Text

Summary of April 14, 2021, Redacted - Partially Closed Observation Meeting with Entergy Operations Inc., LAR to Install Digital Upgrade in Accordance with Digital I&C Interim Staff Guidance No. 06, Revision 2 (EPID L-2020-LLA-0164) (Non-Pro
	ML21112A249
Person / Time
Site:	Waterford
Issue date:	05/07/2021
From:	Perry Buckberg; Plant Licensing Branch IV
To:	; Entergy Operations
	Buckberg P
References
	EPID L-2020-LLA-0164
	Download: ML21112A249 (43)
	v • d • e

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION May 7, 2021 LICENSEE:

Entergy Operations, Inc.

FACILITY:

Waterford Steam Electric Station, Unit 3

SUBJECT:

SUMMARY

OF APRIL 14, 2021, PARTIALLY CLOSED OBSERVATION MEETING WITH ENTERGY OPERATIONS, INC. REGARDING LICENSE AMENDMENT REQUEST TO INSTALL DIGITAL UPGRADE IN ACCORDANCE WITH DIGITAL INSTRUMENTATION AND CONTROL INTERIM STAFF GUIDANCE NO. 06, REVISION 2, LICENSING PROCESSES (EPID L-2020-LLA-0164)

On April 14, 2021, the U.S. Nuclear Regulatory Commission (NRC) staff held a virtual Observation public meeting with representatives from Entergy Operations, Inc. (the licensee) and its contract support staff. The purpose of the meeting was to discuss the licensees amendment request dated July 23, 2020, as supplemented by letters dated January 22, January 29, March 5, andMarch 19, 2021 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML20205L588, ML21024A005, ML21029A156, ML21064A535, and ML21082A393, respectively), for the Waterford Steam Electric Station, Unit 3, regarding a replacement to an existing digital core protection calculator system (CPCS).

The upgrade, if approved, would replace the existing CPCS with a Common Q-based system.

The meeting notice and agenda, dated December 22, 2020, are available in ADAMS under Accession No. ML21036A265. A list of attendees is provided in Enclosure 1.

During the meeting, the NRC staff discussed its open items list, which is a list of NRC staff questions and informal licensee responses regarding the license amendment request for the NRC staff to track and eventually disposition as requests for additional information, requests for confirmation of information, audits, or as needing no additional action. The proprietary version of the open items list, which is being withheld from public disclosure, is in Enclosure 2.

A redacted copy of the open items list is in Enclosure 3.

to this letter contains proprietary information. When separated from, this document is DECONTROLLED.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION During the public portion of the meeting, the NRC staff and licensee discussed the status of Open Item Nos. 24, 26, 33, 38, and 40. The NRC staff clarified that for Open Item No. 38, the staff will need information submitted on the docket that describes the licensees change control process for the vendor oversight plan. The licensee discussed its plans to supplement its amendment request. The licensee also clarified that, regarding statements in the amendment request regarding Institute of Electrical and Electronic Engineers (IEEE) standards (Stds), the modification will meet IEEE Std 603-1991, Criteria for Safety Systems for Nuclear Power Generating Stations, to be consistent with Interim Staff Guidance (ISG)-06, Revision 2, Licensing Process, dated December 2018 (ADAMS Accession No. ML18269A259) but that its licensing basis requires the licensee to meet IEEE Std 279-1971, Criteria for Protection Systems for Nuclear Power Generating Stations.

During the closed portion of the meeting, the licensee and its contract staff gave a presentation on a corrective action related to the CPCS design that does not affect the CPCS safety function.

The NRC staff and licensee also discussed NRCs audit of the licensees vendor oversight plan implementation of the requirements phase.

The NRC staff has determined that the open items list contains proprietary information pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 2.390, Public inspections, exemptions, requests for withholding. The proprietary information is indicated by bold text enclosed within ((double brackets)). The proprietary version of the open items list is provided as Enclosure 2. Accordingly, the NRC staff has also prepared a nonproprietary version of the open items list which is provided as Enclosure 3.

The NRC staff did not make any regulatory decisions or commitments at the meeting. No members of the public identified themselves on the teleconference. The NRC staff did not receive any Public Meeting Feedback forms.

