ML13168A547

Issuance of Amendment to Facility Operating License, Proposed No Significant Hazards Consideration Determination, and Opportunity for Hearing (Exigent Circumstances)

ML13168A547
Person / Time
Site:	Watts Bar
Issue date:	06/22/2013
From:	Andrew Hon Plant Licensing Branch II
To:	James Shea Tennessee Valley Authority
Hon A NRR/DORL/LPL2-2
References
TAC ME1875
Download: ML13168A547 (16)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 June 22, 2013 Mr. Joseph W. Shea Vice President, Nuclear Licensing Tennessee Valley Authority 3R Lookout Place 1101 Market Street, LP 3D-C Chattanooga, TN 37402-2801

SUBJECT:

WATIS BAR NUCLEAR PLANT, UNIT 1 - ISSUANCE OF AMENDMENT REGARDING APPLICATION FOR ONE-TIME CHANGE TO TECHNICAL SPECIFICATION UNDER EXIGENT CIRCUMSTANCES TO ALLOW FOR CONTAINMENT SPRAY PUMP REPAIR (TAC NO. MF1875)

Dear Mr. Shea:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 93 to Facility Operating License No. NPF-90 for Watts Bar Nuclear Plant, Unit 1. This amendment consists of a change Technical Specification (TS) 3.6.6, "Containment Spray System," in response to your application dated May 22, 2013, as supplemented by letter dated June 12, 2013.

The requested change was for a one-time extension of the Completion Time (CT) for TS Limiting Condition for Operation 3.6.6 Required Action A.1 to restore containment spray (CS) train to OPERABLE status from 72 hours3 days <br />0.429 weeks <br />0.0986 months <br /> to 7 days for an inoperable CS train B. This change would allow sufficient time for replacing an intermittently leaking mechanical seal on CS pump 1B-B. The pump repair is currently scheduled for the week of June 24, 2013 to June 30, 2013, therefore, the revised CT would only be applicable from June 24,2013 to June 30, 2013. TVA requested this proposed T8 change under exigent circumstances in accordance with Title 10, Code of Federal Regulations Section 50.91(a)(6).

J. Shea -2 A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice. If you have any questions regarding this letter, please contact me at (301 )415-8480.

Sincerely, Andrew Hon, Project Manager Plant licensing Branch 11-2 Division of Operating Reactor licensing Office of Nuclear Reactor Regulation Docket No. 50-390

Enclosures:

1. Amendment No. 93 to NPF-90

2. Safety Evaluation cc w/encls: Distribution via listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-390 WATTS BAR NUCLEAR PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 93 License No. NPF-90

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

A. The application for amendment by the Tennessee Valley Authority (the licensee) dated May 22,2013, as supplemented by letter dated June 12, 2013, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in Title 10 of the Code of Federal Regulations (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 Oi) that such activities will be conducted in compliance with the Commission's 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 Commission's regulations and all applicable requirements have been satisfied.

-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 Facility Operating License No. NPF-90 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. 93 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. TVA shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of the date of its issuance, and shall be implemented on June 24, 2013.

FOR THE NUCLEAR REGULATORY COMMISSION

~~ ~.~ +or Jessie F. Quichocho, Chief Plant Licensing Projects Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Operating License and the Technical Specifications Date of Issuance: June 22,2013

ATTACHMENT TO LICENSE AMENDMENT NO. 93 FACILITY OPERATING LICENSE NO. NPF-90 DOCKET NO. 50-390 Replace Page 3 of Operating License NPF-90 with the attached Page 3.

Replace the following page of the Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

REMOVE INSERT 3.6-18 3.6-18

-3 (4) TVA, 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, instrument calibration, or other activity associated with radioactive apparatus or components; and (5) TVA, 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 facility.

C. This 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 TVA is authorized to operate the facility at reactor core power levels not in excess of 3459 megawatts thermal.

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 93 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. TVA shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) Safety Parameter Display System (SPDS) (Section 18.2 of SER Supplements 5 and 15)

Prior to startup following the first refueling outage, TVA shall accomplish the necessary activities, provide acceptable responses, and implement all proposed corrective actions related to having the Watts Bar Unit 1 SPDS operational.

