Text

Transmittal of Amendment No. 26 to the Safety Analysis Report
	ML16132A490
Person / Time
Site:	Arkansas Nuclear
Issue date:	04/28/2016
From:	Jeremy G. Browning; Entergy Operations
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML16132A517	List: 2CAN041601, Transmittal of Amendment No. 26 to the Safety Analysis Report;
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	2CAN041601
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~Entergy 2CAN041601 April 28, 2016 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

Amendment 26 to the ANO Unit 2 Safety Analysis Report Arkansas Nuclear One, Unit 2 Docket No. 50-368 License No. NPF-6

Dear Sir or Madam:

Entergy Operations, Inc.

1448 S.R. 333 Russellville, AR 72802 Tel 479-858-3110 Jeremy G. Browning Vice President - Operations Arkansas Nuclear One In accordance with 10 CFR 50.71(e) and 10 CFR 50.4(b)(6), enclosed is Amendment 26 of the Arkansas Nuclear One, Unit 2 (AN0-2) Safety Analysis Report (SAR). Included with this update are the current AN0-2 Technical Requirements Manual (TRM) and the current AN0-2 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 AN0-2 refueling outage (2R24) which ended November 14, 2015. Summaries of changes to the AN0-2 TRM and TS Bases are included in Attachments 1 and 2 of this letter for the period beginning December 10, 2014, and ending April 28, 2016.

In accordance with NEI 98-03, Appendix A, Section A6, a list and short description of information removed from the SAR should be included with each SAR update submittal. For this reporting period, the following figures were removed from the AN0-2 SAR as a result of reviews performed associated with Regulatory Issue Summary (RIS) 2015-17, "Review and Submission of Updates to Final Safety Analysis Reports, Emergency Preparedness Documents, and Fire Protection Documents."

Figure 2.4-5 Figure 3.8-19 Figure 3.8-21 Figure 3.8-22 Figure 3.8-23 Figure 3.8-24 Figure 3.8-25 Section View Intake Structure Other Category 1 Structures - Auxiliary Building Plan & Section Other Category 1 Structures - Intake Structure Plan Other Category 1 Structures - Intake Structure Section Other Category 1 Structures Plans (included Diesel Fuel Oil Vault)

Other Category 1 Structures Sections (included Diesel Fuel Oil Vault)

Other Category 1 Structures - Emergency Cooling Pond Pipe Inlet and Pipe Outlet Structures

2CAN041601 Page 2 of 3 The above figures contained excessive details that might be useful to an adversary and are unnecessary for retention in the AN0-2 SAR. The AN0-2 SAR continues to contain sufficient detail, absent these figures, to permit an understanding of system designs and the associated relationship to safety evaluations, as required by NEI 98-03, "Guidelines for Updating Final Safety Analysis Reports."

The significant upgrade of the AN0-2 SAR was completed in the 2005 timeframe, associated in part with post 9/11/2001 response related to security sensitive information. In light of the previous upgrade and the removed figures listed above, Entergy has determined that the enclosed AN0-2 SAR does not contain information required to be withheld from public disclosure with respect to RIS 2015-17.

In accordance with 10 CFR 50.37(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 1 O CFR 50.59 evaluations associated with AN0-2 Licensing Basis Documents (LBDs) is normally included with the required SAR submittal or within 30 days thereafter. Attachment 3 contains a summary of 10 CFR 50.59 evaluations performed for both AN0-2, and any evaluation common between ANO, Unit 1, and AN0-2, over the aforementioned reporting period. Attachment 4 includes a copy of each of these evaluations.

If you have any questions or require additional information, please contact Stephenie Pyle at 479-858-4704.

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 April 28, 2016.

Sincerely, ORIGINAL SIGNED BY RONALD A. BARNES for JEREMY G. BROWNING JGB/dbb Attachments:

1. Summary of AN0-2 TRM Changes

2. Summary of AN0-2 TS Bases Changes

3. Summary of AN0-2 and Common 10 CFR 50.59 Evaluations

4. 10 CFR 50.59 Evaluations - December 9, 2014, through April 28, 2016

2CAN041601 Page 3 of 3

Enclosures:

