ML15100A046

SQN U1 Renewed License v2

ML15100A046
Person / Time
Site:	Sequoyah
Issue date:	09/28/2015
From:	Bill Dean, Sayoc E Division of License Renewal
To:
Sayoc E,NRR/DLR,415-4084
References
Download: ML15100A046 (15)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-327 SEQUOYAH NUCLEAR PLANT, UNIT 1 RENEWED FACILITY OPERATING LICENSE Renewed License No. DPR-77

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

A. The application for renewed licenses filed by the Tennessee Valley Authority (the licensee or TVA) complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I, and all required notifications to other agencies or bodies have been duly made; B. Construction of the Sequoyah Nuclear Plant, Unit 1 (the facility), has been substantially completed in conformity with Provisional Construction Permit No. CPPR-72 and the application, as amended, the provisions of the Act and the regulations of the Commission; C. The facility will operate in conformity with the application, as amended, the provisions of the Act, and the regulations of the Commission; D. There is reasonable assurance: (i) that the activities authorized by this renewed operating license can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the regulations of the Commission set forth in 10 CFR Chapter I; E. The Tennessee Valley Authority is technically and financially qualified to engage in the activities authorized by this renewed operating license in accordance with the Commission's regulations set forth in 10 CFR Chapter I; F. The Tennessee Valley Authority has satisfied the applicable provisions of 10 CFR Part 140, "Financial Protection Requirements and Indemnity Agreements", of the Commission's regulations; G. The issuance of this renewed license will not be inimical to the common defense and security or to the health and safety of the public; Renewed License No. DPR-77

H. After weighing the environmental, economic, technical, and other benefits of the facility against environmental and other costs and considering available alternatives, the issuance of renewed Facility Operating License No. DPR-77, subject to the conditions for protection of the environment set forth herein, is in accordance with 10 CFR Part 50, Appendix D*, of the Commission's regulations and all applicable requirements have been satisfied; and I. The receipt, possession, and use of source, byproduct and special nuclear material as authorized by this renewed license will be in accordance with the Commission's regulations in 10 CFR Parts 30, 40 and 70.

J. Actions have been identified and have been or will be taken with respect to (1) managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21(a)(1), and (2) time-limited aging analyses that have been identified to require review under 10 CFR 54.21(c), such that there is reasonable assurance that the activities authorized by the renewed operating license will continue to be conducted in accordance with the current licensing basis, as defined in 10 CFR 54.3, for the facility, and that any changes made to the facilitys current licensing basis in order to comply with 10 CFR 54.29(a) are in accordance with the Act and the Commissions regulations.

2. Renewed Facility Operating License No. DPR-77 is hereby issued to Tennessee Valley Authority to read as follows:

A. This renewed license applies to the Sequoyah Nuclear Plant, Unit 1, a pressurized water nuclear reactor and associated equipment (the facility), owned by the Tennessee Valley Authority. The facility is located in Hamilton County, Tennessee, about 9.5 miles northeast of Chattanooga, and is described in TVA's Final Safety Analysis Report as supplemented and amended, and the Final Environmental Statement prepared by the Tennessee Valley Authority.

B. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses the Tennessee Valley Authority:

(1) Pursuant to Section 104(b) of the Act and 10 CFR Part 50, "Licensing of Production and Utilization Facilities", to possess, use, and operate the facility at the designated location in Hamilton County, Tennessee, in accordance with the procedures and limitations set forth in this renewed license, (2) Pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended;

• See 10 CFR § 51.56 Renewed License No. DPR-77

(3) 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 monitoring equipment calibration, and as fission detectors in amounts as required; (4) 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 (5) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the Sequoyah and Watts Bar Unit 1 Nuclear Plants.

