ND-19-1033, Unit 4 - Notice of Uncompleted ITAAC 225-days Prior to Initial Fuel Load Item 2.5.01.02a (Index Number 506)

Unit 4 - Notice of Uncompleted ITAAC 225-days Prior to Initial Fuel Load Item 2.5.01.02a (Index Number 506)

ML19259A307
Person / Time
Site:	Vogtle
Issue date:	09/16/2019
From:	Yox M Southern Nuclear Operating Co
To:	Document Control Desk, Office of New Reactors
References
ND-19-1033
Download: ML19259A307 (13)

Text

Michael J. Vox 7825 River Road Sk Southern Nuclear Regulatory Affairs Director Waynesboro, GA 30830 Vogtle 3 & 4 706-848-6459 tel SEP 1 6 2019 Docket Nos.: 52-025 52-026 ND-19-1033 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 2.5.01.02a (Index Number 5061 Ladies and Gentlemen:

Pursuant to 10 CFR 52.99(c)(3), Southern Nuclear Operating Company hereby notifies the NRC that as of September 6, 2019, Vogtle Electric Generating Plant(VEGP) Unit 3 and Unit 4 Uncompleted Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) Item 2.5.01.02a

[Index Number 506] 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.

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 2.5.01.02a [Index Number 506]

MJY/TMM/sfr

U.S. Nuclear Regulatory Commission ND-19-1033 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. J. Vox Mr. A. S. Parton Ms. K. A. Roberts Mr. T. G. Petrak Mr. C. T. Defnall Mr. C. E. Morrow Mr. J. L. Hughes Ms. K. M. Stacy 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. 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 Qqiethorpe Power Corporation Mr. R. B. Brinkman Mr. E. Rasmussen Municipal Electric Authorltv of Georgia Mr. J. E. Fuller Mr. S. M. Jackson

U.S. Nuclear Regulatory Commission ND-19-1033 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. Hosier, Bechtel Power Corporation Ms. L. Matis, Tetra Tech NUS, Inc.

Dr. W. R. Jacobs, Jr., Ph.D., GDS 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, Batch Bingham

U.S. Nuclear Regulatory Commission ND-19-1033 Enclosure Page 1 of 10 Southern Nuclear Operating Company ND-19-1033 Enclosure Vogtle Electric Generating Plant(VEGP) Unit 3 and Unit 4 Completion Plan for Uncompleted ITAAC 2.5.01.02a [Index Number 506]

U.S. Nuclear Regulatory Commission ND-19-1033 Enclosure Page 2 of 10 ITAAC Statement Design Commitment 2.a) The DAS provides an automatic reactor trip on low wide-range steam generator water level, or on low pressurizer water level, or on high hot leg temperature, separate from the RMS.

2.b) The DAS provides automatic actuation of selected functions, as identified in Table 2.5.1-1, separate from the RMS.

2.c) The DAS provides manual initiation of reactor trip, and selected functions, as identified in Table 2.5.1-2, separate from the RMS. These manual initiation functions are implemented in a manner that bypasses the controi room multiplexers, if any; the RMS cabinets; and the signal processing equipment of the DAS.

2.d) The DAS provides MCR displays of selected plant parameters, as identified in Table 2.5.1-3, separate from the RMS.

3.f) The DAS is powered by non-Class 1E uninterruptible power supplies that are independent and separate from the power supplies which power the RMS.

3.g) The DAS signal processing cabinets are provided with the capability for channel testing without actuating the controlled components.

I nspections/Tests/Analvses Electrical power to the RMS equipment will be disconnected and an operational test of the as-built DAS will be performed using real or simulated test signals.

Electrical power to the RMS equipment will be disconnected and an operational test of the as-built DAS will be performed using real or simulated test signals.

Electrical power to the control room multiplexers, if any, and RMS equipment will be disconnected and the outputs from the DAS signal processing equipment will be disabled. While in this configuration, an operational test of the as-built system will be performed using the DAS manual actuation controls.

Electrical power to the RMS equipment will be disconnected and inspection will be performed for retrievability of the selected plant parameters in the MCR.

Electrical power to the RMS equipment will be disconnected. While in this configuration, a test will be performed by providing simulated test signals in the non-Class 1E uninterruptible power supplies.

Channel tests will be performed on the as-built system.

U.S. Nuclear Regulatory Commission ND-19-1033 Enclosure Page 3 of 10 Acceptance Criteria The generator field control relays (contained in the control cabinets for the rod drive motor-generator sets) open after the test signal reaches the specified limit.

Appropriate DAS output signals are generated after the test signal reaches the specified limit.

i) The generator field control relays (contained in the control cabinets for the rod drive motor-generator sets) open after reactor and turbine trip manual initiation controls are actuated.

ii) DAS output signals are generated for the selected functions, as identified in Table 2.5.1-2, after manual initiation controls are actuated.

