ND-19-1311, Unit 4 - Resubmittal Notice of Uncompleted ITAAC 225-days Prior to Initial Fuel Load Item 2.5.02.10 (Index Number 549)

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Unit 4 - Resubmittal Notice of Uncompleted ITAAC 225-days Prior to Initial Fuel Load Item 2.5.02.10 (Index Number 549)
ML19301D014
Person / Time
Site: Vogtle  Southern Nuclear icon.png
Issue date: 10/28/2019
From: Yox M
Southern Nuclear Operating Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
ND-19-1311
Download: ML19301D014 (14)
v • d • e


Text

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Southern Nuclear OCT 2 8 2019 Docket Nos.: 52-025 52-026 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555-0001 Michael J. Vox Regulatory Affairs Director Vogtle 3 & 4 7825 River Road Waynesboro, GA 30830 706-848-6459 tel ND-19-1311 10 CFR 52.99(c)(3)

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 October 16, 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-19-0137

[ML19073A263] dated March 14, 2019. This resubmittal supersedes ND-19-0137 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. Vox 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-19-1311 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. 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 Qqlethorpe Power Corporation Mr. R. B. Brinkman Mr. E. Rasmussen

U.S. Nuclear Regulatory Commission ND-19-1311 Page 3 of 3 Municipal Electric Authority of Georgia Mr. J. E. Fuller Mr. 8. M. Jackson Daiton Utilities Mr. T. Bundros Westinqhouse Electric Company. LLC Dr. L. Oriani (w/o enclosures)

Mr. D. C. 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., 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-19-1311 Enclosure Page 1 of 11 Southern Nuclear Operating Company ND-19-1311 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-19-1311 Enclosure Page 2 of 11 ITAAC Statement Deslon Commitment

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

InsDectlons/Tests/Anaivses 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 API 000 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-19-1311 Enclosure Page 3 of 11 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 7),

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, "AP1000 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 4 ITAAC 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-19-1311 Enclosure Page 4 of 11 References ^available for NRG inspection^

1. 2.5.02.10-U3-SumRep-Rev X "Unit 3 Protection and Safety Monitoring System Setpoint Methodology Summary Report"
2. 2.5.02.10-U4-SumRep-Rev X "Unit 4 Protection and Safety Monitoring System Setpoint Methodology Summary Report"
3. WCAP-16361-P, February 2011 "Westinghouse Setpoint Methodology 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-19-1311 Enclosure Page 5 of 11 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.

Automatic Reactor Trip Bistable Trip Logic Test Cases Reactor Trip Bistable Description Div A Test Case SV3/SV4 Div B Test Case SV3/SV4 Div C Test Case SV3/SV4 Div D Test Case SV3/SV4 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 T rip 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

U.S. Nuclear Regulatory Commission ND-19-1311 Enclosure Page 6 of 11 Automatic Reactor Trip Bistable Trip Logic Test Cases Reactor Trip Bistabie Description Div A Test Case SV3/SV4 Div B Test Case SV3/SV4 Div C Test Case SV3/SV4 Div D Test Case SV3/SV4 Reactor Coolant Pump High-2 Bearing Water Temperature Trip RCP2A TPS01A-12.3 TPS01B-12.3 TPS01C-12.3 TPS01D-12.3 Reactor Coolant Pump Hlgh-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 TPS01 B-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

U.S. Nuclear Regulatory Commission ND-19-1311 Enclosure Page 7 of 11 Automatic Reactor Trip Bistable Trip Logic Test Cases Reactor Trip Bistable Description Div A Test Case SV3/SV4 Div B Test Case SV3/SV4 Div C Test Case SV3/SV4 Div D Test Case SV3/SV4 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 Hjgh-3 Steam Generator Water Level Trip SGI TPS01A-17.1 TPS01B-17.1 TPS01C-17.1 TPS01D-17.1 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-19-1311 Enclosure Page 8 of 11 The following test reports document the logic for engineered safety feature trip bistables.

APP-PMS-T2R-007 "AP1000 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.

ESF Bistable Test Cases Instrument Bistable Division A Division B Division C Division D PZR Pressure Low TPS02A-01.1 TPS02B-01.1 TPS02C-01.1 TPS02D-01.1 PZR Water Level High-3 TPS02A-08.2 TPS02B-08.2 TPS02C-08.2 TPS02D-08.2 RCP1A Bearing Water High TPS02A-06 TPS02B-06 TPS02C-06 TPS02D-06 RCP 1B Bearing Water High TPS01A-12.2 TPS01B-12.2 TPS01C-12.2 TPS01D-12.2 RCP 2A Bearing Water High TPS01A-12.3 TPS01B-12.3 TPS01C-12.3 TPS01D-12.3 RCP 2B Bearing Water High TPS01A-12.4 TPS01B-12.4 TPS01C-12.4 TPS01D-12.4 Steam Generator 1 Level High-3 TPS02A-28 TPS02B-07.1 TPS02C-09.2 TPS02D-07.1 Steam Generator 2 Level High-3 TPS02A-09.2 TPS02B-09.2 TPS02C-09.2 TPS02D-09.2