1. AN0-2 SAR Amendment 26 (CD Rom)

2. AN0-2 TRM (CD Rom)

3. AN0-2 TS Table of Contents and TS Bases (CD Rom) cc:

Mr. Marc L. Dapas Regional Administrator U.S. Nuclear Regulatory Commission Region IV 1600 East Lamar Boulevard Arlington, TX 76011-4511 NRG Senior Resident Inspector Arkansas Nuclear One P. 0. Box 310 London, AR 72847 U.S. Nuclear Regulatory Commission Attn: Mr. Stephen Koenick MS 0-8B1A One White Flint North 11555 Rockville Pike Rockville, MD 20852 Mr. Bernard R. Bevill Arkansas Department of Health Radiation Control Section 4815 West Markham Street Slot #30 Little Rock, AR 72205 2CAN041601 Summary of AN0-2 TRM Changes to 2CAN041601 Page 1 of 1 Summary of AN0-2 TRM Changes The following changes to the Arkansas Nuclear One, Unit 2 (AN0-2) 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 58 59 60 61 62 Section 3.3.6, 3.7.2, 3.7.4, 3.7.5, B 3.3.3, B 3.7.1, B 3.7.2, B 3.7.3, B 3.7.4, B 3.7.5, B 3.8.6 3.7.5, B 3.7.5 3.10.1, B 3.10.1, 3.10.2, B3.10.2 3.7.2, 3.7.5, B 3.3.6, B 3.7.2, B 3.7.4, B 3.7.5 I

B 3.10.2 Summary TS Amendment 300, "Transition to NFPA 805" and Licensing Basis Document Change LBDC 15-025, "Clarification of Reporting Requirements for Inoperable Seismic Monitors" Condition Report CR-AN0-2-2015-2511, "Clarification of Inoperable Detector Actions for Fire Suppression Systems Not Required by TRMs 3.7.2 or 3.7.3." Editorial corrections are also made.

Engineering Change ECN-54966, "Installation of SFP FLEX Instrumentation and Equipment Connections" Condition Report CR-AN0-1-2015-2649, "Fire Barriers Separating Redundant Trains", Condition Report CR-AN0-2-2015-3215, "Testing of Containment Firewater Isolation Valves", and Engineering Change EC-48711, "CEDM/Computer Room Incipient Fire Detection Installation" Condition Report CR-ANO-C-2016-0850, "Clarification of FLEX Equipment, Connections, and Flow Paths" and CR-AN0-2-2016-0629, "Add Note for Sprinkler Systems not having Detection" 63 Table 3.3.6-1, B 3.3.6 Condition Report CR-AN0-2-2016-1099, "Clarification of Acronyms B

CEDM FLEX NFPA SFP TS CEDM Room Incipient Detection Requirements" (TRM) Bases Control Element Drive Mechanism Flexible and Diverse Coping Strategy National Fire Protection Association Spent Fuel Pool Technical Specification l

2CAN041601 Summary of AN0-2 TS Bases Changes to 2CAN041601 Page 1 of 1 Summary of AN0-2 TS Bases Changes The following changes to the Arkansas Nuclear One, Unit 2 (AN0-2) Technical Specification (TS) Bases were implemented in accordance with the provisions of 10 CFR 50.59 and the Bases Control Program of AN0-2 TS 6.5.14. Because these changes were implemented without prior NRC approval, a description is provided below:

Revision 54 55 56 57 58 59 60 Acronyms Section LCO B 3.0.4, SR B 3.0.3, B 3.8 B 3.6.1.3 B 3.3.2.1, B 3.3.3.1, B 3.4.1, B 3.6.1.3, B 3.6.1.4, B 3.6.2.1, B 3.6.2.3, B 3.6.3.1, B 3.7.3.1, B 3.7.6.1, B 3.8.1.1, B 3.8.2.3 B 4.0.2, B 3.4.9.1 B 3.6.1.3, B 3.9.2 B 3.7.6.1 B 3.3.1.1 B

(TS) Bases DC Direct Current RCS Reactor Coolant System RG Regulatory Guide Summary Licensing Basis Document Change LBDC 14-033, "Correct RG 1.182 Reference to RG 1.160" and TS Amendment 297, "DC Electrical Rewrite" TS Amendment 299, "Escape Hatch Seal Contact Check" TS Amendment 301, "Mode 4 End State" Condition Reports CR-AN0-2-2015-1793, "Clarification of SR 3.0.2 Applicability to Staggered Test Basis Frequencies" and CR-ANO-C-2015-2101, "Clarification of Step Change in RCS Temperature" Licensing Basis Document Change LBDC 15-045, "Clarify Mode 6 Source Range TS Function Bases" and Licensing Basis Document Change LBDC 15-045, "AN0-2 TS Bases 3.6.1.3 Editorial Correction" Condition Report CR-ANO-C-2016-0595, "Remove Reference to Staggered Test Basis" Condition Report CR-ANO-C-2015-1925, "Clarify Settings for Low/Degraded Voltage Protection" SR Surveillance Requirement 2CAN041601 Summary of AN0-2 and Common 10 CFR 50.59 Evaluations to 2CAN041601 Page 1 of 1 Summary of 10 CFR 50.59 Evaluations 50.59 #