The selected plant parameters can be retrieved in the MCR.

A simulated test signal exists at the DAS equipment when the assigned non-Class 1E uninterruptible power supply is provided the test signal.

The capability exists for testing individual DAS channels without propagating an actuation signal to a DAS controlled component.

ITAAC Completion Description Multiple inspections and tests are performed to verify the Diverse Actuation System (DAS) initiates reactor trip, actuates selected functions, and provides plant information to the operator.

The inspections and tests for DAS verify that:

• The DAS provides an automatic reactor trip separate from the Protection and Safety Monitoring System (RMS)on each of the following signals:

o low wide-range steam generator water level o low pressurizer water level o high hot leg temperature

• The DAS provides automatic actuation of selected functions, as identified in Combined License (COL)Appendix C Table 2.5.1-1 (Attachment A), separate from the RMS.

• The DAS provides manual initiation of reactor trip (the generator field control relays open after a reactor and turbine trip manual initiation). The manual initiation function bypasses the RMS cabinets and the signal processing equipment of the DAS.

• The DAS provides manual initiation of selected functions, as identified in Table 2.5.1-2, (Attachment B)other than manual reactor and turbine trip. The manual initiation function bypasses the RMS cabinets and the signal processing equipment of the DAS.

• The DAS provides MCR displays of selected plant parameters, as identified in Table 2.5.1-3, (Attachment C)separate from the RMS.

• The DAS is powered by non-Class 1E uninterruptible power supplies that are independent and separate from the power supplies which power the RMS.

• The as-built DAS signal processing cabinets are provided with the capability for channel testing without actuating the controlled components.

U.S. Nuclear Regulatory Commission ND-19-1033 Enclosure Page 4 of 10 The Generator field control relays (contained in the control cabinets for the rod drive motor-generator sets) open after the test sional reaches the specified limit.

Testing is performed in accordance with the Unit 3 and Unit 4 preoperational test procedures 3-DAS-ITPP-501 and 4-DAS-ITPP-501 (References 1 and 2, respectively) to confirm that with electrical power to the Protection and Safety Monitoring System (PMS)equipment removed, an operational test of the as-built DAS using simulated test signals, provides an automatic reactor trip separate from the PMS for each of the following signals:

• low wide-range steam generator water level

• low pressurizer water level

• high hot leg temperature Initially, PMS is de-energized and the generator field control relays for the control rod motor-generator(MG)sets are verified to be closed through local verification. The DAS instruments have simulated signals generated for each of the following parameters:

• wide-range steam generator level

• pressurizer level

• hot leg temperature When the required setpoints and logic for instrumentation actuation have been met, the generator field control relays of the control rod MG sets are locally verified to trip open. This testing verifies that exceeding the actuation setpoint will generate a reactor trip for each of these distinct parameters. The DAS Processor Cabinet 1 is utilized to confirm reactor trip status.

The completed test results (References 1 and 2)confirm that the generator field control relays (contained in the control cabinets for the rod drive MG sets) open after the test signal reaches the specified limit.

Appropriate DAS output signals are generated after the test signal reaches the specified limit.

Testing is performed in accordance with References 1 and 2 to confirm that with electrical power to the PMS equipment removed, the DAS provides automatic actuation of selected functions, as identified in Combined License (COL) Appendix C Table 2.5.1-1 (Attachment A),

separate from the PMS.

Initially, PMS is de-energized and test instruments are connected to the DAS instruments to provide simulated signals. Initial conditions of components involved for the following selected functions are verified prior to test initiation.

• Initially, the Reactor Trip relay and Turbine Trip relay are verified in the non-actuated condition. Reactor and Turbine Trip are simulated on each of:

o Low Wide-range Steam Generator Water Level o Low Pressurizer Water Level o High Hot Leg Temperature Once the simulated test signal reaches the specified limit, actuation of the automatic Reactor Trip relay and Turbine Trip relay are verified locally.

U.S. Nuclear Regulatory Commission ND-19-1033 Enclosure Page 5 of 10 Initially, solenoid actuation valves for the Passive Residual Heat Removal(PRHR) Heat Exchanger(HX) Outlet valves and the In-containment Refueling Water Storage Tank (IRWST) Gutter Isolation valves are verified in the non-actuated condition. PRHR Actuation and IRWST Gutter Isolation are simulated on each of:

o Low Wide-range Steam Generator Water Level o High Hot Leg Temperature Once the simulated test signal reaches the specified limit, actuation of the solenoid actuation valves for the PRHR HX Outlet valves and the IRWST Gutter Isolation valves are verified locally.