U.S. Nuclear Regulatory Commission ND-19-1311 Enclosure Page 9 of 11 ESF Bistable Test Cases Instrument Bistable Division A Division B Division 0 Division D Steam Generator 1 Steamline Pressure Low-2 TPS02A-01.4 TPS02B-01.4 TPS02C-01.4 TPS02D-01.4 Steam Generator 2

Steamline Pressure Low-2 TPS02A-01.5 TPS02B-01.5 TPS02C-01.5 TPS02D-01.5 Steam Generator 1 Steamline pressure rate High Negative Rate TPS02A-11 TPS02B-11.1 TPS02C-11 TPS02D-11.2 Steam Generator 2

Steamline pressure rate High Negative Rate TPS02A-11 TPS02B-11.3 TPS02C-11 TPS02D-11.3 RCS Tcold Low-2 TPS02A-01.6 TPS02B-01.6 TPS02C-01.6 TPS02D-01.6 ROS Thot High TPS02A-04.3 TPS02B-04.3 TPS02C-04.3 TPS02D-04.3 Tavg Low-1 TPS02A-07.2 TPS02A-07.2 TPS02A-07.2 TPS02D-07.3 Tavg Low-2 TPS02A-07.2 TPS02A-07.2 TPS02A-07.2 TPS02D-07.3 Containment Pressure Low TPS02A-27 TPS02B-11.4 TPS02C-27 TPS02D-11.4 Containment Pressure Low-2 TPS02A-27 TPS02B-11.4 TPS02C-27 TPS02D-11.4 Containment Pressure High-2 TPS02A-01.7 TPS02B-01.7 TPS02C-01.7 TPS02D-01.7 RCS Hot Leg 1 Level 160A Low-2 N/A N/A TPS02C-23 N/A

U.S. Nuclear Regulatory Commission ND-19-1311 Enclosure Page 10 of 11 ESF Bistable Test Cases Instrument Bistable Division A Division B Division C Division D RGB Hot Leg 1 Level Low-4 N/A TPB02B-23 N/A N/A RGB Hot Leg 2 Level 1608 Low-2 N/A N/A TPB02G-05.5 N/A RGB Hot Leg 2 Level Low-4 N/A TPB02B-05.5 N/A N/A GMT A Level Upper Low-3 TPB02A-05.3 TPB02B-05.3 TPB02G-05.3 TPB02D-05.3 GMT B Level Upper Low-3 TPB02A-05.3 TPB02B-05.3 TPB02G-05.3 TPB02D-05.3 GMT A Level Lower Low-6 TPB02A-05.6 TPB02B-05.6 TPB02G-05.6 TPB02D-05.6 GMT B Level Lower Low-6 TPB02A-05.6 TPB02B-05.6 TPB02G-05.6 TPB02D-05.6 Spent Fuel Pool Level Low-2 TPB02A-22 TPB02B-22 TPB02G-22 N/A RGB Wide Range Pressure Low TPB02A-05.4 TPB02B-05.4 TPB02G-05.4 TPB02D-05.4 Startup Feedwater Flow BG 1 Low-2 N/A TPB02B-08.3 N/A TPB02D-08.3 Startup Feedwater Flow BG2 Low-2 N/A TPB02B-08.3 N/A TPB02D-08.3 BR Neutron Flux Doubling TPB02A-15.1 TPB02B-15 TPB02G-15.1 TPB02D-15.1

U.S. Nuclear Regulatory Commission ND-19-1311 Enclosure Page 11 of 11 ESF Bistable Test Cases Instrument Bistable Division A Division B Division 0 Division D MGR Diff Pressure N/A TPS02B-18.4 TPS02G-18.4 N/A MGR Particulate High-2 N/A TPS02B-18.2 TPS02G-18.2 N/A MGR Iodine High-2 N/A TPS02B-18.2 TPS02G-18.2 N/A Containment Radiation High-1 TPS02A-20 TPS02B-20 TPS02G-20 TPS02D-20 Containment Radiation High-2 TPS02A-16.2 TPS02B-21.2 TPS02G-20 TPS02D-16.2 IRWSTWR Level Low N/A N/A TPS02G-22.2 TPS02D-22.2 IRWST Lower MR Level Low-3 TPS02A-10.2 TPS02B-10.2 TPS02G-10.2 TPS02D-10.2

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