ML21313A008

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Issuance of Amendments to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections
ML21313A008
Person / Time
Site: Arkansas Nuclear, Waterford  Entergy icon.png
Issue date: 12/08/2021
From: Siva Lingam
Plant Licensing Branch IV
To: Halter M
Entergy Services
Lingam S, NRR/DORL/LPL4, 415-1564
References
EPID L-2021-LLA-0122
Download: ML21313A008 (37)
v • d • e


Text

December 8, 2021

Mrs. Mandy Halter Vice President, Regulatory Assurance Entergy Services, LLC M-ECH-29 1340 Echelon Parkway Jackson, MS 39213

SUBJECT:

ARKANSAS NUCLEAR ONE, UNITS 1 AND 2 AND WATERFORD STEAM ELECTRIC STATION, UNIT 3 - ISSUANCE OF AMENDMENTS REGARDING REVISION OF TECHNICAL SPECIFICATIONS TO ADOPT TSTF-577, REVISED FREQUENCIES FOR STEAM GENERATOR TUBE INSPECTIONS (EPID L-2021-LLA-0122)

Dear Mrs. Halter:

The U.S. Nuclear Regulatory Commission ( NRC, the Commission) has issued amendments consisting of changes to the Technical Specifications (TSs) in response to your application dated July 1, 2021, for Arkansas Nuclear One, Units 1 and 2 (ANO-1 and ANO-2, respectively) and Waterford Steam Electric Station, Unit 3 (Waterford 3). The following amendments are enclosed:

Amendment No. 273 to Renewed Facility Operating License No. DPR-51 for ANO-1 Amendment No. 326 to Renewed Facility Operating License No. NPF-6 for ANO-2 Amendment No. 262 to Renewed Facility Operating License No. NPF-38 for Waterford 3

The amendments revise the TSs related to steam generator tube inspections and reporting based on operating history. The proposed changes are based on Technical Specifications Task Force (TSTF) Traveler TSTF-577, Revision 1, Revised Frequencies for Steam Generator Tube Inspections, dated March 1, 2021. The NRC issued a final safety evaluation approving TSTF-577 on April 14, 2021.

M. Halter

A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions monthly Federal Register notice.

Sincerely,

/RA/

Siva P. Lingam, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Docket Nos. 50-313, 50-368, and 50-382

Enclosures:

1. Amendment No. 273 to DPR-51
2. Amendment No. 326 to NPF-6
3. Amendment No. 262 to NPF-38
4. Safety Evaluation
5. Notice and Environmental Finding

cc: Listserv

ENTERGY OPERATIONS, INC.

DOCKET NO. 50-313

ARKANSAS NUCLEAR ONE, UNIT 1

AMENDMENT TO RENEWED FACI LITY OPERATING LICENSE

Amendment No. 273 Renewed License No. DPR-51

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Entergy Operations, Inc. (EOI), dated July 1, 2021, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I;

B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission;

C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations;

D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and

E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

Enclosure 1

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.c.(2) of Renewed Facility Operating License No. DPR-51 is hereby amended to read as follows:

(2) Technical Specifications

The Technical Specifications contained in Appendix A, as revised through Amendment No. 273, are hereby incorporated in the renewed license. EOI shall operate the facility in accordance with the Technical Specifications.

3. This amendment is effective as of its dat e of issuance and shall be implemented within 30 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

Jennifer L. Dixon-Herrity, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. DPR-51 and the Technical Specifications

Date of Issuance: December 8, 2021

ATTACHMENT TO LICENSE AMENDMENT NO. 273

RENEWED FACILITY OPERATING LICENSE NO. DPR-51

ARKANSAS NUCLEAR ONE, UNIT 1

DOCKET NO. 50-313

Replace the following pages of the Renewed Facility Operating License No. DPR-51 and Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Renewed Facility Operating License

REMOVE INSERT 3 3

Technical Specifications

REMOVE INSERT 5.0-12 5.0-12 5.0-13 5.0-13 5.0-14 5.0-14 5.0-23 5.0-23

--- 5.0-23a

(5) EOI, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components;

(6) EOI, pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

c. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level

EOI is authorized to operate the facility at steady state reactor core power levels not in excess of 2568 megawatts thermal.

(2) Technical Specifications

The Technical Specifications contained in Appendix A, as revised through Amendment No. 273, are hereby incorporated in the renewed license.

EOI shall operate the facility in accordance with the Technical Specifications.