50.59 Summary 2015-001 Engineering Change EC-53837, "AN0-2 Cycle 25 Reload Analysis Report and Core Operating Limits Report" 2016-001 Engineering Change EC-57218, "Flooding Analysis" 2CAN041601 10 CFR 50.59 Evaluations - December 9, 2014, through April 28, 2016

ANO 50.59 Evaluation 2015-001

NUCLEAR QUALllY RELATED EN-Ll-101 I REV.12

~Entergy MANAGEMENT MANUAL INFORMATIONAL USE PAGE 1 OF6 10 CFR 50.59 Evaluations ATTACHMENT 9.1 50.59 EVALUATION FORM I.

OVERVIEW I SIGNATURES1 Facility: Arkansas Nuclear One Unit 2 Evaluation #I Rev. #: 2015-001 I 0 Proposed Change I Document: Cycle 25 Reload Evaluation RAR and COLR Description of Change:

This evaluation addresses the Arkansas Nuclear One Unit 2 Cycle 25 Reload that resulted from the Cycle 25 core design performed by Westinghouse [Reference 1]. The Cycle 25 PAD identified the following change needing further evaluation under 10 CFR 50.59:

Change to the Core Operating Limits Report (COLR) Figures 4 & 5 DNBR Operating Limits All Cycle 25 Reload specific results met the applicable acceptance criteria and remain bounded by the results of Analyses of Re~ord (AOR).

Summary of Evaluation:

COLR Figures 4 & 5 DNBR Operating Limits Due to the core power distributions in Cycle 25 being different than those assumed in the bounding Thermal Hydraulic Analysis of Record (AOR), the calculated CETOP-D overpower penalty/credit factors are slightly lower than those described in the AOR. The overpower penalty/credit factors are input to various transient and setpoint analyses. Lower overpower penalty/credit factors required several transient analyses to be reviewed for impacts. This review resulted in revised corrected Required Overpower Margin (ROPM) values for Cycle 25 for the Loss of Coolant Flow, Over Cooling, Asymetric SIG Transient, Full Power Bank Withdrawal, and Single Control Element Assembly (CEA) Withdrawal within CPCS Deadband events. There were no impacts to the transient analyses and all transient analysis AORs remained bounding. The revised corrected ROPM, when incorporated into the setpoint analyses

[Reference 2], resulted in the need to change the Core Operating Limits Supervisory System (COLSS) out-of-service (OOS) limit lines for the DNBR Operating Limits in.the COLR. The COLSS OOS DNBR limit lines are modified for Cycle 25 to reflect the results of the setpoint analyses and assure the AORs remain bounding over the entire range of operating conditions.

References:

1.

Letter, A. L. Miller (Westinghouse) to R. E. Griffith (Entergy), "Arkansas Nuclear One Unit 2 Cycle 25 Final Reload Analysis Report," NF-AN0-15-23, June 22, 2015.

2.

Letter, A. L. Miller (Westinghouse) to R. E. Griffith (Entergy), "Startup Test and Setpoints Transmittal (STST) for AN0-2 Cycle 25," NF-AN0-15-40, September 22, 2015.

The printed name, company, department, and date must be included on the form. Signatures may be obtained via electronic processes (e.g., PCRS, ER processes), manual methods (e.g., ink signature), e-mail, or telecommunication. If using an e-mail or telecommunication, attach it to this form.

EN-Ll-101-ATI-9.1, Rev.12

NUCLEAR QUALITY RELATED EN-Ll-101 I REV.12 MANAGEMENT

~Entergy MANUAL INFORMATIONAL USE PAGE 20F 6 10 CFR 50.59 Evaluations ATTACHMENT 9.1 50.59 EVALUATION FORM Is the validity of this Evaluation dependent on any other change?

D Yes 12?J No If "Yes," list the required changes/submittals. The changes covered by this 50.59 Evaluation cannot be implemented without approval of the other identified changes (e.g., license amendment request). Establish an appropriate notification mechanism to ensure this action is completed.

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

Preparer:

Chris G. Eastus I See EC 53837 I ESI I Fuels-South I 9-11-15 Name (print) I Signature I Company I Department I Date D Yes Reviewer:

Daniel H. Williams I See EC 53837 I EOI /ANO Eng-Design I 9-25-15 Name (print) I Signature I Company I Department I Date OSRC:

Ronald A. Barnes I See EC 53837/10-14-15 Chairman's Name (print) I Signature I Date OSRC-2015-019 OSRC Meeting #

l2?J No EN-Ll-101-ATT-9.1, Rev. 12

NUCLEAR QUALITY RELATED EN-Ll-101 I REV.12 MANAGEMENT

~Entergy MANUAL INFORMATIONAL USE PAGE 30F 6 10 CFR 50.59 Evaluations ATTACHMENT 9.1 50.59 EVALUATION FORM II.