C. This renewed license shall be deemed to contain and is subject to the conditions specified in the Commission's 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 The Tennessee Valley Authority is authorized to operate the facility at reactor core power levels not in excess of 3455 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 333 are hereby incorporated into this renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

(3) Initial Test Program The Tennessee Valley Authority shall conduct the post-fuel-loading initial test program (set forth in Section 14 of Tennessee Valley Authority's Final Safety Analysis Report, as amended), without making any major modifications of this program unless modifications have been identified and have received prior NRC approval. Major modifications are defined as:

a. Elimination of any test identified in Section 14 of TVA's Final Safety Analysis Report as amended as being essential;

b. Modification of test objectives, methods, or acceptance criteria for any test identified in Section 14 of TVA's Final Safety Analysis Report as amended as being essential; Renewed License No. DPR-77

c. Performance of any test at a power level different from there described; and

d. Failure to complete any tests included in the described program (planned or scheduled for power levels up to the authorized power level).

(4) Essential Raw Cooling Water (ERCW) Intake Protection (Section 2.2)*

TVA shall provide analyses on the vulnerability of the ERCW intake structure by January 1, 1981. These analyses shall include the probability of a collision by a barge at full speed from any credible direction, including a tow proceeding in the upstream direction, and the ability of the ERCW intake to withstand such collisions, including those by barges carrying flammable cargoes (including liquid natural gas (LNG)). Operation of the new ERCW intake for Unit 2 is not permitted until this matter is resolved.

(5) Onsite Meteorological Measurements Program (Section 2.3.3)

TVA shall be required to make any design modifications that the staff may deem necessary to meet the requirements of NUREG-0654, Appendix 2 and Criterion 4 of Section 2.3.3.

(6) Seismic Design Margin (Section 2.5)

No later than March 1, 1982, TVA shall provide the results of their seismic design margin review conducted in accordance with an NRC approved seismic design margin review program for this facility.

(7) Monitoring Settlement Markers (Section 2.6.3)

Prior to January 1, 1981, TVA shall report to the NRC on their continuing monitoring of settlement markers along the ERCW conduit for the new intake structure. The magnitude of the settlement must be determined and this matter resolved to NRC's satisfaction prior to using the new ERCW for Unit 2.

(8) Low Pressure Turbine Disc Inspection (Section 3.5.1)

Prior to start-up after the second refueling outage, TVA shall reinspect the low pressure turbine discs for cracks to assure that turbine integrity will not be jeopardized. A report shall be submitted to the NRC staff 30 days after the inspection is completed.

• Referenced sections in these conditions refer to appropriate sections in the Safety Evaluation Report (NUREG-0011) and its supplements on this facility.

Renewed License No. DPR-77

(9) Steam Generator Inspection (Section 5.3.1)

(a) Prior to March 1, 1981, TVA shall provide to the NRC the results of its tests to determine the feasibility of using a steam generator camera device.

(b) Prior to start-up after the first refueling, TVA must install inspection ports in each steam generator if the results of the camera device inspection are not satisfactory to the NRC; (c) Prior to start-up after the first refueling, TVA will plug Row 1 of the steam generator tubes, if required by NRC.

(10) Water Chemistry Control Program (Section 5.3.2)

This requirement has been deleted.

(11) Negative Pressure in the Auxiliary Building Secondary Containment Enclosure (ABSCE) (Section 6.2.3)

After the final ABSCE configuration is determined, TVA must demonstrate to the satisfaction of the NRC that a negative pressure of 0.25 inches of water gauge can be maintained in the spent fuel storage area and in the ESF pump room.

(12) Environmental Qualification (Section 7.2.2)

(a) No later than November 1, 1980, TVA shall submit information to show compliance with the requirement of NUREG-0588, "Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment,"

for safety-related equipment exposed to a harsh environment.

Implementation shall be in accordance with NUREG-0588 by June 30, 1982.

(b) By no later than December 1, 1980, complete and auditable records must be available and maintained at a central location which describe the environmental qualification method used for all safety-related electrical equipment in sufficient detail to document the degree of compliance with the DOR Guidelines or NUREG-0588. Thereafter, such records should be updated and maintained current as equipment is replaced, further tested, or otherwise further qualified to document complete compliance by June 30, 1982.