(4) Vehicle Bomb Control Program (Section 13.6.9 of SSER 20)

During the period of the exemption granted in paragraph 2.D.(3) of this license, in implementing the power ascension phase of the approved initial test program, TVA shall not exceed 50% power until the requirements of 10 CFR 73.55(c)(7) and (8) are fully implemented. TVA shall submit a letter under oath or affirmation when the requirements of 73.55(c)(7) and (8) have been fully implemented.

Facility License No. NPF-90 Amendment No. 93

Containment Spray System 3.6.6 3.6 CONTAINMENT SYSTEMS 3.6.6 Containment Spray System LCO 3.6.6 Two containment spray trains and two residual heat removal (RHR) spray trains shall be OPERABLE.

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N0 TE------------------------------------------

The RHR spray train is not required in MODE 4.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One containment spray train A.1 Restore containment spray 72 hours3 days <br />0.429 weeks <br />0.0986 months <br /> {*}

inoperable. train to OPERABLE status.

B. One RHR spray train B.1 Restore RHR spray train to 72 hours3 days <br />0.429 weeks <br />0.0986 months <br /> inoperable. OPERABLE status.

C. Required Action and C.1 Be in MODE 3. 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br /> associated Completion Time not met. AND Co2 Be in MODE 5. 84 hours3.5 days <br />0.5 weeks <br />0.115 months <br />

• For the week commencing June 24,2013 (expiring on June 30, 2013), containment spray pump 1B-B may be inoperable for a period not to exceed 7 days for mechanical seal repair.

Watts Bar-Unit 1 3.6-18 Amendment 93

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 93 TO FACILITY OPERATING LICENSE NO. NPF-90 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT, UNIT 1 DOCKET NO. 50-390

1.0 INTRODUCTION

By letter to the Nuclear Regulatory Commission (NRC) dated May 22, 2013 (Agencywide Documents Access and Management System (ADAMS), Accession No. ML13143A166) as supplemented by information in its letter dated June 12, 2013 (ADAMS ML13163A221),

Tennessee Valley Authority (TVA, the licensee) requested an amendment to the Watts Bar Nuclear Plant, Unit 1 (WBN-1) Technical Specification (TS) 3.6.6. Specifically, the requested change was for a one-time extension of the Completion Time (CT) for TS Limiting Condition for Operation (LCO) 3.6.6 Required Action A.1 to restore containment spray (CS) train to OPERABLE status from 72 hours3 days <br />0.429 weeks <br />0.0986 months <br /> to 7 days for an inoperable CS train B. This change would allow sufficient time for replacing an intermittently leaking mechanical seal on CS pump (CSP)1 B-B. The pump repair is currently scheduled for the week of June 24, 2013. TVA requested this proposed TS change under exigent circumstances in accordance with Title 10, Code of Federal Regulations (10 CFR) Section 50.91 (a)(6). This requested change along with commitments for maintaining operability of the other containment heat removal and pressure control features, containment spray train A, residual heat removal for supplemental containment spray, air return system, and the ice condenser during the time CSP 1B-B will be inoperable are described in TVA's letter dated May 22,2013. The revised CT would only be applicable from June 24, 2013 to June 30, 2013.

The supplemental letter dated June 12, 2013, provided additional information that clarified the application, did not expand the scope of the application as originally noticed and did not change the original proposed no significant hazard consideration determination as published in the Federal Register (78 FR 33117) on June 3, 2013.

The containment is designed to assure that an acceptable upper limit of leakage of radioactive material is not exceeded under design basis accident conditions. The CS system which sprays cool water into the containment atmosphere, thereby limiting the pressure peak, particularly in the long term. The CS system consists of two separate trains of equal capacity, with each train independently capable of meeting the system heat removal requirements. This system can be supplemented with two Residual Heat Removal (RHR) system pumps and two RHR heat exchangers in parallel, with associated piping, valves, and individual spray headers in the upper

-2 containment volume. Each CS train includes a pump, heat exchanger, ring header with nozzles, isolation valves and associated piping, and instrumentation and controls. Partial flow from an RHR system pump through its associated heat exchanger can be used to supplement each CS train. During normal operation, all of the equipment is idle and the associated isolation valves are closed. Upon system activation during a loss of coolant accident (LOCA) or other high energy line break, adequate containment cooling is provided by the CS system.

