2CAN042501, Amendment 32 to Safety Analysis Report

Amendment 32 to Safety Analysis Report

ML25112A321
Person / Time
Site:	Arkansas Nuclear
Issue date:	04/22/2025
From:	Pehrson D Entergy Operations
To:	Office of Nuclear Reactor Regulation, Document Control Desk
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ML25112A320	List: 2CAN042501, Amendment 32 to Safety Analysis Report
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2CAN042501
Download: ML25112A321 (1)

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SECURITY RELATED INFORMATION - WITHHOLD UNDER 10 CFR 2.390 This letter is decontrolled when separated from Enclosure 1 Entergy Operations, Inc., 1448 SR 333, Russellville, AR 72802 10 CFR 50.4(b)(6) 10 CFR 50.59(d)(2) 10 CFR 50.71(e) 10 CFR 54.37(b) 2CAN042501 April 22, 2025 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Amendment 32 to the ANO Unit 2 Safety Analysis Report Arkansas Nuclear One, Unit 2 NRC Docket No. 50-368 Renewed Facility Operating License No. NPF-6 In accordance with 10 CFR 50.71(e) and 10 CFR 50.4(b)(6), enclosed is Amendment 32 of the Arkansas Nuclear One, Unit 2 (ANO-2) Safety Analysis Report (SAR). Included with this update are the current ANO-2 Technical Requirements Manual (TRM) and the current ANO-2 Technical Specification (TS) Bases. The TS Bases file also includes the Table of Contents which outlines the contents of both the TSs and the TS Bases, since the Table of Contents is revised by the licensee under 10 CFR 50.59. Pursuant to 10 CFR 50.71(e)(4), these documents are being submitted within six months following the previous ANO-2 refueling outage (2R30) which ended October 28, 2024. Summaries of changes to the ANO-2 TRM and TS Bases are included in Attachments 1 and 2 of this letter for the period beginning November 16, 2023, and ending April 22, 2025.

In accordance with NEI 98-03, "Guidelines for Updating Final Safety Analysis Reports,"

Appendix A, Section A6, a list and short description of information removed from the SAR should be included with each SAR update submittal. For this reporting period, the following information was removed from the SAR which meets the criteria of Appendix A, Section A4 or A5 of NEI 98-03, which requires reporting in accordance with Appendix A, Section A6:

MET Tower Generator Actual Capacity (2.3.3.4) - Removed excessive detail that stated exact rating of generator when listing minimum capacity was more appropriate.

Turbine Missile Probabilities (3.5.2.2.2.1, 3.5.2.2.2.2, 3.5.2.2.2.3, 3.5.2.2.2.4, Table 3.5-2, Table 3.5-3, 10.2.3) - Adopted Regulatory Guide 1.115, "Protection Against Missiles" Rev. 2 which states that only the turbine missile genesis probability (P1) needs to be evaluated to ensure safe plant operation. Removed unneeded information concerning trajectory (P2),

damage (P3), and total hazard probability (P4). Information concerning the "hypothetical missile" was removed since it was deprecated. Discussion regarding low pressure (LP) turbine wheels was also removed as it is not applicable to the mono-block designs used on ANO-2.

Douglas E. Pehrson Site Vice President Arkansas Nuclear One Tel 479-858-3110

SECURITY RELATED INFORMATION - WITHHOLD UNDER 10 CFR 2.390 This letter is decontrolled when separated from Enclosure 1 2CAN042501 Page 2 of 3 Margin Management Program (8.3.2.1.1) - Removed reference to the "Margin Management Program" since it was retired in 2016.

Associated in part with the post September 11, 2001 response related to security sensitive information, Entergy has reviewed the ANO-2 SAR and determined that the items in the following paragraph contain information required to be withheld from public disclosure with respect to NRC Regulatory Issue Summary (RIS) 2015-17, "Review and Submission of Updates to Final Safety Analysis Reports, Emergency Preparedness Documents, and Fire Protection Documents."

The following information is located on SAR Pages 2.8-1 through 2.8-10:

SAR Section 2.8.1, "Flood Related Information" SAR Section 2.8.1.1, "Probable Maximum Flood Combined with Wind Wave Action" SAR Section 2.8.1.2, "Probable Maximum Flood Combined with Ozark Dam Failure" SAR Section 2.8.1.3, "Probable Maximum Flood on Streams and Rivers" SAR Section 2.8.1.3.1, "Probable Maximum Precipitation" SAR Section 2.8.1.3.2, "Precipitation Losses" SAR Section 2.8.1.3.3, "Runoff Model" SAR Section 2.8.1.3.4, "Probable Maximum Flood Flow" SAR Section 2.8.1.3.5, "Water Level Determinations" SAR Section 2.8.1.3.6, "Coincident Wind Wave Activity" SAR Section 2.8.1.3.7, "Site Drainage System" SAR Section 2.8.1.4, "Potential Dam Failures (Seismically Induced)"

SAR Section 2.8.1.5, "Design Basis for Subsurface Hydrostatic Loadings" SAR Section 2.8.2, "Additional Natural Gas Pipeline Information" SAR Section 2.8.3, "Additional New Fuel Storage Information" The above is consistent with currently redacted information from the ANO-2 SAR (reference ML23324A028, ANO-2 SAR Amendment 31). Entergy requests the aforementioned information be withheld from public disclosure in accordance with 10 CFR 2.390. Accordingly, a complete version and a redacted version of the ANO-2 SAR are included as enclosures to this letter.

In accordance with 10 CFR 54.37(b), after a renewed license is issued, the SAR update required by 10 CFR 50.71(e) must include any systems, structures, and components (SSCs) newly identified that would have been subject to an aging management review or evaluation of time-limited aging analyses in accordance with 10 CFR 54.21. The SAR update must describe how the effects of aging will be managed such that the intended function(s) in 10 CFR 54.4(b) will be effectively maintained during the period of extended operation. For this reporting period, no new SSCs that would have been subject to an aging management review or evaluation of time-limited aging analyses in accordance with 10 CFR 54.21 were identified.

A summary of the 10 CFR 50.59 evaluations during the reporting period is normally included with the required SAR submittal or within 30 days thereafter. Attachment 3 contains a summary of the 10 CFR 50.59 evaluations performed for ANO-2 over the aforementioned reporting period. Attachment 4 includes a copy of the evaluations.

SECURITY RELATED INFORMATION-WITHHOLD UNDER 10 CFR 2.390 This letter is decontrolled when separated from Enclosure 1 2CAN042501 Page 3 of 3 includes a list of SAR pages that were updated during the period beginning November 16, 2023, and ending April 22, 2025.

If you have any questions or require additional information, please contact Riley Keele, Manager, Regulatory Assurance, at 479-858-7826.

The information contained in the above Licensing Basis Documents accurately reflects changes made since the previous submittal. The changes to these documents reflect information and analyses submitted to the Commission, prepared pursuant to Commission requirements, or made under the provisions of 10 CFR 50.59. I declare under penalty of perjury that the foregoing is true and correct. Executed on April 22, 2025.

Respectfully, Douglas E. Pehrson ANO Site Vice President DEP/mar Attachments:

1. Summary of ANO-2 TRM Changes

2. Summary of ANO-2 TS Bases Changes

3. Summary of ANO-2 10 CFR 50.59 Evaluations

4. 10 CFR 50.59 Evaluations - From November 16, 2023 through April 22, 2025

5. List of Affected SAR Pages

Enclosures:

1. ANO-2 SAR Amendment 32-Un-redacted Version

2. ANO-2 SAR Amendment 32 - Redacted Version

3. ANO-2 TRM

4. ANO-2 TS Table of Contents and TS Bases cc:

NRC Region IV Regional Administrator NRC Senior Resident Inspector - Arkansas Nuclear One NRC Project Manager - Arkansas Nuclear One Designated Arkansas State Official

2CAN042501 Summary of ANO-2 TRM Changes 2CAN042501 Page 1 of 1 Summary of ANO-2 TRM Changes The following changes to the Arkansas Nuclear One, Unit 2 (ANO-2) Technical Requirements Manual (TRM) were implemented in accordance with the provisions of 10 CFR 50.59.

Because these changes were implemented without prior NRC approval, a description is provided below:

Revision Section Summary 89 3.7.5, TR 4.3.4.2.a, B 3.3.4 LBDCR 23-034 - Engineering Change, EC-93373, "ANO-2 Turbine Missile Probability and TV/GV testing frequency change" LBDCR 23-038 - "Add note to Doors 185 and 186 in Table 3.7.5-1 that tells user that the ANO-2 diesel generator fuel oil vault doors are required fire barrier doors for both units" 90 3.7.5, TR 4.1.2.c, B 3.3.3 LBDCR 24-011 - "Delete DR-387 from TRM Table 3.7.5-1 Since Door is Not Contained Within a Regulatory Required Fire Barrier and Is Not Required to be in the TRM" LBDCR 24-031 - "Extend Frequency of Testing Boron Injection Flow Paths (TR 4.1.2.c) from 18 Months to 36 Months" LBDCR 24-035 - "Revise Unit 2 TRM 3.3.3, 'Seismic Instrumentation' Basis to Remove Discussion of Reportability" Acronyms B

Basis GV Governor Valve LBDCR License Basis Document Change Request TR Test Requirement TRM Technical Requirements Manual (TRM)

TV Throttle Valve

2CAN042501 Summary of ANO-2 TS Bases Changes 2CAN042501 Page 1 of 1 Summary of ANO-2 TS Bases Changes The following changes to the Arkansas Nuclear One, Unit 2 (ANO-2) Technical Specification (TS) Bases were implemented in accordance with the provisions of 10 CFR 50.59 and the Bases Control Program of ANO-2 TS 6.5.14. Because these changes were implemented without prior NRC approval, a description is provided below:

Revision Section Summary 87 B 3/4.3.3.6, B 3/4.7.1.2, B 3/4.7.6 License Basis Document Change LBDC 24-002, "Correct overlooked surveillance frequencies in bases from TS Amendment 315, 'TSTF-425, Relocation of Surveillance Frequencies'"

88 B 3/4.3.3.6 License Basis Document Change LBDC 23-044, "ANO-2 TS Basis 3/4.3.3.5, INSTRUMENT 7 and 8 descriptions are reversed on the two tables" 89 B 3/4.5.2, B 3/4.6.2.1, B 3/4.8, B 3/4.6.3, B 3/4.6.2.3, B 3/4.3, B 3/4.7.1.2 TS Amendment 333, "ADOPT TSTF-505, Revision 2,

'Provide Risk-Informed Extended Completion Times -

RITSTF Initiative 4b'"

Acronyms B

Basis LBDC License Bases Document Change SR Surveillance Requirement TS Technical Specifications TSTF Technical Specification Task Force

2CAN042501 Summary of ANO-2 10 CFR 50.59 Evaluations 2CAN042501 Page 1 of 1 Summary of ANO-2 10 CFR 50.59 Evaluations 50.59 #

50.59 Summary 2023-003 Engineering Change EC-93379, "Evaluation of Extending Main Turbine Valve Testing Intervals" 2023-004 Engineering Change EC-95520 "Add Switch in Series with Recirculation Actuation Signal (RAS) to Avoid Lifted Lead Temporary Modification" 2024-002 Engineering Change EC-54161217 "Evaluation to Support 2CV-1050-2, 'Emergency Feedwater (EFW) Green Train Steam Supply Valve' Normally Closed" 2024-003 Engineering Change EC-87349 "ANO-2 Inadequate Core Cooling (ICC) Reactor Vessel Level Monitoring System (RVLMS) / Core Exit Thermocouple (CET) /

Subcooling Margin Monitoring (SMM) Data Acquisition System (DAS) Replacement"

2CAN042501 10 CFR 50.59 Evaluations - From November 16, 2023, through April 22, 2025 (36 Pages)

ANO 50.59 Evaluation 2023-003

ATTACHMENT 9.1 50.59 EVALUATION FORM Sheet 1 of 3 EN-LI-101 R22 I.