(3) Safety Analysis Report

The licensees SAR supplement submitted pursuant to 10 CFR 54.21(d),

as revised on March 14, 2001, describes certain future inspection activities to be completed before the period of extended operation. The licensee shall complete these activities no later than May 20, 2014.

(4) Physical Protection

EOI shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans, including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains Safeguards Information protected under 10 CFR 73.21, is entitled: Arkansas Nuclear One Physical Security Plan, Training and Qualifications Plan, and Safeguards Contingency Plan, as submitted on May 4, 2006.

Renewed License No. DPR-51 Amendment No. 273 Revised by letter dated July 18, 2007 Programs and Manuals 5.5

5.0 ADMINISTRATIVE CONTROLS

5.5 Programs and Manuals

5.5.9 Steam Generator (SG) Program

An SG Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the SG Program shall include the following:

a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the as found condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The as found condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes.

Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.

b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.
1. Structural integrity performance criterion: All in-service SG tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down), all anticipated transients included in the design specification, and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary to secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary to secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.
2. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed 1 gpm.
3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, RCS Operational LEAKAGE.

ANO-1 5.0-12 Amendment No. 215,224,239,250,258, 273 Programs and Manuals 5.5

5.0 ADMINISTRATIVE CONTROLS

5.5 Programs and Manuals

c. Provisions for SG tube plugging criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.
d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube plugging criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. A degradation assessment shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.
1. Inspect 100% of the tubes in each SG during the first refueling outage following SG installation.
2. After the first refueling outage following SG installation, inspect 100%

of the tubes in each SG at least every 96 effective full power months, which defines the inspection period.

3. If crack indications are found in any SG tube, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the crack indication shall be at the next refueling outage. If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.
e. Provisions for monitoring operational primary to secondary LEAKAGE.

ANO-1 5.0-13 Amendment No. 215,224,239,250,258, 273 Programs and Manuals 5.5

5.0 ADMINISTRATIVE CONTROLS

5.5 Programs and Manuals

5.5.10 Secondary Water Chemistry

This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation. The program shall include:

a. Identification of a sampling schedule for the critical variables and control points for these variables;
b. Identification of the procedures used to measure the values of the critical variables;
c. Identification of process sampling points;
d. Procedures for the recording and management of data;
e. Procedures defining corrective actions for all off control point chemistry conditions; and
f. A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events required to initiate corrective action.

ANO-1 5.0-14 Amendment No. 215,224,239,250,258, 273 Reporting Requirements 5.6

5.0 ADMINISTRATIVE CONTROLS

5.6 Reporting Requirements

5.6.6 Reactor Building Inspection Report

Any degradation exceeding the acceptance criteria of the containment structure detected during the tests required by the Containment Inspection Program shall undergo an engineering evaluation within 60 days of the completion of the inspection surveillance. The results of the engineering evaluation shall be reported to the NRC within an additional 30 days of the time the evaluation is completed. The report shall include the cause of the condition that does not meet the acceptance criteria, the applicability of the conditions to the other unit, the acceptability of the concrete containment without repair of the item, whether or not repair or replacement is required and, if required, the extent, method, and completion date of necessary repairs, and the extent, nature, and frequency of additional examinations.

5.6.7 Steam Generator Tube Inspection Report

A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.9, "Steam Generator (SG) Program." The report shall include:

a. The scope of inspections performed on each SG;
b. The nondestructive examination techniques utilized for tubes with increased degradation susceptibility;
c. For each degradation mechanism found:
1. The nondestructive examination techniques utilized;
2. The location, orientation (if linear), measured size (if available), and voltage response for each indication. For tube wear at support structures less than 20 percent through-wall, only the total number of indications needs to be reported;
3. A description of the condition monitoring assessment and results, including the margin to the tube integrity performance criteria and comparison with the margin predicted to exist at the inspection by the previous forward-looking tube integrity assessment; and
4. The number of tubes plugged during the inspection outage.
d. An analysis summary of the tube integrity conditions predicted to exist at the next scheduled inspection (the forward-looking tube integrity assessment) relative to the applicable performance criteria, including the analysis methodology, inputs, and results;

ANO-1 5.0-23 Amendment No. 215,224,239,250,258, 273 Reporting Requirements 5.6

5.0 ADMINISTRATIVE CONTROLS

5.6 Reporting Requirements

e. The number and percentage of tubes plugged to date, and the effective plugging percentage in each SG; and
f. The results of any SG secondary side inspections.

ANO-1 5.0-23a Amendment No. 273 ENTERGY OPERATIONS, INC.