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

Does the proposed Change:

1.

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

BASIS:

D Yes

[2]

No D

Yes

[2]

No The Cycle 25 reload fuel does not initiate any accidents evaluated in the SAR. There are no changes to the Cycle 25 fuel design or fabrication that would have an impact on the frequency of occurrence of a fuel handling or mis-loading accident. *None of the barriers to a fuel mis-loading are affected by the Cycle 25 Reload. The lower CETOP-D overpower pena[ty/credit factors are inputs to the setpoint analyses and will not increase the frequency of any accident. Therefore, the Cycle 25 Reload will not result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in thj'l 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 important to safety previously evaluated in the

[2]

No UFSAR?

BASIS:

Since fuel design for the Cycle 25 core is essentially identical to the fuel design present in the Cycle 24 core, the Cycle 25 reload fuel will not require any equipment important to safety to be operated in a different manner or with a higher duty. As a result there is no more than a minimal increase in the likelihood bf a fuel failure. Therefore, the probability of a malfunction of a structure, system or component important to safety is not impacted due to the introduction of the Cycle 25 core.

EN-Ll-101-ATT-9.1, Rev. 12

NUCLEAR QUALITY RELATED EN-Ll-101 I REV.12 MANAGEMENT

~Entergy MANUAL INFORMATIONAL USE PAGE 40F 6 10 CFR 50.59 Evaluations ATTACHMENT9.1 50.59 EVALUATION FORM

3.

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

BASIS:

D Yes 1:8:1 No Cycle 25 reload safety analyses were evaluated/performed to assure that acceptance criteria are met for fuel performance, thermal-hydraulic performance, post-LOCA Emergency Core Cooling System (ECCS) performance, and non-LOCA event responses. Acceptable results from these analyses confirm that the core can be operated safely and meets all license requirements for accident response.

Changes to the Cycle 25 reload analyses were evaluated for possible impacts on the current Analysi~ of Record (AOR) results. For the lower CETOP-D overpower penalty/credit factors all of the accident analyses were shown to be bounded by their existing licensing basis AORs.

The COLR change ensures that ANO is operated in a manner that is consistent with the analysis assumptions. The change to the DNBR Operating Limits of COLR Figures 4 & 5 assures the accident analysis and their consequences as described in the SAR remaining bounding over the entire range of operation.

As a result of the COLR changes, the respective AORs remain bounding for Cycle 25, thus, there is no increase in the consequences of an accident previously evaluated in the SAR due to the Cycle 25 Reload.

4.

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

BASIS:

D Yes 1:8:1 No Cycle 25 reload safety analyses were performed to assure that acceptance criteria are met for fuel performance, thermal-hydraulic performance, post-LOCA ECCS performance and non-LOCA transient response. These analyses confirm that the Cycle 25 core can be operated safely and can be expected to meet license requirements for accident response.

The function and duty of SSCs important to safety as assumed in the safety analyses are not altered.

The Cycle 25 analyses do not place greater reliance on any specific plant system, structure, or component, including the fuel itself, to perform a safety function. No changes in the assumptions concerning equipment availability or failure modes have been made and none are necessary to implement Cycle 25.

The COLR change ensures that ANO is operated in a manner that is consistent with the capability of the SSCs. The change to the DNBR Operating Limits of COLRFigures 4 & 5 assures the accident analysis and their consequences as described in the SAR remaining bounding over the entire range of operation.

As a result of the COLR changes, the respective AORs remain bounding for Cycle 25, thus, there is no increase in the consequences of a malfunction of a structure, system, or component important to safety previously evaluated in the SAR by the Cycle 25 Reload.

EN-Ll-101-ATT-9.1, Rev. 12

QUALITY RELATED EN-Ll-101 REV.12

'='Entergy NUCLEAR MANAGEMENT MANUAL INFORMATIONAL USE PAGE 50F 6 10 CFR 50.59 Evaluations ATTACHMENT 9.1

5.

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

BASIS:

D Yes

~ No Cycle 25 Reload does not introduce any new operating conditions, plant configurations or failure modes that could lead to a credible accident of a different type than any previously evaluated in the SAR. No accident initiator is impacted by Cycle 25 Reload. Therefore, the possibility of an accident of a different type than any previously evaluated in the SAR will not be created.

6.