Initially, the solenoid actuation valve for the Core Makeup Tank(CMT)outlet valves and the Reactor Coolant Pumps(RCP) Non-1 E beakers are verified in the non-actuated condition.

CMT outlet valves actuation and Trip of RCP breakers are simulated on each of:

o Low Wide-Range Steam Generator Water Level o Low Pressurizer Water Level Once the simulated test signal reaches the specified limit, actuation of the solenoid actuation valve for the CMT outlet valves and trip of the RCP switchgear Non-1 E circuit breakers are verified locally.

Initially, the PCS outlet motor-operated valve is verified closed and the solenoid actuation valves for the Passive Containment Cooling System (PCS) outlet valves and selected Containment Isolation valves identified in Combined License(COL) Appendix C Table 3.7-1 (Attachment D) are verified in the non-actuated condition. Isolation of selected Containment Penetrations and Initiation of PCS are simulated on High Containment Temperature. Once the simulated test signal reaches the specified limit, the PCS outlet motor-operated valve is verified open and actuation of the solenoid actuation valves for PCS outlet valves and selected Containment Isolation valves identified in Attachment D are verified locally.

The completed test results (References 1 and 2)confirm that the appropriate DAS output signals are generated after the test signal reaches the specified limit.

i) The generator field control relavs (contained in the control cabinets for the rod drive motor-generator sets) open after reactor and turbine trip manual initiation controls are actuated.

Testing is performed in accordance with References 1 and 2 to confirm that the DAS provides manual initiation of reactor trip while the PMS cabinets and the signal processing equipment of the DAS are bypassed.

The preoperational test de-energizes the PMS cabinet power supplies (no multiplexer exists) and prevents DAS automatic output actuation by placing DAS in Master Test. These conditions ensure that the actuations tested are caused by the DAS manual functions. Testing is performed by initially verifying the generator field control relays are closed locally and then manually actuating the DAS manual functions to trip the reactor and turbine. During the manual reactor and turbine trip, the generator field control relays are verified to open locally by inspection at the motor-generator set control cabinets.

U.S. Nuclear Regulatory Commission ND-19-1033 Enclosure Page 6 of 10 The completed test results (References 1 and 2)confirm that the generator field control relays (contained in the control cabinets for the rod drive motor-generator sets) open after reactor and turbine trip manual initiation controls are actuated.

ID DAS output sionals are generated for the selected functions, as identified in Table 2.5.1-2.

after manual initiation controls are actuated.

Testing is performed in accordance with References 1 and 2 to confirm that the DAS provides manual initiation of selected DAS manual functions identified in Combined License(COL)

Appendix C Table 2.5.1-2(Attachment B) while the RMS cabinets and the signal processing equipment of the DAS are bypassed.

The preoperational test de-energizes the RMS cabinet power supplies (no multiplexer exists) and prevents DAS automatic output actuation by placing DAS in Master Test. These conditions ensure that the actuations tested are caused by the DAS manual functions. Testing is performed by manually actuating selected DAS manual functions identified in Attachment B.

During the testing of the manual functions, the DAS output signal is verified for each selected function by relay contact inspection in the DAS cabinets.

The completed test results (References 1 and 2)confirm that the DAS output signals are generated for the selected functions, as identified in Table 2.5.1-2 after manual initiation controls are actuated.

The selected plant parameters can be retrieved in the MGR.

Testing is performed in accordance with References 1 and 2 to confirm that the DAS provides MCR displays of selected plant parameters, as identified in Combined License(COL)Appendix C Table 2.5.1-3(Attachment C), separate from the RMS.

The preoperational test includes inspections (checks) that with electrical power to the RMS equipment disconnected, analog signals provide an indication at the DAS panel in the MCR for the Sensors and Displays as identified in Attachment C. Testing is performed by simulating an analog signal and then inspection is performed to check that appropriate indication is retrieved in the MCR.

The completed test results (References 1 and 2)confirm that the selected plant parameters can be retrieved in the MCR.

A simulated test sional exists at the DAS eauioment when the assioned non-Class 1E uninterruptible power suppIv is provided the test sional.

Testing is performed in accordance with the Unit 3 and Unit 4 component test packages SNCXXXXXX and SNCYYYYYY (References 3 and 4, respectively) to verify that the non-Class 1E power supply to DAS is independent from the Class 1E power supplied to the RMS.

The RMS cabinet power supplies are de-energized and testing is performed by injecting a signal to the non-Class 1E DAS power supply and then checking that the test signal is present at the DAS equipment.

U.S. Nuclear Regulatory Commission ND-19-1033 Enclosure Page 7 of 10 The completed test results (References 3and 4)confirm that a simulated test signal exists at the DAS equipment when the assigned non-Class 1E uninterruptible power supply is provided the test signal.

The capabilitv exists for testing individual DAS channels without Dropaaating an actuation sional to a DAS controlled component.