Please direct any inquiries to me at 301-415-1383 or by e-mail to Perry.Buckberg@nrc.gov.

/RA Audrey L. Klett for/

Perry H. Buckberg, Senior Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-382

Enclosures:

1. List of Attendees

2. Open Items List (Proprietary)

3. Open Items List (Non-proprietary) cc w/o Enclosure 2: Listserv

List of Attendees

LIST OF ATTENDEES APRIL 14, 2021, VIRTUAL PUBLIC MEETING WITH ENTERGY OPERATIONS, INC., ET AL.

WATERFORD STEAM ELECTRIC STATION, UNIT 3, LICENSE AMENDMENT REQUEST TO INSTALL DIGITAL UPGRADE U.S. Nuclear Regulatory Commission Entergy Operations, Inc.

Brent Ballard, NRR1/DORL2/LPL13 Jacob Champagne Perry Buckberg, NRR/DORL/LPL44 Phil Couture Samir Darbali, NRR/DEX5/ELTB6 Remy DeVoe Jennifer Dixon-Herrity, NRR/DORL/LPL4 Ron Gaston DaBin Ki, NRR/DRO7/IOLB8 Loren Miller Audrey Klett, NRR/DORL/LPL1 Roger Rucker Shiattin Makor, RIV9/DRS10 John Schrage Richard Stattel, NRR/DEX/EICB11 Christopher Talazac Tarico Sweat, NRR/DSS12/STSB13 William Truss Dan Warner, NSIR14/DPCP15/CSB16 Tom Wengert, NRR/DORL/LPL4 Jensen Hughes, Inc.

Deanna Zhang, NRR/DRO/IQVB17 Alan Harris Jack Zhao, NRR/DEX/EICB Sargent and Lundy Members of the Public Pareez Golub None introduced Westinghouse Electric Company, LLC Warren Odess-Gillett John Wiesemann 1 Office of Nuclear Reactor Regulation 2 Division of Operating Reactor Licensing 3 Plant Licensing Branch I 4 Plant Licensing Branch IV 5 Division of Engineering and External Hazards 6 Long Term Operations and Modernization Branch 7 Division of Reactor Oversight 8 Operator Licensing and Human Factors Branch 9 Region IV 10 Division of Reactor Safety 11 Instrumentation and Controls Branch 12 Division of Safety Systems 13 Technical Specifications Branch 14 Office of Nuclear Security and Incident Response 15 Division of Physical and Cyber Security Policy 16 Cyber Security Branch 17 Quality Assurance and Vendor Inspection Branch

(Non-proprietary)

Open Items List Proprietary information pursuant to Section 2.390 of title 10 of the Code of Federal Regulations has been redacted from this document.

Redacted information is identified by blank space enclosed within ((double brackets)).

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION No.

IMS ID Topic &

(Reviewer)

LAR/LTR Section LAR/ LTR Page NRC Comment / Open Item Description Licensee Response Status Audit, RAI or RCI No.

Acronyms and abbreviations are defined on the last page of this document.

Certrec IMS Request ID Format (second column of this table)

A-Audit (Generic/Multiple Documents)

CCF-Common Cause Failure/D3 EQ-Equipment Qualification HFE - Human Factors Engineering PSAI-Plant Specific Action Items RC-Regulatory Commitments RT-Response Time SA-System Architecture SDOE-Secure Development and Operational Environment ST-Surveillance Testing/Self-Diagnostics/SR Elimination SDP-System Development Processes, including SPM PSAIs TS-Technical Specifications VOP-Vendor Oversight Plan Updated by Entergy on 10/19/20 Proprietary Documents will be uploaded to the Westinghouse Sharepoint site at the below address

((

))

1 ST-01 Self-Tests (Jack Zhao, Richard Stattel, Samir Darbali)

B.2.5 B-5 The BTP 7-17 Evaluation conclusion states that It is not possible to test self-diagnostics as part of surveillance testing because it would require creating destructive faults within the I&C system, such as Random-Access Memory (RAM) errors.