2.0 REGULATORY EVALUATION

Section 182a of the Atomic Energy Act (the Act) requires applicants for nuclear power plant operating licenses to include TSs as part of the license. The TSs ensure the operational capability of structures, systems, and components that are required to protect the health and safety of the public. The regulatory requirements related to the content of the TSs are contained in 10 CFR 50.36. That regulation requires that the TSs include items in the following specific categories: (1) Safety limits, limiting safety system settings, and limiting control settings (50.36(c)(1 >>; (2) limiting conditions for operation (50.36{c){2>>; (3) Surveillance requirements (50.36(c)(3>>; (4) Design features (50.36(c)(4>>; and (5) Administrative controls (50.36(c)(5>>.

Section 50.36(c)(2) of 10 CFR states in part:

limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met.

Required Action A.1 of LCO 3.6.6 is the remedial action permitted by the TS, when one CS train is inoperable. It requires the licensee to restore the CS train to OPERABLE status within 72 hours3 days <br />0.429 weeks <br />0.0986 months <br />. The licensee has requested a one-time change to the CT for Required Action A.1 from 72 hours3 days <br />0.429 weeks <br />0.0986 months <br /> to 7 days to repair an intermittent seal leakage for CSP 1B-B online.

Applicable General Design Criteria (GDC) are described in the WBN-1 Updated Final Safety Analysis Report (UFSAR) Section 3.1.2, WBN Conformance With GDC. WBN-1 was designed to meet the intent of the uProposed General Design Criteria for Nuclear Power Plant Construction Permits" published in July 1967. The WBN-1 construction permit was issued in January 1973. This UFSAR, however, addresses the NRC GDC published as Appendix A to 10 CFR Part 50 in July 1971, including Criterion 4 as amended October 27, 1987.

Criterion 16 - Containment deSign Reactor containment and associated systems shall be provided to establish an essentially leaktight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.

The containment spray system sprays cool water into the containment atmosphere, thereby limiting the pressure peak, particularly in the long term.

-3 Criterion 38 - Containment heat removal A system to remove heat from the reactor containment shall be provided. The system safety function shall be to reduce rapidly, consistent with the functioning of other associated systems, the containment pressure and temperature following any loss of coolant accident and maintain them at acceptably low levels. Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for on site electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

The CS system sprays cool water into the containment atmosphere to remove heat following any LOCA, particularly in the long term.

Criterion 50 - Containment design basis The reactor containment structure, including access openings, penetrations, and the containment heat removal system shall be designed so that the containment structure and its internal compartments can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from any loss-of-coolant accident. This margin shall reflect consideration of (1) the effects of potential energy sources which have not been included in the determination of the peak conditions, such as energy in steam generators and as required by § 50.44 energy from metal-water and other chemical reactions that may result from degradation but not total failure of emergency core cooling functioning, (2) the limited experience and experimental data available for defining accident phenomena and containment responses, and (3) the conservatism of the calculational model and input parameters.

The CS system sprays cool water into the containment atmosphere, thereby limiting the pressure peak, particularly in the long term.

3.0 TECHNICAL EVALUATION

WBN-1 TS LCO 3.6.6 requires, among other things, two CS trains to be OPERABLE in MODES 1 through 4; (MODE 1 is power operation, MODE 2 is start up, MODE 3 is hot standy and MODE 4 is hot shutdown.) The RHR spray train is not required in MODE 4. The licensee plans to perform a mechanical seal replacement maintenance activity on the 1B-B CSP while the unit is at power. The activity will require the licensee to enter Condition A of TS 3.6.6, one containment spray train inoperable. The licensee stated that the expected time required for the maintenance activity is 68 hours2.833 days <br />0.405 weeks <br />0.0931 months <br />. However, the expected maintenance time does not include time that may be necessary to address any additional issues that may be encountered during the seal replacement. The licensee requested to change the CT for Required Action A.1 to 7 days. This request represents an extension of current TS CT by 4 days and therefore, a relaxation of TS requirements. The licensee presented multiple justifications for the extension of the CT.