Testing is performed in accordance with References 1 and 2 to confirm that the DAS signal processing cabinets are provided with the capability for channel testing without actuating the controlled components.

The preoperational test places channels in bypass and then simulates process variables exceeding setpoint and verifies no end device operation. The end device verification is performed locally by measuring contact resistance or voltage on the actuating components. The channels are returned to normal and the next process variable is tested in a similar manner.

This testing is repeated untii the RCS High Hot Leg Temperature, Low Wide-range Steam Generator Water Level, Low Pressurizer Water Level, and High Containment Temperature variables in all DAS channels have been tested.

The completed test results (References 1 and 2)confirm that the capability exists for testing individual DAS channels without propagating an actuation signal to a DAS controlled component.

References 1,2, 3 and 4 are available for NRG inspection as part of Unit 3 and Unit 4ITAAC 2.5.01.02a Completion Package (References 5 and 6).

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 finding review found there are no relevant ITAAC findings associated with this ITAAC.

References favaiiabie for NRG inspection)

1. 3-DAS-ITPP-501,"Diverse Actuation System Preoperational Test Procedure"

2. 4-DAS-ITPP-501,"Diverse Actuation System Preoperational Test Procedure"

3. SNCXXXXXX,"DAS Component Power Verification Test- ITAAC: SV3-2.5.01.02a Item 3.f"

4. SNCYYYYYY,"DAS Component Power Verification Test- ITAAC: SV4-2.5.01.02a Item 3.f"

5. 2.5.01.02a-U3-CP-Rev0, ITAAC Completion Package

6. 2.5.01.02a-U4-CP-Rev0, ITAAC Completion Package

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

U.S. Nuclear Regulatory Commission ND-19-1033 Enclosure Page 8 of 10 Attachment A COL Appendix 0 Table 2.5.1-1 Table 2.5.1-1 Functions Automatically Actuated by the DAS

1. Reactor and Turbine Trip on Low Wide-range Steam Generator Water Level or Low Pressurlzer Water Level or High Hot Leg Temperature

2. Passive Residual Heat Removal(PRHR)Actuation and in-containment Refueling Water Storage Tank (IRWST) Gutter Isolation on Low Wide-range Steam Generator Water Level or on High Hot Leg Temperature

3. Core Makeup Tank(CMT)Actuation and Trip All Reactor Coolant Pumps on Low Wide-Range Steam Generator Water Level or Low Pressurlzer Water Level

4. Isolation of Selected Containment Penetrations and Initiation of Passive Containment Cooling System (PCS)on High Containment Temperature Attachment B COL Appendix C Table 2.5.1-2 Table 2.5.1-2 Functions Manually Actuated by the DAS

1. Reactor and Turbine Trip

2. PRHR Actuation and IRWST Gutter Isolation

3. CMT Actuation and Trip All Reactor Coolant Pumps

4. First-stage Automatic Depressurization System (ADS) Valve Actuation

5. Second-stage ADS Valve Actuation

6. Third-stage ADS Valve Actuation

7. Fourth-stage ADS Valve Actuation

8. PCS Actuation

9. Isolation of Selected Containment Penetrations

10. Containment Hydrogen Igniter Actuation

11. IRWST Injection Actuation

12. Containment Recirculation Actuation

13. Actuate IRWST Drain to Containment

U.S. Nuclear Regulatory Commission ND-19-1033 Enclosure Page 9 of 10 Attachment 0 COL Appendix 0 Table 2.5.1-3 Table 2.5.1-3 DAS Sensors and Displays Equipment Name Tag Number Reactor Coolant System (RCS) Hot Leg Temperature RCS-300A RCS Hot Leg Temperature RCS-300B Steam Generator 1 Wide-range Level SGS-044 Steam Generator 1 Wide-range Level SGS-045 Steam Generator 2 Wide-range Level SGS-046 Steam Generator 2 Wide-range Level SGS-047 Pressurizer Water Level RCS-305A Pressurizer Water Level RCS-305B Containment Temperature VCS-053A Containment Temperature VCS-053B Core Exit Temperature IIS-009 Core Exit Temperature IIS-013 Core Exit Temperature IIS-030 Core Exit Temperature IIS-034 Rod Control Motor Generator Voltage PLS-001 Rod Control Motor Generator Voltage PLS-002

U.S. Nuclear Regulatory Commission ND-19-1033 Enclosure Page 10 of 10 Attachment D

^Excerpted from COL Appendix 0 Table 3.7-1 Table 3.7-1 Risk-Significant Components

• Equipment Name *Tag Number CVS-PL-V045,-V047 Containment Isolation Valves Controlled by DAS VFS-PL-V003, -V004, -V009, -V010 WLS-PL-V055,-V057