OVERVIEW / SIGNATURES1 Facility: Arkansas Nuclear One, Units 1 & 2 Evaluation # / Rev. #: 1 Proposed Change / Document: EC-93373 Description of Change: EC-93373 performs the engineering evaluation to support the extension of the test intervals associated with the ANO-1&2 turbines, specifically:

ANO-1 throttle valves (TVs) and governor valves (GVs);

ANO-2 stop valves (SVs), control valves (CVs), and combined intermediate valves (CIVs).

These valves are key elements of the turbine overspeed protection system (OPS) for their respective units due to their function and use in terminating steam flow to the high pressure (HP) and low pressure (LP) turbines. Prompt and effective termination of steam flow to the main turbine during a load separation event leading to a turbine overspeed conditions limits the potential that turbine missiles can be generated as a result of turbine rotor components failure.

The engineering evaluation is homogenous for both units: that is, to follow the methodology set forth in Regulatory Guide 1.115, Rev. 2, which states that safe and reliable operation is ensured if the probability of missile genesis from each units turbine (P1) is kept below an acceptance criteria of 1E-5 per year. New P1 values have been developed for each units turbine in accordance with previously existing methodology; a sensitivity study is then performed to reduce existing margin between each units P1 value and the acceptance criteria to allow for the extension of test interval.

The basis of this sensitivity study is to elongate the period between test intervals and examine the effect it has on P1 via reliability theory.

For ANO-1, the GVs and TVs are currently tested at a six month interval; this work allows for an extension of these tests up to eighteen months. No attempt is made in this work to alter the current ANO-1 test interval for these valves - but this activity lays the groundwork to pursue this extension at a later date.

For ANO-2, the SVs, CVs, and CIVs are tested at a four month interval; this work allows for an extensions of these tests up to eight months. This work goes on to alter the current ANO-2 test interval for these valves, pursuing the appropriate procedure changes and a change to the current ANO-1 Technical Requirements Manual (ANO-1 TRM).

In accordance with industry best practices, the test interval extensions are expected to be staggered, i.e. while an eight month extension is demonstrated to be acceptable with this work, an increase to six months must be had prior to ensure performance degradation does not occur. The same is valid for ANO-1, when and if that extension is pursued.

This activity will not impact the ANO-1 SAR, which addresses turbine overspeed qualitatively in 14.1.2.9, referencing the original Cycle 1 turbine overspeed analysis, concluding turbine overspeed is not a credible accident sequence. Work in 1R8 and 1R14 added additional 1

The printed name should be included on the form when using electronic means for signature or if the handwritten signature is illegible. Signatures may be obtained via electronic authentication, manual methods (e.g., ink signature), e-mail, or telecommunication. Signing documents with indication to look at another system for signatures is not acceptable such as See EC or "See Enterprise Asset Management (EAM) Application." Electronic signatures from other systems are only allowed if they are included with the documentation being submitted for capture in eB (e.g., if using an e-mail, attach it to this form; if using Enterprise Asset Management (EAM) Application, attach a screenshot of the electronic signature(s); if using Corrective Action Program (CAP) Application, attach a copy of the completed corrective action).

ANO-2 ANO-2 condition limits the potential that EC-93379, "Evaluation of Extending Main Turbine Valve Testing Intervals" FFN-2023-003

ATTACHMENT 9.1 50.59 EVALUATION FORM Sheet 2 of 3 EN-LI-101 R22 conservatism to this analysis with fully/partially integrated rotor upgrades to the 1 and 2 section LP rotors. This work is revisited each cycle during the ANO-1 reload, via the ANO-1 Reload Technical Document.

This activity does, however, impact the ANO-2 SAR, which addresses turbine overspeed quantitatively in Section 3.5.2.2 & 10.2.3, specifically stating missile probabilities; the test interval is also explicitly defined in the ANO-2 TRM. As a result, an LBDCR is issued for this work.

Revision 1 of this 50.59 Evaluation addresses a concern from ANO Licensing regarding the consequences of valve failure, discussed in Questions 3 & 4specifically, the hypothetical scenario that the unit is operating with Technical Specification limits for primary activity and primary-to-secondary leakage. That position would render the secondary fluid radioactive, and therefore, an unmitigated turbine overspeed event would yield a radiological consequence. Questions 3 & 4 now address that. The conclusion of this evaluation is unchanged.

Summary of Evaluation:

The eight questions in this evaluation were answered in the negative. Therefore, prior NRC approval for this change is not required.

Is the validity of this Evaluation dependent on any other change?

Yes No If Yes, list the required changes/submittals. The changes covered by this 50.59 Evaluation cannot be implemented without approval of the other identified changes (e.g., license amendment request). Establish an appropriate notification mechanism to ensure this action is completed.

N/A Based on the results of this 50.59 Evaluation, does the proposed change require prior NRC approval?

Yes No Preparer2:

Matt Montgomery / / EOI / Integrated Risk / See Sign.

Name (print) / Signature / Company / Department / Date Reviewer2:

Lindsey McConnell / / EOI / Major Projects / See Sign.

Name (print) / Signature / Company / Department / Date Independent Review3:

Name (print) / Signature / Company / Department / Date Responsible Manager Concurrence:

Name (print) / Signature / Company / Department / Date 50.59 Program Coordinator Concurrence:

Name (print) / Signature / Company / Department / Date 2

Either the Preparer or Reviewer will be a current Entergy employee.

3 If required by Section 5.1[2].

N/A T. Hatfield/ /EOI/Central Design/9-27-23 Michael Hall/

/EOI/Regulatory Assurance/9-27-23 Digitally signed by Matthew T.

Montgomery DN: cn=Matthew T. Montgomery, c=US, o=Integrated Risk, ou=Entergy, email=mmontg4@entergy.com Date: 2023.09.26 14:54:36 -05'00' cn=Lindsey McConnell, c=US, o=Engineering Supervisor, ou=Entergy ANO, email=lmcconn@entergy.com 2023.09.27 07:30:18 -05'00' Lindsey McConnell Digitally signed by Thomas A. Hatfield II DN: cn=Thomas A. Hatfield II, c=US, o=Entergy - ANO, ou=Design Engineering, email=thatfie@entergy.com Date: 2023.09.27 08:14:07 -05'00' Thomas A.

Hatfield II Digitally signed by Michael Hall DN: cn=Michael Hall, c=US, o=Entergy, ou=ANO Regulatory Assurance, email=mhall10@entergy.com Date: 2023.09.27 09:24:11 -05'00' Michael Hall

ATTACHMENT 9.1 50.59 EVALUATION FORM Sheet 3 of 3 EN-LI-101 R22 OSRC:

Chairmans Name (print) / Signature / Date [GGNS P-33633, P-34230, & P-34420; W3 P-151]

OSRC Meeting #

II.

50.59 EVALUATION [10 CFR 50.59(c)(2)]

Does the proposed Change being evaluated represent a change to a method of evaluation ONLY? If Yes, Questions 1 - 7 are not applicable; answer only Question 8.

If No, answer all questions below.

Yes No Does the proposed Change:

1.

Result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the UFSAR?

Yes No BASIS:

Decreasing the frequency of turbine valve tests increases the unreliability of the turbine overspeed protection system; this is applicable to both ANO-1 & 2. High reliability of the overspeed protection system is required to ensure that the risk of turbine missiles remains low. Production of turbine missiles in a turbine overspeed transient is an accident described in ANO-1 SAR Chapter 14, while it is an event described in ANO-2 SAR Chapters 3 & 10.

The analyses that support the proposed change in the test intervals, CALC-ANO1-ME-23-00001 for ANO-1 and CALC-ANO2-ME-22-00007 for ANO-2, do not calculate a new missile probability. These analyses estimate the impact on existing turbine missile probability P1 calculated by the OEM vendors. The results of these analyses indicate that the acceptance criterion of RG 1.115, Rev. 2, of 1E-5 per year will continue to be met with extended test intervals. Section 4.3.1 of NEI 96-07 indicates that a proposed activity would not result in more than a minimal increase in the frequency of an accident previously evaluated in the UFSAR provided that the plant-specific frequency threshold is not exceeded as a result of the change. Given we continue to meet the acceptance criterion of 1E-5 per year with margin, no, this change does not result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the UFSAR.

OSRC-2023-023

ATTACHMENT 9.1 50.59 EVALUATION FORM Sheet 4 of 3 EN-LI-101 R22

2.

Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the UFSAR?

Yes No BASIS:

For ANO-1, the GVs and TVs are used to terminate steam flow to the HP turbines, whereas for ANO-2, the SVs, CVs, and CIVs are used to terminate steam flow to both the HP and LP turbines. For both units, these valves serve to assist in reducing the probability of a turbine overspeed event that could ultimately lead to turbine missile generation. The OPS functions to mitigate the effects of an overspeed event by signaling closure of these valves, and periodic testing of these valves demonstrates that they will satisfactorily perform this design function.

NEI 96-07 Section 4.3.2 provides guidance on a more than minimal increase in the likelihood of a malfunction of a structure, system, or component important to safety previously evaluated in the UFSAR by stating that NRC approval would be necessitated in the event that the change in likelihood of occurrence of a malfunction is calculated in support of the evaluation and increases by more than a factor or two, and this factor of two should be applied at the component level.

To ensure this factor of two is not exceeded, the turbine test intervals are to be implemented in a phased approach; that is, for ANO-1, the current six month testing interval for the GVs and TVs are to be extended in three month intervals, i.e. to 9 months, then to 12 months, then 15 months, then 18 months (if desired to align with LP valve testing), while the ANO-2 four month testing interval for the SVs, CVs, and CIVs is to be extended in two month intervals, i.e. to 6 months, then 8 months. This phased implementation has been analyzed in CALC-ANO1-ME-23-00001 and CALC-ANO2-ME 00007 to not increase valve unreliability by more than a factor of two.

Furthermore, this is in alignment with best industry practices for turbine valve test extensions pursued at similar sites as ANO. As defense-in-depth, valve stroke timing and EHC fluid quality are to be trended and monitored for degradation. Therefore, no, this change does not result in more than a minimal increase in the occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the either units UFSAR.

ATTACHMENT 9.1 50.59 EVALUATION FORM Sheet 5 of 3 EN-LI-101 R22

3.

Result in more than a minimal increase in the consequences of an accident previously evaluated in the UFSAR?