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

BASIS:

D Yes

~ No The Cycle 25 core design does not m,odify the design or operation of structures, systems, or components important to safety beyond the fuel itself. The reload core will not require any structure, system or component, including the fuel itself, important to safety to be operated in a different manner or with a higher duty. Structures, systems and components important to safety, including the fuel itself will function in the same manner in Cycle 25 as in previous cycles with the same NGF design. The core design is essentially the same between Cycle 24 and Cycle 25. The malfunction of the fuel in Cycle 25 will result in a fuel failure, the same result as in Cycle 24. The changes in core characteristics do not change any parameter that would affect the function of structures, systems or components important to safety. There are no new methods of failure associated with any of the changes associated with the Cycle 25 core.

7.

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

BASIS:

D Yes

~ No Cycle 25 reload safety analyses were performed to assure that acceptance criteria are met for fuel performance, thermal-hydraulic performance, post-LOCA ECCS performance and non-LOCA response. These analyses confirm that the core can be operated safely and can be expected to meet license requirements for accident response. The Cycle 25 reload safety analyses were performed to demonstrate compliance with the existing design basis limits for the fuel cladding, RCS pressure boundary and containment fission product barriers. For the fuel cladding, analyses have demonstrated that all design basis limits are met for the Cycle 25 core. The RCS pressure boundary and containment design basis limits are not affected by the Cycle 25 fuel.

8.

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

BASIS:

D Yes

~ No There were no changes to any of the methodologies, described in the SAR, used in establishing the design bases or in the safety analyses for Cycle 25 Reload. There is no requirement for any Technical Specifications (TS) changes as a result of Cycle 25 Reload. All the analyses that require NRC approved methods were performed using NRC approved methods.

EN-Ll-101-ATT-9.1, Rev. 12

NUCLEAR QUALITY RELATED EN-Ll-101 I REV.12

~Entergy MANAGEMENT MANUAL INFORMATIONAL USE

- PAGE 60F6 10 CFR 50.59 Evaluations ATTACHMENT 9.1

50.59 EVALUATION FORM 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.

EN-Ll-101-ATT-9.1, Rev. 12

ANO 50.59 Evaluation 2016-001

NUCLEAR QUALITY RELATED EN-Ll-101 I REV.12

~Entergy MANAGEMENT MANUAL INFORMATIONAL USE PAGE 1OF8 10 CFR 50.59 Evaluations ATTACHMENT 9.1 50.59 EVALUATION FORM I.

OVERVIEW I SIGNATURES1 Facility: AN0-1 and AN0-2 Evaluation# I Rev.#: 2016-001 I 0 Proposed Change I Document:

EC 57218 - External Flooding Description of Change:

The proposed change (EC 57218) is an engineering evaluation which performs an aggregate review of the external flood protection design basis at ANO. The results of the review are documented in the form of a new engineering report, a new series of flood boundary drawings, and EC mark-ups to a number of existing documents..1, "External Flooding", to the new engineering report, CALC-ANOC-CS-15-00003, "ANO Flood Protection Design Basis", provides a consolidated account of the external flood protection design basis requirements in a single engineering document. Attached to the report is a new flood protection features list, which provides a controlled list of credited flood protection features, such as watertight doors and penetrations located in flood barriers.

The new flood protection boundary drawings provide an easy visual reference which clearly identifies the credited external flood protection boundaries.

The existing documents to be marked up are: penetration drawings, a controlled list of station doors and hatches, the flooding upper level document, the AN0-1 general structural design guide, the operating procedures for natural emergencies, and the AN0-1 and AN0-2 SARs.

EC 57218 found that the external flood protection boundaries would cause certain safety-related components located below the design basis flood level to become submerged during an external flooding event. EC 57218 determined that these components would continue to be capable of performing their function(s) during an external flooding event.

The Process Applicability Determination (PAD) found that the proposed activity is changing the flood protection design and licensing basis such that the flood protection features currently credited in the SARs with protecting safety-related systems and equipment from the effects of an external flooding event will no longer be credited with protecting certain safety-related components located below the design basis flood level from external flood events. This aspect of the proposed activity adversely affects a method of controlling a SAR-described design function (configuring the structures and associated penetrations to protect safety-related structures, systems, and components (SSCs) from the effects of external flooding). In determining that it was acceptable for these components tc:> be submerged during an external flooding event, this change fundamentally alters the existing means of performing or controlling the function of protecting these safety-related components from the effects of external flooding.

The printed name, company, department, and date must be included on the form. Signatures may be obtained via electronic processes (e.g., PCRS, ER processes), manual methods (e.g., ink signature), e-mail, or telecommunication. If using an e-mail or telecommunication, attach it to this form.

