2CAN102001, Transmittal of Amendment 29 to Safety Analysis Report

Transmittal of Amendment 29 to Safety Analysis Report

ML20294A313
Person / Time
Site:	Arkansas Nuclear
Issue date:	10/05/2020
From:	Dinelli J Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
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ML20294A311	List: 2CAN102001, Transmittal of Amendment 29 to Safety Analysis Report
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2CAN102001
Download: ML20294A313 (50)

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SECURITY RELATED INFORMATION SAR SECTION 2.8 OF ENCLOSURE 1 TO BE WITHHELD IN ITS ENTIRETY FROM PUBLIC DISCLOSURE IN ACCORDANCE WITH 10 CFR 2.390

~Entergy Entergy Operations, Inc.

1448S.R. 333 Russellville, AR 72802 Tel 479-858-3110 John C. Dinelli ANO Site Vice President 10 CFR 50.71 (e) 2CAN102001 October 5, 2020 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

Amendment 29 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

Dear Sir or Madam:

In accordance with 10 CFR 50.71(e) and 10 CFR 50.4(b)(6), enclosed is Amendment 29 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 (2R27) which ended April 21, 2020. 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 May 8, 2019, and ending October 5, 2020.

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, Figure 3.10-7, which depicted a general instrument rack and was a copy of drawing M-2511-J-G104-1, was removed from the SAR. The design and construction of safety-related instrument racks is discussed in SAR Sections 3.10.1 and 3.10.2.2.10. In addition, simplified drawings associated SECURITY RELATED INFORMATION SAR SECTION 2.8 OF ENCLOSURE 1 TO BE WITHHELD IN ITS ENTIRETY FROM PUBLIC DISCLOSURE IN ACCORDANCE WITH 10 CFR 2.390

SECURITY RELATED INFORMATION SAR SECTION 2.8 OF ENCLOSURE 1 TO BE WITHHELD IN ITS ENTIRETY FROM PUBLIC DISCLOSURE IN ACCORDANCE WITH 10 CFR 2.390 2CAN102001 Page 2 of4 with instrument rack designs remain in the SAR (see Figures 3.10-8 and 3.10-9). Figure 2.4-7 (Discharge Canal Plan, Profile & Details) was also removed from the SAR as it was largely illegible and contained excess detail not required to provide a 3rd party with an understanding of general plant design and operation. This figure also contained SUNSI information in that it depicted the natural gas line layout. Figure 2.8-2 remains within the SAR to provide details of the discharge canal. With respect to NEI 98-03, "Guidelines for Updating the Final Safety Analysis Reports," these figures may be removed from the SAR because the information "exceeds that necessary to present the plant design bases, safety analyses and appropriate UFSAR description."

No other "removal of detail" from the SAR occurred during this reporting period.

Associated in part with post September 11, 2001, response related to security sensitive information, Entergy has reviewed the ANO-2 SAR and determined that the following items 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 2.8.1.3, "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" SECURITY RELATED INFORMATION SAR SECTION 2.8 OF ENCLOSURE 1 TO BE WITHHELD IN ITS ENTIRETY FROM PUBLIC DISCLOSURE IN ACCORDANCE WITH 10 CFR 2.390

SECURITY RELATED INFORMATION SAR SECTION 2.8 OF ENCLOSURE 1 TO BE WITHHELD IN ITS ENTIRETY FROM PUBLIC DISCLOSURE IN ACCORDANCE WITH 10 CFR 2.390 2CAN102001 Page 3 of 4 The above is consistent with currently redacted information from the ANO-2 SAR (reference ML19141A164, ANO-2 SAR Amendment 28). 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 on the enclosed compact disc (CD).

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. includes a list of SAR pages that were updated during the period beginning May 8, 2019, and ending October 5, 2020.

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

I hereby certify that to the best of my knowledge and belief, 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. Executed on October 5, 2020.

SECURITY RELATED INFORMATION SAR SECTION 2.8 OF ENCLOSURE 1 TO BE WITHHELD IN ITS ENTIRETY FROM PUBLIC DISCLOSURE IN ACCORDANCE WITH 10 CFR 2.390

SECURITY RELATED INFORMATION SAR SECTION 2.8 OF ENCLOSURE 1 TO BE WITHHELD IN ITS ENTIRETY FROM PUBLIC DISCLOSURE IN ACCORDANCE WITH 10 CFR 2.390 2CAN102001 Page 4 of 4 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 - May 8, 2019, and ending October 5, 2020

5. List of Affected SAR Pages Enclosures (CD Rom):

1. ANO-2 SAR Amendment 29 - Un-redacted Version

2. ANO-2 SAR Amendment 29 - 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 SECURITY RELATED INFORMATION SAR SECTION 2.8 OF ENCLOSURE 1 TO BE WITHHELD IN ITS ENTIRETY FROM PUBLIC DISCLOSURE IN ACCORDANCE WITH 10 CFR 2.390

Attachment 1 2CAN102001 Summary of ANO-2 TRM Changes to 2CAN102001 Page 1 of 1 Summary of AN0-2 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.

Licensing Basis Document Change LBDC 19-018, 78 3.3.4 "Revise MTG Control Valve Test Frequency" Licensing Basis Document Change LBDC 19-035, 79 3.7.1, B 3.7.1 "Modification of Hot Work Suspension" Licensing Basis Document Change LBDC 20-008, 80 3.8.6, B 3.8.6 "Appendix R Emergency Lighting No Longer Required to be Maintained per NFPA-805" Engineering Change EC-84345, "Removal of Chlorine 81 3.3.7, B 3.3.7 Monitors from the Control Room Ventilation System" Licensing Basis Document Change LBDC 20-026, 82 6.5.1, B 3.7.7 "Update 1ST and Snubber Programs to 5th Interval" Conditi6n Report CR-ANO-2-2020-2117, "Add Hatches Table 3.7.5-2, 3.3.6, 83 to Fire Barrier TRM" and Condition Report CR-ANO-C-B 3.3.6 2018-2844, "Remove 50% Allowance of Fire Detection" Acronyms 1ST In-Service Testing MTG Main Turbine Generator NFPA National Fire Protection Association

Attachment 2 2CAN0102001 Summary of ANO-2 TS Bases Changes to 2CAN102001 Page 1_of 1 Summary of AN0-2 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 ofANO-2 TS 6.5.14. Because these changes were implemented without prior NRC approval, a description is provided below:

Revision Section Summarv Condition Report CR-ANO-C-2019-1332, "Clarify ASME OMN-20 Code Frequency Extension B 3.4.11, 72 Requirements" and Licensing Basis Document Change B 3.8.1.1 LBDC 19-017, "Correct Typographical SR Reference Error in ANO-2 TS Bases 3,8.1.1" B 4.0.2, B 3.1.3, B 3.3.1.1, B 3.3.2.1, B 3.4.6.2, B 3.4.8, TS Amendment 315, "TSTF-425, Relocation of 73 B 3.6.1.1, B 3.6.2.1, Surveillance Frequencies" B 3.6.2.2, B 3.6.3.1, B 3.7.1.2, B 3.7.6.1, B 3.8.2.3, B 3.8.3 B 3.0.2, B 3.0.3, TS Amendment 316, "TSTF-529, Clarify Use and 74 B 3.0.4, B 3.0.6, Application Rules" SR B 4.0.2, SR B 4.0.3 75 B 3.6.2.3, B 3.6.4.1 TS Amendment 318, "Containment Sump TS Addition" Licensing Basis Document Change LBDC 19-019, 76 B 3.1.3.1 "Clarify CEA Operability Impacts" Engineering Change EC-75937, "Manual RCS Flow 77 B 3.3.1.1 Calculation Using COLSS Pump dP Methods" Acronyms ASME American Society of Mechanical Engineers CEA Control Element Assembly COLSS Core Operating Limit Supervisory System dP Differential Pressure OMN ASME Code for Operation and Maintenance of Nuclear Power Plants RCS Reactor Coolant System SR (TS) Surveillance Requirement TSTF Technical Specification Task Force

Attachment 3 2CAN102001 Summary of ANO.:.2 10 CFR 50.59 Evaluations to 2CAN102001 Page 1 of 1 Summary of 10 CFR 50.59 Evaluations 50.59 # 50.59 Summary 2020-002 Engineering Change EC-84175, "Safety Injection NozzleThermal Sleeve" 2020-003 OP-2203.012F, "Annunciator 2K06 Corrective Action"

Attachment 4 2CAN102001 10 CFR 50.59 Evaluations - May 8, 2019, and ending October 5, 2020 (33 Pages)

ANO 50.59 Evalua tion 2020-002

-~Entergy NUCLEAR MANAGEMENT MANUAL QUALITY RELATED INFORMATIONAL USE 10 CFR 50.59 Evaluations EN-Ll-101 I

PAGE 1 OF 22 REV. 20 ATTACHMENT 9.1 50.59 EVALUATION FORM I. OVERVIEW/ SIGNATURES 1 Facility: Arkansas Nuclear One - Unit 2 (ANO-2) Evaluation#/ Rev.#: FFN-2020-002 / 0 Proposed Change/ Document: EC-84175, "Safety Injection Nozzle Thermal Sleeve" Description of Change:

Background

Based on Operating Experience (OE) regarding thermal sleeves becoming dislodged in other Combustion Engineering (CE) units [Ref. 6], ultrasonic (UT) inspections were performed on all four safety injection (SI) nozzles during ANO-2 refueling outage 2R23 in May 2014 to validate the presence of the thermal sleeves. Results were consistent among the four SI nozzles, indicating thermal sleeves were present at that time.

In June 2014 during 2R23 startup, alarms indicating confirmed metallic noises occurred on the Vibration and Loose Parts Monitoring System (VLPMS) while the reactor was at 17-18% power. Analysis of the limited available VLPMS data was inconclusive due to insufficient data. Since there was no notable impact to RCS flow or other system parameters, startup continued and ANO-2 operated for 3 more cycles before the condition of the missing thermal sleeve was discovered in 2R26 when the core support barrel (CSB) was removed. [Ref.7, 8]

During 2R26 in October 2018 a planned reactor vessel (RV) internal inspection was performed. During the inspection, two large pieces of debris from the dislodged "B" (300°) cold leg (CL) SI nozzle thermal sleeve were discovered near the bottom of the RV between the flow skirt (AKA flow baffle) and RV bottom head, adjacent to the 205° and 280° core stop lugs. Additional debris was discovered on the top surface of the core support plate. During the inspection and subsequent examinations, damage was discovered in several locations on RV internals. It was concluded the dislodged "B" thermal sleeve caused the damage, and had fragmented into two larger and several smaller pieces of debris. Most of the debris was collected during Foreign Object Search and Retrieval (FOSAR) efforts, but not all debris could be positively accounted for, resulting in the possibility that pieces of the "B" thermal sleeve remain in the RCS. [Ref. 8, 9, 1O]

The following damage and abnormal conditions were identified during 2R26:

• Anomaly in the RV bottom head cladding between the 205° and 250° core stop lugs, accompanied by signs of corrosion of the low alloy carbon steel vessel.

• Abrasions on the CSB upper and center cylinder near the 300° CL inlet.

• Scratching and appreciable wear on the 205°core stop lug.

• Fretting on RV cladding and the flow skirt

• Partially blocked openings in the flow skirt from dislodged sleeve (flow skirt blockage was removed during 2R26).

• Remaining "B" thermal sleeve fragments not collected during FOSAR.

• "B" thermal sleeve missing.

• "A" and "D" thermal sleeves rotated approximately 90°.

