip) TPS01B-22.1 TPS01C-22.1 TPS01D-22.1 Manual Safeguards Actuation DDS-HS318 SV3/SV4-PMS-T2R-007 TPS01A-19.1 TPS01B-19.1 TPS01C-19.1 TPS01D-19.1 Manual GMT Actuation DDS-HS315 SV3/SV4-PMS-T2R-008 TPS02A-04.1 TPS02B-04.1 TPS02C-04.1 TPS02D-04.1 Manual Automatic Depressurizatlon DDS-HS301/ SV3/SV4-PMS-T2R-008 TPS02A-05.1 System (ADS) Stages 1, 2, and 3 DDS-HS302 TPS02B-05.1 Actuation TPS02C-05.1 TPS02D-05.1 Manual ADS Stage 4 Actuation DDS-HS303/ SV3/SV4-PMS-T2R-008 TPS02A-05.4 DDS-HS304 TPS02B-05.4 TPS02C-05.4 TPS02D-05.4 Manual PRHR Actuation DDS-HS323 SV3/SV4-PMS-T2R-008 TPS02A-08.1 TPS02B-08.1 TPS02D-08.1 Manual Containment Cooling DDS-HS317 SV3/SV4-PMS-T2R-008 TPS02A-13 Actuation TPS02B-13 TPS02C-13 Manual IRWST Injection Actuation DDS-HS307/ SV3/SV4-PMS-T2R-008 TPS02A-03 DDS-HS308 TPS02B-03 TPS02C-03 TPS02D-03 Manual Containment Recirculation DDS-HS305/ SV3/SV4-PMS-T2R-008 TPS02A-10.1 Actuation DDS-HS306 TPS02B-10.1 TPS02C-10.1 TPS02D-10.1 Manual Containment Isolation DDS-HS327 SV3/SV4-PMS-T2R-008 TPS02A-02 TPS02B-02 TPS02C-02 TPS02D-02 Manual Main Steam Line Isolation DDS-HS321 SV3/SV4-PMS-T2R-008 TPS02B-11.1 TPS02D-11.1

U.S. Nuclear Regulatory Commission ND-19-1272 Enclosure Page 12 of 12 Attachment C*

Minimum Inventory of Controls at the RSW Description* Control Tag Test Report Test Case Number Manual Feedwater Isolation DDS-HS331 SV3/SV4-PMS-T2R-008 TPS02B-07.1 TPS02D-07.1 Manual Containment Hydrogen Igniter Soft Control 3-DDS-ITPP-520 N/A (Nonsafety-related)'^'

Manual Containment Vacuum Relief DDS-HS344 SV3/SV4-PMS-T2R-008 TPS02A-27 TPS02C-27 Excerpt from COL Appendix C Table 2.5.4-1

Retrieved from "https://kanterella.com/w/index.php?title=ND-19-1272,_Unit_4_-_Notice_of_Uncompleted_ITAAC_225-days_Prior_to_Initial_Fuel_Load_Item_2.5.04.02.1_(Index_Number_557)&oldid=2522762"