EN-Ll-101-ATT-9.1, Rev. 12

NUCLEAR QUALITY RELATED EN-Ll-101 I REV.12

~Entergy MANAGEMENT MANUAL INFORMATIONAL USE PAGE 20F 8 10 CFR 50.59 Evaluations ATTACHMENT 9.1 50.59 EVALUATION FORM Summary of Evaluation:

In response to an impending external flood condition, existing plant procedures require a shutdown of both units. The frequency of occurrence of accidents that could occur is not affected. In addition, there is no increase in the likelihood of a malfunction of any safety-related SSC whose malfunction might contribute to the initiation of an accident. Although non-safety related SSCs may not be capable of performing their function(s), the loss of such functions would not contribute to the initiation of an accident, nor are these components credited for accident mitigation.

An engineering review of cable specifications, manufacturer's test results, the environmental qualification engineering report, and industry tests and evaluations for low-voltage power and control cables concluded that - since the service conditions for the postulated external flood (five days of submergence at temperatures less than 50 °C, outside containment) are much less severe than the loss-of-coolant-accident (LOCA) and other harsh conditions during these tests - the affected safety-related cables are qualified for their service conditions, including submergence during an external flood event. Likewise, the safety-related mechanical components could be submerged during an external flood event with no effect on their ability to perform their functions. Therefore, the change in the design requirements of the identified safety-related components to include submergence during an external flood event will not result in more than a minimal effect on the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the SAR.

Existing plant procedures require a shutdown otboth units in response to an impending external flood condition. No other accidents are assumed to occur during the external flooding event. As long as safe shutdown conditions are achieved and maintained, there are no radiological consequences from an external flooding event. As discussed previously, safety-related equipment associated with this change has been determined to remain unaffected and operable. Since the -non-safety-related equipment affected by the reduction in flood protection features is not required for safe shutdown and safety-related equipment remains operable, the proposed changes do not affect the radiological consequences of an accident or of a malfunction of equipment important to safety.

The inability of non-safety-related SSCs to perform their function(s) does not contribute to the initiation of any accident that might combine with the external flooding event to create the possibility for an accident of a different type. The modified flood protection boundaries will allow equipment formerly maintained dry to become submerged. Although safety-related equipment in such areas will continue to be capable of performing the associated specified safety fi.Jnction(s), including safe shutdown functions, non-safety-related equipment in such areas may be rendered non-functional. However, the loss of function of non-safety-related SSCs does not represent a malfunction with a different result.

Since safe shutdown conditions are achieved and maintained, no fission product barrier or fission product barrier limit is affected.

Methods of evaluation related to the submergence of safety-related equipment are not described in the AN0-1 or AN0-2 SARs. The proposed change being evaluated does not result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses for accident prevention or mitigation.

EN-Ll-101-ATT-9.1, Rev. 12

QUALITY RELATED EN-Ll-101 REV.12

~Entergy NUCLEAR MANAGEMENT MANUAL INFORMATIONAL USE PAGE 30F8 10 CFR 50.59 Evaluations ATTACHMENT 9.1 50.59 EVALUATION FORM 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 require prior NRC approval?

Preparer:

Scott Boeing I see attached email I S&L I Engineering I 01-21-16 Name (print) I Signature I Company I Department I Date D Yes

~ No Reviewer:

David Bice I see attached email I EOI I Regulatory Assurance I 01-21-16 Name (print) I Signature I Cor:npany I Department I Date OSRC:

Ronald A. Barnes I ORIGINAL SIGNED BY RONALD A. BARNES I 01-21-16 Chairman's Name (print) I Signature I Date EMAILS OSRC-2016-002 OSRC Meeting #

From: SCOTT.R.BOEING@sargentlundy.com [1]

Sent: Thursday, January 21, 2016 3:20 PM To: BICE, DAVID B (ANO)

Cc: ROBERT.L.MARSH@sargentlundy.com

Subject:

EC 57218 50.59 Evaluation Attached is my prepared 50.59 Evaluation for EC 57218 with comment incorporated. Please review and concur.

From: BICE, DAVID B (ANO)

Sent: Thursday, January 21, 2016 3:25 PM To: 'SCOTT.R.BOEING@sargentlundy.com' Cc: ROBERT.L.MARSH@sargentlundy.com

Subject:

RE: EC 57218 50.59 Evaluation I have reviewed the subject 5059 and concur. This concurrence may be considered the Reviewer signature on the 5059.

EN-Ll-101-ATT-9.1, Rev. 12

' I NUCLEAR QUALITY RELATED EN-Ll-101 REV.12 MANAGEMENT

~Entergy MANUAL INFORMATIONAL USE PAGE 40F8 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 evaluation ONLY? If "Yes," Questions 1 - 7 are not applicable; answer only 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 previously evaluated in the SAR?