-~ Entergy NUCLEAR MANAGEMENT MANUAL QUALITY RELATED INFORMATIONAL USE 10 CFR 50.59 Evaluations EN-Ll-101 PAGE 2 OF 22 REV. 20 ATTACHMENT 9.1 50.59 EVALUATION FORM A steam generator primary side inspection was performed during 2R26 to satisfy Westinghouse Nuclear Safety Advisory Letter NSAL-12-1 [Ref 18]. No SG bowl abnormalities or foreign material were noted in the insgection report. The inspection included a 100% eddy current inspection with no SG tube abnormalities noted.

A review of the Paperless Condition Reporting System (PCRS) shows no evidence of failed fuel at ANO-2 since 2R23.

EC-79912 [Ref. 1OJ was developed to document the Westinghouse Phase 1 evaluation which provides a qualitative assessment of the damage and conditions described above to establish acceptability of returning the RCS and emergency core cooling system (ECCS) to service. That evaluation included a loose parts assessment; structural evaluation of the "B" SI nozzle, hydraulic evaluation of the RCS; structural assessment of CSB damage, structural assessment of RCS component damage, and assessment of missing cladding in RCS indications including corrosion analysis and Section XI evaluation. The Phase 1 evaluation supports continued operation with these systems in Operable -

Degraded Nonconforming (OP-DNC) status for one operating cycle, which ends with refueling outage 2R27 in the spring of 2020.

Summary of Change The change evaluated herein and authorized under EC-84175 restores the RCS and ECCS to operable status by accepting the damage and abnormal conditions identified during 2R26 as-is, without modification or repair. The engineering change (EC) also considers potential loose parts effects resulting from future SI thermal sleeves becoming dislodged as described under the individual tasks below. The EC documents acceptability of continued safe operation of ANO-2 through the end of the current extended operating license period (July 2038). This is accomplished by incorporating into the ANO-2 license and design bases the Westinghouse Phase 2 evaluation consisting of six tasks documented in References 1 - 4. In addition to the base Phase 2 scope, an Environmentally Assisted Fatigue evaluation was developed as documented in Reference 16. These documents are summarized in EC-84175, which comprises the scope of this 50.59 Evaluation.

The Task 1 loose parts assessment (LPA) in Reference 1 provides an operability assessment that addresses long term effects of a combination of un-retrieved debris from the "B" thermal sleeve plus postulated debris from the other three sleeves conservatively assumed to dislodge during subsequent operating cycles.

The "B" thermal sleeve did not remain intact, resulting in multiple pieces being discovered both upstream and downstream of the flow skirt. The following summarizes loose parts discovery:

• No loose parts were identified in the hot legs (HL) or Cls.

• Two large pieces were found on the upstream side of the flow skirt.

• Many pieces were found on the core support plate (CSP), having passed through the flow skirt and core support structure bottom plate.

• Any material identified in fuel assemblies during refueling was removed prior to reload, but none was specifically characterized as pieces of the thermal sleeve and is not included in retrieved material estimates.

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10 CFR 50.59 Evaluations ATTACHMENT 9.1 50.59 EVALUATION FORM

• No loose parts were identified in the steam generator (SG) inlet and outlet plenums.

• In retrospect, loose parts discovered in outages between 2R23 and 2R26 were likely sleeve fragments, but these items were not quantified and are conservatively not included in retrieved material estimates.*

Conservative estimates of material retrieved during 2R26 are less than the total calculated area of the "B" thermal sleeve, resulting in the possibility that sleeve fragments remain in the RCS. Material that was accounted for and removed during 2R26 consisted of two large pieces of the therIT)al sleeve on the .

upstream side of the flow skirt and many pieces of material that had passed through the flow skirt and collected on the core support plate. Unaccounted for material from the "B" sleeve considered as loose parts remaining in the RCS is 122 in 2 of the 3/16" thick, SB-168 lnconel sleeve material, with a Rockwell B Hardness of 75 - 95 [Ref. 17]. This quantity is considered conservative since the FOSAR activity completed in 2R26 effectively left no portions of the primary system uninspected and no additional loose parts were discovered, yet the entire unaccounted for material is assumed to remain. Reference 1 conservatively assumes the entire unaccounted for material from the "B" thermal sleeve (122 in 2) plus an assumed equivalent amount of material from the three remaining sleeves (370 in 2) passes through the core and circulates in the RCS. The total assumed area of fragments in the RCS would be approximately 492 in 2 with a weight of about 28 lbs. This assumed quantity could effectively circulate through the RCS for a period of time until being removed by CVCS or settling in low flow regions or connected auxiliary systems with a decreasing likelihood of consequences over time. It is important to note that multiple conservatisms are applied to result in this estimated quantity of material in the RCS.

The conservative premise of this evaluation is that a SI nozzle thermal sleeve can fragment into pieces small enough to pass through the core and circulate in the RCS. Fragment sizes are limited by openings in the various RVI components through which debris must pass. In order to pass through the flow skirt, the largest dimension must be less than the 2.72" diameter holes in the skirt. For reference, the largest debris fragment found downstream of the flow skirt on the CSP was 2.858" by 2.513" by 3/16" thick. Per Reference 1, the largest fragment that could pass through the core would be no larger than 0.1 O" in any smaller dimension by 1" long, with an approximate weight of 0.003 pounds.

This consolidated loose parts assessment [Ref. 1] supports a conclusion of safe continued operation with postulated loose parts from four thermal sleeves in the RCS and auxiliary systems for the remaining licensed life of ANO-2:

• Thermal sleeve debris poses little to no risk of degrading Departure from Nuclear Boiling (DNB) performance.

• Risk of debris related fuel cladding leaks is increased, but is considered low due to effectiveness of the fuel design in mitigating the risk for debris causing fuel damage.

• Structural impacts to RV components downstream of the flow skirt are insignificant.

• Potential to impact Control Element Assembly (CEA) operation is remote

• Likelihood of impacts to in-core instrumentation is remote.

• Pressurizer heater performance is not expected to be affected.

• Pressurizer instrumentation is not susceptible to debris blockage.

• If it is postulated that debris impairs pressurizer spray flow, auxiliary spray would be utilized to support safe shutdown.

NUCLEAR MANAGEMENT QUALITY RELATED EN-Ll-101 I REV. 20 MANUAL INFORMATIONAL USE PAGE 4 OF 22 10 CFR 50.59 Evaluations ATTACHMENT 9.1 50.59 EVALUATION FORM

• Reactor Coolant Pump (RCP) internals will receive no substantial damage.

• Possible RCP seal degradation could occur and would be addressed through normal procedure actions.

• RCP safety related functions are unaffected and normal RCP operation is unlikely to be affected.

• Postulated fragments will result in minimal Steam Generator (SG) tube degradation and no impact on structural integrity.

• The SI system, Safety Injection Tanks (SIT), Containment Spray System (CSS), Shutdown Cooling SOC), Chemical and Volume Control System (CVCS), and RCS chemistry are unaffected by any debris in the RCS. *

• Debris transported to the Spent Fuel Pool (SFP) during defueling would settle to the pool bottom and not affect SFP cooling.

• When less than four RCPs are in operation, reverse flow in a CL is possible. It is shown that damage to an idle RCP is unlikely and any significant damage to a CL Temperature Element (TE)

Thermowell (TW) would be detected.

As documented in Reference 17 and stated where applicable herein, in support of the loose parts determinations made in Reference 1, representative material hardnesses of various impacted components are considered versus the loose part thermal sleeve materials. The various impacted components are of a similar hardness as the sleeve materials and the damage impacts therefore are consistent with the conclusions drawn in Reference 1.

The Task 2 RCS hydraulic assessment in Reference 2 is based on the unlikely case the remaining three thermal sleeves become dislodged, travel to the flow skirt, and unroll intact, blocking the maximum practicable flow area. It considers SG tube plugging and concludes the resulting RCS flow with maximum practicable flow skirt blockage would remain above the Technical Specification (TS) 3.2.5, "RCS Flow Rate," minimum limit of 120.4 x 10 6 lbm/hr, stating that four RCP normal operation scenario bounds lower flow scenarios including less than four RCP operation, SOC operation, and natural circulation. The assessment also demonstrates flow skirt blockage to be more limiting than plausible blockage at other locations in the RCS flow path including fuel bundle entry and blockage between fuel rods.

The Task 3 Evaluation of CSB damage in Reference 2 evaluates upper and center CSB cylinder known damage from the "B" thermal sleeve, as well as postulated damage due to the possibility of more thermal sleeve failures. The evaluation concludes stresses due to existing damage and stresses due to postulated additional damage are less than allowable stresses. Note that should additional sleeves become dislodged, resulting damage must be validated to determine if it is within the bounds of the postulated damage evaluation.

The Task 4 ASME Section Ill Design Analysis Report for the damaged RV bottom head and core stop lugs in Reference 3 concludes the RV bottom head and core stop lug damage is acceptable "as isl' for continued operation for 14 future fuel cycles based on calculated flaw depth, which is based on conservative corrosion rates and extrapolated to the end of the 60 year design life.

I The Task 5 ASME Section Ill Design Analysis Report for a sleeveless SI nozzle in Reference 4 evaluates a nozzle without a thermal sleeve and notes the original analysis of record continues to apply to nozzles with thermal sleeves remaining in place. The analysis concludes a SI nozzle without a thermal sleeve is

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I REV. 20 MANUAL INFORMATIONAL USE PAGE 5 OF 22 10 CFR 50.59 Evaluations ATTACHMENT9.1 50.59 EVALUATION FORM acceptable for continued operation for the duration of the existing extended operating license. Results also apply to SI nozzles with rotated thermal sleeves where the bottom of the nozzle is exposed to SI flow. This thermal and fatigue analysis confirms that the case of a SI nozzle without a thermal sleeve bounds the case of a SI nozzle with a rotated thermal sleeve.

• The Task 6 ASME Section XI RV bottom head flaw analysis in Reference 2 concludes the 20-year flaw evaluation of the RV bottom head cladding anomaly shows acceptable fatigue crack growth. Primary local stress meets the ASME stress limit considering wall thickness reduction and corrosion. Further, the document states it is reasonably assumed the potential bottom head damage from additional dislodged sleeves would be bounded by the existing evaluation.

• The operability evaluation associated with Reference 7 concludes fretting on RV cladding and the flow skirt are minor and inconsequential. These conditions are not evaluated further.

Reference 16 provides an updated Environmentally Assisted Fatigue (EAF) evaluation to address RV bottom head damage and SI nozzles without thermal sleeves. The EAF evaluation shows that bottom head and SI nozzle fatigue usage factors are acceptable. It concludes that the CUFs for the RV bottom head and the SI nozzle without a thermal sleeve are equal to 0.697 and 0.434, respectively. By comparison, the ANO-2 License Renewal Application (LRA) concluded EAF CUFs of 0.0075 for the RV bottom head (undamaged) and 0.6534 for the SI nozzle (with a thermal sleeve). For the SI nozzle, the increase in fatigue due to a loss of a thermal sleeve is offset by the use of 60-year projected cycles and application of more modern analytical techniques in evaluating the design condition; therefore, there is a net reduction in EAF CUFs relative to the ANO-2 LRA. Both EAF CUFs remain below the ASME Code CUF acceptance criteria of 1.0, and are acceptable.

ANO-2 Operations normal monitoring of plant conditions such as VLPMS indications and RCS flow will provide indication of additional thermal sleeves becoming dislodged. Appropriate action under existing plant procedures will be taken should evidence of additional dislodged sleeves be identified.

As documented in the EC-84175 Process Applicability Determination (PAD), the change does not adversely affect a method of performing or controlling a design function, nor does it adversely affect a method of evaluation that demonstrates a design function will be accomplished. The damage and abnormal conditions have actual and potential adverse effects on design functions as described in the PAD including:

• Potential flow blockage through the flow skirt under normal operation, SOC operation, and accident conditions should additional thermal sleeves become dislodged.