DOCKET NO. 50-368

ARKANSAS NUCLEAR ONE, UNIT 2

AMENDMENT TO RENEWED FACI LITY OPERATING LICENSE

Amendment No. 326 Renewed License No. NPF-6

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Entergy Operations, Inc. (the licensee), dated July 1, 2021, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I;

B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission;

C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations;

D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and

E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

Enclosure 2

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-6 is hereby amended to read as follows:

(2) Technical Specifications

The Technical Specifications contained in Appendix A, as revised through Amendment No. 326, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications

3. This amendment is effective as of its dat e of issuance and shall be implemented within 30 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

Jennifer L. Dixon-Herrity, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. NPF-6 and the Technical Specifications

Date of Issuance: December 8, 2021

ATTACHMENT TO LICENSE AMENDMENT NO. 326

RENEWED FACILITY OPERATING LICENSE NO. NPF-6

ARKANSAS NUCLEAR ONE, UNIT 2

DOCKET NO. 50-368

Replace the following pages of the Renewed Facility Operating License No. NPF-6 and Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Renewed Facility Operating License

REMOVE INSERT Technical Specifications

REMOVE INSERT 6-8 6-8 6-9 6-9 6-14 6-14 6-22 6-22

--- 6-22a

3

(4) EOI, pursuant to the Act and 10 CFR Parts 30, 40 and 70 to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitori ng equipment calibration, and as fission detectors in amounts as required;

(5) EOI, pursuant to the Act and 10 CFR Parts 30, 40 and 70 to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and

(6) EOI, pursuant to the Act and 10 CFR Parts 30 and 70 to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This renewed license shall be deemed to contain and is subject to conditions specified in the following Commission regulations in 10 CFR Chapter I; Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level

EOI is authorized to operate the facility at steady state reactor core power levels not in excess of 3026 megawatts thermal. Prior to attaining this power level EOI shall comply with the conditions in Paragraph 2.C.(3).

(2) Technical Specifications

The Technical Specifications contained in Appendix A, as revised through Amendment No. 326, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

Exemptive 2nd paragraph of 2.C.2 deleted per Amendment 20, 3/3/81.

(3) Additional Conditions

The matters specified in the following conditions shall be completed to the satisfaction of the Commission within the stated time periods following issuance of the renewed license or within the operational restrictions indicated.

The removal of these conditions shall be made by an amendment to the renewed license supported by a favorable evaluation by the Commission.

2.C.(3)(a) Deleted per Amendment 24, 6/19/81.

Renewed License No. NPF-6 Amendment No. 326 ADMINISTRATIVE CONTROLS

6.5.9 Steam Generator (SG) Program

An SG Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the SG Program shall include the following:

a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the as found condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The as found condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.
b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational leakage.
1. Structural integrity performance criterion: All in-service SG tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down), all anticipated transients included in the design specification, and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary to secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary to secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.
2. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed 1 gpm through any one SG.
3. The operational leakage performance criterion is specified in LCO 3.4.6.2, "Reactor Coolant System Operational Leakage."
c. Provisions for SG tube plugging criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

ARKANSAS - UNIT 2 6-8Amendment No. 255,266,307, 326 ADMINISTRATIVE CONTROLS

6.5.9 Steam Generator (SG) Program (continued)

d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube plugging criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. A degradation assessment shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.
1. Inspect 100% of the tubes in each SG during the first refueling outage following SG installation.
2. After the first refueling outage following SG installation, inspect 100% of the tubes in each SG at least every 96 effective full power months, which defines the inspection period.
3. If crack indications are found in any SG tube, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the crack indication shall be at the next refueling outage. If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.
e. Provisions for monitoring operational primary to secondary leakage.

ARKANSAS - UNIT 2 6-9 Amendment No. 255,266,307, 326 Next Page is 6-14 ADMINISTRATIVE CONTROLS

6.5.10 Secondary Water Chemistry

This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation. The program shall include:

a. Identification of a sampling schedule for the critical variables and control points for these variables;
b. Identification of the procedures used to measure the values of the critical variables;
c. Identification of process sampling points;
d. Procedure for the recording and management of data;
e. Procedures defining corrective actions for all off control point chemistry conditions; and
f. A procedure identifying the authority responsible for the interpretation of the data, and the sequence and timing of administrative events required to initiate corrective action.