Renewed License No. DPR-77

(c) By no later than June 30, 1982, all safety-related electrical equipment in the facility shall be qualified in accordance with the provisions of:

Division of Operating Reactors "Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors" (DOR Guidelines); or, NUREG-0588, "Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment,"

December 1979. Copies of these documents are attached to the Order for Modification of License DPR-77 dated November 6, 1980.

(13) Loss of Non-Class IE Instrumentation and Control Room System Bus During Operation (Section 7.10)

Prior to exceeding five percent power, TVA must complete revisions to plant emergency procedures to the satisfaction of the NRC.

(14) Engineering Safety Feature (ESF) Reset Controls (Section 7.11)

In conformance with IE Bulletin 80-06, TVA shall test the system to identify any further areas of concern, and TVA shall review the control schemes to determine that they are the best in terms of equipment control and plant safety. The results of these test and review efforts shall be provided to the NRC in accordance with the bulletin.

(15) This specification has been deleted (16) Fire Protection TVA shall implement and maintain in effect all provisions of the approved fire protection program referenced in Sequoyah Nuclear Plant's Final Safety Analysis Report and as approved in NRC Safety Evaluation Reports contained in NUREG-0011, Supplements 1, 2, and 5, NUREG-1232, Volume 2, NRC letters dated May 29 and October 6, 1986, and the Safety Evaluation issued on August 12, 1997, for License Amendment No. 227, subject to the following provision:

TVA may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

(17) Acccident Analysis (Steam and Feedwater Line Breaks) (Section 15.2)

If requested, TVA shall provide plant specific inputs to NRC for an independent audit of steam and feedwater line break analyses. TVA shall implement any modifications that may become necessary as a result of these analyses or audit.

Renewed License No. DPR-77

(18) Requirements for Modification To Or Addition Of Instrumentation And Controls (a) Within 18 months after issuance of this license, instrument downscale failure alarms shall be installed for the effluent monitoring instrumentation channels for radioactive gaseous and radioactive liquid effluents. Also, appropriate modifications to procedures and Technical Specifications 3.3.3.9 and 3.3.3.10 shall have been completed.

(b) Within six months from issuance of this license, TVA shall submit for NRC review the basis for the values for each Reactor Protection System and Engineered Safety Feature instrumentation channel including:

(1) Technical Specification trip setpoint value; (2) Technical Specification allowable value (the Technical Specification trip setpoint plus the instrument drift assumed in the accident analysis);

(3) The instrument drift assumed to occur during the interval between Technical Specification surveillance tests; (4) The components of the cumulative instrument bias; and (5) The minimum margin between the Technical Specification trip setpoint, the allowable value, and the trip value assumed in the accident analysis.

(c) At the first outage of sufficient duration but no later than startup following the second refueling outage, TVA shall have installed, demonstrated operable, proposed appropriate Technical Specifications, and received NRC approval for an additional level of over/undervoltage protection acceptable to the NRC staff. The level of protection from the effects of power transients on safety-related equipment provided by Part I of the staff's "Degraded Grid Voltage Position," or equivalent, is required.

Renewed License No. DPR-77

(19) Mechanical Snubbers This condition is deleted.

(20) Low Temperature Overpressure Protection (Section 5.2.2)

At the first outage of sufficient duration but no later than startup following the second refueling outage, TVA shall install an overpressure mitigation system which meets NRC requirements.

(21) Control Rod Guide Thimble (Section 4.2)

Prior to startup after first refueling, TVA shall submit the details of the inspection program for control rod guide thimble tube wall wear for NRC approval.

(22) TMI Action Plan Full Power Conditions Each of the following conditions shall be completed to the satisfaction of the NRC by the times indicated:

A. Safety Engineering Group (Section 22.2.I.B.1.2)

This condition is deleted.

B. Short-Term Accident Analysis and Procedure Revision (Section 22.2.I.C.1)

Within thirty effective full-power days, TVA shall revise Emergency Operating Procedures and brief the operators on the revision.