• Potential flow blockage through individual fuel assemblies under normal operation, SOC operation, and accident conditions due to postulated remaining debris from the previously dislodged "B" thermal sleeve or from additional thermal sleeves become dislodged.

• Structural effects in those areas damaged by the previously dislodged "B" thermal sleeve.

• Potential damage to fuel cladding due to existing and potential future loose parts from dislodged thermal sleeves.

• Increased nozzle stresses due to the absence of thermal sleeves in the SI nozzles.

In addition to the designated Entergy 50.59 evaluation reviewer, this evaluation was reviewed by Jeffrey Erickson in the ENERCON regulatory affairs organization.

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PAGE 6 OF 22 REV. 20 ATTACHMENT 9.1 50.59 EVALUATION FORM

References:

1. CALC-ANO2-ME-20-00003: LTR-SEE-19-131, Rev. 0-A "Operability Assessment for Primary Side Loose Parts due to the Dislodged Safety Injection Nozzle B Thermal Sleeve at Arkansas Nuclear One, Unit2"

a. Task 1: Detailed Loose Parts Assessment

2. CALC-ANO2-ME-20--00004: VVCAP-18512-P, Rev. 0-A "BackgrQund ~ng Technical Basil?

Supporting Long-term Operation for ANO-2 with Dislodged Safety Injection Nozzle Thermal Sleeves and Associated Damage"

a. Task 2: Hydraulic Evaluation Considering Additional Dislodged Thermal Sleeves

b. Task 3: Detailed Evaluation of Core Support Barrel Damage

c. Task 6: ASME Section XI Reactor Vessel Bottom Head Flaw Analysis

3. CALC-86-E-0036-274: DAR-SDA-19-2, Rev. 0-A "Addendum to the ANO-2 Reactor Vessel and Core Stop Lugs Design Report for Damage due to a Dislodged Thermal Sleeve"

a. Task 4: ASME Section Ill Design Analysis Report for the Damaged Reactor Vessel Bottom Head and Core Stop Lugs

4. CALC-86-E-0036-275: DAR-SDA-19-3, Rev. 0-A "Addendum to the ANO-2 Piping Design Report for Missing Thermal Sleeve on Safety Injection Nozzle"

a. Task 5: ASME Section Ill Design Analysis Report for a Sleeveless Safety Injection Nozzle

5. CR-ANO-2-2013-01190, IER L3-13-2 Identification of Dislodged Thermal Sleeve at a Similar Combustion Engineering Unit

6. CR-ANO-2-2014-02035, Vibration and Loose Parts Monitor Alarm During 2R23 Startup

7. CR-ANO-2-2018-02976, Discovery of Foreign Objects in ANO-2 Reactor Vessel

8. CR-ANO-2-2018-03127, Scrapes and Abrasions on ANO-2 Core Support Barrel Outer Surface Axial Welds

9. CR-ANO-2-2018-03252, Scrapes and Abrasions on ANO-2 Core Support Barrel Outer Surface Middle Girth Section

10. EC-79912, Westinghouse Phase I Evaluation Supporting Justification for Continued Operation at

~M . .

11. CALC-86-E-0036-143, RCP Design Report Addendum To Byron Jackson TCF-1018 & Analysis of 36x36x42 DFSS Primary Coolant Pmps

12. EN-NF-102, Corporate Fuel Reliability

13. STM 2-03, Reactor Coolant System Training Manual

14. UFSAR, Updated Final Safety Analysis Report

15. ULD-2-SYS-24, Inadequate Core Cooling

16. CALC-ANO2-ME-20-00007: CN-SDA-20-2, Environmentally-Assisted Fatigue Evaluations of the ANO-2 Safety Injection Nozzle and Reactor Vessel Bottom Head

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PAGE 7 OF 22 REV. 20 ATTACHMENT 9.1 50.59 EVALUATION FORM

17. EC 84175, Attachment 7.011: Material Hardness Comparison

18. NSAL-12-1, Westinghouse Nuclear Safety Advisory Letter - SG Channel Head Degradation Summary of Evaluation:

The evaluation documented in the responses to Questions 1 through 8 is based on the Phase 2 assessments and analyses summarized in Section I. The results of the evaluation show acceptability of operating ANO-2 with the damage and abnormc1I conditions described above through the end of the current license period without additional approvals by the NRC. The evaluations performed as documented herein demonstrate that ANO can be safely operated until the end of its current license with the damage and abnormal conditions identified in Section I. The evaluations also show that should additional thermal sleeves become detached or damaged in the future, there is a reasonable expectation of operability. However, if VLPMS, RCS flow, or other parameters indicate ari additional thermal sleeve becomes detached, an evaluation will be performed to ensure that all results are within the bounds assumed in the referenced evaluations. If referenced evaluations are not bounding, additional actions will be taken in accordance with existing procedures.

Is the validity of this Evaluation dependent on any other change? D 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 D Yes ~ No require prior NRC approval?

f')Entergy NUCLEAR MANAGEMENT QUALITY RELATED EN-Ll-101 I REV. 20 MANUAL INFORMATIONAL USE PAGE 8 OF 22 10 CFR 50.59 Evaluations ATTACHMENT 9.1 50.59 EVALUATION FORM Preparer1*2 : Danny Hughes / ORIGINAL SIGNED BY DANNY HUGHES / ENERCON / Engineering / 4-2-20 Name (print) / Signature / Company / Department / Date Reviewer1*2 : Dave MacPhee / ORIGINAL SIGNED BY DAVID MACHPEE /Entergy/ Engineering/ 4-2-20 Name (print) / Signature / Company / Department / pate Independent NA Review1*3 : Name (print) / Signature / Company / Department / Date Responsible Manager Tom Hatfield/ ORIGINAL SIGNED BY TOM HATFIELD/ Entergy/ Engineering/ 4-3-20 Concurrence 1 : Name (print) / Signature / Company / Department / Date 50.59 Program Coordinator Michael Hall/ ORIGINAL SIGNED BY MICHAEL HALL/ Entergy/ Engineering / 4-3-20 Concurrence 1 : Name (print) / Signature / Company / Department / Date ORIGINAL SIGNED BY BRYAN DAIBER / 4-3-20 Chairman's Name (print) / Signature I Date [GGNS P-33633, P-34230, & P-34420; W3 P-151]

OSRC-2020-005 OSRC Meeting #

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 Asset Suite." 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 Asset Suite, attach a screens hot of the electronic signature(s); if using PCRS, 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].

-Entergy NUCLEAR MANAGEMENT QUALITY RELATED EN-Ll-101 I REV. 20 MANUAL INFORMATIONAL USE PAGE 9 OF 22 10 CFR 50.59 Evaluations ATTACHMENT9.1 50.59 EVALUATION FORM II. 50.59 EVALUATION [10 CFR 50.59(c)(2)]

Does the proposed Change being evaluated represent a change to a method of

• D Yes evaluation ONLY? If "Yes," Questions 1 - 7 are not applicable; answer only igj No Question 8. If "No," answer all questions below.

Does the proposed Change:

1. Result in more than a minimal increase in the frequency of occurrence of an accident D Yes previously evaluated in the SAR? igj No BASIS:

The SAR accident analyses were reviewed to determine what, if any, previously evaluated accidents may be associated with damage and abnormal conditions described in Section I. Results follow with discussion as to whether the activity will result in a change in frequency of occurrence of a relevant accident previously evaluated:

15.1.3, CEA Misoperation A CEA misoperation is any event which could result from a single malfunction in the reactivity control system, except certain CEA withdrawal accidents considered in 15.1.1 and 15.1.2. As described in Reference 1, the potential for loose parts in the RCS to impact CEA operation is remote. The AN0-2 reactor is controlled in part by 81 CEAs, each weighing 69 pounds and consisting of 5 individual control elements attached together as a single unit via an upper assembly known as a spider. Each finger moves inside one of five guide tubes integral to the associated fuel assembly.

The guide tubes guide the CEA through its entire length of travel inside the fuel assembly. The

.CEAs are mechanically positioned for normal reactor control and drop into the core automatJcally by gravity during a reactor trip.

Reference 1 explains that the potential for a loose fragment to adversely affect CEA operation is remote. Clearances between the CEA and CEA guide tube are too large to trap a fragment of the size postulated to pass through or around the core. It also shows that.several holes exist in the guide tubes which could serve as pathways for potential debris entry and possible interaction with a CEA. However, the small holes are perpendicular to the reactor coolant flow path through a fuel assembly, making it unlikely debris would be able to turn and enter the guide tube. In the unlikely event debris does enter the guide tube, it would likely drop to a benign location at the bottom of the guide tube since upward flow velocity is low. Any debris of low enough density to transport up through the guide tube would not be expected to impede CEA movement based on the small size.

In the very unlikely event CEA movement is impeded, maneuvering of the CEA is expected to clear the obstruction with no significant damage to the CEA finger or guide tube, based on OE. Therefore, it is concluded there is no more than a minimal increase in the frequency of occurrence of a CEA misoperation accident.

-"'=":!>Entergy NUCLEAR MANAGEMENT MANUAL QUALITY RELATED INFORMATIONAL USE 10 CFR 50.59 Evaluations EN-Ll-101 I

PAGE 10 OF 22 REV. 20 ATTACHMENT 9.1 50.59 EVALUATION FORM 15.1.5, Total and Partial Loss of Reactor Coolant Forced Flow Causes of this accident are limited to mechanical and electrical failures resulting in loss of one or more reactor coolant pumps (RCP). As described in Reference 1, any remaining debris from a dislodged thermal sleeve that could circulate through the RCS and enter a RCP would not affect the safety related function of the RCPs and would be unlikely to affect normal operation through end of licensed plant life due to small size and mass. Any debris that might circulate through the RCS and reach the RCP through normal coolant circulation would be expected to be carried through the hydraulic passages of the RCP causing no substantial damage to the pump internals or adverse effects to the function of the RCP. This is due to the very small relative mass of the loose parts compared to the much larger rotating impeller and stationary hydraulics. Since RCP internals are not expected to receive any substantial damage, weekly RCS flow surveillances are not affected.

During plant conditions when less than four RCPs are in operation, reverse flow at a fraction of the normal RCS flow may be present in CLs with an idle RCP. Should a sleeve become dislodged in a CL with an idle RCP, it may be possible for it to transport in the reverse flow stream back to the RCP discharge. This would only be postulated to occur prior to the sleeve deforming and traveling down the annulus between the RV and CSB. Should a dislodged sleeve transport to the RCP discharge, the sleeve would have no effect on rotating parts of the pump and would therefore not impede pump start or operation. This is because the Byron Jackson RCPs at ANO-2 are model DFSS pumps with discharge diffuser vanes [Ref 1, 11 ]. Spacing between the robust vanes would prevent an object the size of a thermal sleeve from entering pump rotating equipment.

Therefore, it is concluded this change will not cause more than a minimal increase in frequency of occurrence of the Total and Partial Loss of Reactor Coolant Forced Flow accident.

15.1.13, Major Rupture of Pipes Containing Reactor Coolant up to and Including Double-ended Rupture of the Largest Pipe in the RCS - Loss of Coolant Accident (LOCA)

The relationships between the change and this accident include the following three scenarios, effects of missing or rotated sleeves on nozzle integrity, effects of damage from a dislodged sleeve, effects of CL TW damage due to reverse flow.