ARKANSAS - UNIT 2 6-14Amendment No. 255,307,326 ADMINISTRATIVE CONTROLS

6.6.6 Containment Inspection Report

Any degradation exceeding the acceptance criteria of the containment structure detected during the tests required by the Containment Tendon Surveillance Program shall undergo an engineering evaluation within 60 days of the completion of the inspection surveillance. The results of the engineeri ng evaluation shall be reported to the NRC within an additional 30 days of the time the evaluation is completed. The report shall include the cause of the condition that does not meet the acceptance criteria, the applicability of the conditions to the other unit, the acceptability of the concrete containment without repair of the item, whether or not repair or replacement is required and, if required, the extent, method, and completion date of necessary repairs, and the extent, nature, and frequency of additional examinations.

6.6.7 Steam Generator Tube Inspection Report

A report shall be submitted within 180 days after the initial entry into HOT SHUTDOWN following completion of an inspection performed in accordance with the Specification 6.5.9, "Steam Generator (SG) Program." The report shall include:

a. The scope of inspections performed on each SG;
b. The nondestructive examination techniques utilized for tubes with increased degradation susceptibility;
c. For each degradation mechanism found:
1. The nondestructive examination techniques utilized;
2. The location, orientation (if linear), measured size (if available), and voltage response for each indication. For tube wear at support structures less than 20 percent through-wall, only the total number of indications needs to be reported;
3. A description of the condition monitoring assessment and results, including the margin to the tube integrity performance criteria and comparison with the margin predicted to exist at the inspection by the previous forward-looking tube integrity assessment; and
4. The number of tubes plugged during the inspection outage.
d. An analysis summary of the tube integrity conditions predicted to exist at the next scheduled inspection (the forward-looking tube integrity assessment) relative to the applicable performance criteria, including the analysis methodology, inputs, and results;
e. The number and percentage of tubes plugged to date, and the effective plugging percentage in each SG; and
f. The results of any SG secondary side inspections.

ARKANSAS - UNIT 2 6-22 Amendment No. 255,257,262,266, 307, 326 ADMINISTRATIVE CONTROLS

6.6.8 Specific Activity

The results of specific activity analysis in which the primary coolant exceeded the limits of Specification 3.4.8. The following information shall be included: (1) Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (2)

Results of the last isotopic analysis for radioiodine performed prior to exceeding the limit, results of analysis while limit was exceeded the results of one analysis after the radioiodine activity was reduced to less than limit. Each result should include date and time of sampling and the radioiodine concentrations; (3) Clean-up system flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (4) Graph of the I-131 concentration and one other radioiodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radioiodine limit.

ARKANSAS - UNIT 2 6-22a Amendment No. 326 ENTERGY OPERATIONS, INC.

DOCKET NO. 50-382

WATERFORD STEAM ELECTRIC STATION, UNIT 3

AMENDMENT TO RENEWED FACI LITY OPERATING LICENSE

Amendment No. 262 Renewed License No. NPF-38

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Entergy Operations, Inc. (EOI), dated July 1, 2021, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I;

B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission;

C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations;

D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and

E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

Enclosure 3

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.2 of Renewed Facility Operating License No. NPF-38 is hereby amended to read as follows:
2. Technical Specifications and Environmental Protection Plan

The Technical Specifications contained in Appendix A, as revised through Amendment No. 262, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. EOI shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of its date of issuance and shall be implemented within 30 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

Jennifer L. Dixon-Herrity, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. NPF-38 and the Technical Specifications

Date of Issuance: December 8, 2021 ATTACHMENT TO LICENSE AMENDMENT NO. 262

TO RENEWED FACILITY OPERATING LICENSE NO. NPF-38

WATERFORD STEAM ELECTRIC STATION, UNIT 3

DOCKET NO. 50-382

Replace the following pages of the Renewed Facility Operating License No. NPF-38 and Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Renewed Facility Operating License

REMOVE INSERT Technical Specifications

REMOVE INSERT 6-7a 6-7a 6-7b 6-7b 6-7c ---

6-7d ---

6-17a 6-17a

--- 6-17b

the NRC of any action by equity investors or successors in interest to Entergy Louisiana, LLC that may have an effect on the operation of the facility.

C. This renewed license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

1. Maximum Power Level

EOI is authorized to operate the facility at reactor core power levels not in excess of 3716 megawatts thermal (100% power) in accordance with the conditions specified herein.

2. Technical Specifications and Environmental Protection Plan

The Technical Specifications contained in Appendix A, as revised through Amendment No. 262, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. EOI shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. Antitrust Conditions

(a) Entergy Louisiana, LLC shall comply with the antitrust license conditions in Appendix C to this renewed license.

(b) Entergy Louisiana, LLC is responsible and accountable for the actions of its agents to the extent said agent's actions contravene the antitrust license conditions in Appendix C to this renewed license.