Though this is a quote out of the Vogtle LAR safety evaluation, it is a statement made by the licensee and not the NRC to address this criterion in BTP 7-17, self-test functions should be verified during periodic functional tests. The interpretation being made that the BTP criterion calls for complete functional testing of the self-diagnostic functions is incorrect. Instead, the BTP states that the licensee should confirm the execution of self-diagnostic tests during plant operation and the NRC staff believes that it is possible to do so by implementing the following necessary plant monitoring activities as already included in the Enclosure for this LAR.

The licensee (Waterford) has addressed this in the LAR as follows:

Post installation, CPCS operability will be verified using 1) the automated diagnostics credited in this LAR (i.e., as described in LTR Appendix B), 2)

Technical Requirements Manual (TRM) 3/4.3.1, "Reactor Protective Instrumentation" and associated surveillance procedures; and 3) Waterford TS 6.5.1.8, "Surveillance Frequency Control Program (SFCP). A failure of credited automated diagnostics to detect a fault will be either detected by other diagnostics in the system or by checker(s) of diagnostics. This condition will be alarmed and displayed on the main control room (MCR) operator modules (OM) and/or the main control room annunciators. Upon receipt of an alarm or abnormal conditions, the station operating procedures will require the operators to perform system checks and verify operability of the CPCS deviation / function. The procedure will direct the operator to dispatch a maintenance technician to determine the source of the alarm as needed.

(Entergy 11/3/20 Update)

The LAR Enclosure Section 2.3, Reason for the Proposed Changes, will be revised as follows:

Crediting Self-Diagnostics for TS Surveillance Requirement Elimination The Common Q design also provides additional reliability and operational margin via the self-diagnostics. These self-diagnostics are continually monitoring the health of the hardware and software.

Appendix B to the Licensing Technical Report (LTR) (Attachment 4) and the Waterford System Engineer and Operations Actions Supporting TS SR Reduction (LAR Enclosure Section 3.4) provides the justification to remove selected SRs.

Note: "and the Waterford System Engineer and Operations Actions Supporting TS SR Reduction (LAR Section 3.4) provides the justification to remove selected SRs is new inserted text.

The LAR Enclosure Section 2.4, Description of the Proposed TS Changes, for TS 3.3.1/Table 4.3-1, will be revised as follows:

For row TS 3.3.1/Table 4.3-1, the sentence "LTR Appendix B provides the detailed justification that demonstrates that the self-diagnostics meet the requirements of 10 CFR 50.36 for the CPCS..."

with "LTR Appendix B along with the Waterford System Engineer and Operations Actions Supporting TS SR Reduction (LAR Enclosure Section 3.4), provides the detailed justification that demonstrates that Closed (V)

This is a proposed change to the LAR enclosure and not to the WCAP.

Therefore I am unable to verify changes as of 2/25/2021.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION (W3F1-2020-0038 Page 18 of 27)

The NRC staff agrees with the licensees proposed actions in the LAR. So, for this LAR the licensee should cite both the credited self-diagnostic functions in Appendix B and proposed monitoring activities to justify the SR elimination in Section 2.2 of the LAR, not just the Appendix B. However, since Section 2.2 of the LAR cites Appendix B to WCAP-18464 as the sole justification for SR elimination (see Enclosure W3F1-2020-0038, Page 5 of 27) and Appendix B does not include any plant monitoring activity, it could lead to the misunderstanding that if the NRC accepts this LAR, it would also be accepting Appendix B as the only basis for the SR elimination. In addition, the LAR says on Page 18 of 27, in part, that while LTR Appendix B states that monitoring is not required in order to credit self-diagnostic features. The NRC staff does not agree with this statement to address the above criterion in BTP 7-17.

Furthermore, Appendix B says to leverage the Vogtle LAR for the SR elimination. But, the Vogtle LAR included plant monitoring activities as one of bases for the SR elimination. Therefore, the SR Elimination basis in both Section 2.2 of this LAR and Appendix B will need to include the licensees commitment to perform self-diagnostic monitoring activities and the appendix B interpretations should be revised to establish consistency with the LAR.

the self-diagnostics meet the requirements of 10 CFR 50.36 for the CPCS..."

2 ST-02 Self-Tests (Jack Zhao, Richard Stattel, Samir Darbali)

B.2.5 B-6 The bullet item on this page states the following:

((

))

(Entergy 11/3/20 Update) PROPRIETARY RESPONSE A.