Effects of missing or rotated thermal sleeves on SI nozzle structural integrity:

As discussed in Reference 2, thermal sleeves are provided in the SI nozzles as a thermal barrier to address temperature extremes that occur when relatively cold water is injected via the SI system into the much higher temperature RCS. Reference 4, an addendum to the ANO-2 piping design report, evaluates SI nozzle structural integrity without a thermal sleeve and with a rotated sleeve. It concludes an SI nozzle without a thermal sleeve is acceptable as-is for continued operation for the duration of the extended license. It also concludes a nozzle with a rotated sleeve is acceptable.

The SI nozzle missing a thermal sleeve is evaluated for primary plus secondary stress, thermal stress ratchet and fatigue, including cladding stresses as appropriate. For both the missing sleeve and rotated sleeve case, SI flow is applied to piping or nozzle surfaces as appropriate to model the conditions present during various transients. Upon determination of limiting stress paths, ASME stress calculations were performed using both total and linearized stresses from all analysis cases.

In all cases, stresses are below prescribed limits. Note that the dynamic analysis was not updated since removal of the sleeve weight of approximately 45 pounds, which is much less than the 2000-pound nozzle forging, is considered insignificant.

-~ Entergy NUCLEAR MANAGEMENT MANUAL QUALITY RELATED INFORMATIONAL USE 10 CFR 50.59 Evaluations EN-Ll-101 PAGE 11 OF 22 REV. 20 ATTACHMENT 9.1 50.59 EVALUATION FORM The rotated sleeve condition also assesses primary stress due to the rotated sleeve creating hydraulic loads on the affected nozzle. These loads react at the evenly spaced bosses that position a sleeve and are satisfied by applicable bearing and fatigue stress evaluations, assuming cycling loads due to vortex shedding and pump pulsation, Effects of damage from a dislodged sleeve:

Damage described in Reference 2 is limited to RVI components and does not directly or indirectly translate to the major_ RCS pipe rupture accident. ,

Effects of potential damage to CL TWs during a reverse flow condition:

Should a dislodged sleeve move forward or backward in the region of the cold leg pipe between the RCP outlet and the SI nozzle, the thermal sleeve could potentially strike and damage a resistance temperature detector (RTD) thermowell. There are three TWs in the RCP outlet nozzle and near the top of CL piping. Impact to a thermowell is considered a worst case occurrence and is not supported by industry OE. However, it is appropriate to consider this unlikely scenario.

It is possible an impact could result in consequences ranging from deformation of the TW, to RTD failure, to a fractured or sheared TW, resulting in pressure boundary leakage. In the event the unlikely worst case condition were to occur, the postulated leak would not exceed the makeup capability of the CVCS. Should RTD failure or TW leakage occur, the lost inventory would be

• sensed by the pressurizer level control circuit, which would compensate by starting the available charging pump(s) and reducing letdown flow rate. That compensatory action by the control circuit would alert the plant Operators to take corrective.action. In addition, the indication of any lost RTD channels on the control room boards would support Operator response, decisions and actions.

Based on this discussion, it is concluded a dislodged or rotated thermal sleeve does not more than minimally increase the frequency of occurrence of a major RCS pipe rupture accident.

15.1.18, Steam Generator Tube Rupture with or without a Concurrent Loss of AC Power The relationship between the change and this accident relates to possible transport of dislodged sleeve debris through the RCS to the SG tubes, resulting in possible tube damage As described in Reference 1, there are four potential SG tube damage mechanisms associated with thermal sleeve debris transport including debris impacts at the tube to tubesheet welds, debris impacts to a tube plug, debris becoming lodged in SG tubes, and debris passing through a tube an exiting the SG via the CL. Per UFSAR 5.5.2.2, SG tubes are 0.6875" diameter and are manufactured from nickel-chromium-iron Alloy 690, with a Rockwell B hardness of 85 [Ref. 17] as compared to a sleeve fragment hardness of 75 - 95.

Debris impacts at the tube to tubesheet welds:

Reference 1 discusses the potential that small thermal sleeve fragments could repeatedly impact the bottom surface of tube to tubesheet welds. Based on the maximum estimated weight of a loose part passing through the RV of 0.003 pounds, and the estimated flow rate in the SG primary head, Reference 1 concludes dents of approximately 0.5 mils might occur, resulting in a small amount of peening to the weld and possibly tube inside diameter. This effect of denting and peening is not expected to induce any adverse effects, including primary water stress corrosion cracking PWSCC.

NUCLEAR QUALITY RELATED EN-Ll-101 REV. 20 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 12 OF 22 10 CFR 50.59 Evaluations ATTACHMENT 9.1 50.59 EVALUATION FORM Debris impacts to a tube plug:

Debris impacts to the tube plugs are identical to those discussed above for tube to tubesheet welds.

Reference 1 states no plug loosening, ejection or other adverse effects are expected, and plugs are expected to continue to perform their design function.

Debris becoming lodged in SG tubes:

Reference 1 states the postulated-largest fragmentthat could-pass through the core would be no larger than 0:1 O" in any smaller dimension by 1" long could pass through a large radius U-bend SG tube and exit the SG through the CL. It also states debris of this size could pass through a small radius U-bend, but it would be possible to become lodged in the U-bend region of the tube.

Reference 1 states that should this unlikely and hypothetical scenario occur, wear in the U-bend due to vibration is unlikely due to the lack of postulated relative motion between the debris and tube because of the likelihood of the debris becoming firmly lodged into place. The small size of the postulated fragment is such that tube flow induced vibration characteristics would not change significantly. Further, the effects of expansion and contraction around lodged debris has been shown to result in stresses well below the tube material endurance limit.

Per UFSAR 5.2.7.1.4.A, primary to secondary leakage via a steam generator is detected through various process monitors, assuring any such leakage is identified, allowing appropriate mitigating action to be taken.

Therefore, it is concluded the frequency of occurrence of this accident is not more than minimally increased.

15.1.25, Fuel Cladding Failure Combined with Steam Generator Leak The relationship between the change and this accident relates to debris transport and possible damage to fuel cladding and steam generator tubes, as well as RC flow blockage. As discussed under 15.1.18 above, debris resulting from a dislodged sleeve is unlikely to result in SG tube damage. Therefore, the SG tube leak aspect of this accident is unaffected. However, Reference 1 concludes there may remain a raised risk of a debris related leaking fuel rod in subsequent operating cycles. This is because it is possible sleeve related debris could become lodged within a fuel bundle, resulting in potential long term damage to fuel cladding. Three fuel rod cladding configurations are described in UFSAR 4.2.1.2 and Table 4.2-3: 0.374" OD Optimized ZIRLO',

0.382" OD ZIRLO', 0.382" OD Zircaloy-4, each with a Rockwell B hardness of 89 - 95 [Ref. 17] as compared to a sleeve fragment hardness of 75 - 95.

Reference 1 goes on to state that although possible, the likelihood of long slender pieces passing the fuel GUARDIAN grid that might cause fuel fretting is unlikely since the GUARDIAN grid is specifically designed to trap debris of such configurations against the solid fuel rod end cap to prevent fuel clad fretting. Further, operating history of the GUARDIAN grid has been excellent. Only one confirmed debris induced leaking fuel rod has occurred for over 7200 GUARDIAN equipped assemblies installed.

f!JEntergy NUCLEAR MANAGEMENT QUALITY RELATED EN-Ll-101 I REV. 20 MANUAL INFORMATIONAL USE PAGE 13 OF 22 10 CFR 50.59 Evaluations ATTACHMENT 9.1 50.59 EVALUATION FORM The eves process radiation monitor is a trend monitor and its primary purpose is to indicate the possibility of fuel clad failure. If an alarm is received and the lodine-131 activity has increased and remains significantly above the prior steady state level, additional fuel failure can be assumed to have occurred.

UFSAR 15.1.25 states accident consequences are based upon a continuous one gpm primary to secondary SG tube leak concurrent with an RCS specific activity equivalent to one percent failed fuel. Although the frequency of long-term fuel cladding damage may increase, the frequency of an accident-exceeding the RCS specific activity limits is not affected since any potential effects would occur over time and RCS activity is routinely monitored.

Regarding flow blockage effects, Reference 1 shows that Departure from Nucleate Boiling Ratio (DNBR) is not significantly affected by any postulated RC flow blockage from sleeve fragments. Any relatively large area blockag~ such as at the flow skirt or fuel bundle inlet does not significantly affect DNB based on a representative test with 36% of the flow area between rods blocked. Flow redistributes downstream within approximately one grid span. Another test with Westinghouse fuel shows full flow recovery within 30", even with total blockage of the fuel bundle inlet. Since minimum DNBR typically occurs at higher elevations in the core, these postulated blockages have no significant effect on minimum DNBR. Further, it is highly unlikely debris would accumulate in one area of the core as evidenced by location of sleeve fragments spread about the core support plate, leaving a small likelihood that one or more fuel assemblies would see 36% flow blockage.

Based on this discussion, it is concluded the frequency of _occurrence of this accident is not more than minimally increased.

15.1.27, Failure or Overpressurization of Low Pressure Residual Heat Removal System The relationship between the change and this accident relate to potential effects on failure related to SI nozzle integrity from missing or rotated sleeves, and any potential damage from debris transport.

The SOC system provides residual heat removal for ANO-2 and takes suction on nozzles attached to the RCS HLs. Reactor coolant is circulated by a Low Pressure Safety Injection (LPSI) pump through a SOC heat exchanger to the LPSI header, returning to the RCS through the SI nozzles. As described under 15.1.13 above, SI nozzles are shown to be structurally acceptable for continued operation with a missing or rotated thermal sleeve. Regarding damage from debris transport, Reference 1 states any fragments small enough to pass through the fuel could reach the HL SOC nozzles. Loose fragments of the size small enough to pass through the core would not cause failures in the SOC system that would prevent it from performing its design function. The system has large diameter piping and large port valves that are expected to allow any postulated loose parts to pass. The single stage SOC pumps have relatively large flow passages that would be unaffected by any such debris.

UFSAR Table 9.3-24 shows SOC heat exchangers are of the u-tube design. Tubes are 0.75" OD and are manufactured from super austenitic stainless steel, SB-676 (N08367), with a Rockwell B hardness of 89 [Ref. 17] as compared to a sleeve fragment hardness of 75 - 95. In the unlikely event the postulated 0.1" x 1" sleeve fragment enters a SOC heat exchanger, it would likely pass through unimpeded. Considering the Reference 1 discussion regarding such a fragment, it is unlikely but possible a fragment could become lodged in the u-tube bend, but no deleterious effects are expected.

f!)Entergy NUCLEAR MANAGEMENT QUALITY RELATED EN-Ll-101 I REV. 20 MANUAL INFORMATIONAL USE PAGE 14 OF 22 10 CFR 50.59 Evaluations ATTACHMENT 9.1 50.59 EVALUATION FORM Per UFSAR 5.2. 7 .1.4. B, Service Water (SW) exiting the SOC heat exchangers is equipped with process radiation monitors to detect any leakage from the RCS to SW, assuring any such leakage is identified, allowing appropriate mitigating action to be taken.

Therefore, it is concluded the frequency of occurrence of this accident is not more than minimally increased.

2. Result in more than a minimal increase in the likelihood of occurrence of a malfunction D Yes of a structure, system, or cqmponent (SSC) important to saf(lty previously eyaluated in [8] No the SAR?

BASIS:

The change described in Section I potentially affects the following SSCs:

Reactor Vessel and Internals In describing effects of sleeve fragments on RV internals and RCS components downstream of the flow baffle, Reference 1 shows that the kinetic energy of the bounding size fragment at RCS velocity will have insignificant structural effect on the CSB, shroud, upper guide structure, RCS pipe inside surfaces and lower support structure (LSS). Minor damage to the bottom RV head cladding could occur, but due to the small fragment size downstream of the flow skirt, damage should be less significant than actual damage identified in the CSB / RV annulus, which was deemed acceptable.