C. Control Room Design (Section 22.2.I.D.1)

TVA shall consider the benefits of installing data recording and logging equipment in the control room to correct the deficiencies associated with the trending of important parameters on strip chart recorders used in the control room as part of the Detailed Control Room Design Review.

Implementation shall be carried out in accordance with SECY 82-111B.

Renewed License No. DPR-77

D. Hydrogen Control Measures (Section 22.2, II.B.7)

(1) Four additional igniter units shall be installed in the containment upper containment compartment in locations acceptable to the NRC staff prior to startup following the second refueling outage.

(2) Additional tests shall be performed on the Tayco igniter to demonstrate that the igniters will initiate combustion in a spray environment such as that expected in the upper compartment of the ice condenser containment.

E. Auxiliary Feedwater (Section 22.2, II.E.1.1)

Prior to exceeding five percent power, auxiliary feedwater pump endurance tests will be completed and a report shall be submitted to NRC within 30 days after all tests are completed.

F. Radiation Monitors (Section 22.2.II.E.4.2)

TVA will install Radiation Monitors for isolation of fluid lines carrying potential radioactivity outside of containment at the earliest practical date consistent with scheduled or forced plant outages but prior to operation following the first refueling.

G. Emergency Preparedness Plan (Section 22.2.III.A.1.1 And Appendix E)

(a) TVA shall maintain in effect an emergency plan that meets the regulatory requirement of 10 CFR Part 50, Appendix E, and the operator Planning Objectives of NUREG-0654, "Criteria for Preparation and Evaluation of Radiological Preparedness in Support of Nuclear Power Plants," January 1980.

(b) No later than 90 days from the date of issuance of this license, TVA shall report to the NRC the status of any items related to emergency preparedness identified by FEMA or the NRC as requiring further action.

Renewed License No. DPR-77

(23) TMI Action Plan Dated Conditions Each of the following conditions shall be completed to the satisfaction of the NRC by the times indicated:

A. Shift Technical Advisor (Section 22.3, I.A.1.1)

This condition is deleted.

Items for completion by January 1, 1981:

B. Plant Shielding (Section 22.3, II.B.2)

TVA shall complete modification to assure adequate access to vital areas and protection of safety equipment following an accident resulting in a degraded core.

C. Auxiliary Feedwater Initiation and Indication (Section 22.3, II.E.1.2)

(a) TVA shall upgrade, as necessary, automatic initiation of the auxiliary feedwater system to safety-grade quality.

(b) TVA shall upgrade, as necessary, the indication of auxiliary feedwater flow to each steam generator to safety grade quality.

D. Additional Accident Monitoring Instrumentation (Section 22.3, II.F.1)

(1) TVA shall install interim noble gas monitors at the first outage of sufficient duration.

(2) At the first outage of sufficient duration but no later than startup following the second refueling outage, TVA shall install the following qualified monitoring instrumentation:

(a) Integrated monitoring assembly which will accomplish particulate, iodine and noble gas monitoring.

(b) Containment high range radiation monitor.

(c) Containment pressure monitor.

(d) Containment water level monitor.

(e) Containment hydrogen monitor.

E. Reactor Coolant System Vents (Section 22.3, II.B.1)

At the first outage of sufficient duration, but no later than startup following second refueling outage, TVA shall install reactor coolant system and reactor vessel head highpoint vents that are remotely operable from the control room.

Renewed License No. DPR-77

F. Post Accident Sampling (Section 22.3, II.B.3)

This condition has been deleted.

H. Instruments for Inadequate Core Cooling (Section 22.3, II.F.2)

(1) By January 1, 1982, TVA shall install a backup indication for incore thermocouples. This display shall be in the control room and cover the temperature range of 200 F - 2000 F.

(2) At the first outage of sufficient duration but no later than startup following the second refueling outage, TVA shall install reactor vessel water level instrumentation which meets NRC requirements.