((

))

B.

((

))

Closed Verified these changes are present in Revision 1 of WCAP 18484 on share point 3/11/2021.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION (a.1) The last sentence of the response states that However, Entergy understands how the wording of the statement is ambiguous in this respect.

Please explain if the LAR and LTR will be revised to address the ambiguous wording.

(a.1) Yes, the LAR Enclosure will be revised per the response to SA-01a.

Yes, the LTR Section 3.3.4, System Requirements Documentation (D.2.3.3 and D.2.3.3.1), will be revised as follows:

From: "Reference 2 is the CPCS System Requirements Document. It is the system requirements specification for the reference design for the Common Q CPCS. The reference design system requirements is based on two requirements documents that define the legacy CPCS functionality:

- Functional Design Requirements for a Core Protection Calculator (Reference 36) and

- Functional Design Requirements for a Control Element Assembly Calculator (Reference 37)

The Common Q CPCS reference design system requirements specification (Reference 2) was developed to migrate the functional requirements of References 36 and 37) to a Common Q CPCS architecture. The result was the Palo Verde CPCS implementation.

The existing Waterford CPCS is based on the same two functional design requirements documents (References 36 and 37). Therefore, the CPCS reference design is also applicable to the Waterford CPCS replacement plus additional changes to accommodate plant interface differences, requested licensee improvements, and changes in technology in the Common Q platform."

To: "Reference 2 is the CPCS System Requirements Document for the reference design for the Common Q CPCS. The reference design system requirements is based on two requirements documents that define the legacy CPCS functionality:

- Functional Design Requirements for a Core Protection Calculator (Reference 36) and

- Functional Design Requirements for a Control Element Assembly Calculator (Reference 37)

The Common Q CPCS reference design system requirements specification (Reference 2) was developed to migrate the functional requirements of References 36 and 37 to a Common Q CPCS architecture. The result was the Palo Verde CPCS implementation.

Note that Revision 7 of Reference 2 (ML032830027) was reviewed by the NRC.

The existing Waterford CPCS is based on the same two functional design requirements documents (References 36 and 37). Therefore, the CPCS reference design is also applicable to the Waterford CPCS replacement plus additional changes to accommodate plant interface differences, requested licensee improvements, and changes in technology in the Common Q platform. Reference 2 is the current

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION link section

6. Section 2.1.2.2.4.1: added CEA positions to items transmitted to CEAPD.

7. Section 2.1.2.2.4.3: Removed reference to CEAPD SysRS.

8. Section 2.2.1.4.4: Added requirement for CEA trip snapshot page with live CEA position data.

9. Section 2.2.1.4.6: added CRC value to Addressable Constants page.

10. Section 2.2.1.4.7: added CRC value to Change Addressable constants page.

11. Section 2.2.1.4.12: added missing colon for "Page 3".

12. Section 2.2.1.4.19: defined CEA inputs to be displayed as SUBGRPx on this page.

13. Section 2.2.1.4.20: Corrected spelling of capability.

14. Section 2.2.1.5.2.1.2: clarified trouble alarm occurs for loss of other display.

15. Section 2.2.1.5.2.2: Added alarm icon label to sentence.

16. Section 2.2.1.5.2.2.1 and 2, added OM and MTP CRCs do not agree to trouble list.

17. Section 2.2.2.4: changed heading text and changed requirements for AI calibration testing for CPC, CPP1, and CPP2 functional tests.

18. Section 2.2.2.4: Removed requirement to enable the Exit Functional test icons only if the associated AI calibration is complete.

This section was modified to reflect the as implemented software.

19. Section 2.2.2.4.6: added section to describe functional test interlock requirements.

20. Section 2.3: corrected CEAPD description and removed reference.

21. Section 2.3.4.1.3: added missing period to end of sentence.

22. Section 2.3.4.4.3.2: corrected description since CEAPD does not use trip buffer data.

23. Section 3.1.1.1.6.3.1: added "minimum" to description.

24. Section 3.1.1.1.9.13: removed reference to CEAPD and added reference to applicable sections.

25. Section 3.1.1.1.9.13.1: Clarified data being sent to CEAPD.

26. Section 3.1.1.1.10.3: defined the CEA position data being sent to CEAPD and usage.

27. Section 3.1.1.1.10.8: added CEA positions to CEAPD cross channel comparison information.

Appendix Changes:

1. Corrected Table of contents to remove "symbol" link after Sec.

3.2.5.6.