AMENDMENT NO. 262 ADMINISTRATIVE CONTROLS

6.5.8 DELETED

6.5.9 STEAM GENERATOR (SG) PROGRAM

An SG Program shall be established and implemented to ensure that SG tube integrity is maintained.

In addition, the SG Program shall include the following:

a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.
b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational leakage.
1. Structural integrity performance criterion: All in-service SG tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down), all anticipated transients included in the design specification, and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary to secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary to secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.
2. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Primary to secondary leakage is not to exceed 540 gpd through any one SG.
3. The operational leakage performance criterion is specified in LCO 3.4.5.2, "Reactor Coolant System Operational Leakage."
c. Provisions for SG tube plugging criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

WATERFORD - UNIT 3 6-7a AMENDMENT NO. 189,204,236, 250, 262 ADMINISTRATIVE CONTROLS

STEAM GENERATOR (SG) PROGRAM (Continued)

d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed.

The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet and that may satisfy the applicable tube plugging criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. A degradation assessment shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

1. Inspect 100% of the tubes in each SG during the first refueling outage following SG installation.
2. After the first refueling outage following SG installation, inspect 100% of the tubes in each SG at least every 96 effective full power months, which defines the inspection period.
3. If crack indications are found in any SG tube, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the crack indication shall be at the next refueling outage. If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.
e. Provisions for monitoring operational primary to secondary leakage.

WATERFORD - UNIT 3 6-7b AMENDMENT NO. 204,207, 236, 262 ADMINISTRATIVE CONTROLS

ANNUAL REPORTS (Continued)

(1) Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded;

(2) Results of the last isotopic analysis for radioiodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radioiodine activity was reduced to less than limit. Each result should include date and time of sampling and the radioiodine concentrations;

(3) Clean-up system flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded;

(4) Graph of the I-131 concentration and one other radioiodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above steady-state level; and

(5) The time duration when the specific activity of the primary coolant exceeded the radioiodine limit.

6.9.1.5 STEAM GENERATOR TUBE INSPECTION REPORT

A report shall be submitted within 180 days after the initial entry into HOT SHUTDOWN following completion of an inspection performed in accordance wi th the Specification 6.5.9, "Steam Generator (SG) Program." The report shall include:

a. The scope of inspections performed on each SG;
b. The nondestructive examination techniques utilized for tubes with increased degradation susceptibility;
c. For each degradation mechanism found:
1. The nondestructive examination techniques utilized;
2. The location, orientation (if linear), measured size (if available), and voltage response for each indication. For tube wear at support structures less than 20 percent through-wall, only the total number of indications needs to be reported;
3. A description of the condition monitoring assessment and results, including the margin to the tube integrity performance criteria and comparison with the margin predicted to exist at the inspection by the previous forward-looking tube integrity assessment; and
4. The number of tubes plugged during the inspection outage.

WATERFORD - UNIT 3 6-17a AMENDMENT NO. 8,116,188,202,204, 207, 236, 262 ADMINISTRATIVE CONTROLS

d. An analysis summary of the tube integrity conditions predicted to exist at the next scheduled inspection (the forward-looking tube integrity assessment) relative to the applicable performance criteria, including the analysis methodology, inputs, and results;
e. The number and percentage of tubes plugged to date, and the effective plugging percentage in each SG; and
f. The results of any SG secondary side inspections.

WATERFORD - UNIT 3 6-17b AMENDMENT NO. 262 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 273 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-51 AMENDMENT NO. 326 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-6 AMENDMENT NO. 262 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-6 ENTERGY OPERATIONS, INC.

ARKANSAS NUCLEAR ONE, UNIT 1 ARKANSAS NUCLEAR ONE, UNIT 2 WATERFORD STEAM ELECTRIC STATION, UNIT 3 DOCKET NOS. 50-313, 50-368, AND 50-382

Application Safety Evaluation Date July 1, 2021, ADAMS Accession December 8, 2021 No. ML21182A158 Principal Contributors to Safety Evaluation Matthew Hamm

1.0 PROPOSED CHANGE

S

Entergy Operations, Inc. (the licensee) requested changes to the technical specifications (TSs) for Arkansas Nuclear One, Units 1 and 2 (ANO-1 and ANO-2) and Waterford Steam Electric Station, Unit 3 (Waterford 3) by license amendment request (application). In its application, the licensee requested that the U.S. Nuclear Regulatory Commission (NRC, the Commission) process the proposed amendments under the Consolidated Line Item Improvement Process (CLIIP). The proposed changes would revise the three units Steam Generator (SG) Program and the Steam Generator Tube Inspection Repo rt TSs based on Technical Specifications Task Force (TSTF) Traveler TSTF-577, Revision 1, Revised Frequencies for Steam Generator Tube Inspections, dated March 1, 2021 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML21060B434), and the associ ated NRC staff safety evaluation (SE) of TSTF-577, dated April 14, 2021 (ADAMS Accession No. ML21098A188).