BASIS:

D Yes IZI No D Yes IZI No The exclusion of equipment from existing external flood protection provisions does not affect the frequency of occurrence of an external flood.

A design basis accident is not assumed to occur concurrent with an external flood. Existing plant procedures require a shutdown of both units in response to an impending external flood condition.

EC 57218 demonstrates that certain safety-related cables and other identified safety-related components which may be submerged during an external flooding event will remain capable of performing their safety functions. Because the subject safety related cabling has been shown to not more than minimally affect the potential for equipment failure, the frequency of occurrence of accidents or events that could occur is not appreciably affected. Although associated non-safety-related SSCs may be rendered non-functional, the loss of such functions would not contribute to the initiation of an accident.

Therefore, the exclusion of certain Class/Category 1 equipment from external flood boundary protection 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 a structure, system, or component important to safety previously evaluated in the SAR?

BASIS:

D Yes IZI No Certain safety-related components are located below the design basis flood level in areas that are not within the modified flood-protected boundaries and, therefore, may be submerged during an external flooding event. These are low voltage power and control cables manufactured by Anaconda, Okonite, or Rockbestos. In addition, there are two service water manual valves in Room 2223 and a manual valve and an expansion joint in Room 76.

Two safety-related cables associated with the diesel fuel oil transfer pumps are located in AN0-2 Room 2223 of the Auxiliary Building Extension (ABE). There are no terminal end points or cable splices for these cables in the ABE. One hundred and fifty-three safety-related cables pass through AN0-1 Room 76, including cable RPB5253A 1, which runs from MCC B-52 to terminal box TB691 in Room 76, and cable RPB5253A (which runs from TB691 to the sluice gate operator for motor-operated valve SG-1). However, terminal box TB691 is located above the design-basis flood level in Room 76 and, therefore, is not affected by an external flood.

EN-Ll-101-ATT-9.1, Rev. 12

NUCLEAR QUALITY RELATED EN-Ll-101 I REV.12

-=-Entergy MANAGEMENT MANUAL INFORMATIONAL USE PAGE 50F 8 10 CFR 50.59 Evaluations ATTACHMENT9.1 50.59 EVALUATION FORM Walkdowns have determined that, for the cables of interest, there are no conduit seals, terminal boxes, or other devices which could allow water to contact conductors located below the design-basis flood level.

An engineering review of ANO cable specifications, manufacturer's test results, the environmental qualification engineering report, and industry tests and evaluations was performed. These test and evaluations included manufacturer's long-term water immersion tests, IEEE-383 LOCA submergence tests, an EPRI evaluation of submerged cables in conduit during LOCA testing (EPRI TR 107386), and an evaluation of the results of heat aging followed by submergence testing (NUREG/CR-5655).

Based on the results of these tests and evaluations, and the fact that the service conditions for the

postulated-external flood (five days of submergence at temperatures less than 50 °C, outside containment) are much less severe than the LOCA and other harsh conditions during these tests, the engineering review concluded that the cables are qualified for their service conditions, including submergence during an external flood event.

Therefore, the change in the design requirements of the identified safety-related cables in AN0-1 Room 76 and AN0-2 Room 2223 to include submergence during an external flood event will not more than minimally affect the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the SAR.

The two manual valves in Room 2223 are a normally closed vent valve (2SW-1185) on the 30" service water return line to the emergency pond and a normally closed drain valve (2SW-1220) on that same line used to support cleaning of the line during outages. EC 57218 concluded that these valves could be submerged during an external flood event with no effect on their ability to perform their functions to remain closed.

In Room 76, the manual valve is a normally closed valve (AC-20010) on a unit cooler (VUC-140) supply drain line, and the expansion joint (XJ-42) is on the emergency feedwater turbine exhaust line. EC 57218 concluded that these components could be submerged during an external flood event with no effect on their ability to perform their functions of remaining closed or intact.

AN0-1 SAR Section 5.1.6 and AN0-2 SAR Section 3.4.4 indicate that safety-related equipment is either located above the flood level or is protected. The availability of safety-related equipment during an external flood ensures that safe shutdown of both units can be accomplished. EC 57218 demonstrates that certain safety-related cables and other identified safety-related components which may be submerged during an external flooding event will remain capable of performing their safety functions. As a result, both units remain capable of being safety shut down during an external flood event using safety-related SSCs. Therefore, the proposed change does not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the SAR.

EN-Ll-101-ATT-9.1, Rev. 12

QUALITY RELATED EN-Ll-101 REV.12

"='Entergy NUCLEAR MANAGEMENT MANUAL INFORMATIONAL USE PAGE 6 OF 8 10 CFR 50.59 Evaluations ATTACHMENT9.1 I

3.