2. Pg 116: Added IRPC decision statement to reflect text description.

3. Pg 217: Added definition of CEAIW.

4. Pg 217, 219: Moved all variable definitions to end of section 3.2.6.1.1

5. Pg 220: Clarified that CPOS(i,1) is the CEA position of the current execution cycle.

Revision 11 Change Summary:

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION Text Main Body Changes

1. Pg 59 clarified the conditions for taking the CEAC snapshot.

2. This revision incorporated changes to the Reactor Power Cutback detection algorithm in Appendix A.

Revision 12 Change Summary:

Appendix Changes

1. Pg A224, added footnote for starting the RPC timer.

Revision 13 Change Summary:

Text Main Body Changes

1. Page 150, incorporated CAPs Commitment 07-285-W006.02 for both CEACs inoperable.

Revision 14 Change Summary:

Text Main Body Changes

1. Re-numbered Sections to match Table of Contents per CAPAL 100074239.

Appendix Changes:

1. Correct QHOT definition in Sections 3.2.4.5 & 3.2.4.16 of Appendix A per CAPS #08-315-W001.

11/16/20 Update:

Entergy did not perform a regression analysis between the 00000-ICE 30158 Revision 7 and 00000-ICE 30158 Revision 14 documents.

Entergy performed a lower level regression analysis audit of the Palo Verde CPCS software changes between the initial release of the software that was approved by the NRC and the current baseline of the Palo Verde CPCS software. This VOP audit included all software change requests for the Palo Verde CPCS software. These software changes in some cases required a revision to the 00000-ICE 30158.

This regression analysis audit is documented in an Entergy regression analysis audit report (AUD-WF3-2019-236-CA058).

There were no hardware design changes to the CPCS since NRC approval.

(c) Reference SA-01a and SA-01b 00000-ICE-30158, Rev 14, System Requirements Specification for the Common Q Core Protection Calculator System, is the basis document for WNA-DS-04517-CWTR3, System Requirements Specification for the Core Protection Calculator System. WNA-DS-04517-CWTR3 is the WF3 delta document for WF3. Requirements traceability is to WNA-DS-04517-CWTR3. When WNA-DS-04517-CWTR3, Rev 0, was reviewed and approved for owners acceptance per procedure EN-DC-149, the applicable sections of 00000-ICE-30158, Rev 14, were reviewed. Based on the regression analysis for

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION Please clarify the methods that will be used for performing the reload analysis.

(3.1) Follow-up question to OI 17(3): The first sentence of the response states that The response times calculated in WNA-CN-00572-CWTR3 for the CPCS are bounded by the current response time requirements specified in the reference design (00000-ICE-30158).

Please clarify the adequacy of the response time requirements specified in the reference design in terms of the thermal limits (i.e.,

DNBR and LHGR) calculation.

(1.1) The actual WF3 CPCS calculated response times will be used as input for the reload analysis.

(2)

The identification and justification for the CPCS delay time values in the thermal margin estimate for each applicable transient and accident listed in Table 3.2.6-1 is documented in Westinghouse document LTR-GIC-20-003, Waterford 3 CPCS Response Time Information for FSAR and Technical Specification. A 2nd document, WNA-CN-00572-CWTR3, Core Protection Calculator System Response Time Calculation provides the response time calculation for the WF3 CPCS. Both of these can be submitted to the NRC.

WNA-CN-00572-CWTR3, Core Protection Calculator System Response Time Calculation was attached to the LAR. LTR-GIC 003, Waterford 3 CPCS Response Time Information for FSAR and Technical Specification is on the Westinghouse document portal.

See OI 26 (h)

(3)

The response times calculated in WNA-CN-00572-CWTR3 for the CPCS are bounded by the current response time requirements specified in the reference design (00000-ICE-30158). The response time testing conducted during FAT and post installation testing will confirm that the system meets these response time criteria.