A very small loose part might possibly enter gaps between RVI components such as the RV keyway, CSB snubber log to RV core stabilizing lug and fuel alignment plate to core shroud guide lug gaps.

Any such occurrence would likely result in crushing of the debris with minor surface effect on RVI components. Reference 2 evaluates upper and center CSB cylinder known damage from the "B" thermal sleeve, as well as postulated damage due to the possibility of more thermal sleeve failures.

The evaluation concludes stresses due to existing damage and stresses due to postulated additional damage are less than allowable. As a result, structural performance of the Reactor Vessel and Internals will not be impaired and there is no associated malfunction because all performance indicators comply with all applicable design requirements, even though margins between calculated values and allowable values may have been reduced in some cases. Note that should additional sleeves become dislodged, resulting damage must be validated to determine if it is within the bounds the postulated damage evaluation. References 2, 3, and 4 find the RV bottom head cladding damage, CSB upper and center cylinder damage and core stop lug damage to be acceptable as-is. Reference 3 concludes the RV bottom head and core stop lug damage is acceptable "as is" based on calculated flaw depth, which is based on conservative corrosion rates and extrapolated to the end of the 60 year design life. These acceptable evaluation results indicate no malfunctions are associated with these conditions, so the likelihood of a malfunction of the RV and RVI is unaffected.

CEAs As explained in the response to Question 1, Accident 15.1.3, the potential for CEA malfunctions associated with thermal sleeve debris is negligible and there is by definition not more than a minimal increase in likelihood.

-~Entergy NUCLEAR MANAGEMENT MANUAL QUALITY RELATED INFORMATIONAL USE 10 CFR 50.59 Evaluations EN-Ll-101 I

PAGE 15 OF 22 REV. 20 ATTACHMENT 9.1 50.59 EVALUATION FORM Pressurizer and Surge Line Should thermal sleeve fragments carry through the RV. by reactor coolant flow, they could enter the HL containing the pressurizer surge line. Postulated fragment sizes could enter the surge line during an in-surge condition. Reference 1 states, if parts are conservatively assumed to enter the line rather than continue in the high RCS flow, any impact forces are negligible due to the small fragment size. In the unlikely event pieces are carried to the pressurizer, they would settle to the bottom head and remain there having no expected effect on the pressurizer heaters due to the small mass, and no effect on the vertically oriented nozzles to pressurizer level instrumentation. These acceptable evaluation results indicate no malfunctions are associated with these conditions, so the likelihood of a malfunction of the Pressurizer, Surge Line and Spray Line is unaffected.

Pressurizer Spray As stated in Reference 1, if loose parts are postulated to enter pressurizer spray piping, the fragments must first pass through the core, hot legs, SGs, and RCPs before entering the two CLs supplying pressurizer spray. If it is further postulated that debris impairs pressurizer spray flow, either high or low, there are no serious consequences according to Reference 14. Pressurizer pressure could be controlled via auxiliary spray or pressurizer venting concurrent with automatic reactor trips. Debris passing through the spray valves would also pass through the spray nozzle and

• would be expected to settle to the bottom of the pressurizer as described in the preceding pressurizer discussion. The potential for Pressurizer Spray malfunctions as described in this hypothetical scenario are negligible and by definition do not constitute a more than minimal increase in likelihood.

As explained in the response to Question 1, Accident 15.1.5, the potential for RCP malfunctions associated with thermal sleeve debris is negligible and there is by definition not more than a minimal increase in likelihood.

As explained in the response to Question 1, Accident 15.1.18, the potential for SG tube malfunctions associated with thermal sleeve debris is negligible. If postulated fragments enter a SG primary chamber, impacts to the tubesheet, channel head and divider plate could occur. The divider plate is manufactured from lnconel and the tubesheet and channel head are covered by cladding. Any thermal sleeve fragment impact indentations would be less than the thickness of the cladding. It is anticipated that insignificant damage would occur since fragments are small and impact force is limited. The potential for SG malfunctions associated with thermal sleeve debris is negligible and there is by definition not more than a minimal increase in likelihood.

RCS Instrumentation and RV Level Instrumentation RCS instrumentation includes the hot leg and cold leg temperature measurement RTDs and the pressurizer level and pressure instrumentation, including static differential pressure sensors attached to pressurizer instrument nozzles. The hot leg and cold leg RTDs are isolated from direct contact with debris since they are inserted in TWs in the hot leg and cold leg piping. The postulated fragment size would have passed through the core and should TW contact occur, it would be with negligible force. Pressurizer instrument lines experience no flow and are not susceptible to

ff)Entergy NUCLEAR MANAGEMENT QUALITY RELATED EN-Ll-101 I REV. 20 MANUAL INFORMATIONAL USE PAGE 16 OF 22 10 CFR 50.59 Evaluations ATTACHMENT 9.1 50.59 EVALUATION FORM blockage. The reactor vessel level measuring system (RVLMS) [Ref. 15] includes two level measurement instruments inserted into instrument guide tubes in the RV. This system supplies reactor vessel level and other information to operators during certain accident scenarios.

Reference 1 states openings in instrument guide tubes are oriented go to the primary flow direction 0

with the exception of a flow hole. For debris to enter holes oriented go it must travel perpendicular 0

to the direction of flow. The flow hole is 0.040" diameter and any very small debris that would enter that opening would not cause adverse impact to instruments.

Safety Injection System (SIS), Safety Injection Tanks (SIT), Containment Spray System (CSS)

As described in Question 1, 15.1.13 there are no malfunctions related to SI nozzle structural integrity. The SIS and CSS initially take suction from the refueling water tank (RWf) and do not draw inventory from the RCS, so RCS debris is not a concern. If it is postulated that debris enters the RWf during refueling cavity drain down, fragments would be expected to settle to the bottom of the tank. Due to the suction strainer and short standpipe suction configuration, fragments would not be expected to enter the SIS or CSS during RWf injection. As RWf drains down, suction is transferred to the containment sump which is equipped with robust filtration. Since safety injection actuation is the result of a RCS loss of coolant accident (LOCA), any debris in the RCS could be discharged into containment via the leak. When the RWf water supply is depleted, SIS and CSS suction transfer to the containment sump in recirculation mode. The sump is provided with a filtration system that will capture debris of a size that could affect SIS and CSS performance. Any postulated debris that might actually transport to the sump would be of a quantity so small as to not affect sump performance. Therefore, any postulated thermal sleeve debris discharged during a LOCA would not affect SIS or CSS performance.

The SITs are self-contained and are filled from the High Pressure Safety Injection System (HPSI) or HPSI Pressurization System. No sleeve fragments would be present in these systems since the suction is from the filtered and purified eves. Therefore, the potential for SIS, SIT, and CSS malfunctions associated with thermal sleeve debris is negligible and there is by definition not more than a minimal increase in likelihood.

soc As explained in the response to Question 1, Accident 15.1.27, the potential for SOC malfunctions associated with thermal sleeve debris is negligible and there is by definition not more than a minimal increase in likelihood.

The eves takes RCS letdown, cools it to a temperature appropriate for downstream equipment, provides filtration and purification, temporary storage, and preheating prior to providing the purified coolant to the RCS via the charging function. It is the function of the letdown and purification portion of the system to remove contaminants, so that portion is not explicitly evaluated in Reference 1.

Upstream of the filtration system are the containment isolation valves. It is possible sleeve debris could become trapped on the valve seats resulting in possible leakage. However, because of isolation valve redundancy it is unlikely both valves would leak by simultaneously. The charging portion of the system takes suction from the volume control tank, which contains the filtered and purified letdown, as well as from the RWf. As discussed previously, debris exiting the RWf is unlikely. Therefore, the potential for eves malfunctions associated with thermal sleeve debris is negligible and there is by definition not more than a minimal increase in likelihood.

f')Entergy NUCLEAR MANAGEMENT QUALITY RELATED EN-Ll-101 I REV. 20 MANUAL INFORMATIONAL USE PAGE 17 OF 22 10 CFR 50.59 Evaluations ATTACHMENT9.1 50.59 EVALUATION FORM Spent Fuel Pool (SFP)

During refueling, it is possible any fragments present in the fuel could transfer with the fuel from the refueling canal to the SFP. Should fragments then become dislodged from the fuel, they would settle to the bottom of the storage rack or pool floor. Due to the high suction point for SFP cooling system, fragments would not enter SFP cooling. Any fragments on the pool bottom would not be expected to re-introduce to fuel assemblies due to the low natural circulation flow. Therefore, the potential for SFP malfunctions associated with thermal-sleeve debris is negligible and there is by definition not more than a minimal increase in likelihood.

RCS Chemistry and materials The displaced thermal sleeves are original plant design and are a normal component in the RCS.

As such, their location in the RCS has no effect on RCS chemistry and materials. There are no malfunctions relative to RCS chemistry and materials, so the likelihood of a malfunction is unaffected.

Fu~I Reference 1 shows that postulated blockage from large or small debris will not adversely affect DNB.

Reference 1 shows that DNBR is not significantly affected by any postulated RC flow blockage from sleeve fragments. Any relatively large area blockage such as at the flow skirt or fuel bundle inlet does not significantly affect DNB based on a representative test with 36% of the flow area between rods blocked. Flow redistributes downstream within approximately one grid span. Another test with Westinghouse fuel shows full flow recovery within 30", even with total blockage of the fuel bundle inlet. Since minimum DNBR typically occurs at higher elevations in the core, these postulated blockages have no significant effect on minimum DNBR. Further, it is highly unlikely debris would accumulate in one area of the core as evidenced by location of sleeve fragments spread about the core support plate, leaving a small likelihood that one or more fuel assemblies would see 36% flow blockage. The Task 2 RCS hydraulic assessment in Reference 2 is based on the unlikely case the remaining three thermal sleeves become dislodged, travel to the flow skirt, and unroll intact, blocking the maximum practicable flow area. It considers SG tube plugging and concludes the resulting RCS flow with maximum practicable flow skirt blockage would remain above the Technical Specification (TS) 3.2.5, "RCS Flow Rate," minimum limit of 120.4 x x10 6 lbm/hr. The assessment also demonstrates flow skirt blockage to be more limiting than plausible blockage at other locations in the RCS flow path. Further, Reference 2 states the four RCP normal operation scenario bounds lower flow scenarios including less than four RCP operation, SOC operation, and natural circulation.

Therefore, there is not a more than minimal increase in the likelihood of occurrence of a fuel related malfunction as a result of this change.

NUCLEAR QUALITY RELATED EN-Ll-101 REV. 20 f!)Entergy MANAGEMENT MANUAL INFORMATIONAL USE PAGE 18 OF 22 10 CFR 50.59 Evaluations ATTACHMENT 9.1 50.59 EVALUATION FORM

3. Result in more than a minimal increase in the consequences of an accident previously D Yes evaluated in the SAR? 1:8:1 No BASIS:

NEI 96-07 clarifies that for the purposes of 10CFR50.59, consequences means dose. The six accidents deemed relevant to th.is change are discussed in detail in Question 1 to provide a basis for response to that question. That discussion, as appropriate to consequences, is considered in the response to Question 3. The accident consequence discussion is also relevant to two additional accidents, 15.1.2 and 15.1.14, where DNBR initial condition sensitivity is a consideration .

..;..15

.....___1___.2....,_.........