Yes No BASIS:

A turbine overspeed event is not postulated to yield any radiological consequence to the public or operators in the control room, by extension of the fact that the water-steam mixture in the secondary is not normally radioactive. This is valid for both ANO-1 & 2. In the event either unit is operating at the Tech Spec limit for primary activity and primary-to-secondary leakage such that the secondary has a non-negligible degree of radioactivity, an existing Chapter 14/15 accident analysis bounds the scenario. Therefore, no, this change does not result in a more than minimal increase in the consequences of an accident previously evaluated in either units UFSAR.

4.

Result in more than a minimal increase in the consequences of a malfunction of a structure, system, or component important to safety previously evaluated in the UFSAR?

Yes No BASIS:

As stated in the response to Question 3 above, radiological consequence is not postulated to occur as a result of a turbine overspeed transient. A hypothetical failure of the subject valves could potentially lead to a release of secondary steam, but a dose concern is not credible. As stated above, the water-steam mixture in the secondary is not normally radioactive, but if postulated to be, due to either unit operating at TS limits for primary activity and primary-to-secondary leakage, an existing Chapter 14/15 accident analysis would bound the transient. Therefore, no, this change does not result in more than a minimal increase in the consequences of a malfunction of a structure, system, or component important to safety previously evaluated in either units UFSAR.

5.

Create a possibility for an accident of a different type than any previously evaluated in the UFSAR?

Yes No BASIS:

No new accident can be introduced as a result of this change; the turbine overspeed accident remains bounding. This is valid for both ANO-1 & 2. Therefore, no, this change does not create a possibility for an accident of a different type than any previously evaluated in either units UFSAR.

6.

Create a possibility for a malfunction of a structure, system, or component important to safety with a different result than any previously evaluated in the UFSAR?

Yes No BASIS:

The turbine valves and other overspeed protection system components must actuate to an overspeed condition to trip the turbine. Less frequent testing of the steam turbine valves does not change the existing failure modes or introduce new failure modes of these components. This is valid for ANO-1 & 2. Therefore, no, this change does not create a possibility for a malfunction of a structure, system, or component important to safety with a different result than any previously evaluated in the UFSAR.

ATTACHMENT 9.1 50.59 EVALUATION FORM Sheet 6 of 3 EN-LI-101 R22

7.

Result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered?

Yes No BASIS:

This change involves the test intervals associated with turbine valves in the secondary. These valves are not credited fission product barriers in the UFSAR. This is valid for both ANO-1 & 2. Therefore, no, this change does not result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered.

8.

Result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses?

Yes No BASIS:

The OEM vendor evaluation for the missile genesis probability, P1, is as-built with this change; Westinghouse/Siemens provided this input for ANO-1, while GE provided this input for ANO-2. The vendor has utilized the same evaluation method described in the UFSAR. A sensitivity study has been performed atop these analyses that manipulate an input to the method - the interval of valve testing - and examine the relative impact of different valve test intervals using reliability theory.

Following the guidance of NEI 96-07, the NRCs approval of our analysis of the turbine overspeed event was not contingent on a fixed, understood value of the interval of valve testing, only that the overall probability of the event remains low. As such, this input can be manipulated, so long as the acceptance criterion of RG 1.115, Rev. 2, continues to be met. This is valid for ANO-1 & 2. Therefore, no, this change does not result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses.

If any of the above questions is checked Yes, obtain NRC approval prior to implementing the change by initiating a change to the Operating License in accordance with NMM Procedure EN-LI-103.

ANO 50.59 Evaluation 2023-004

EN-LI-101 R22 ATTACHMENT 9.1 50.59 EVALUATION FORM Sheet 1 of 3 I.

OVERVIEW / SIGNATURES1 Facility: _Arkansas Nuclear One - Unit 2____

Evaluation # / Rev. #: __FFN-2023-004/1 Proposed Change / Document: __ ADD SWITCH IN SERIES WITH RAS RELAY TO AVOID LIFTED LEAD TEMP MOD SIPD 8349 / EC95520 Description of Change: This change installs a keylock switch to prevent the Recirculation Actuation Signal (RAS), from tripping the Low Pressure Safety Injection, pumps (LPSI) during shutdowns, Mode 5 or 6. When the switch is in BYPASS position, the RAS trip is bypassed for the corresponding LPSI pump and there will be a Green indicating light. The switch will be keylocked to prevent inadvertent actuation. The use of the switch will be procedurally controlled to ensure use at the proper conditions. The function of the RAS is to change the source of water for the high-pressure safety injection (HPSI) and the containment spray pumps to transfer to the sump. The LPSI pumps are tripped to prevent damage due to low suction head. This modification will replace the use of lifted leads during the implementation of procedure OP-2104.004, Shutdown Cooling System during testing while shutdown cooling is in service. During unit operations the keylock switch will be in the NORMAL position and the RAS LPSI pump trip will be active. This modification adds two additional failure modes into the circuit. In a case where the logic contacts of the switch were to fail to change position when the switch position is changed, either from BYPASS to NORMAL or from NORMAL to BYPASS. In the BYPASS to NORMAL case the contact would remain open and the LPSI pump would fail to trip on a RAS signal. In the NORMAL to BYPASS case the RAS signal would still trip the LPSI pump when it is being used for shutdown cooling with an inadvertent RAS signal. Loss of shutdown cooling would potentially allow reactor coolant temperature to rise during a shutdown. If an inadvertent RAS signal tripped the pump, manual action would be required to restart the pump in an expeditious manner. The indicating light section of the circuit will be fused and the light will only be energized when the switch is in BYPASS. The light section of the circuit was determined to have no adverse effects in the PAD for EC 95520.

Summary of Evaluation:

The eight questions to determine if the modification to add Bypass switches in the circuit to trip LPSI pump on a RAS signal were all answered as No. The addition of 1

The printed name should be included on the form when using electronic means for signature or if the handwritten signature is illegible. Signatures may be obtained via electronic authentication, manual methods (e.g., ink signature), e-mail, or telecommunication. Signing documents with indication to look at another system for signatures is not acceptable such as See EC or "See Enterprise Asset Management (EAM) Application." Electronic signatures from other systems are only allowed if they are included with the documentation being submitted for capture in eB (e.g., if using an e-mail, attach it to this form; if using Enterprise Asset Management (EAM) Application, attach a screenshot of the electronic signature(s); if using Corrective Action Program (CAP) Application, attach a copy of the completed corrective action).

EN-LI-101 R22 the Bypass switches, one for each train, will not present an unreviewed safety question and do not require review and approval by the NRC prior to implementation.

Is the validity of this Evaluation dependent on any other change?

Yes No If Yes, list the required changes/submittals. The changes covered by this 50.59 Evaluation cannot be implemented without approval of the other identified changes (e.g., license amendment request). Establish an appropriate notification mechanism to ensure this action is completed._________________________________________________________________________

Based on the results of this 50.59 Evaluation, does the proposed change require prior NRC approval?

Yes No Preparer2:

Richard Blagbrough/ /Entergy/Engineering/

Name (print) / Signature / Company / Department / Date Reviewer2:

Brad Miller/ /Entergy/Engineering/

Name (print) / Signature / Company / Department / Date Independent Review3:

Not Required Name (print) / Signature / Company / Department / Date Responsible Manager Concurrence:

Name (print) / Signature / Company / Department / Date 50.59 Program Coordinator Concurrence:

Name (print) / Signature / Company / Department / Date OSRC:

Chairmans Name (print) / Signature / Date [GGNS P-33633, P-34230, & P-34420; W3 P-151]

___OSRC-2023-027_______________________________

OSRC Meeting #

II.

50.59 EVALUATION [10 CFR 50.59(c)(2)]

Does the proposed Change being evaluated represent a change to a method of evaluation ONLY? If Yes, Questions 1 - 7 are not applicable; answer only Question 8.

If No, answer all questions below.

Yes No 2

Either the Preparer or Reviewer will be a current Entergy employee.

3 If required by Section 5.1[2].

/Entergy/Engineering

/Entergy/Regulatory Assuance Discussed again at OSCR-2024-001 Digitally signed by Richard Blagbrough DN: cn=Richard Blagbrough, c=US, o=Engineering, ou=EFIN, email=rblagbr@entergy.com Date: 2024.02.05 11:37:58 -06'00' Richard Blagbrough Digitally signed by Brad Miller DN: cn=Brad Miller, c=US, o=ANO Electrical Design, ou=Entergy Operations Inc., email=jmille3@entergy.com Date: 2024.02.05 11:42:11 -06'00' Brad Miller Digitally signed by Scott Kerins DN: cn=Scott Kerins, c=US, email=skerins@entergy.com Date: 2024.02.05 12:03:09 -

06'00' Scott Kerins Digitally signed by Michael Hall DN: cn=Michael Hall, c=US, o=Entergy, ou=ANO Regulatory Assurance, email=mhall10@entergy.com Date: 2024.02.05 13:30:19 -06'00' Michael Hall

EN-LI-101 R22 Does the proposed Change:

1.

Result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the UFSAR?

Yes No BASIS: The addition of a keylock bypass switch in the 2P-60A or 2P-60B, LPSI pumps, stop circuit will not add an accident initiator or change an accident initiator for the accidents analyzed in the UFSAR during operations while the switch is in the NORMAL position. The function of the RAS signal is to switch the source of cooling water from the Refueling Water Tank (RWT) to the sumps, and min flow valves that return to the RWT are closed, ref. UFSAR 6.2.3.2.2.2. Containment Spray System (CCS) and HPSI pumps are used after the RAS signal is received and LPSI pumps are tripped. According to section 7.3.1.1.11.6 in the UFSAR The RAS automatically stops the low pressure safety injection pumps. Per UFSAR 6.2.2.2.1.C RAS automatically causes LPSI pumps to stop. This is a design function of the RAS as stated in the UFSAR. The failure of the logic circuit contact while in BYPASS position during shutdown cooling with an inadvertent RAS signal could initiate a loss of LPSI for shutdown cooling. Shutdown cooling is discussed in UFSAR section 9.3.6 and the loss of shutdown cooling is discussed in UFSAR section 9.3.6.3. The switches will be keylocked to prevent inadvertent actuation.

The use of the switches will be procedurally controlled to ensure use at the proper conditions. The RAS circuit is not an initiator to UFSAR accidents as listed in Chapter 15 of the UFSAR. However, the RAS signal is used to help mitigate the effects of the LOCA(Loss of Coolant Accident) as described in UFSAR 15.1.13.

Since neither of the failure modes of the switch will cause the increase in the frequency of an accident initiation, the addition of these bypass switches in the LPSI trip circuit will not result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the UFSAR.

EN-LI-101 R22 2.

Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the UFSAR?