The tubes within an SG function as an integral part of the reactor coolant pressure boundary and, in addition, isolate fission products in the primary coolant from the secondary coolant and the environment. SG tube integrity means that the tubes are capable of performing this safety function in accordance with the plant design and licensing basis.

All three units SGs have Alloy 690 thermally treated (Alloy 690TT) tubes.

Enclosure 4

1.1 Proposed TS Changes to Adopt TSTF-577

In accordance with NRC staff-approved TSTF-577, the licensee proposed changes that would revise ANO-1 TS 5.5.9, ANO-2 TS 6.5.9, and Waterford 3 TS 6.5.9, all named Steam Generator (SG) Program, and ANO-1 TS 5.6.7, ANO-2 TS 6.6.7, and Waterford 3 TS 6.9.1.5, all named Steam Generator Tube Inspection Report. Specifically, the licensee proposed the following changes to adopt TSTF-577:

ANO-1 TS 5.5.9, ANO-2 TS 6.5.9, and Waterford 3 TS 6.5.9:

The term Steam Generator would be replaced with SG in the first paragraph of ANO-1 TS 5.5.9, ANO-2 TS 6.5.9, and Waterford 3 TS 6.5.9 and the first sentence of ANO-1 TS 5.5.9.b.1, ANO-2 TS 6.5.9.b.1, and Waterford 3 6.5.9.b.1. The first paragraph of ANO-1 TS 5.5.9, ANO-2 TS 6.5.9, and Waterford 3 TS 6.5.9 would begin with An instead of A.

ANO-1 TS 5.5.9.d.2, ANO-2 TS 6.5.9.d.2, and Waterford 3 TS 6.5.9.d.2 would be revised by deleting the requirement to base inspection frequency on the more restrictive metric between either the effective full power months (EFPM) or refueling outage and to use just the EFPM metric.

ANO-1 TS 5.5.9.d.2, ANO-2 TS 6.5.9.d.2, and Waterford 3 TS 6.5.9.d.2 would be revised by deleting the allowance to extend the inspection period by 3 months and by deleting the discussion of prorating inspections.

ANO-1 TS 5.5.9.d.2, ANO-2 TS 6.5.9.d.2, and Waterford 3 TS 6.5.9.d.2 would be revised by deleting the requirement to inspect 100 percent of the tubes during each period in paragraphs d.2.a, d.2.b, d.2.c, and d.2.d (144, 120, 96, and 72 EFPM, respectively) and by adding the requirement to inspect 100 percent of the tubes every 96 EFPM.

The first sentence in ANO-1 TS 5.5.9.d.3, ANO-2 TS 6.5.9.d.3, and Waterford 3 TS 6.5.9.d.3 would be revised by changing shall not exceed 24 effective full power months or one refueling outage (whichever results in more frequent inspections) to shall be at the next refueling outage.

ANO-1 TS 5.6.7, ANO-2 TS 6.6.7, and Waterford 3 TS 6.9.1.5:

Existing reporting requirement b. would be renumbered as c. and be revised by editorial and punctuation changes.

New reporting requirement b. would be added to require the nondestructive examination techniques utilized for tubes with increased degradation susceptibility be reported.

Existing reporting requirement c. would be renumbered as c.1. and be revised by editorial and punctuation changes.

Existing reporting requirement d. would be renumbered as c.2. and be revised to note that the location, orientation (if linear), measured size (if available), and voltage response do not need to be reported for tube wear indications at support structures that are less than 20 percent through-wall. However, the total number of tube wear indications at support structures that are less than 20 percent through-wall would be reported.

New reporting requirement d. would be added to require an analysis summary of the tube integrity conditions predicted to exist at the next scheduled inspection relative to the applicable performance criteria, including the analysis, methodology, inputs, and results.

Existing reporting requirement e. would be renumbered as c.4. and be revised by editorial and punctuation changes.

Existing reporting requirement f. would be renumbered as e. and be revised by editorial and punctuation changes.

New reporting requirement f. would be added to require the results of any SG secondary side inspections be reported.