(3.1) It is LTR-GIC-20-003 that correlates the response time calculated in WNA-CN-00572-CWTR3 to the various CPCS trips.

LTR-GIC-20-003 describes the adequacy of the new response time requirements. After further investigation, it was determined that the revised calculated response times are not bounded by the reference design, and the WF 3 SyRS, WNA-DS-04517-CWTR3, needs to specify these new response time requirements. A Westinghouse Corrective Action Issue Report (IR-2020-11971) was issued accordingly. A new revision 5 of WNA-DS-04517-CWTR3 with the new response time requirements is now issued.

The Licensing Technical Report is not impacted by this revision because the LTR only referred to the Palo Verde response times and stating that WF3 specific response times would be calculated.

Resolution. WNA-DS-04517-CWTR3, Revision 5 will be docketed by 12/31/2020.

(4)

LTR Section 3.2.6 states, As part of the normal fuel reload process, Waterford runs the safety analysis of record with the WF3 CPCS calculated response times to validate that acceptable margin is maintained. It is the fuel reload process performed under 10 CFR 50.59 that evaluates the results of the rerun of the safety analysis prior to core reload. If the results become more limiting, the analyses results will be evaluated against the 10CFR50.59 criteria. If the 10CFR50.59 criteria requires NRC approval, then a new submittal will be generated. Based upon previous analysis impacts, it is expected

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION 10/28/2020 Update:

(d.1) The second paragraph in LAR Section 3, sub-section 2, Licensing Technical Report (LTR), refers to LTR Section 3.2.18 and the ANO-2 and PVNGS evaluations. Please explain if this paragraph will also be revised.

11/10/2020 Update:

(c.1) The response to item (c) suggests that conformance with BTP 7-19 is not required. However, LAR Section 4.1 Applicable Regulatory Requirements/Criteria lists BTP 7-19. LTR Section 3.2.18 also identifies BTP 7-19. Please clarify if the LAR and LTR will be revised to remove references to BTP 7-19?

3. In NUREG 0787 Supplement 5, Section 4.4.2 dated June 1983, the NRC indicated that the CPCS/CEACs are essentially the same as the ANO-2 Cycle-2 CPCs and since the ANO-2 CPC/CEAC were approved by the NRC staff (July 21, 1981 Memorandum), the review of the Waterford 3 CPC/CEAC concentrated on the software modifications and its implementation. Because there is no additional documentation in any of the subsequent supplements, the implicit conclusion is that the acceptability of the CPCs failing to meet design function at ANO-2 also applies to Waterford 3.

4. FSAR Section 7.2.1 describes the Reactor Protection System (RPS). FSAR Section 7.2.1.1.8 describes how the system is designed to eliminate credible multiple channel failures originating from a common cause. This section applies to all of the RPS, which includes CPCs. This section is unchanged since Revision 0 of the FSAR (circa 1985)

Discussion The CPC digital upgrade project does not alter how the diversity within the RPS is achieved, as described in the FSAR 7.2.1.1.8.

However, industry and regulatory developments over the past 35 years have provided further improvements to address reactor protection systems common cause failures. The most noteworthy is the Anticipated Transient Without Scram (ATWS) rule (i.e., 10 CFR 50.62). Implementation of the ATWS Mitigation System is described in FSAR section 7.8. The system is designed to mitigate the consequences of Anticipated Operational Occurrences (AOOs) coupled with a failure of the RPS to trip the reactor. The Diverse Reactor Trip System (DRTS) provides an independent means of sensing a high pressurizer pressure and then de-energizing the MG set output contactor coils that provide the power to the Control Element Drive Mechanisms, and subsequently trip the reactor.

The NRC provided the acceptance for the Waterford 3 ATWS mitigating systems design in the Safety Evaluation dated September 8, 1989 (ML8909180108). The NRC inspection of Compliance with the 10 CFR 50.62 (ATWS Rule) is documented in Inspection Report 89-39 dated December 5, 1989 (ML8912110063). There were no violations or deviations noted in the report concerning the implementation of the ATWS system at Waterford 3.