-4 Normally the licensee performed seal replacement for this pump during refueling outages. In this online maintenance, the licensee planned to complete the planned maintenance described in 68 hours2.833 days <br />0.405 weeks <br />0.0931 months <br />, thus, within the allowed time of 72 hours3 days <br />0.429 weeks <br />0.0986 months <br />, but it might take up to additional 126 hours5.25 days <br />0.75 weeks <br />0.173 months <br /> to resolve other issues that may be identified with the pump during the maintenance task. The licensee, in its letters of May 22 and June 12, 2013, described the maximum possible impact of removing this spray pump from service on the plant's response to a design basis accident LOCA with concurrent loss of offsite power and the failure of the emergency diesel generator or electrical supply bus that would power the remaining CSP.

The containment maximum capability pressure is 15 pound square inch - gauge (psig). At the request of the staff, the licensee performed additional sensitivity studies of the minimum emergency core cooling system (ECCS) response and the additional loss of the remaining spray pump. The study showed:

• Containment pressure exceeded 15 psig in about 2.5 hours0.208 days <br />0.0298 weeks <br />0.00685 months <br /> after the start of the accident and reaching a peak of about 32 psig at about 5 hours0.208 days <br />0.0298 weeks <br />0.00685 months <br /> before slowly declining.

• Recovering use of the remaining CSP at any time while pressure was above 15 psig would allow reduction in containment pressure to below 15 psig in about 5 minutes. Design and construction of the containment in accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code involved a design factor of safety of three.

Thus, a structural failure of the containment would not be expected until internal pressure exceeded three times the maximum capability pressure, as documented in NUREG/CR-6906, "Containment Integrity Research at Sandia National Laboratories."

With respect to the leakage behavior of containment penetrations, the licensee's letter of June 12, 2013, indicated that the last integrated leakage rate test was performed on October 12,2012, and demonstrated a leakage rate of 42.7 percent of the allowable leakage of 246 standard cubic feet per hour (that assumed in the dose analysis). The last local leak rate test results summation (penetration leakage) was performed May 23,2013, and showed minimum penetration pathway leakage to be less than 15 percent of the allowable leakage. A conservative linear extrapolation of containment leakage to the 32 psig maximum calculated design basis accident containment pressure for the single train of ECCS without the CSP would still be within the assumed leakage of the dose analysis.

In addition, WBN-1 does not credit CS for the removal of fission products in any design basis dose consequence analyses performed to ensure the acceptance criteria of 10 CFR Part 100 are satisfied.

Like most Westinghouse pressurized reactors, the CT of 72 hours3 days <br />0.429 weeks <br />0.0986 months <br /> for the CSP inoperability was established historically by engineering judgment as a tradeoff between 1) not fully meeting the single active failure criterion for a design basis LOCA with a loss of offsite power, and 2) requiring frequent plant shutdown and startup transients (shutdown/startup) when most small to moderate maintenance actions on a component could be completed in a reasonably short time to restore operability, and that the possibility of an accident occurring during that short time was

-5 very small. Thus, the one-time extension of the CT for TS LCO 3,6,6 Required Action A,1 for an additional 4 days would not be expected to significantly affect the dose consequence analyses results should an accident occur while the spray pump was out of service. Therefore, the NRC staff determined that the proposed change is acceptable, 4,0 EXIGENT CIRCUMSTANCES

Background

The NRC's regulations contain provisions for issuance of amendments when the usual 30-day public comment period cannot be met These provisions are applicable under exigent circumstances. Consistent with the requirements in 10 CFR 50,91 (a)(6), exigent circumstances exist when: (1) a licensee and the NRC must act quickly; (2) time does not permit the NRC to publish a Federal Register notice allowing 30 days for prior public comment; and (3) the NRC determines that the amendment involves no significant hazards considerations, As discussed in the licensee's application dated May 22, 2013, the licensee requested that the proposed amendment be processed by the NRC on an exigent basis.