U__n__c__ ontrolled CEA Withdrawal from Critical Conditions UFSAR 15.1.2.1 states a continuous withdrawal of CEAs could result from a malfunction in the Control Element Drive Mechanism Control System. The relationship between the change and this accident relates to accident sensitivity to DNBR and any potential DNBR effects. Reference 1 shows that DNBR is not significantly affected by any postulated RC flow blockage from sleeve fragments. Any relatively large area blockage such as at the flow skirt or fuel bundle inlet does not significantly affect DNB based on a representative test with 36% of the flow area between rods blocked. Flow redistributes downstream within approximately one grid span. Another test with Westinghouse fuel shows full flow recovery within 30", even with total blockage of the fuel bundle inlet. Since minimum DNBR typically occurs at higher elevations in the core, these postulated blockages have no significant effect on minimum DNBR. Further, it is highly unlikely debris would accumulate in one area of the core as evidenced by location of sleeve fragments spread about the core support plate, leaving a small likelihood that one or more fuel assemblies would see 36% flow blockage.

No failure causing a CEA withdrawal or insertion can prevent CEA insertion upon a trip signal. The Reactor Protective System (RPS) pressurizer safety valves assure safety limits are not exceeded.

The change does not affect initial conditions, does not alter, prevent, or degrade the effectiveness of actions in the UFSAR, nor does it alter any assumptions used to evaluate radiological consequences. Therefore, consequences of this accident are not affected by this change .

..;..15__._1__.3_,_______ C___E___A.......Misoperation UFSAR 15.1.3.1 states protection for CEA misoperation events is provided either by a Core Protection Calculator System (CPCS) trip or by providing adequate DNBR and local power density margin. The consequences of this accident are not affected by this change. The change does not alter, prevent, or degrade the effectiveness of actions in the UFSAR, nor does it alter any assumptions used to evaluate radiological consequences. Further, the change does not serve a direct role in mitigating accident consequences.

""1__5._1~.5_,__T~o .....t=a-I and Partial Loss of Reactor Coolant Forced Flow UFSAR 15.1.5.2.2.1 states protection against exceeding the DNBR limit due to a decrease in coolant flow is provided by initial steady state thermal margin maintained by adhering to TS requirements for DNBR margin and by RPS trip on low DNBR. The consequences of this accident are not affected by this change. The change does not alter, prevent, or degrade the effectiveness of

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PAGE 19 OF 22 REV. 20 ATTACHMENT 9.1 50.59 EVALUATION FORM actions in the UFSAR, nor does it alter any assumptions used to evaluate radiological consequences. Further, the change does not serve a direct role in mitigating accident consequences.

15.1.13, Major Rupture of Pipes Containing Reactor Coolant up to and Including Double Ended Rupture of the Largest Pipe in the RCS - Loss of Coolant Accident (LOCA)

The consequences of this accident are not affected by this change. The change does not alter, prevent, or degrade the effectiveness of actions in the UFSAR, nor does it alter any assumptions used to evaluate radiological consequences. Further, the change does not serve a direct role in mitigating accident consequences.

15.1.14, Major Secondary System Pipe Breaks with or without a Concurrent Loss of AC Power The relationship between the change and this accident relates to accident sensitivity to DNBR and any potential DNBR effects resulting from the change. Reference 1 shows that DNBR is not significantly affected by any postulated RC flow blockage from sleeve fragments. Any relatively large area blockage such as at the flow skirt or fuel bundle inlet does not significantly affect DNB based on a representative test with 36% of the flow area between rods blocked. Flow redistributes downstream within approximately one grid span. Another test with Westinghouse fuel shows full flow recovery within 30", even with total blockage of the fuel bundle inlet. Since minimum DNBR typically occurs at higher elevations in the core, these postulated blockages have no significant effect on minimum DNBR. Further, it is highly unlikely debris would accumulate in one area of the core as evidenced by location of sleeve fragments spread about the core support plate, leaving a small likelihood that one or more fuel assemblies would see 36% flow blockage. The change does not affect initial conditions, does not alter, prevent, or degrade the effectiveness of actions in the UFSAR, nor does it alter any assumptions used to evaluate radiological consequences.

15.1.18, Steam Generator Tube Rupture with or without a Concurrent Loss of AC Power As described in Question 1, debris resulting from a dislodged sleeve is unlikely to result in SG tube damage. This fact assures the second barrier to radionuclide release is preserved.

UFSAR 15.1.18.1 states radioactivity released is based on a double-ended break of one SG tube.

Therefore, consequences of this accident are not affected by this change. The change does not alter, prevent, or degrade the effectiveness of actions in the UFSAR, nor does it alter any assumptions used to evaluate radiological consequences. Further, the change does nof serve a direct role in mitigating accident consequences.

15.1.25, Fuel Cladding Failure Combined with Steam Generator Leak As described in Section I, debris resulting from a dislodged sleeve is unlikely to result in SG tube damage. It is possible debris could become lodged within a fuel bundle, resulting in potential long-term damage to fuel cladding, the first barrier to radionuclide release. Regarding potential flow blockage effects on tube cladding, Reference 1 shows that DNBR is not significantly affected by any postulated RC flow blockage from sleeve fragments. Any relatively large area blockage such as at the flow skirt or fuel bundle inlet does not significantly affect DNB based on a representative test with 36% of the flow area between rods blocked. Flow redistributes downstream within approximately one grid span. Another test with Westinghouse fuel shows full flow recovery within 30", even with

fl)Entergy NUCLEAR MANAGEMENT QUALITY RELATED EN-Ll-101 I REV. 20 MANUAL INFORMATIONAL USE PAGE 20 OF 22 10 CFR 50.59 Evaluations ATTACHMENT 9.1 50.59 EVALUATION FORM total blockage of the fuel bundle inlet. Since minimum ONBR typically occurs at higher elevations in the core, these postulated blockages have no significant effect on minimum ONBR. Further, it is highly unlikely debris would accumulate in one area of the core as evidenced by location of sleeve fragments spread about the core support plate, leaving a small likelihood that ~me or more fuel assemblies would see 36% flow blockage. UFSAR 15.1.25 states consequences for this accident are based upon a continuous one gpm primary to secondary SG tube leak concurrent with an RCS specific activity equivalent to one percent failed fuel. As mentioned above, it is possible the conservatively postulated sleeve debris could result in long term damage to fuel cladding. However, the absence of indications of fuel damage to date and conservatisms in the evaluation indicate the possibility of 1% failed fuel is highly unlikely. Further, RCS activity is routinely monitored to assure compliance with applicable regulations. The change does not alter, prevent, or degrade the effectiveness of actions in the UFSAR, nor does it alter any assumptions used to evaluate radiological consequences. Further, the change does not serve a direct role in mitigating accident consequences. It is therefore concluded radiological dose consequence mitigators are not altered or affected and therefore consequences are not altered or affected.

15.1.27. Failure or Overpressurization of Low Pressure Residual Heat Removal System The relationship of the change to this accident is the potential for tube leakage allowing RCS to leak into SW. However, there is no discussion in the UFSAR of this potential condition being a contributor to this accident. That said, the Low Pressure Residual Heat Removal System, or Shutdown Cooling (SOC) system, has provisions to minimize radiological consequences in the unlikely event a SOC heat exchanger tube leak. A process radiation monitor is provided in the SW discharge from the SOC heat exchangers with an alarm in the control room to warn of RCS leakage into SW. The consequences of this accident are not affected by this change. The change does not alter, prevent, or degrade the effectiveness of actions in the UFSAR, nor does it alter any assumptions used to evaluate radiological consequences. Further, the change does not serve a direct role in mitigating accident consequences.

4. Result in more than a minimal increase in the consequences of a malfunction of an SSC D Yes important to safety previously evaluated in the SAR? igj No BASIS:

As described in Section I, the evaluation of damage and abnormal conditions show all affected components and systems, all of which are important to safety (although not necessarily credited in the accident analyses), are within applicable acceptance criteria and are acceptable for continued use through the end of the current license. Risk of debris related fuel cladding leaks is increased, but is considered low due to effectiveness of the fuel design in mitigating the risk for debris causing fuel damage. There are no impacted consequence mitigating SSCs beyond those described in Section I and/or in response to Question 2 above. Therefore, it is concluded there is not more than a minimal increase in the consequences of an SSC malfunction.

QUALITY RELATED EN-Ll-101 REV. 20 NUCLEAR MANAGEMENT MANUAL INFORMATIONAL USE PAGE 21 OF 22 10 CFR 50.59 Evaluations ATTACHMENT 9.1 50.59 EVALUATION FORM

5. Create a possibility for an accident of a different type than any previously evaluated in D Yes the SAR?

~ No BASIS:

The proposed activity documents a use-as-is evaluation for the damage and abnormal conditions, as well as applicable aspects of future failures, described in Section I. SSCs affected by these conditions have been evaluated and 'are within applicable design limits as discussed in Section I.

These system conditions are deemed acceptable and as such are not accident initiators. Therefore, since analysis and evaluation has shown the configuration is acceptable with respect to the original design, the design change does not create a possibility for an accident of a different type than any previously evaluated in the SAR.

6. Create a possibility for a malfunction of an SSC important to safety with a different result D Yes than any previously evaluated in,the SAR?

~ No BASIS:

The proposed activity documents a use-as-is evaluation for the damage and abnormal conditions, as well as applicable aspects of future failures, described in Section I. SSCs affected by these conditions have been evaluated and are within applicable design limits as discussed in Section I.

These system conditions are deemed acceptable and as such do not create the possibility of a malfunction. Therefore, since analysis and evaluation has shown the configuration is acceptable with respect to the original design, the design change does not create a possibility of a malfunction of an SSC important to safety with a different result than previously evaluated in the SAR.

7. Result in a design basis limit for a fission product barrier as described in the SAR being D Yes exceeded or altered?

~ No BASIS:

The NRC has defined a design basis limit for a fission product barrier as the controlling numerical value for a parameter established during the licensing review as presented in the SAR for any parameter(s) used to determine the integrity of the barrier. For power reactors, these barriers are generally limited to the fuel cladding, the RCS pressure boundary, and the containment building.

The fuel cladding, and RCS pressure boundary are shown, as discussed in Section I, to meet existing design limits. Containment is not affected by this change. Further, design basis limits are not changed for any plant system. Therefore, this change does not result in a design basis limit for a fission product barrier as described in the SAR being exceeded or altered.

-Entergy NUCLEAR MANAGEMENT QUALITY RELATED EN-Ll-101 I REV. 20 MANUAL INFORMATIONAL USE PAGE 22 OF 22 10 CFR 50.59 Evaluations ATTACHMENT 9.1 50.59 EVALUATION FORM

8. Result in a departure from a method of evaluation described in the SAR used in D Yes establishing the design bases or in the safety analyses? [8] No BASIS:

The methods of evaluation identified in the Westinghouse Phase 2 evaluations are as follows:

• Reference 3 evaluation methods are consistent with the Analyses of Record (AOR) using the criteria of the ASME Code, 1968 Edition with Addenda up to and including Summer 1970 or alternate methods are justified to be equivalent."

• Reference 4 evaluation methods utilized are consistent with the AOR using the criteria of ASME 1971."

• Reference 2 section 2.2 evaluations of the effect of thermal sleeve flow blockages on RCS flow are performed for the historical system condition of ANO-2 in late November 2011 at the time following replacement of one of RCP impellers.

• Reference 2 section 3.2.1 states that the analyses of record (AOR) for the reactor internals do not contain detailed stress intensity calculations. Rather, these analyses perform a margin assessment based on prior analyses and load changes in a conservative manner. For this evaluation, the detailed stress intensity calculations are performed and compared to acceptance criteria, thus removing the inherent conservatism in the AOR.