The ATWS system at Waterford 3 is a more rigorous backup to a postulated common cause failure of the CPCS relative to reliance on the NRC's evaluation of a similar system at ANO-2. The ATWS system at Waterford 3 is plant specific, incorporated in the design basis, and is continually evaluated as the overall plant design evolves. Both the extended power uprate and replacement steam generator projects resulted in evaluations of the ATWS mitigating systems to ensure the major plant changes did not negatively impact the ATWS systems (SGT-LTR-TDA-09-20, Evaluation of Anticipated Transients Without Scram (ATWS) Rule Compliance for Waterford 3 with RSGs and a Full Core of NGF Fuel

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION Conclusion The function of the CPCS to calculate and provide LPD and DNBR trip signals to the RPS to prevent fuel damage during AOOs is unchanged. The features of the RPS which provide analog trips as a backup to failure of the CPCS to cause trips is unchanged as a result of the CPCS digital upgrade project. However, since initial startup of Waterford 3, the implementation of the ATWS rule provides complete protection of the fuel for AOOs that should result in the RPS tripping the reactor. The ATWS systems are independent from the RPS, have been inspected by the NRC and continually evaluated for impacts as the plant design evolves.

(c) Entergy Update 11/3/20 The technical and licensing basis for the existing CPCS are the following sections of the WF3 UFSAR:

Chapter 7.2 (Since the CPCS is an integral part of the Reactor Protective System, the CPCS basis is described throughout the section. Note Section 7.2.1.1.8 establishes the licensing basis for diversity against a predictable common failure mode)

Appendix 4.3A.5.2 & 4.3A.5.3 To summarize what is described in UFSAR Chapter 7.2.1.1.2.5, the basic architecture for the CPCS is a four channel computer system (i.e., Core Protection Calculator [CPC]) that calculates these parameters and initiates reactor trip signals to the analog reactor protection system. This basic architecture also includes two computers (CEAC 1 and CEAC 2) that calculate a CEA position penalty factor used by all four CPC computers.

The WF3 I&C architecture mirrors the echelons of defense described in NUREG 6303, Method for Performing Diversity and Defense-in-Depth Analyses of Reactor Protection Systems, to protect the health and safety of the public. The first echelon is the non-safety control systems which controls the nuclear plant process within its technical specification limits. The second echelon of defense is the plant protection system to automatically shutdown reactivity and provide heat removal in case of an accident. And the third echelon of defense is the manual indications and controls to allow operators to manually control the plant. In addition to these echelons of defense, there is an ATWS system to protect the health and safety of the public should an anticipated transient occur without a scram.

This plant modification only impacts the second echelon of defense, the plant protection system, and in particular the reactor protection system. The WF3 operating license allows for a computerized digital system to calculate and initiate a reactor trip on low DNBR and High LPD in support of the WF3 accident analysis, as described in the WF3 UFSAR Chapter 7.2.1.1.2.5. As summarized above and

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION described in detail in WF3 UFSAR Chapter 7.2.1.1.2.5, the basic architecture for this aspect of the reactor protection system is a four channel computer system (i.e., CPC) that calculates these parameters and initiates reactor trip signals to the analog reactor protection system. This basic architecture also includes two computers (CEAC 1 and CEAC 2) that calculate a CEA position penalty factor used by all four CPC computers. This plant modification does not invalidate the diversity claims in UFSAR Section 7.2.1.1.8.

The Common Q CPCS upgrade preserves this basic architecture but improves upon it by multiplying the number of CEAC computers from two to eight (2 in each channel) to improve system reliability. There are still four independent CPC channels calculating DNBR and LPD as in the existing architecture. Therefore the D3 strategy for WF3 is not impacted by this plant modification.

There are no plans at this time to replace any of the non-safety plant control systems with the Common Q platform which could potentially impact the WF3 D3 strategy. Should the PPS be replaced with a digital system, then compliance to BTP 7-19 would be required.

(d) Attachment 4 of the Enclosure to Entergy letter number W3F1-2020-0038 (WCAP-18484-P, Licensing Technical Report for the Waterford Steam Electric Station Unit 3 Common Q Core Protection Calculator System"), Section 3.2.18 will be revised to delete reference to the ANO-2 diversity analysis and refer to the LAR for the D3 assessment for the Common Q CPCS.