Existing reporting requirement g. would be renumbered as c.3. and be revised to add the requirements to report the margin to the tube integrity performance criteria and a comparison with the margin predicted to exist at the inspection by the previous forward-looking tube integrity assessment. In addition, the requirement to report the results of tube pulls and in situ testing would be deleted.

1.2 Additional Proposed TS Changes

1.2.1 ANO-1

In addition to the changes proposed consistent with the traveler discussed in Section 1.1, the licensee proposed the following variations:

The current ANO-1 TS 5.6.7 is titled, "Steam Generator Tube Inspection Reports" (plural). The licensee proposed renaming TS 5.6.7 to "Steam Generator Tube Inspection Report" (singular) for consistency with TSTF-577 and the Standard Technical Specifications (STSs) in NUREG-1430.1

The licensee also noted that the first paragraph in the current ANO-1 TS 5.6.7 references the title of TS 5.5.9, "Steam Generator (SG) Program," without placing the title in quotes. The licensee proposed placing the title of TS 5.6.7 in the first paragraph of TS 5.5.9 in quotes for consistency with TSTF 577 and the STSs.

1 U.S. Nuclear Regulatory Commission, Standard Technical Specifications, Babcock and Wilcox Plants, NUREG-1430, Volume 1, Specifications, and Volume 2, Bases, Revision 4, April 2012 (ADAMS Accession Nos. ML12100A177 and ML12100A178, respectively).

1.2.2 ANO-2

In addition to the changes proposed consistent with the traveler discussed in Section 1.1, the licensee proposed the following variations:

The licensee noted that ANO-2 TSs have different numbering than STSs in NUREG-1432.2

The licensee proposed placing the title of TS 3.4.6.2 as it appears in the current ANO-2 TS 6.5.9 paragraph b.3 in quotes and removing the existing italic format of the text.

The licensee proposed placing the title of TS 6.5.9 as it appears in the current ANO-2 TS 6.6.7 introductory paragraph in quotes and removing the existing italic format of the text.

1.2.3 Waterford 3

In addition to the changes proposed consistent with the traveler discussed in Section 1.1, the licensee proposed the following variations.

The licensee noted that Waterford 3 TSs have different numbering than STSs in NUREG-1432.

The licensee proposed placing the title of TS 6.5.9 as it appears in the current Waterford 3 TS 6.9.1.5 introductory paragraph in quotes.

2.0 REGULATORY EVALUATION

The regulations in Title 10 of the Code of Federal Regulations (10 CFR) paragraph 50.36(c)(5),

Administrative controls, state that [a]dministrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner. Each licensee shall submit any reports to the Commission pursuant to approved technical specifications as specified in

[10 CFR] 50.4. TS Section 5.0, Administrative Controls, requires that an SG program be established and implemented to ensure that SG tube integrity is maintained. Programs established by the licensee, including the SG program, are listed in the administrative controls section of the TS to operate the facility in a safe manner.

The NRC staffs guidance for the review of TSs is in NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor]

Edition (SRP), Chapter 16.0, Technical Specifications, Revision 3, dated March 2010 (ADAMS Accession No. ML100351425). As described therein, as part of the regulatory standardization effort, the NRC staff has prepared STSs for each of the LWR nuclear designs.

Accordingly, the NRC staffs review includes consideration of whether the proposed changes are consistent with NUREG-1430 for ANO-1 and NUREG-1432 for ANO-2 and Waterford 3 as modified by NRC-approved travelers.

2 U.S. Nuclear Regulatory Commission, Standard Technical Specifications, Combustion Engineering Plants, NUREG-1432, Volume 1, Specifications, and Volume 2, Bases, Revision 4, April 2012 (ADAMS Accession Nos. ML12102A165 and ML12102A169, respectively).

TSTF-577 revised the STSs related to SG tube inspections and SG tube inspection reporting requirements. The NRC approved TSTF-577, under the CLIIP on April 14, 2021 (ADAMS Package Accession No. ML21099A086).

3.0 TECHNICAL EVALUATION

3.1 Proposed TS Changes to Adopt TSTF-577

The NRC staff compared the licensees proposed TS changes in Section 1.1 of this SE against the changes approved in TSTF-577. In accordance with SRP Chapter 16.0, the NRC staff determined that the STS changes approved in TSTF-577 are applicable because ANO-1, ANO-2, and Waterford 3 are pressurized-water reactor (PWR) design plants and the NRC staff approved the TSTF-577 changes for PWR designs. The NRC staff finds that the licensees proposed changes to the ANO-1, ANO-2, and Waterford 3 TSs in Section 1.1 of this SE are consistent with those found acceptable in TSTF-577.