Yes No BASIS: According to section 7.3.1.1.11.6 in the UFSAR The RAS automatically stops the low pressure safety injection pumps. Per 6.2.2.2.1.C RAS automatically causes LPSI pumps to stop. This is a design function of the RAS as stated in the UFSAR. With this design function any increase in a failure mode probability should be addressed. Section 6.3.2.14 bullet C in the UFSAR states Assuming the single active failure of the B Train LPSI failing to trip after RAS. This is also discussed in UFSAR Section 6.3.2.20.4. UFSAR Section 7.4.2.2 also describes conformance to IEEE 308. The UFSAR states that, The electrical circuitry associated with safe shutdown systems conforms to IEEE 308 In IEEE 308-1971 (Class 1E Electrical Systems for Nuclear Power Generating Stations) a common failure mode is defined as, a mechanism by which a single design basis event can cause redundant equipment to be inoperable. Furthermore, IEEE 308-1971 describes class 1E Electrical systems (which applies to the LPSI stop circuitry) sufficient physical separation, electrical isolation, and redundancy shall be provided to prevent the occurrence of common failure mode in Class 1E systems. The change involves the addition of one switch per circuit and each circuit is redundant. The switch will be keylocked to prevent inadvertent actuation. The use of the switch will be procedurally controlled to ensure use at the proper conditions. The failure of one switch to trip the LPSI pump during RAS actuation would not cause the opposite LPSI pump to fail to trip. Therefore, the criteria described in IEEE 308-1971 is still met. The failure mode of the pump not tripping is currently affected by one relay.

This modification will add one switch. The relay can fail from a coil failure or a contact failure. The switch can only fail from a contact failure. The addition of the bypass switch does not affect the PRA analysis and the increase in the failure modes probability will not increase by a factor of two.

The other failure mode is when one set of contacts of the bypass switch fail while the switch is in BYPASS position during shutdown cooling and a loss of LPSI shutdown cooling could occur if RAS actuated. The intent is to block RAS from tripping a LPSI pump during shutdown. The RAS signal may be blocked below Mode 4 when RCS pressure is below the bypass permissive per Section 6.3.5.1.2 of the UFSAR Section 9.3.6.3 of the UFSAR discusses the loss of cooling event.

These switches are being procured as safety related, 1E, and will be manufactured per an Appendix B program to current standards. The failure of the switch manufactured to 1E standards and constructed with a minimal number of contacts is a negligible increase in the likelihood of occurrence of a malfunction. It is an improvement to the current method of lifting leads while testing is performed during testing. Both methods could initiate the loss of shutdown cooling event. With this consideration and evaluation of the failure mode and effect discussed above in the UFSAR the changes made by adding a bypass switch in the LPSI trip circuit by EC-95520 does not result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the UFSAR.

EN-LI-101 R22 3.

Result in more than a minimal increase in the consequences of an accident previously evaluated in the UFSAR?

Yes No BASIS: The failure mode from this bypass switch in the LPSI pump circuit would be that the LPSI pump would fail to trip when required by the actuation of the RAS signal on low RWT level during Modes 1-4. The failure of a LPSI pump to trip is already encompassed by the accident analysis as discussed in the UFSAR sections 6.3.2.14.C and 6.3.2.20.4. This failure mode would not affect the other actions performed by the RAS signal. Since this failure of the LPSI pump to trip will not affect the valve operations, there will be no change to the dose in the Chapter 15 accident scenarios and the dose on or off site will not be increased. The failure mode of concern in modes 5 and 6 is the loss of shutdown cooling. The Loss of Shutdown Cooling is listed in UFSAR Section 15.1.27, which only directs the reader back to UFSAR Section 6.3, which discusses the loss of piping or pressure boundaries which would not be affected by this modification. The scenario where a pump is lost is discussed in UFSAR Section 9.3.6.3. The failure mode that this modification makes would only affect the operation of the pump and would not affect the dose analysis. Therefore, this modification will not result in more than a minimal increase in the consequences of an accident previously evaluated in the UFSAR.

4.

Result in more than a minimal increase in the consequences of a malfunction of a structure, system, or component important to safety previously evaluated in the UFSAR?

Yes No BASIS: This change installs a keylock switch which would be placed in the BYPASS position when the unit is in shutdown, Mode 5 or 6, and the LPSI pump was running to prevent a trip of the pump due to a RAS signal. One failure mode of concern would be the switch logic contact remaining open when the unit is in Modes 1 to 4 and the switch is in the normal position. This would prevent the RAS signal from tripping the LPSI pump. The other failure mode would be logic contact of the switch remaining closed when the switch in in the BYPASS position. This would trip the LPSI pump in shutdown cooling mode with an inadvertent RAS signal. Both of these failure modes are covered by the bounding conditions in the UFSAR. The consequence of the failure of a LPSI pump to trip with RAS actuation is discussed in UFSAR section 6.3.2.14.C. The loss of shutdown cooling is bounded by the current analysis in the UFSAR, section 9.3.6.3. Since the consequence of the malfunction is bounded by the current analysis, which includes the failure of a LPSI pump to trip during a RAS and the loss of shutdown cooling. This will not increase the consequence of malfunction of a structure, system or component important to safety previously evaluated in the UFSAR.

EN-LI-101 R22 5.

Create a possibility for an accident of a different type than any previously evaluated in the UFSAR?

Yes No BASIS: The accidents previously evaluated in the UFSAR bound all the events that could be caused by the failure in this circuit. This change installs a keylock switch which would be placed in the BYPASS position when the unit is in shutdown, Mode 5 or 6, and the LPSI pump was running to prevent a trip of the pump due to a RAS signal. One failure mode of concern would be the switch logic contact remaining open when the unit is in Modes 1 to 4 and the switch is in the NORMAL position.

This would prevent the RAS signal from tripping the LPSI pump. The other failure mode would be the logic contact of the switch remaining closed when the switch is in the BYPASS Position. This would trip the LPSI pump in shutdown cooling mode when a RAS signal was received. Both of these failure modes are covered by the bounding conditions in the UFSAR. Since both effects have been analyzed and are bounded by the analysis in the UFSAR, the addition of a bypass switch in the LPSI trip circuit will not create a possibility for an accident or a different type than any previously evaluated in the UFSAR.

6.

Create a possibility for a malfunction of a structure, system, or component important to safety with a different result than any previously evaluated in the UFSAR?

Yes No BASIS: The failure modes of failing to trip a LPSI pump during a RAS signal or the tripping of a LPSI pump in shutdown cooling with inadvertent RAS signal when the switch is in BYPASS position during shutdown are both bounded by the analyses in the UFSAR. Both of these events are within the bounds of the analysis of the UFSAR as discussed in sections 6.3.2.14.C and 9.3.6.3. The installation of this bypass switch will not create a malfunction of a system structure or component important to safety with a different result than any previous evaluated in the UFSAR.

EN-LI-101 R22 7.

Result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered?

Yes No BASIS: The addition of a keylock switch to bypass the LPSI pump trip on a RAS signal will not affect the design basis limit for a fission product barrier. The existing analyzed accidents are bounding, as discussed in UFSAR section 6.3.2.14.C and 6.3.2.20.4 for the lack of a LPSI trip during Modes 1-4. The RAS signal would still perform its function other than tripping the one LPSI pump and cooling would be provided in the Recirculation mode using the HPSI system. The failure mode during shutdown cooling, where the logic contact remains closed resulting in a LPSI pump trip during RAS testing, is encompassed by the discussion in UFSAR section 9.3.6.3, where the initiation of alternate means of shutdown cooling are discussed.

The limiting scenarios for shutdown cooling require alignment of other systems.

This is not a new failure mode of the system as the failure of the relay or bad lifted lead would cause this failure as it exists and is analyzed currently. The analysis in Section 9.3.6.3 of the UFSAR would still be more conservative and bounding. There is no change to failure modes of the systems as currently in place. Since there is no change to the existing limiting accident scenarios and conditions there would be no change to the limiting scenarios considered in the evaluation of ECCS per 10CFR50.46 fission product barriers. Therefore, this change of installing a bypass switch in the LPSI trip circuit will not result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered.

EN-LI-101 R22 8.

Result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses?

Yes No BASIS: This change will change the logic in the LPSI motor trip circuit. This modification adds a bypass switch preventing a RAS trip of the LPSI pump. This bypass switch will be used during testing while the unit is shutdown. The addition of a logic bypass switch will have no effect on analysis methods described in the UFSAR. This logic change will not change any method of evaluation as described in the UFSAR.

If any of the above questions is checked Yes, obtain NRC approval prior to implementing the change by initiating a change to the Operating License in accordance with NMM Procedure EN-LI-103.

ANO 50.59 Evaluation 2024-002

EN-LI-101 R22 ATTACHMENT 9.1 50.59 EVALUATION FORM Sheet 1 of 3 I.

OVERVIEW / SIGNATURES1 Facility: Arkansas Nuclear One-Unit 2 Evaluation # / Rev. #: FFN-2024-002 Proposed Change / Document: EC-54161217, Evaluation to Support 2CV-1050-2 Normally Closed Description of Change:

EC-54161217 evaluates changing the emergency feedwater (EFW) green train motor operated steam supply valve, 2CV-1050-2, from normally open to normally closed. Additionally, EC-54161217 will add an Emergency Feedwater Actuation Signal (EFAS) to the 2CV-1050-2 circuitry to open the valve automatically when required. The intent of this change is by maintaining the green train valve closed, the cycling differential pressures between the steam generators during normal operation can be avoided and the subsequent valve chattering of 2MS-39A and 2MS-39B will be eliminated (CR-ANO-2-2024-01748). EC-54161217 thus affects a method of performing or controlling the design function of EFW and its associated valves downstream to the steam generator (SGs) thereby leading to the 10 CFR 50.59 evaluation.

The EFW system will be enhanced with a modification which will limit the cycling of 2MS-39A and 2MS-39B by closing 2CV-1050-2. 2MS-39A and 2MS-39B are spring loaded open and close when there is backflow out of the steam header. The check valves serve to maintain pressure in the EFW steam supply lines in case of an upstream failure. 2CV-1050-2 is in the Main Steam (MS) supply piping header that supplies steam to the turbine driven EFW pump turbine (2K-3). 2CV-1050-2 is directly upstream of nozzle check valve 2MS-39B that has previously experienced degradation due to continuous cycling of the valve at full power operation.

If 2CV-1050-2 is to remain normally closed, then an automatic function to open must be added to comply with the Final Safety Analysis Review (FSAR) design requirements. EFAS controls the EFW actuation and a new signal will be added to open 2CV-1050-2. This will be accomplished by replacing the control room hand switch 2HS-1050-2 from a 3-position spring return-to-center style operator to a 3-position maintained operator hand switch. The new switch will be located in the same location. An existing EFAS relay with extra contacts will be connected to 2HS-1050-2 to provide an EFAS #2 (EFAS-2) signal to acuate the open logic. An existing cable will be used between control room cabinets. Additional wiring will be added inside the control room cabinets.

2HS-1050-2 will be able to override the open signal by placing the hand switch to the closed position for isolation.

As discussed the FSAR, 2K04-F2 currently alarms when 2CV-1050-2 is in the closed position using a limit switch to energize auxiliary relay 2ZSX-1050-2. 2ZSX-1050-2 uses a normally open contact that will be closed when the valve is open and opens when the valve is in the closed position, causing 2K04-F2 to annunciate. With 2CV-1050-2 taken to a normally closed position in this modification, this portion of the alarm circuit must be defeated when the valve is closed and no EFAS-2 signal is present. This is accomplished by a spare normally open contact from 2C40-K303 in series with a contact from handswitch 2HS-1050-2, which are both placed in parallel with the existing logic for annunciation. This is acceptable as it drives the alarm as needed for the new configuration of 2CV-1050-2. The original intent of the alarm function when EFAS-2 is actuated or when the valve is manually opened is maintained.