• Reference 2 section 4.1 evaluates the structural impact of thermal sleeves caught in the flow baffle region and considers the structural integrity of the flow baffle, RV bottom head, and core stop lugs from a qualitative perspective.

• Reference 2 section 5.1 states that the RV bottom head flaw analysis is being performed in accordance with ASME Code. The Code year for the Section 111 analysis is 1968 ASME Code with Addenda through Summer 1970, and the Code year for the Section XI analysis is 2007 ASME Code with Addenda through 2008.

• Reference 16 states that the EAF evaluation is to be performed based on NUREG / CR-6717 per the ANO-2 LRA. This is accomplished with finite element analysis (FEA) and site projected cycle data instead of the design cycle values.

The methods of evaluation identified above do not depart from the methods used in the applicable AORs or the ASME codes and do not result in a departure from a method of evaluation described in the SAR for the safety analysis and in establishing the design basis.

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-Ll-103.

• --1 I

I ANO 50.59 Evaluation 2020-003

~Entergy NUCLEAR MANAGEMENT MANUAL QUALITY RELATED INFORMATIONAL USE 10 CFR 50.59 Evaluations EN-Ll-101 I

PAGE 1 OF 9 REV. 20 ATTACHMENT 9.1 50.59 EVALUATION FORM I. OVERVIEW/ SIGNATURES 1 Facility: Arkansas Nuclear One, Unit 2 (ANO-2) Evaluation#/ Rev.#: 2020-003 / 0 Proposed Change I Document: OP-2203.012F, Annunciator 2K06 Corrective Action Description of Change:

This activity evaluates the acceptability of crediting operator action to maintain the operability of the "B" Containment Spray System (CSS) train with the outlet of the "B" Shutdown Cooling (SOC) heat exchanger (2SW-11 B) closed or significantly throttled. The activity constitutes an adverse change to the method of performing or controlling the design function of the CSS train.

Service Water (SW) is designed to automatically align to the respective CSS train SOC heat exchanger upon receipt of a Recirculation Actuation Signal (RAS). RAS is initiated following the injection phase of an accident upon a low Refueling Water Tank (RWT) level of 6 +/- 0.5%. The ANO-2 Safety Analysis Report (SAR) assumes a RAS will be received no earlier than 30 minutes following the onset of an accident (reference SAR Section 6.3.2.2.1 ). The automatic SW alignment to the "B" SOC heat exchanger is temporarily being replaced with a manual action until the heat exchanger's inlet motor operated valve (MOV) 2CV-1456-2 can be repaired.

Currently, the SW inlet valve to the "B" SOC heat exchanger (2CV-1456-2) is degraded and subsequently secured in the full open position (fail safe position). However, this configuration draws significant SW flow from the rest of the SW system, which also supplies Auxiliary Cooling Water (ACW) loads. The non-safety related ACW loads are required to maintain normal plant power operations. Due to the elevated outside temperatures during the summer months, maintaining 2CV-1456-2 in the full open position has significantly reduced cooling margin for the components supplied by ACW and, in turn, present a challenge to steady-state plant operation. Therefore, this activity proposes to significantly throttle or close, if necessary, the "B" SOC heat exchanger outlet valve (2SW-11 B) to restore cooling capacity in support of continued steady-state plant operation until 2CV-1456-2 can be repaired.

The configuration of the "B" CSS train SOC heat exchanger until maintenance can be performed will be controlled with the in MOV secured in the full open position (design function met), but with the outlet valve 2SW-11 B closed or significantly throttled (does not meet design function without operator action).

In accordance with Section 08.05 of IMC 326, "Operability Determinations," manual action cannot be substituted for automatic action if such involves the protection of a safety limit. Safety limits are required by 10 CFR 50.36, "Technical Specifications," and for ANO-2, are contained in Section 2 of the station TSs. This activity does not impact a safety limit (or limiting safety system setting) as described in Section 2 of the ANO-2 TSs but is related to meeting the specified safety function of the CSS as described in TS 3.6.2.1, "Containment Spray System."

In accordance with Section 06.08 of IMC 326, the purpose of the 10 CFR 50.59 (heretofore referred to as "5059") evaluation is to assess potential adverse impacts on other plant structures, systems, or components (SSCs), not.the degraded SSC itself. Nevertheless, impacts to both the "B" CSS train and any other plant SSC are addressed herein.

'l)Entergy NUCLEAR MANAGEMENT QUALITY RELATED EN-Ll-101 I REV. 20 MANUAL INFORMATIONAL USE PAGE 2 OF 9 10 CFR 50.59 Evaluations ATTACHMENT 9.1 50.59 EVALUATION FORM The above discussion, along with the information provided in the Summary of Evaluation section below and the response to the eight questions which follow, complies with the requirements of EN-OP-104, "Operability Determination Process," Step 8.10.7, "Crediting Operator Actions for Operability" (which is based on the guidance of IMC 0326).

Summary of Evaluation:

In accordance with NEI 96-07, Step 4.3.2, the example excerpted (Example 4) and included in response to Question 2 below illustrates a case where there would not be more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety. The responses to the eight 5059 questions herein support the conclusion that this change does not require prior NRC approval. Based on this evaluation, the proposed activity does not present a significant adverse impact to nuclear or public safety and meets the NRG-endorsed guidance of NEI 96-07.

Is the validity of this Evaluation dependent on any other change? [8] Yes 0 No If "Yes," list the required changeslsubmittals. 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.

This 5059 assumes the following have been completed or established prior to relying on operator action to open 2SW-11 B:

• Revision to OP-2203.012F, "Annunciator 2K06 Corrective Action"

• Installation of red placards on the low and low-low RWT level alarm windows.

• Establishing a standing order of the subject matter and placed in the Control Room.

• Designation of an operator that is not part of the required normal shift staffing.

Based on the results of this 50.59 Evaluation, does the proposed change D Yes [8] No require prior NRC approval?

f')Entergy NUCLEAR MANAGEMENT QUALITY RELATED EN-Ll-101 I REV. 20 MANUAL INFORMATIONAL USE PAGE 3 OF 9 10 CFR 50.59 Evaluations ATTACHMENT 9.1 50.59 EVALUATION FORM Preparer1*2 : David Bice / see email below / EOI / Regulatory Assurance / 07-21-2020 Name (print) / Signature / Company / Department / Date Reviewer 1*2 : Phil Couture Ill / see email below / EOI / Fleet Licensing Programs I 07-21-2020 Name (print) / Signature / Company / Department / bate Independent _N_/_A_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

Review1*3 : Name (print) / Signature / Company / Department / Date Responsible Manager Riley Keele Jr / see email below / EOI / Regulatory Assurance / 07-22-2020 Concurrence 1 : Name (print) / Signature :J Company / Department / Date 50.59 Program David Bice / see email below / EOI / Regulatory Assurance / 07-21-2020 Coordinator Name (print) / Signature / Company I Department / Date Concurrence 1 :

Bryan Daiber / see OSRC 2020-009 Meeting Minutes / 07-22-2020 Chairman's Name (print) / Signature / Date [GGNS P-33633, P-34230, & P-34420; W3 P-151]

2020°009 OSRC Meeting #

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 Asset Suite." 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 Asset Suite, attach a screenshot of the electronic signature(s); if using PCRS, 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].

"='Entergy NUCLEAR MANAGEMENT MANUAL QUALITY RELATED INFORMATIONAL USE 10 CFR 50.59 Evaluations EN-Ll-101 PAGE 4OF9 I REV. 20 ATTACHMENT 9.1 50.59 EVALUATION FORM EMAILS From: BICE, DAVID B (ANO) <DBICE@entergy.com>

Sent: Tuesday, July 21, 2020 3:43:06 PM To: Couture 111, Philip <pcoutur@entergy.com>

Subject:

FW: Peer Check and Validation - SPTA and EOP Revs Proposed for 2CV-1456-2 and 2SW-11 B I have performed the attached PAD and 5059 associated with manual operation of the "B" SOC Hx outlet valve 2SW-11 B. Please review and provide concurrence with these two documents. 'Your concurrence via reply email will be considered the Reviewer signature on the PAD and 5059 forms. Thanks.

From: Couture Ill, Philip <pcoutur@entergy.com>

Sent: Tuesday, July 21, 2020 4:13 PM To: BICE, DAVID B (ANO) <DBICE@entergy.com>

Subject:

Re: Peer Check and Validation - SPTA and EOP Revs Proposed for 2CV-1456-2 and 2SW-11 B You can sign me off as reviewer for the PAD and 50.59 per this email.

From: BICE, DAVID B (ANO) <DBICE@entergy.com>

Sent: Wednesday, July 22, 2020 7:16 AM To: Keele Jr, Riley D <rkeele@entergy.com>

Subject:

"B" SOC Hx 2SW-11 B 5059 Please review and provide manager concurrence via reply email with the subject 5059. Thanks.

From: Keele Jr, Riley D <rkeele@entergy.com>

Sent: Wednesday, July 22, 2020 7:23 AM To: BICE, DAVID B (ANO) <DBICE@entergy.com>

Subject:

RE: "B" SOC Hx 2SW-11 B 5059 I concur with the attached 50.59 regarding manual closure of 2SW-11 B due to 2CV-1456-2 degradation.

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 Asset Suite." 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 Asset Suite, attach a screenshot of the electronic signature(s); if using PCRS, 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].

-Entergy NUCLEAR MANAGEMENT QUALITY RELATED EN-Ll-101 I REV. 20 MANUAL INFORMATIONAL USE PAGE 5 OF 9 10 CFR 50.59 Evaluations ATTACHMENT 9.1 50.59 EVALUATION FORM II. 50.59 EVALUATION [10 CFR 50.59(c)(2)]

Does the proposed Change being evaluated represent a change to a method of D Yes evaluation ONLY? If "Yes," Questions 1 - 7 are not applicable; answer only ~ No Question 8. If "No," answer all questions below.

Does the proposed Change:

1. Result in more than a minimal increase in the frequency of occurrence of an accident D Yes previously evaluated in the SAR? ~ No BASIS:

The CSS is an accident mitigation feature and is unrelated to any accident initiator. In addition, SW supply to the SOC heat exchanger cannot cause an accident, whether the SW supply is aligned or secured. Therefore, this change does not result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the SAR.

2. Result in more than a minimal increase in the likelihood of occurrence of a malfunction of D Yes a structure, system, or component (SSC) important to safety previously evaluated in the ~ No SAR?

BASIS:

In accordance with NEI 96-07, "Guidelines for 10 CFR 50.59 Implementation," as endorsed by the NRC via Regulatory Guide (RG) 1.187, "Guidelines for Implementing of 10 CFR 50.59, changes, Tests, and Experiments," manual operator action can be temporarily credited in lieu of automatic action provided the following is met (from NEI 96-07, Section 4.3.2, Example 4):

The change involves a new or modified operator action that supports a design function credited in safety analyses provided:

• The action (including required completion time) is reflected in plant procedures and operator training programs

• The licensee has demonstrated that the action can be completed in the time required considering the aggregate affects, such as workload or environmental conditions, expected to exist when the action is required

• The evaluation of the change considers the ability to recover from credible errors in performance of manual actions and the expected time required to make such a recovery

• The evaluation considers the effect of the change on plant systems.

-Entergy NUCLEAR MANAGEMENT QUALITY RELATED EN-Ll-101 I REV. 20 MANUAL INFORMATIONAL USE PAGE 6 OF 9 10 CFR 50.59 Evaluations ATTACHMENT 9.1 50.59 EVALUATION FORM The action to sufficiently open 2SW-11 B will commence, for all practical purposes, at the onset of an accident. Accident initiation is easily and promptly recognized by Control Room personnel. .