(d.1) Yes, LAR Enclosure Section 3.2, Licensing Technical Report (LTR) will be revised as part of a LAR Supplement. The following paragraph will be deleted:

"LTR Section 3.2.18 describes the NRC evaluation of the first CPCS at Arkansas Nuclear One, Unit 2 (ANO-2) in NUREG-0308, "Safety Evaluation Report Related to the Operation of Arkansas Nuclear One, Unit 2," Supplement 1 (i.e., the ANO-2 NRC SER) in regards to CPCS Common Cause Failure (CCF). This was also the evaluation the NRC staff referred to in their PVNGS safety evaluation for the Common Q CPCS upgrade license amendment (Reference 6.10, Section 3.4.6.11). The NRC cited the ANO-2 evaluation to conclude, in part, that CCF is adequately addressed for the Common Q CPCS replacement for PVNGS. The Waterford LTR included this as part of the reference design licensing precedence."

(c.1) In LAR Section 4.1, "Applicable Regulatory Requirements/Criteria", under the bullet, "The applicable portions of the following branch technical positions within NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition" (SRP), Chapter 7, "Instrumentation and Controls," as follows:"; the sub-bullet "Branch

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION

f.

Westinghouse organization chart, as referenced in LTR Section 5.2.12, Software V&V Processes

g. Control Panel 7 & 2 Cyber Security Door Lock Plan, ENT-WF3-CPC-115

h. Document that identifies and justifies the values of the CPCS delay times used in the thermal margin estimate for each of the applicable transients and accidents listed in Table 3.2.6-1 of Attachment 4.

i.

N/A Software Safety Plan for the Core Protection Calculator System Upgrade

j.

SPEC-10-00001-MULTI, 73.55 Fleet Strategy Implementation -

Fiber Optic Cable Common-Procurement Specification (Reference 40 of the LTR)

k. AC160 CPU Loading Restrictions, Document Number AN03007Sp (SyRS Reference 1.4.2.12)

l.

Project Management Plan for the Waterford 3 Core Protection Calculator Upgrade, GPEP-PMP-2019-000020, Revision 1

m. WF3 Project Quality Plan

n. Subsequent EQSR (see open item 23)

o. Waterford Unit 3 Common Q Implementation - Non-LOCA Evaluation of Updated CPCS Response Times, LTR-TA-20-4, Revision 0 (LTR Reference 24)

p. PO 10587546 - CPC, CEAC, CEAPDS Single Channel and Four Channel Components

q. PO 10591996 - Input / Output (I/O) Simulator Components

r. SPEC-18-00005-W, Rev 0

s. CPCS Replacement Project Critical Procurement Project (CPP),

CPP-WF3-2019-002 (WTWF3-2019-00236)

t.

EN-MP-100, Critical Procurements

u. EN-DC-115, Engineering Change Process

v. EN-IT-104, Software Quality Assurance Program

w. 00000-ICE-36369, Rev. 02, CPC Timing Analysis for the Common Q Core Protection Calculator System

x. EN-DC-149, Acceptance of Vendor Documents

y. Waterford 3 Core Protection Calculator System Safety Function Table, LTR-TA-19-154, Revision 0

z. Entergy Quality Assurance Program Manual aa. Entergy Specification SPEC-18-00005-W, Revision 0, Core Protection Calculator Purchase Specification, April 2, 2019.

bb. Westinghouse Letter CWTR3-19-21, Revision 2, Transmittal of Westinghouse Final Compliance Matrix for SPEC-18-00005-W, June 28, 2019.

cc. Westinghouse Calculation Note CN-EQT-19-6, Revision 0, Determination of In-Equipment Response Spectra for Waterford Unit 3 Core Protection Calculator System, August 12, 2019.

dd. Westinghouse Document WCAP-16166-P Supplement 1-E09, Revision 1, Equipment Qualification Report for AC160 Platform - AI687 in Section 3 of the SPM. So, there is no WF3 CPC project-specific Software Safety Plan, Section 3, Software Safety Plan, of the Common Q Software Program Manual is followed for the project.

j.

Located in WEC SharePoint

k.

Provided in the WEC SharePoint

l.

WEC Uploaded to SharePoint (11/3/20 Update)

m.

Provided in the WEC SharePoint

n.

Provided in the WEC SharePoint

o.