The modification will also install a thermal overload bypass for the 2CV-1050-2 breaker control circuit utilizing an existing spare contact from EFASX3-B-2 located in 2B63. Additional wiring will be added in

EN-LI-101 R22 2B63 in accordance with specifications. The overload bypass will be applied when 2CV-1050-2 receives the EFAS-2 signal to open. This EC will also change the torque switch bypass limit switch such that the torque switch will be bypassed for almost the entire valve opening stroke. Both changes satisfy Unit 2 UFSAR criteria as stated in section 8.3.1.1.8.11.11.

Summary of Evaluation:

The condition identified in CR-ANO-2-2024-01748 regarding the failure of the check valve 2MS-39B due to prolonged cross flow in the EFW supply lines will be eliminated. The design function of the EFW system to provide sufficient flow to remove decay heat load after a design bases event by delivering water to the SGs is not impacted by the proposed modification. The design function of 2CV-1050-2 as a GDC 57 closed system isolation valve is maintained. The requirement to automatically operate the EFW system with the addition of an EFAS-2 signal to open 2CV-1050-2 is maintained. An alarm to indicate when 2CV-1050-2 is not in the correct normal position is being maintained. This modification maintains the licensing requirements for the EFW system while enhancing the reliability to maintain system design functions.

Is the validity of this Evaluation dependent on any other change?

Yes No es Evaluation cannot be implemented without approval of the other identified changes (e.g., license amendment request). Establish an appropriate notification mechanism to ensure this action is completed._________________________________________________________________________

Based on the results of this 50.59 Evaluation, does the proposed change require prior NRC approval?

Yes No Preparer2:

Chris Walker / / EOI / Safety Analysis / See Signature Name (print) / Signature / Company / Department / Date Reviewer2:

Matt Montgomery / / EOI / Integrated Risk / See Signature Name (print) / Signature / Company / Department / Date Independent Review3:

N/A Name (print) / Signature / Company / Department / Date Responsible Manager Concurrence:

Name (print) / Signature / Company / Department / Date 50.59 Program Coordinator Concurrence:

Name (print) / Signature / Company / Department / Date 1 The printed name should be included on the form when using electronic means for signature or if the handwritten signature is illegible. Signatures may be obtained via electronic authentication, manual methods (e.g., ink signature),

e-mail, or telecommunication. Signing documents with indication to look at another system for signatures is not Enterprise Asset Management (EAM) Application." Electronic signatures from other systems are only allowed if they are included with the documentation being submitted for capture in eB (e.g., if using an e-mail, attach it to this form; if using Enterprise Asset Management (EAM) Application, attach a screenshot of the electronic signature(s); if using Corrective Action Program (CAP) Application, attach a copy of the completed corrective action).

2 Either the Preparer or Reviewer will be a current Entergy employee.

3 If required by Section 5.1[2].

/EOI/Design Engineering/See signature Digitally signed by Matthew T. Montgomery DN: cn=Matthew T. Montgomery, c=US, o=Integrated Risk, ou=Entergy, email=mmontg4@entergy.com Date: 2024.10.22 16:24:21 -05'00' Digitally signed by Vincent Bond DN: cn=Vincent Bond, c=US, o=Project Engineering, ou=Engineering Manager, email=vbond@entergy.com Date: 2024.10.22 17:03:12 -05'00' Vincent Bond Digitally signed by Michael Hall DN: cn=Michael Hall, c=US, o=Entergy Operations Inc., ou=ANO Regulatory Assurance, email=mhall10@entergy.com Date: 2024.10.22 17:13:07 -05'00' Michael Hall

EN-LI-101 R22 ATTACHMENT 9.1 50.59 EVALUATION FORM Sheet 2 of 3 OSRC:

[GGNS P-33633, P-34230, & P-34420; W3 P-151]

OSRC Meeting #

II.

50.59 EVALUATION [10 CFR 50.59(c)(2)]

Does the proposed Change being evaluated represent a change to a method of evaluation ONLY 7 are not applicable; answer only Question 8.

Yes No Does the proposed Change:

1.

Result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the UFSAR?

Yes No BASIS:

The design function of steam supply valve 2CV-1050-2 is to ensure the capability of supplying cooling water to the SGs during emergency conditions and to close for SG isolation. Changing the position of valve 2CV-1050-2 from normally open to normally close does not affect the design function of the valve or the EFW system. By modifying the initial position of 2CV-1050-2 to closed during normal plant operations, the cyclic flow between SGs will be prevented which will reduce the strain on the 2MS-39A and 2MS-39B check valves, increasing their reliability and the reliability of the system as a whole. During a loss of offsite power, the Motor Operated Valve (MOV) receives power from an emergency diesel generator. The modification introduces an open signal upon EFAS-2 actuation to automatically open 2CV-1050-2, consistent with other valves in the EFW system required to change state during upset conditions. The addition of the EFAS signal does not change the design function of the EFAS, which is to monitor plant operating conditions and to initiate engineered safety features (ESF) operation in the event of an accident. The modification will utilize an existing EFAS relay with spare contacts and existing spare cable between the control room cabinets. No change is made to the location of the 2CV-1050-2 hand switch. A thermal overload bypass will be added in accordance with FSAR utilizing an existing relay. Additional connecting wires will be added inside the control room cabinets and 2B63. The alarm circuitry will be modified from the current logic to alarm if CV-1050-2 is not in the correct position. 2CV-1050-2 will still be closed manually from the control room overriding the EFAS-2 signal if needed to isolate the 2E-24B.

The modification will not increase the failure rate of any accident evaluated in Chapter 15 of the FSAR.

==

Conclusion:==

The modification is being performed to comply with requirements outlined in the FSAR for the steam supply lines valves. The changes to the plant are minimal and do not impact the design function of the turbine driven EFW system. The upgrade is intended to increase the reliability of the overall system. The design function of the EFW system is to respond to design basis accident and not create an accident. The enhancement to the EFW system does not result in more than a minimal increase in the frequency of occurrence of accidents evaluated in the FSAR.

OSRC-2024-012 Digitally signed by Riley Keele DN: cn=Riley Keele, c=US, email=rkeele@entergy.com Reason: I am approving this document Date: 2024.10.24 11:44:26 -05'00' Riley Keele

EN-LI-101 R22

2.

Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the UFSAR?

Yes No BASIS:

The design function of steam supply valve 2CV-1050-2 is to ensure the capability of supplying cooling water to the Steam Generators during emergency conditions and to close for SG isolation.

Changing position of valve 2CV-1050-2 from normally open to normally close does not affect the design function of the valve or the EFW system. 2CV-1050-2 is designed to operate in either the open or closed position and can fail in either position. Closing the valve during normal plant operations will eliminate the cyclic flow between SG through the steam supply lines. Preventing this flow will reduce the strain and wear on the 2MS-39A and 2MS-39B components increasing their reliability. 2MS-39A and 2MS-39B are spring loaded open and close when there is backflow out of the steam header. With 2CV-1000-1 remaining open during normal operations, pressure and temperature are allowed to equalize in the turbine driven EFW pump supply line minimizing thermal and steam shock to the EFW pump turbine. There are existing steam traps in the system which prevent water buildup around and past 2CV-1050-2. Additionally, by preventing water accumulation in the line, the concerns raised by Generic Letters (GL) 89-10

-Related Motor Operated Valve and GL 95-07, "Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves" regarding potential binding of power operated gate valves is adequately addressed. The valve closing time requirements will be unaffected because the valve is assumed to be open when required for isolation. The valve opening times will satisfy the requirement for opening during a design basis event. The postulated pipe rupture of the EFW turbine steam line downstream of 2CV-1050-2 flow rates would be reduce if 2CV-1050-2 is closed.

The MOV and power cabling outside the control room will remain unchanged with the valve remaining diesel backed. The operation to open 2CV-1050-2 is already included in the emergency diesel loading schemes. The EFAS system already provides the signals to actuate and control the EFW system. The modification introduces an open signal upon EFAS-2 actuation to automatically open 2CV-1050-2. Providing an open signal provides assurance at least one of the steam supply valves is open (considering the possibility of valve misposition this increases the defense-in-depth of the system). The modification complies with the operating requirements outlined in the FSAR.

The modification EFAS circuitry will be performed without changing the design function of the EFAS which is to monitor plant operating conditions and to automatically initiate engineered safety features operation in the event of an accident. The modification will utilize an existing EFAS relay with spare contacts and existing spare cable between the control room cabinets. The 2CV-1050-2 hand switch will be located in the same location within the control room. Additional connecting wiring will be added inside the control room cabinets and 2B63F1 in accordance with specifications. The alarm circuitry will be modified to alarm if CV-1050-2 is not in the correct position.

Evaluating the change at the component level, the 2020 Component Unreliability Data maintained by Idaho National Laboratory, INL/EXT-21-65055, states that the US nuclear industry has had 190 failures of an MOV to open in 593,626 demands between 2006 and 2020. Hence, the frequency of failure at a component level will increase by 3.43E-4 (mean) failures per demands. The same dataset states that MOVs failed to close 123 times in 593,626 demands between 2006 and 2020, or 2.28E-4 (mean) failures per demands. The replacement of one failure mode for the other, therefore, would not constitute an increase by more than a factor of two as prescribed in NEI 96-07.

From a probabilistic standpoint, incorporation of this new failure mode provides a negligible increase in overall plant risk, as described in the quantitative risk evaluation provided in EC-54161217. This evaluation yielded an estimated impact of 3.51E-8 delta CDF and 2.80E-10 delta LERF, considering internal events, internal flooding, and fire. This is well within the limitations imposed in RG 1.174, Rev. 3.

==

Conclusion:==

EN-LI-101 R22 The changes to the plant do not impact the design function of the turbine driven EFW system or the EFAS system. The ability to remotely close 2CV-1050-2 from the control room is unaffected. The upgrade is intended to increase the reliability of the overall system and decreasing the potential for an EFW system failure. The enhancement to the EFW system does not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC evaluated in the FSAR.

3.

Result in more than a minimal increase in the consequences of an accident previously evaluated in the UFSAR?

Yes No BASIS:

Section 15 of the ANO-2 FSAR describes the accidents for which the EFW and EFAS Systems were designed to mitigate. The EFW system is relied upon for most accident scenarios if there is a reactor trip or loss of off-site power. The bounding accidents are discussed below.

FSAR Section 15.1.8 describes the Loss of Normal Feedwater Flow (LOFW) where feedwater flow is completely stopped and SG inventory starts to reduce. The EFW system is required to respond to the event within 97.4 seconds. It has two independent steam supplies to the turbine driven EFW pump. With 2CV-1000-1 open, steam will already be available to 2K-3 EFW pump turbine. With this modification, 2CV-1050-2 will automatically open with an EFAS-2 signal within 33.183 seconds (31.9 second stroke open + 1.283 second signal delay). This does not consider the 15 second diesel load delay. Both steam supplies will be available to satisfy the cooling requirements. The modification will have no impact on the LOFW accident.