Although not.all accidents require a mitigating response from the CSS, this action will be added to OP-2203.012F, "Annunciator 2K06 Corrective Action," (heretofore referred to as "ACA") such that it will be initiated very early following the onset of an accident. As stated above, the SW system (if required) will initially automatically realign to its emergency 'mode of operation if the initiated accident requires this realignment (i.e., no SW flow through the "B" SOC heat exchanger is automatically aligned until a RAS is received no less than 30 minutes later). The ACA will direct a designated operator (an operator in addition to normal shift compliment)-to access the valve locally- and begin opening 2SW-11 B. This ensures sufficient SW flow will be provided through the "B" SOC heat exchanger prior to receipt of a RAS such that the CSS design function can be met even upon a single failure of the "A" CSS train. Note that the SW system emergency alignment assumes flow to the SOC heat exchanger at any point during an accident; therefore, aligning SW to the SOC heat exchanger prior to RAS initiation has no impact on the safety function of the SW system.

In support of the above action, a standing order will be maintained in the ANO-2 Control Room and a red placard will be placed on the associated ACA alarm windows (low and low-low RWT level). As is normal practice, one of the control board operators will immediately address these alarms following the onset of an accident while the remaining control board operator will perform Standard Post Trip Actions (SPTAs) at the direction of the Control Room Supervisor (CRS). It is important to recognize that the 30-minute time to receipt of a RAS is based on worst-case loss of coolant accident (LOCA) with the RWT being at minimum level permitted by the Technical Specifications (91.7%) at accident onset. The RWT is maintained well above the minimum level as a normal practice (> 95%) and is currently at 95.8%. Therefore, more than 30-minutes is expected to be available between accident onset and the need for the RAS.

Given current RWT level, a simulator session was performed for the large break LOCA to assist in validating the response time of the designated operator. The first aiarm (low RWT level) was received two minutes from accident onset. Although both the operating crew and the designated operator have been pre-briefed on actions that will be required, it is conservatively assumed that five additional minutes will elapse for the control board operator to refer to the ACA and notify the designated operator to open 2SW-11 B. A designated operator was pre-staged-at one of the farthest points within the protected area from the Auxiliary Building. The operator was then timed, walking a normal pace fr.om the pre-stage area to the valve location (lowest point in Auxiliary Building and through simulating the full opening of the valve (using a stopwatch, a total time of 9 minutes and 35 seconds). The overall time is, therefore, assumed to be < 20 minutes from accident onset to the establishment of SW flow through the "B" SOC heat exchanger. The low-low RWT alarm was received 25 minutes following accident onset at which time the ACA directs the control board operator to validate SW flow through the "B" SOC heat exchanger. Although no credible error is expected to prevent the initial action to open 2SW-11 B from completing, this 2 nd alarm provides an additional opportunity of ensuring flow has been established.

Operator training programs currently include significant training with respect to valve locations and local valve operations. Nevertheless, upon taking the watch, the designated operator will demonstrate knowledge of the physical location of the valve and describe the actions necessary to access and open the valve. The designated operator's relief (next shift) will likewise demonstrate these attributes in the presence of the operator being relieved.

-Entergy NUCLEAR MANAGEMENT QUALITY RELATED EN-Ll-101 I REV. 20 MANUAL INFORMATIONAL USE PAGE 7 OF 9 10 CFR 50.59 Evaluations ATTACHMENT 9.1 50.59 EVALUATION FORM The preceding four paragraphs meet the requirements of the 1st NEI bullet denoted above.

Based on the designated operator being located within the protected area and the simplicity of the task (rotating the valve handwheel in the counter-clockwise direction a sufficient number of revolutions to establish necessary SW flow), the action can be completed sufficiently before a RAS would be received (conservative estimated total time of approximately 20 minutes). Being designated, the operator will not have any additional workload/responsibilities during the accident or be exposed to adverse environmental conditions that would delay this action. For example, a line break in the room where 2SW-11 B is located would be a single failure of the "B" train (whether-a break in the CSS or an emergency core cooling pump). In such an event, the remaining CSS train will provide the specified safety function (additional failures are not assumed) and no operator action will be required. Although area radiological levels will begin to increase following receipt of a RAS, the operator is expected to have completed the action and cleared the area prior to the RAS; therefore, no challenge to any Entergy or federal dose limits is credible. This paragraph meets the 2nd bullet denoted above.

Because the operator will be designated and demonstrate knowledge and understanding of the expected task upon taking the watch, it is not credible that the operator could fail to perform the task or open the wrong valve. In addition, an unforeseen separation of the valve disc from the valve stem is not assumed. The Control Room verification will take place when notified by the designated operator (local flow indication is available to the designated operator) that the action has been completed, but no later than when the low-low RWT level alarm is received (expected 25 minutes following accident onset). Therefore, there exists no credible error that could prevent the action from being achieved prior to RAS initiation. This meets the 3 rd bullet denoted above.

With respect to other SSCs, th.e closure or throttling of 2SW-11 B will provide additional cooling flow from the SW system to associated plant loads, providing margin to operating limits with respect to controlling SSC temperatures. This .is the normal configuration for all operations where a RAS is not present (i.e., SW isolated to* the SOC heat exchanger) and, therefore, has no impact on the operation of other plant SSCs. Should an associated accident occur in this configuration, all non-essential loads are shed from the SW system. Therefore, the planned configuration will have no adverse impact on any other plant SSCs. This meets the 4 th bullet denoted above.

Based on the establishment of a non-crew designated operator, the ease of the task to be performed, the prompt notification to initiate the task from the ACA procedure, and the ample time (up to at least 30 minutes) to complete the task, the likelihood of not establishing SW flow in support of the CSS long-term recirculation design function is not significantly impacted. This is also consistent with the .conclusion of Step 4.3.2 of NRG-endorsed NEI 96-07 as described in the "Summary of Evaluation" section of this 5059 evaluation. Therefore, this change does not result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component (SSC) important to safety previously evaluated in the SAR.

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PAGE 8 OF 9 REV. 20 ATTACHMENT 9.1 50.59 EVALUATION FORM

3. Result in more than a minimal increase in the consequences of an accident previously D Yes evaluated in the SAR? ~ No BASIS:

As provided in response to Question 2 above, there is no credible error that could prevent establishment of SW flow to the "B" SOC heat exchanger in support of meeting the design function of the CSS during the recirculation mode of operation, even with the assumed failure of the "A" CSS train. With the CSS fulfilling meeting its specified safety function, there cannot be more than a minimal increase in the consequences of an accident previously evaluated in .the SAR which relies on CSS performance.

4. Result in more than a minimal increase in the consequences of a malfunction of an SSC D Yes important to safety previously evaluated in the SAR? ~ No BASIS:

As discussed above, the manual action to initiate SW flow through the "B" SOC heat exchanger will not prevent fulfillment of the CSS's specified safety function. Therefore, this change will not result in more than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the SAR.

5. Create a possibility for an accident of a different type than any previously evaluated in the D Yes SAR? ~ No BASIS:

As described in response to Question 1, the CSS supports accident mitigation and is unrelated to any accident initiator. Because there is no credible error that would prevent alignment of the SW system to the "B" SOC heat exchanger, the respective specified safety function will continue to be met even should a single failure of the "A" CSS occur at accident onset. Therefore, this change does not create a possibility for an accident of a different type than any previously evaluated in the SAR.

6. Create a possibility for a malfunction of an SSC important to safety with a different result D Yes than any previously evaluated in the SAR? ~ No BASIS:

The SAR assumes a single failure of one train of a given system, dependent on the accident being evaluated. This change does not present a credible failure mode that would prevent the "B" CSS train from performing its specified safety function. Any other failure of the "B" CSS train unrelated to the operator action addressed by this 5059 evaluation would result in the "A" CSS train performing the required safety function. Therefore, this change does not create the possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the SAR.

~Entergy NUCLEAR MANAGEMENT MANUAL QUALITY RELATED INFORMATIONAL USE 10 CFR 50.59 Evaluations EN-Ll-101 I

PAGE 9 OF 9 REV. 20 ATTACHMENT 9.1 50.59 EVALUATION FORM

7. Result in a design basis limit for a fission product barrier as described in the SAR being D Yes exceeded or altered? [8l No BASIS:

With no credible error being prevalent that could prevent the "B" CSS train from performing its specified safety function solely in relation to this activity (i.e., alignment of SW to the "B" SOC heat exchanger), either or both CSS trains will remain available to ensure no fission product barrier (fuel clad, Reactor Coolant System pressure boundary, Containment Building) limit is challenged, exceeded, or altered.

8. Result in a departure from a method of evaluation described in the SAR used in D Yes establishing the design bases or in the safety analyses? [8l No BASIS:

This change does not alter any computer code, calculation, or other method used in evaluating the impacts of a design bases accident as described in the SAR. Therefore, this change does not result in a departure from a method of evaluation described in the SAR 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-Ll-103.

Attachment 5 to 2CAN102001

.List of Affected SAR Pages to 2CAN102001 Page 1 of 1 List of Affected SAR Pages The following is a list of SAR pages revised in Amendment 29 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 4.5-1 9.5-21 15.3-16 2.2-1 4.7-6 9.5-24 15.3-17 2.2-7 4.7-14 9.5-25 15.3-26 2.4-1 4.7-22 9.5-26 15.3-45 2.4-5 Figure 4.3-1 Figure 9.2-1 B 15.3-53 2.8-1 Figure 4.3-1A Figure 9.2-6 15.3-58 2.8-2 Figure 4.3-1C Figure 9.3-1 15.3-66 2.8-10 Figure 4.3-1D 10.4-13 15.3-67 Figure 2.2-2 Figure 4.3-1E 10.4-21 15.3-104 Figure 2.2-3 Figure 4.3-2 10.5-6 15.3-107 Figure 2.4-7 Figure 4.3-3 Figure 11.2-1 15.3-130 Figure 2.5-102 Figure 4.3-4 15.1-5 Figure 15.1.0-1 F Figure 2.8-1 Figure 4.3-5 15.1-6 Figure 15.1.1-9 Figure 2.8-2 Figure 4.3-6 15.1-19 Figure 15.1.1-10 Figure 2.8-3 Figure 4.3-7 15.1-20 Figure 15.1.1-11 3.4-2 Figure 4.3-8 15.1-24 Figure 15.1.1-12 3.5-14 Figure 4.3-9 15.1-28 Figure 15.1.5-28 3.10~13 Figure 4.3-10 15.1-57 Figure 15.1.7-11 Figure 3.2-3 6.2-43 15.1-62 Figure 15.1.7-12 Figure 3.10-7 7.1-2 15.1-63 F[gure 15.1.7-13 Figure 3.10-8 8.1-1 15.1-66 Figure 15.1.7-14 Figure 3.10-9 8.3-6 15.1-76 Figure 15.1.7-15 4.2-18 Figure 8.3-6 15.1-111 4.2-72 Figure 8.3-17 15.1-119 4.2-81 Figure 8.3-18 15.1-130 4.2-86 9.2-13 15.1-145 4.4-1 9.5-16 15.3-1 4.4-26 9.5-20 15.3-3

Enclosure 2 to 2CAN102001 ANO-2 SAR Amendment 29 - Redacted Version (CD Rom)

(4286 Pages) -

Enclosure 3 to 2CAN102001 ANO-2 TRM (CD Rom)

(144 Pages)

Enclosure 4 to 2CAN102001 ANO-2 TS Table of Contents and TS Bases (CD Rom)

(136 Pages)

Enclosure 1 to 2CAN102001 ANO-2 SAR Amendment 29 - Un-redacted Version (CD Rom)

(4286 Pages)