I. Upgrade Emergency Support Facilities (Section 22.3, II.A.1.2)

(1) At the first outage of sufficient duration, but no later than startup following the second refueling outage, TVA shall update the Technical Support Facilities to meet NRC requirements.

(2) TVA shall maintain interim emergency support facilities (Technical Support Center, Operations Support Center and the Emergency Operations Facility) until the final facilities are complete.

J. Relief and Safety Valve Test Requirements (Section 22.2, II.D.1)

TVA shall conform to the results of the EPRI test program. TVA shall provide documentation for qualifying (a) reactor coolant system relief and safety valves, (b) piping and supports, and (c) block valves in accordance with the review schedule given in SECY 81-491 as approved by the Commission.

(24) Compliance with Regulatory Guide 1.97 TVA shall implement modifications necessary to comply with Revision 2 of Regulatory Guide 1.97, "Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following An Accident,"

dated December 1980 by startup from the Unit 2 Cycle 4 refueling outage.

(25) Mixed Core DNBR Penalty TVA will obtain NRC approval prior to startup for any cycle's core that involves a reduction in the departure from nucleate boiling ratio initial transition core penalty below that value stated in TVA's submittal on Framatome fuel conversion dated April 6, 1997.

Renewed License No. DPR-77

(26) Control Room Air Conditioning System Maintenance TVA commits to the use of a portable chiller package and air-handling unit to provide alternate cooling if both trains of the control room air condition system become inoperable during the maintenance activities to upgrade the compressors and controls or immediately enter Technical Specification 3.0.3.

(27) Mitigation Strategy License Condition Develop and maintain strategies for addressing large fires and explosions and that include the following key areas:

(a) Fire fighting response strategy and with the following elements:

1. Pre-defined coordinated fire response strategy and guidance

2. Assessment of mutual aid fire fighting assets

3. Designated staging areas for equipment and materials

4. Command and control

5. Training of response personnel (b) Operations to mitigate fuel damage considering the following:

1. Protection and use of personnel assets

2. Communications

3. Minimizing fire spread

4. Procedures for implementing integrated fire response strategy

5. Identification of readily-available pre-staged equipment

6. Training on integrated fire response strategy

7. Spent fuel pool mitigation measures (c) Actions to minimize release to include consideration of:

1.Water spray scrubbing 2.Dose to onsite responders (28) The licensee shall implement and maintain all Actions required by Attachment 2 to NRC Order EA-06-137, issued June 20, 2006, except the last action that requires incorporation of the strategies into the site security plan, contingency plan, emergency plan and/or guard training and qualification plan, as appropriate.

(29) Upon implementation of the amendment adopting TSTF-448, Revision 3, the determination of control room envelope (CRE) unfiltered air inleakage as required by Surveillance Requirement (SR) 4.7.7.h, in accordance with TS 6.17.c.(i), the assessment of CRE habitability as required by Specification 6.17.c.(ii), and the measurement of CRE pressure as required by Specification 6.17.d, shall be considered met. Following implementation:

Renewed License No. DPR-77

(a) The first performance of SR 4.7.7.h, in accordance with Specification 6.17.c.(i), shall be within the specified Frequency of 6 years, plus the 18-month allowance of SR 4.0.2, as measured from May 3, 2004; the date of the most recent successful tracer gas test, as stated in the August 4, 2004, letter response to Generic Letter 2003-01; or within the next 18 months if the time period since the most recent successful tracer gas test is greater than 6 years.

(b) The first performance of the periodic assessment of CRE habitability, Specification 6.17.c.(ii), shall be within 3 years, plus the 9-month allowance of SR 4.0.2, as measured from May 3, 2004; the date of the most recent successful tracer gas test, as stated in the August 4, 2004, letter response to Generic Letter 2003-01; or within the next 9 months if the time period since the most recent successful tracer gas test is greater than 3 years.

(c) The first performance of the periodic measurement of CRE pressure, Specification 6.17.d, shall be within 18 months, plus the 138 days allowed by SR 4.0.2, as measured from May 30, 2007, the date of the most recent successful pressure measurement test, or within 138 days if not performed previously.