FSAR Section 15.1.10 describes Excess Heat Removal Due to Secondary System malfunction. The emergency feed water system could possibly initiate while at power and send cold water from the Q condensate storage tank (QCST) to the SGs. The modification to have 2CV-1050-2 normally closed and automatically open with an EFAS signal will have no effect on the capability of the turbine driven EFW pump. The turbine driven EFW system will function as designed and the redundant steam supply is maintained. Failure of 2CV-1050-2 will not cause the EFW turbine driven pump to operate when not desired.

FSAR Section 15.1.12 describes internal and external events including major and minor fires, floods, storms, and earthquakes. The impacts to each of these events is assessed here:

Fire: The modification makes no physical changes to 2CV-1050-2 aside from its normal position. A small amount of wiring was added to two control room cabinets/2B63 and a slightly different valve control switch will be used. The fire impacts due to additional components were evaluated in accordance with the Fire Protection Program and the impact was determined to be acceptable via EN-DC-128.

Floods: The modification makes no physical changes to 2CV-1050-2 aside from its normal position. The failure of the valve to open with a concurrent break of the EFW steam supply piping would cause the flow out the break to be reduced and prevent both SGs from losing inventory. Once a break is located operations will take both the supply valves to closed with no impact on flooding.

Storms: The bounding storm is a tornado which will have no impact on the changes made to the EFW system. This EC is not rerouting any piping and existing conduit is being used; therefore, no new tornado missile or tornado missile target is being added.

Earthquakes: An existing EFAS relay with small amount of wiring will be added to seismically qualified cabinets within the control room and 2B63. The 2CV-1050-2 control switch will also be updated. These changes will be made in accordance seismic design requirements and will not have any impact on existing control room components.

EN-LI-101 R22 FSAR Section 15.1.13 describes the Major Rupture of Pipes containing Reactor Coolant (LOCA).

2CV-1050-2 is an isolation valve on penetration number 2P2 which meets the requirements of GDC

57. However, this penetration serves an engineered safeguard function and must be un-isolated to support this function; automatic isolation is therefore inappropriate. The changes made to the control system for 2CV-1050-2 will not impact the valve from being closed from the control room, and GDC 57 continues to be met. When taking the valve manually closed with the selector switch the EFAS signal would be overridden.

FSAR Section 15.1.14.1 describes a Main Steam Line Break Accident (MSLB) where a major secondary steam line pipe breaks coincident with a Loss of AC Power (LOAC) and turbine trip. The break can occur either inside or outside containment. The EFW system will respond once inventory in the unaffected SG reaches the low inventory trip. For the overcooling event, the EFW response time is assumed to be relatively fast. If 2E-24A blows down, 2MS-39A will close and prevent the blowdown of 2E-24B and steam flow to the turbine driven EFW pump will be supplied once 2CV-1050-2 opens. If 2E-24B blows down, 2MS-39B will close and prevent the blowdown of 2E-24A, while steam flow to the turbine driven EFW pump will be supplied with an already-open 2CV-1000-1.

Placing 2CV-1050-2 in a normally closed position will increase the reliability of both 2MS-39A and 2MS-39B and ensure that they function properly. EFW continues to perform its design function as required.

FSAR Section 15.1.14.2 describes a Feedwater Line Break Accident (FWLB) where there is a rupture of the main feedwater system pipe with a LOAC. A FWLB quickly reduced SG inventories and EFW must respond within 112.4 seconds. The motor driven EFW pump is assumed to be failed.

Assuming 2E-24A will be isolated the system will rely upon 2CV-1050-2 to open within 48.183 seconds (33.183 response + 15 seconds diesel load). EFW continues to perform its design function as required.

FSAR Section 15.1.18 describes the Steam Generator Tube Rupture (SGTR) where a double ended tube rupture occurs in one of the SG and a LOAC is also assumed. The Reactor Coolant System (RCS) charging system will not be able to maintain RCS pressure. The failed SG will not be isolated until operator action occurs as credited in the analysis. The EFW system will respond once inventory in the unaffected SG reaches the low inventory trip. For the SGTR event the EFW response time is assumed to be relatively fast. Steam flow to the turbine driven EFW pump will be supplied by both trains prior to the affected SG being isolated. Operators will isolate the failed SG as soon as possible, which has been analyzed and validated by operating crews to be less than one hour. The EFW can perform its design function as required.

For its isolation function during an SGTR, 2CV-1050-2 will initially open upon EFAS signal and the valve will be closed within one hour by operators to isolate the affected SG provided the rupture takes place on 2E-24B. In this scenario the valve failing closed is beneficial. However, if the rupture were to take place on 2E-24A and the single active failure is taken to be 2CV-1050-2 failing to open, only contaminated steam would be used to drive 2K-3 for the first hour of the transient, as opposed to a mixture of steam from both SGs in the previous design. This scenario is bounded by the existing SGTR dose analysis.

==

Conclusion:==

Based on reviewing the impacted accident scenarios, placing 2CV-1050-2 in a normally closed position will not increase the consequences of an accident previously evaluated in the FSAR.

4.

Result in more than a minimal increase in the consequences of a malfunction of a structure, system, or component important to safety previously evaluated in the UFSAR?

Yes No BASIS:

EN-LI-101 R22 The modification to the control circuits by adding an EFAS-2 open signal is a reliability enhancement to the existing system as described in the FSAR. The upgrades to the turbine driven EFW system do not change the function or performance requirements of the EFW system described in the FSAR.

The modification does not modify or alter plant operating parameters that would result in and increase challenges to components important to safety. EFAS and EFW function together and adding an EFAS-2 signal to open 2CV-1050-2 does not create any new interface requirement with SSCs important to safety that function to limit the consequences of an accident. If 2CV-1050-2 valve failed to reposition when required, there are alternate means of core cooling besides the turbine driven EFW pump. Consequences of a malfunction of the turbine driven EFW pump have previously been evaluated.

As defense in depth, from a probabilistic standpoint, incorporation of this new failure mode provides a negligible increase in overall plant risk, as described in the quantitative risk evaluation provided in EC-54161217. This evaluation yielded an estimated impact of 3.51E-8 delta CDF and 2.80E-10 delta LERF, considering internal events, internal flooding, and fire. This is well within the limitations imposed in RG 1.174, Rev. 3.

==

Conclusion:==

The upgrade to the turbine driven EFW system steam supply does not result in more than a minimal increase in the consequences of a malfunction of a SSC important to safety.

EN-LI-101 R22 ATTACHMENT 9.1 50.59 EVALUATION FORM Sheet 3 of 3

5.

Create a possibility for an accident of a different type than any previously evaluated in the UFSAR?

Yes No BASIS:

The modification to the EFW system changes the position of the 2CV-1050-2 steam supply valve from normally open to normally closed. EFAS provides the signals to initiate the EFW system by automatically opening the necessary valves which now will include 2CV-1050-2. The failure of 2CV-1050-2 could lead to a loss of one of the redundant steam supply sources to the turbine driven pump, namely 2E-24B. The modification used one of the existing EFAS relays with an existing cable. The 2CV-1050-2 hand switch is being replaced with a very similar design in the same location. A thermal bypass contact will be utilized in 2B63. These changes do not create an avenue where a previously incredible accident is now credible.

==

Conclusion:==

The upgrade to the Unit 2 turbine driven EFW system does not create the possibility of an accident of different type than previously evaluated in the ANO FSAR.

6.

Create a possibility for a malfunction of a structure, system, or component important to safety with a different result than any previously evaluated in the UFSAR?

Yes No BASIS:

Changing 2CV-1050-2 to a normally closed state requiring an EFAS signal to open does not introduce a new limiting structure, system, or component failure not already bounded by an evaluation in the UFSAR. The worst-case credible malfunction for this modification is a loss of steam supply to the EFW turbine pump and existing analyses continue to bound this failure: taking the single failure of the 2CV-1050-2 valve to correctly change position would still maintain the flow path of the red train via 2CV-1000-1. The EFAS system already provides opening signals to the EFW system.

==

Conclusion:==

The upgrade to the turbine driven EFW system in conjunction with the EFAS interface additions does not create a possibility for a malfunction of an SSC important to safety with different results than previously evaluated in the FSAR.

7.

Result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered?

Yes No BASIS:

The review of the Chapter 15 accidents and malfunctions to address Question 3 concluded that all design basis limits for fission product barriers continue to be met as a result of this change.

==

Conclusion:==

The change for the operation of the 2CV-1050-2 does not result in a design basis limit for a fission product barrier as descripted in the FSAR being exceeded or altered.

EN-LI-101 R22

8.

Result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses?

Yes No BASIS:

The change in normal position of valve 2CV-1050-2 from open to close does not impact any evaluations that demonstrate the adequacy of the EFW system or the EFAS system as described in Unit 2 SAR. The proposed modification evaluates design basis calculations for adverse impacts and concludes that none of the analysis assumptions or resulting margins are adversely impacted by the change in normal valve position and the EFAS circuitry.

==

Conclusion:==

This change does not result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses.

change by initiating a change to the Operating License in accordance with NMM Procedure EN-LI-103.

ANO 50.59 Evaluation 2024-003

EN-LI-101 R22 ATTACHMENT 9.1 50.59 EVALUATION FORM Sheet 1 of 8 I.

OVERVIEW / SIGNATURES1 Facility: _ANO Unit 2____________________

Evaluation # / Rev. #: ___________________

Proposed Change / Document: EC 87349 ANO-2 ICC [RVLMS/CET/SMM] DAS REPLACEMENT Description of Change:

Arkansas Nuclear One Unit 2 (ANO-2) is replacing the obsolete Inadequate Core Cooling Monitoring System (ICCMS) Data Acquisition System (DAS) and the Sub-Cooled Margin Monitoring (SMM) system with a Westinghouse-supplied Post Accident Monitoring System (PAMS). The ICCMS DAS and SMM are currently two physically discrete subsystems, each with their own field terminations, processing electronics, local displays, and physical enclosures. As a part of this upgrade, these two subsystems will be "combined" (but remain electrically separated) and all functions (including display) will be performed by the replacement PAMS. The upgrade scope is limited to the aforementioned equipment and does not include any modification to the existing control room panel-mounted meters or recorders, however the replacement PAMS will provide outputs to support them. The customer's field signals will be used as-is in the replacement PAMS.

The replacement PAMS is a safety-related two channel system, which implements the Core Exit Thermocouple (CET) monitoring, Reactor Vessel Level Measurement (RVLM), and the SMM functions.

Each channel of the PAMS is comprised of a single cabinet that will replace the existing ICCMS DAS cabinets (2C388-1&2). The new PAMS cabinets will assume the existing cabinet tag numbers (2C388-1&2). The cabinet containing the SMM (2C336-3&4) will remain in place and the required input signals (2 RCS Hot Leg temps, 2 RCS Cold leg temps, and RCS Pressure) will be rerouted to the replacement PAMS cabinets. Each PAMS cabinet consists of an Advant Controller 160 (AC160) processing rack for I/O and PAMS calculations, a Maintenance and Test Panel (MTP) for local display and datalink output, and associated power supplies and cabling.