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

BASIS:

0 Yes

~ No The modified flood protection boundaries will allow equipment formerly maintained dry to become submerged during an external flooding event. As noted in the response to Question 2, safety-related equipment in such areas will continue to be capable of performing specified safety functions.

Non-safety-related equipment in such areas may be rendered non-functional. However, as noted in the response to Question 1, existing plant procedures require a shutdown of both units in response to an impending external flopd condition 13nd related shutdown accidents/events are not assumed nor impacted by this change. As long as safe shutdown conditions are achieved and maintained, there are no -accident-related radiological consequences from an external flooding event. Since the no-n-safety~related equipment affected by the reduction in flood protection features is not required for safersftutdown and safetY:,-related equipment remains operable, the proposed changes do not affect the radiological consequences of an accident.

Therefore, the proposed activity does not result in more than a minimal increase in the consequences of an accident previously evaluated in the SAR.

4.

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

BASIS:

0 Yes

~ No The modified flood protection boundaries will allow equipment formerly maintained dry to become submerged. As noted in the response to Question 2, safety-related equipment in such areas will continue to be capable of performing specified safety functions, including safe shutdown functions.

Non-safety related equipment in such are-as may be rendered non-functional. However, as noted in the resp~mse -tQ Question 1, existing plant procedures require a shutdown of both units in response to an impending external flood condition. As long as safe shutdown conditions are achieved and maintained, there are no radiological consequences from an external flooding event. Since the non-safety-related equipment affected by the reduction in flood protection features is not used for safe shutdown and safety-related equipment remains operable, the proposed changes do not affect the radiological consequences of a malfunction of equipment important to safety.

Therefore, the proposed activity does 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.

EN-Ll-101-ATT-9.1, Rev. 12

NUCLEAR QUALITY RELATED EN-Ll-101 I REV.12

~Entergy MANAGEMENT MANUAL INFORMATIONAL USE PAGE 70F 8 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 the SAR?

D Yes

[8]

No BASIS:

The modified flood protection boundaries will allow equipment formerly maintained dry to become submerged. As noted in the response to Question 2, safety-related equipment in such areas will continue to be capable of performing specified safety functions, including safe shutdown functions.

Non-safety related equipment in such areas may be rendered non-functional. However, as noted in the response to. Question 1, existing plant procedures require a shutdown of both units in response to an impending external flood condition. The inability of non-safety-related SSCs to perform normal function(s) does not contribute to the initiation of any accident that might combine with the external flooding event to create the possibility for an accident of a different type.

Therefore, the proposed 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 a structure, system, or component important to safety with a different result than any previously evaluated in the SAR?

D Yes

[8]

No

7.

BASIS:

The modified flood protection boundaries will allow equipment formerly maintained dry to become submerged. As noted in the response to Question 2, safety-related equipment in such areas will continue to be capable of performing Specified safety functions, including safe shutdown functions.

Non-safety related equipment in such areas may be rendered non-functional. However, the loss of function of non-safety-related SSCs does not represent a malfunction with a different result.

Therefore, the proposed change does not create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the SAR.

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

BASIS:

D Yes

[8]

No The modified flood protection boundaries will allow equipmel)t formerly maintained dry to become submerged. As noted in the response to Question 2, safety-related equipment in such areas will continue to. be capable of performing specified safety functions, including safe shutdown functions.

As noted in the response to Question 1, existing plant procedures require a shutdown of both units in response to an impending external flood condition. As long as safe shutdown conditions are achieved and maintained, no fission product barrier is affected. In addition, failure of any of the affected (submerged) SSCs has no impact on fuel cladding, the Reactor Coolant System (RCS) pressure boundary, or containment design pressure limits.

Therefore, the proposed changes do not result in a design basis limit for a fission product barrier as described in the SAR being exceeded or altered.

EN-Ll-101-ATT-9.1, Rev. 12

~Entergy ATTACHMENT 9.1 NUCLEAR MANAGEMENT MANUAL QUALITY RELATED INFORMATIONAL USE 10 CFR 50.59 Evaluations EN-Ll-101 I REV.12 PAGE 8 OF 8 50.59 EVALUATION FORM

8.

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

D Yes

~ No BASIS:

Methods of evaluation related to the submergence of safety-related equipment are not described in the AN0-1 or AN0-2 SARs and are not part of the accident analyses methodologies. Therefore, the proposed change being evaluated does not result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses.

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

EN-Ll-101-ATT-9.1, Rev. 12 to 2CAN041601 AN0-2 SAR Amendment 26 (compact disc) to 2CAN041601 AN0-2TRM (compact disc) to 2CAN041601 AN0-2 TS Table of Contents and TS Bases (compact disc)