In the SE of TSTF-577, the NRC staff concluded that the TSTF-577 changes to STS 5.5.9, Steam Generator (SG) Program, and STS 5.6.7, Steam Generator Tube Inspection Report, were acceptable because, as discussed in Section 3.0 of that SE, they continued to ensure SG tube integrity and, therefore, protected the public health and safety. In particular, the structural integrity performance criterion and accident-induced leakage performance criterion (explained in STS 5.5.9.b, items 1 and 2, respectively) will continue to be met with the proposed revised SG inspection intervals (maximum allowable time between SG inspections) and inspection periods (maximum allowable time between 100 percent of SG tubes inspections). Additionally, the proposed changes to the reporting requirements will provide more detailed and consistent information to the NRC. Therefore, the NRC staff found that the proposed changes to the SG program and inspection reporting requirements were acceptable because they continued to meet the requirements of 10 CFR 50.36(c)(5) by providing administrative controls necessary to assure operation of the facility in a safe manner. For these same reasons, the NRC staff concludes that the corresponding proposed changes to the ANO-1, ANO-2, and Waterford 3 TSs in Section 1.1 of this SE continue to meet the requirements of 10 CFR 50.36(c)(5).

3.2 Additional Proposed TS Changes

The NRC staff reviewed the proposed changes described in Section 1.2 of this SE and determined the changes are editorial because they do not substantively alter TS requirements and the changes will make the ANO-1, ANO-2, and Waterford 3 TSs more consistent with STSs, as modified by the TSTF-577 changes. Therefore, the changes are acceptable.

3.3 TS Change Consistency

The NRC staff reviewed the proposed TS changes for technical clarity and consistency with the existing requirements for customary terminology and formatting. The NRC staff finds that the proposed changes are consistent with Chapter 16.0 of the SRP and are therefore acceptable.

4.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

NOTICE AND ENVIRONMENTAL FINDINGS RELATED TO AMENDMENT NO. 273 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-51 AMENDMENT NO. 326 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-6 AMENDMENT NO. 262 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-6 ENTERGY OPERATIONS, INC.

ARKANSAS NUCLEAR ONE, UNIT 1 ARKANSAS NUCLEAR ONE, UNIT 2 WATERFORD STEAM ELECTRIC STATION, UNIT 3 DOCKET NOS. 50-313, 50-368, AND 50-382

Application Safety Evaluation Date July 1, 2021, ADAMS Accession December 8, 2021 No. ML21182A158

1.0 INTRODUCTION

Entergy Operations, Inc. (the licensee) requested changes to the technical specifications (TSs) for Arkansas Nuclear One, Units 1 and 2 (ANO-1 and ANO-2) and Waterford Steam Electric Station, Unit 3 (Waterford 3) by license amendment request (application). In its application, the licensee requested that the U.S. Nuclear Regulatory Commission (NRC, the Commission) process the proposed amendments under the Consolidated Line Item Improvement Process (CLIIP). The proposed changes would revise the Steam Generator (SG) Program and the Steam Generator Tube Inspection Report TSs based on Technical Specifications Task Force (TSTF) Traveler TSTF-577, Revision 1, Revised Frequencies for Steam Generator Tube Inspections, dated March 1, 2021 (Agencywi de Documents Access and Management System (ADAMS) Accession No. ML21060B434), and the asso ciated NRC staff safety evaluation of TSTF-577, dated April 14, 2021 (ADAMS Accession No. ML21098A188).

2.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Arkansas State and the Louisiana State officials were notified of the proposed issuance of the amendments on November 8, 2021. The State officials had no comments.

3.0 ENVIRONMENTAL CONSIDERATION

The amendments relate, in part, to changes in recordkeeping, reporting, or administrative procedures or requirements. The amendments also relate, in part, to changing requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding

Enclosure 5

that the amendments involve no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on August 10, 2021 (86 FR 43690).

Accordingly, the amendments meet the eligibility cr iteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and 51.22(c)(10). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

ML21313A008 *by email OFFICE NRR/DORL/LPL4/PM* NRR/DORL/LPL4/LA* NRR/DSS/STSB/BC (A)*

NAME SLingam PBlechman NJordan DATE 11/8/2021 11/16/2021 11/4/2021 OFFICE OGC* NLO with changes NRR/DORL/LPL4/BC* NRR/DORL/LPL4/PM*

NAME MSpencer JDixon-Herrity SLingam DATE 12/1/2021 12/8/2021 12/8/2021

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