(30) Steam Generator Replacement Project During the Sequoyah Nuclear Plant, Unit 2, refueling outage 18, lifts of heavy loads associated with the steam generator replacement project shall be performed in accordance with the additional conditions provided in Appendix C.

(31) License Renewal License Conditions (a) The information in the Updated Final Safety Analysis Report (UFSAR) supplement, submitted pursuant to 10 CFR 54.21(d), is henceforth part of the UFSAR which will be updated in accordance with 10 CFR 50.71(e).

As such, the licensee may make changes to the programs and activities described in the UFSAR supplement, without prior Commission approval, provided the licensee evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

(b) The licensee's UFSAR supplement submitted pursuant to 10 CFR 54.21(d), as revised during the license renewal application review process, and as revised in accordance with license condition 2.C.(31)(a), describes certain programs to be implemented and activities to be completed prior to the period of extended operation.

Renewed License No. DPR-77

1. TVA shall implement those new programs and enhancements to existing programs no later than March 17, 2020.

2. TVA shall complete those designated inspection and testing activities, as noted in Appendix A of the Safety Evaluation Report Related to the License Renewal of Sequoyah Nuclear Plant, Units 1 and 2, dated January 2015, no later than March 17, 2020, or the end of the last refueling outage prior to the period of extended operation, whichever occurs later.

3. TVA shall notify the NRC in writing within 30 days after having accomplished item (b)1 above and include the status of those activities that have been or remain to be completed in item (b)2 above.

D. Exemptions from certain requirements of Appendices G and J to 10 CFR Part 50 are described in the Office of Nuclear Reactor Regulation's Safety Evaluation Report, Supplement No. 1. These exemptions are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest. The exemptions are, therefore, hereby granted. The granting of these exemptions are authorized with the issuance of the License for Fuel Loading and Low Power Testing, dated February 29, 1980. The facility will operate, to the extent authorized herein, Act, and the regulations of the Commission.

E. Physical Protection (1) The licensee 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 contain Safeguards Information protected under 10 CFR 73.21, is entitled: "Sequoyah Nuclear Plant Security Plan, Training And Qualification Plan, And Safeguards Contingency Plan" submitted by letter dated May 8, 2006.

(2) The licensee shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The licensee CSP was approved by License Amendment No. 329, as amended by changes approved by License Amendment No. 333.

Renewed License No. DPR-77

F. This renewed license is subject to the following additional condition for the protection of the environment:

Before engaging in additional construction or operational activities which may result in an environmental impact that was not evaluated by the Commission, Tennessee Valley Authority will prepare and record an environmental evaluation of such activity.

When the evaluation indicates that such activity may result in a significant adverse environmental impact that was not evaluated, or that is significantly greater than that evaluated in the Final Environmental Statement prepared by the Tennessee Valley Authority and the Environmental Impact Appraisal prepared by the Commission in May 1979 and March 2015, the Tennessee Valley Authority shall provide a written evaluation of such activities and obtain prior approval from the Director, Office of Nuclear Reactor Regulation.

G. If TVA plans to remove or to make significant changes in the normal operation of equipment that controls the amount of radioactivity in effluents from the Sequoyah Nuclear Plants, the Commission shall be notified in writing regardless of whether the change affects the amount of radioactivity in the effluents.

H. Deleted.

I. TVA shall immediately notify the Commission of any accident at this facility which could result in an unplanned release of quantities of fission products in excess of allowable limits for normal operation established by the Commission.

J. TVA shall have and maintain financial protection of such type and in such amounts as the Commission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims.

K. This renewed license is effective as of the date of issuance and shall expire September 17, 2040.

FOR THE NUCLEAR REGULATORY COMMISSION William M. Dean, Director Office of Nuclear Reactor Regulation

Attachment:

Appendices A and B Technical Specifications, and Appendix C Date of Issuance: September 28, 2015 Renewed License No. DPR-77