The functions provided by the upgraded PAMS replicates all functions committed to by ANO-2 in its licensing basis, specifically the display of relevant Post Accident variables in SAR Table 7.5-3. The replacement PAMS, including the MTP are based on the NRC-approved (ML21140A104) Common Qualified (Common Q) Platform, which is described in WCAP-16097-P-A, "Common Qualified Platform Topical Report" (Reference 7). Furthermore, the algorithms used for CET monitoring and SMM are based on the standard Common Q algorithms documented in topical report, WCAP-16097-P-A, Appendix 1, "Common Qualified Platform Post Accident Monitoring Systems" (Reference 8). Relevant differences and enhancements from the existing ICCMS DAS and SMM to the WEC-supplied PAMS, including display schemes and data processing are discussed in the following sections.

PAMS Display The existing displays for Post Accident variables consist of the non-safety related Safety Parameters Display System (SPDS) located on the main control board, the 4-line plasma Keypad Display Unit (KDU) on the outside of ICCMS DAS cabinet for RVLMS and CET monitoring, and the one-line digital display on the outside of the SMM cabinet. The SPDS has been designed for high reliability and provides the RVLMS and CET information to the operators on the main control board through computer-generated graphics, unlike the KDU. As such, the SPDS is considered the "primary display" in ANO-2's NUREG-0737, Item II.F.2 related licensing correspondence. As a part of the upgrade, the PAMS will provide a datalink output to support the existing SPDS and will replace the existing KDU and SMM display with a 1

The printed name should be included on the form when using electronic means for signature or if the handwritten signature is illegible. Signatures may be obtained via electronic authentication, manual methods (e.g., ink signature), e-mail, or telecommunication. Signing documents with indication to look at another system for signatures is not acceptable such as See EC or "See Enterprise Asset Management (EAM) Application." Electronic signatures from other systems are only allowed if they are included with the documentation being submitted for capture in eB (e.g., if using an e-mail, attach it to this form; if using Enterprise Asset Management (EAM) Application, attach a screenshot of the electronic signature(s); if using Corrective Action Program (CAP) Application, attach a copy of the completed corrective action).

FFN2024003

EN-LI-101 R22 ATTACHMENT 9.1 50.59 EVALUATION FORM Sheet 2 of 8 15" color touchscreen flat panel display mounted on the outside of the replacement PAMS cabinet. The Flat Panel Display (FPD) consists of configurable, human factors designed screens that present the same variables as the existing system but offers a more enhanced user experience than can be provided by a 4-line display and keypad.

Reactor Vessel Level Monitoring The RVLMS algorithm provided by the PAMS will replicate the functionality of the existing algorithm with the exception of the reactor core power measurement capability of the existing system. In certain core conditions, the existing level probe has the ability to measure fuel surface heat transfer conditions and therefore measure core power generation. These were additional features/capabilities of the probes but are not functions required to meet NUREG-0737, Clarification of TMI Action Plan Requirements (Reference 4) for ICC detection. These features are not used by the plant operations and will therefore not be retained by the replacement PAMS.

Saturation Margin Monitoring The existing SMM calculates RCS temperature and pressure margin to saturation using RCS temperatures and pressure. The SMM provided in the PAMS replacement also calculates RCS temperature and pressure margin as well as enhanced capabilities to calculate margins using CET temperature and reactor dome temperature for those relative vessel locations. The ANO-2 Safety Analysis Report (SAR) (Reference 3) lists the subcooling margin range (called degrees of subcooling in Table 7.5-3 of the SAR) as 0-200 deg F. However, the current system recorder and replacement system have a subcooling margin range of 0-100 deg F.

Core Exit Thermocouple Monitoring The replacement PAMS replicates the functionality of the existing CET monitoring system by processing all CET values and displaying them in a spatially oriented manner by location in the core, as well as providing the values to the existing SPDS for display and further processing. The temperature range for the CET has changed from 0 - 2300 degrees F to 32 - 2300 degrees F. In addition, the replacement CET algorithm (currently referred to as average CET) provides a Representative CET temperature, which is a statistical average of all valid CETs in the channel and calculates a temperature that is higher than 95%

of the total valid CET temperatures.

Summary of Evaluation:

PAD Screen Question # 1 concluded an evaluation needs to be performed. This is because the ANO-2 PAMS is a digital upgrade using Common Q, which is a microprocessor/software based platform.

Due to this, there exists the possibility of introducing a software common cause failure in the redundant PAMS channels, which could impact the design functions of the system. In addition, the replacement PAMS combines the functionality of the existing ICCMS-DAS and SMM. Therefore, this is a change to an SSC that adversely affects an UFSAR described design function and needs to be evaluated.

All of the eight evaluation questions listed below were addressed and all eight questions were answered no. The discussion for each question is provided. The conclusion of this evaluation is that no License Amendment is required; the changes may be made without NRC approval.

EN-LI-101 R22 ATTACHMENT 9.1 50.59 Evaluation Form Sheet 3 of 8 Is the validity of this Evaluation dependent on any other change?

Yes No If Yes, list the required changes/submittals. The changes covered by this 50.59 Evaluation cannot be implemented without approval of the other identified changes (e.g.,

license amendment request). Establish an appropriate notification mechanism to ensure this action is completed.

Based on the results of this 50.59 Evaluation, does the proposed change require prior NRC approval?

Yes No Preparer2:

Brad Miller/

/EOI/ Central Design Engineering/ 11/18/24 Name (print) / Signature / Company / Department / Date Reviewer2:

Greg Svestka/

/EOI/ Plant Support Engineering RBS/ 11/18/24 Name (print) / Signature / Company / Department / Date Independent Review3:

NA Name (print) / Signature / Company / Department / Date Responsible Manager Concurrence:

Name (print) / Signature / Company / Department / Date 50.59 Program Coordinator Concurrence:

Name (print) / Signature / Company / Department / Date OSRC:

Chairmans Name (print) / Signature / Date [GGNS P-33633, P-34230, & P-34420; W3 P-151]

OSRC Meeting #

2 Either the Preparer or Reviewer will be a current Entergy employee.

3 If required by Section 5.1[2].

OSRC2024014

/EOI/Design Engineering/11/18/24 Digitally signed by Brad Miller DN: cn=Brad Miller, c=US, o=ANO Electrical Design, ou=Entergy Operations Inc.,

email=jmille3@entergy.com Date: 2024.11.18 16:13:48 -

06'00' Brad Miller Digitally signed by Gregory K Svestka DN: cn=Gregory K Svestka, c=US, o=EFIN, ou=Entergy Operations Inc.,

email=gsvestk@entergy.com Date: 2024.11.18 16:20:49 -

06'00' Gregory K Svestka Digitally signed by Vincent Bond DN: cn=Vincent Bond, c=US, o=Project Engineering, ou=Engineering Manager, email=vbond@entergy.com Date: 2024.11.18 17:06:21 -06'00' Vincent Bond Digitally signed by Michael Hall DN: cn=Michael Hall, c=US, o=Entergy Operations Inc.,

ou=ANO Regulatory Assurance, email=mhall10@entergy.com Date: 2024.11.25 10:24:20 -06'00' Michael Hall Date: 11/27/24 8:57:45 AM Paul Butler

EN-LI-101 R22 ATTACHMENT 9.1 50.59 EVALUATION FORM Sheet 4 of 8 II.

50.59 EVALUATION [10 CFR 50.59(c)(2)]

Does the proposed Change being evaluated represent a change to a method of evaluation ONLY? If Yes, Questions 1 - 7 are not applicable; answer only Question 8.

If No, answer all questions below.

Yes No Does the proposed Change:

1.

Result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the UFSAR?

Yes No BASIS:

The accident analysis in Chapter 15 of the ANO-2 SAR (Reference 3) was reviewed. The existing ICCMS-DAS and SMM, and the replacement ANO-2 PAMS are not accident initiators to any of the Chapter 15 events since they are only display systems and do not have any control functions. The ANO-2 PAMS is used in post accident scenarios. However, according to Table 7.5-3 of the ANO-2 SAR (Reference 3), the variables associated with the PAMS (SMM, CET, and RVLMS) are considered RG 1.97 Type A and B variables. The operator does not rely on these Type B variables for any safety critical actions. In the case of the Type A variables (RCS Hot Leg Water Temperature and RCS Pressure), the operator has primary indication (i.e., safety related indicators and recorders) and does not solely rely on the PAMS for this information. Therefore, this activity does not result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the UFSAR.

EN-LI-101 R22 ATTACHMENT 9.1 50.59 EVALUATION FORM Sheet 5 of 8

2.

Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the UFSAR?

Yes No BASIS:

PAD Screen Question #1 concluded that the modification adversely affects a UFSAR described design function because the change involves a digital modification, and because the new PAMS combines the functionality of the SMM calculator and ICCMS-DAS. WNA-LI-00107-CARK2, Qualitative Assessment for ANO-2 Post Accident Monitoring System 10 CFR 50.59 Evaluation (Reference 1), documents a Qualitative Assessment that was performed in accordance with the guidance of RIS 2002-22, Supplement 1, Clarification on Endorsement of Nuclear Energy Institute Guidance In Designing Digital Upgrades in Instrumentation and Control Systems (Reference 2). This Qualitative Assessment concluded that based on design attributes employed, the design process of the ANO-2 PAMS hardware and software, and review of the Common Q platform operating experience, it is reasonable to conclude that the ANO-2 PAMS will exhibit a sufficiently low likelihood of introducing a systematic failure due to a latent design defect. Section 4 of WNA-LI-00107-CARK2 (Reference 1) also discusses how the SMM calculation is still available if the PAMS fails.

The modification is replacing the ICCMS-DAS datalink to the SPDS with a new design as described in Sections 5.1.4.1 and 5.1.4.5 of WNA-LI-00107-CARK2 (Reference 1). Per Section 5.1.4.1 the new design ensures failure or loss of the data link does not prevent the PAMS channel from performing its safety function. Per Section 4 of Reference 1 the PAMS is designed so that any single failure, in either channel, will not prevent proper monitoring, display and alarm action of the other PAMS channels, or inhibit operation of any other system at the system level. The FMEA for this system shows that no single failure will defeat more than one of the two redundant PAMS channels.

A reliability analysis was performed that shows the ANO-2 PAMS meets 99% availability as required by NUREG-0737, Clarification of TMI Action Plan Requirements (Reference 4). The results of the reliability analysis are documented in WNA-AR-01139-CARK2, Post Accident Monitoring System -

Reliability Analysis (Reference 5). The analysis has shown that the estimated unavailability of PAMS is 7.109E-04. The availability is calculated to be 0.9992891 or 99.92891 percent which surpasses the required availability of 99 percent.

Therefore, this activity does not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the UFSAR.

EN-LI-101 R22 ATTACHMENT 9.1 50.59 EVALUATION FORM Sheet 6 of 8

3.

Result in more than a minimal increase in the consequences of an accident previously evaluated in the UFSAR?

Yes No BASIS:

The existing ICCMS-DAS and SMM, and the replacement ANO-2 PAMS are not accident initiators to any of the Chapter 15 events since they are only display systems and do not have any control functions. The existing ICCMS-DAS and SMM, and the replacement ANO-2 PAMS are used in post accident scenarios. However, according to Table 7.5-3 of the ANO-2 SAR (Reference 3), the variables associated with the PAMS (SMM, CET, and RVLMS) are considered RG 1.97 Type A and B variables. The operator does not rely on these Type B variables for any safety critical actions. In the case of the Type A variables (RCS Hot Leg Water Temperature and RCS Pressure), the operator has primary indication (i.e., safety related indicators and recorders) and does not solely rely on the PAMS for this information. Section 4 of WNA-LI-00107-CARK2 (Reference 1) also discusses how the SMM calculation is still available if the PAMS fails and does not result in a failure mode which would create a different result in terms of single failure criterion.

In addition the following ANO-2 Emergency Operating Procedures (EOPs) and Abnormal Operating Procedures (AOPs) were reviewed:

OP-2202.002, "Reactor Trip Recovery" OP-2202.003, "Loss of Coolant Accident" OP-2202.004, "Steam Generator Tube Rupture" OP-2202.005, "Excess Steam Demand" OP-2202.006, "Loss of Feedwater" OP-2202.007, "Loss of Offsite Power" OP-2202.008, "Station Blackout" OP-2202.009, "Functional Recovery" OP-2202.010, "Standard Attachments" OP-2202.011, "Lower Mode Functional Recovery"

• OP-2203.011, "RCS Overcooling" OP-2203.013, "Natural Circulation Operations" OP-2203.014, "Alternate Shutdown" OP-2203.016, "Excess RCS Leakage" OP-2203.029, "Loss of Shutdown Cooling" OP-2203.030, "Remote Shutdown" OP-2203.038, "Primary to Secondary Leakage" These procedures were identified to rely on the ICCMS-DAS variables in post accident scenarios.

The replacement PAMS provides the same variables as the existing ICCMS-DAS, and therefore would support the existing EOPs and AOPs. The HFE evaluation, ANO-2-2024-0005, Transmittal #8 (ANO-2 to WEC) Addressment of PSAIs that require Entergy responses within Westinghouse Qualitative Assessment (Reference 6) addresses additional operating procedures reviewed and identified OP-2202.009, "Functional Recovery" and OP-2203.013, "Natural Circulation Operations" requiring modification.

Therefore, because the changes identified under this 50.59 are not an initiator of any accidents and no failure modes with different results are introduced, the changes do not introduce the possibility of minimal increase in the consequences of an accident previously evaluated in the UFSAR.

EN-LI-101 R22 ATTACHMENT 9.1 50.59 Evaluation Form Sheet 7 of 8

5.

Create a possibility for an accident of a different type than any previously evaluated in the UFSAR?

Yes No BASIS:

The existing ICCMS-DAS and SMM, and the replacement ANO-2 PAMS are not accident initiators since they are only display systems and do not have any control functions. The ANO-2 PAMS is used in post accident scenarios. However, according to Table 7.5-3 of the ANO-2 SAR (Reference 3), the variables associated with the PAMS (SMM, CET, and RVLMS) are considered RG 1.97 Type A and B variables. The operator does not rely on these Type B variables for any safety critical actions. In the case of the Type A variables (RCS Hot Leg Water Temperature and RCS Pressure), the operator has primary indication (i.e., safety related indicators and recorders) and does not solely rely on the PAMS for this information. Therefore, this activity does not create a possibility for an accident of a different type than any previously evaluated in the UFSAR.

6.

Create a possibility for a malfunction of a structure, system, or component important to safety with a different result than any previously evaluated in the UFSAR?

Yes No BASIS:

WNA-LI-00107-CARK2, Qualitative Assessment for ANO-2 Post Accident Monitoring System 10 CFR 50.59 Evaluation (Reference 1), documents a Qualitative Assessment that was performed in accordance with the guidance of RIS 2002-22, Supplement 1 (Reference 2). The Qualitative Assessment provides a discussion on the failure modes and effects analysis for the replacement PAMS and how the system meets the single failure criterion. This Qualitative Assessment concluded that based on design attributes employed, the design process of the ANO-2 PAMS hardware and software, and review of the Common Q platform operating experience, it is reasonable to conclude that the ANO-2 PAMS will exhibit a sufficiently low likelihood of systematic failure due to a latent design defect.

Section 4 of WNA-LI-00107-CARK2 (Reference 1) also discusses how the SMM calculation is still available if the PAMS fails and does not result in a failure mode which would create a different result in terms of single failure criterion. Therefore, this activity does not create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the UFSAR.

4.

Result in more than a minimal increase in the consequences of a malfunction of a structure, system, or component important to safety previously evaluated in the UFSAR?

Yes No BASIS:

PAD Screen Question #1 concluded that the modification adversely affects a UFSAR described design function because the change involves a digital modification, and because the new PAMS combines the functionality of the SMM calculator and ICCMS-DAS. WNA-LI-00107-CARK2, Qualitative Assessment for ANO-2 Post Accident Monitoring System 10 CFR 50.59 Evaluation (Reference 1),

documents a Qualitative Assessment that was performed in accordance with the guidance of RIS 2002-22, Supplement 1, Clarification on Endorsement of Nuclear Energy Institute Guidance In Designing Digital Upgrades in Instrumentation and Control Systems (Reference 2). This Qualitative Assessment concluded that based on design attributes employed, the design process of the ANO-2 PAMS hardware and software, and review of the Common Q platform operating experience, it is reasonable to conclude that the ANO-2 PAMS will exhibit a sufficiently low likelihood of introducing a systematic failure due to a latent design defect. Section 4 of WNA-LI-00107-CARK2 (Reference 1) also discusses how the SMM calculation is still available if the PAMS fails.

Any malfunction depending on information of the PAMS is still within the design basis of the plant because the PAMS does not adversely impact any other SSCs. Therefore, this activity does not result in more than a minimal increase in the consequences of malfunction of an SSC important to safety previously evaluated in the UFSAR.

EN-LI-101 R22 ATTACHMENT 9.1 50.59 Evaluation Form Sheet 8 of 8 References

1.

WNA-LI-00107-CARK2, Revision 1, Qualitative Assessment for ANO-2 Post Accident Monitoring System 10 CFR 50.59 Evaluation, Westinghouse Electric Company LLC.

(CALC-ANO2-IC-23-00006 Attachment 1)

2.

NRC Regulatory Issue Summary 2002-22, Supplement 1, Clarification on Endorsement of Nuclear Energy Institute Guidance In Designing Digital Upgrades in Instrumentation and Control Systems, dated May 2018. United States Nuclear Regulatory Commission.

3.

Arkansas Nuclear One Unit 2 Safety Analysis Report, dated December 2023. Entergy Corporation.

7.

Result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered?

Yes No BASIS:

The existing ICCMS-DAS and SMM, and the replacement ANO-2 PAMS do not change a fission product barrier. The existing ICCMS-DAS and SMM are responsible for displaying Type C variables (RCS Pressure and Core Exit Temp), which are defined in the ANO-2 SAR (Reference 3) as:

These variables provide information to indicate the potential for breach of the barriers to fission product release.

Design basis limits for a fission product barrier are the controlling numerical values established during the licensing review as presented in the UFSAR for any parameter(s) used to determine the integrity of the fission product barrier. The change reviewed under this 50.59 does not affect any design basis limits. The Type C variables (RCS Pressure and Core Exit Temp) do not directly interface with any fission product barrier (fuel cladding, RCS boundary, or containment). The replacement ANO-2 PAMS failure cannot affect a fission product barrier. As established in the basis for questions 1 and 2 above there is no more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the UFSAR and in the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the UFSAR. As such it can be concluded that this activity does not result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered.

8.

Result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses?

Yes No BASIS:

In accordance with the criteria of NEI 96-07, no methods of evaluations described in the UFSAR are being changed as a result of this modification. This activity does not adversely affect a method of evaluation that demonstrates intended design function(s) of an SSC will be accomplished as described in the UFSAR. The replacement of the obsolete Inadequate Core Cooling Monitoring System (ICCMS) Data Acquisition System (DAS) and the Sub-Cooled Margin Monitoring (SMM) system with a Westinghouse-supplied Post Accident Monitoring System (PAMS) will ensure that the replacement components will meet the established design and licensing basis. Thus, this activity does not involve a change to a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses.

If any of the above questions is checked Yes, obtain NRC approval prior to implementing the change by initiating a change to the Operating License in accordance with NMM Procedure EN-LI-103.

EN-LI-101 R22

4.

NUREG-0737, Clarification of TMI Action Plan Requirements, United States Nuclear Regulatory Commission, Nov. 1980.

5.

WNA-AR-01139-CARK2, Revision 1, Post Accident Monitoring System - Reliability Analysis, Westinghouse Electric Company LLC. (CALC-ANO2-IC-23-00004)

6.

ANO-2-2024-0005, Transmittal #8 (ANO-2 to WEC) Addressment of PSAIs that require Entergy responses within Westinghouse Qualitative Assessment, Entergy Corporation.

May 21, 2024.

7.

WCAP-16097-P-A, Revision 5, "Common Qualified Platform Topical Report" (CALC-ANO2-IC-23-00006 Attachment 2)

8.

WCAP-16097-P-A, Appendix 1, Revision 0, "Common Qualified Platform Post Accident Monitoring Systems" (CALC-ANO2-IC-23-00006 Attachment 3)

2CAN042501 List of Affected SAR Pages 2CAN042501 Page 1 of 1 List of Affected SAR Pages The following is a list of SAR pages revised in Amendment 32 to support corrections, modifications, implementation of licensing basis changes, etc., as described in the Table of Contents of each SAR chapter (reference Enclosure 1 of this letter). Information relocated from one page to another in support of the aforementioned revisions is not considered a change; therefore, these pages are not included in the following list. In addition, pages associated with the individual Table of Contents are not listed below as related revisions are administrative only changes.

Cover Page Figure 4.3-1E Figure 10.4-1 2.3-28 Figure 4.3-2 Figure 10.4-2 Figure 2.5-39 Figure 4.3-3 15.1-132 3.1-2 Figure 4.3-4 15.1-134 3.1-3 Figure 4.3-5 18.2-6 3.3-2 Figure 4.3-6 18.2-7 3.5-6 Figure 4.3-7 3.5-7 Figure 4.3-8 3.5-8 Figure 4.3-9 3.9-28 Figure 4.3-10 3.12-4 5.2-50 3.12-5 5.2-51 3.12-6 5.8-2 3.13-1 5.8-51 3.13-16 Figure 5.2-33 3.13-17 Figure 5.2-33A 3.13-18 6.3-9 3.13-21 6.3-30 3.13-42 6.7-11 3.13-43 6.7-12 4.2-16 8.3-45 4.2-18 8.3-74 4.7-6 8.3-79 4.7-14 9.1-14 4.7-22 9.1-15 Figure 4.3-1 9.1-16 Figure 4.3-1A Figure 9.5-1 Figure 4.3-1C 10.2-7 Figure 4.3-1D 10.2-8 2CAN042501 ANO-2 SAR Amendment 32 - Un-redacted Version (4297 Pages) 2CAN042501 ANO-2 SAR Amendment 32 - Redacted Version (4297 Pages)

2CAN042501 ANO-2 TRM (157 Pages)

2CAN042501 ANO-2 TS Table of Contents and TS Bases (151 Pages)