Under the provisions in 10 CFR 50,91 (a)(6), the NRC notifies the public in one of two ways:

(1) by issuing a Federal Register notice providing notice of an opportunity for hearing and allowing at least 2 weeks from the date of the notice for prior public comment; or (2) by using local media to provide reasonable notice to the public in the area surrounding the licensee's facility, In this case, the NRC issued a Federal Register notice 78 FR 33117 on June 3,2013, providing notice of an opportunity for hearing and allowing at least 2 weeks from the date of the notice for prior public comment Coming out of the fall 2012 WBN-1 Refueling Outage 11 on October 29,2012, CSP 1B-B mechanical seal leakage was observed to be very low and continued to be very low (Le" less than 1 drop/minute) until the next CSP 1B-B quarterly performance test in early February 2013, during which seal leakage was observed to have increased to 15 - 20 drops/minute, During a subsequent test, CSP 1B-B seal leakage further increased to approximately 100 drops/minute.

In response, TVA started daily monitoring of seal leakage and provided three separate trigger points and actions, During the day shift on April 25, 2013, seal leakage rate was observed to be substantially higher at 3,180 cubic centimeters/hour (cc/hr). (100 drops/min is approximately 389 cc/hr.) After a followup pump run in the afternoon of April 25, 2013, no visible seal leakage was observed during this run, Furthermore, following the CSP 1B-B run, leakage was observed to be minimal (Le" less than 1 drop/minute). After the conclusion of the pump run, leakage was observed to have returned to less than 1 drop/minute during frequent leak checks conducted throughout the following shift, Even though TVA determined that this provided reasonable assurance that the mechanical seal would not fail catastrophically and render the pump inoperable, the exigency for conducting the mechanical seal replacement online resides in the fact that CSP seal leakage contributes to Reactor Coolant System (RCS) leakage outside containment when the CSP suction is aligned to the containment sump during the recirculation phase of LOCA mitigation, There is a limit of 3,760 cc/hr for total RCS leakage outside containment as specified in UFSAR Table 6,3-6, E-2, This limit is used in WBN's offsite and control room dose analyses, and is associated with the

-6 Primary Coolant Sources Outside Containment Program as required by WBN TS 5.7.2.4, which is a subset of the WBN ASME Section XI System Pressure Test Program. The current CSP 1B-B leakage is required to be tracked by both programs. With respect to the CSP 1B-B seal leakage, the measured leakage should not exceed the UFSAR limit for total leakage of 3,760 cC/hr, minus the cumulative sum of the actual current leakage from the other sources contributing to RCS leakage outside of containment. Even though the CSP 1B-B seal leakage has returned to a manageable level following the last pump run, TVA recognizes the potential for another unexpected increase in CSP 'I B-B seal leakage, which could again challenge the primary coolant sources outside containment leakage limit. As a result of this event, TVA decided to expeditiously replace the CSP 1B-B mechanical seal at the first available opportunity. The licensee is ready to perform this repair the week of June 24, 2013, which is the first available Train B workweek.

Summary Based on the above circumstances, the NRC staff finds that the licensee made a timely application for the proposed amendment following identification of the issue. In addition, the NRC staff finds that the licensee could not avoid the exigency without the continued risk of another unanticipated significant CSP 1B-B seal leakage that would place the plant in an unanalyzed condition and force a plant shutdown due to delayed seal repair. Based on these findings, and the determination that the amendment involves no significant hazards consideration as discussed below, the NRC staff has determined that a valid need exists for issuance of the license amendment using the exigent provisions of 10 CFR 50.91 (a)(6).

6.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION

DETERMINATION The NRC's regulations in 10 CFR 50.92 state that the NRC may make a final determination that a license amendment involves no significant hazards consideration if operation of the facility, in accordance with the proposed amendment would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

As required by 10 CFR 50.91 (a), the licensee's determination no significant hazards consideration is presented below:

The proposed change will provide a one-time change to extend the Completion Time for Technical Specification (TS) Limiting Condition for Operation (LCO) 3.6.6 Required Action A.1 for restoration of Containment Spray (CS) Train B to Operable status from 72 hours3 days <br />0.429 weeks <br />0.0986 months <br /> to 7 days. This change is necessary to support the CSP 1B-B mechanical seal replacement maintenance activity scheduled to commence on June 24, 2013.

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed change does not alter any plant equipment or operating practices in such a manner that the probability of an accident is [significantly] increased. The

-7 proposed change will not alter assumptions relative to the mitigation of an accident or transient event Therefore, this proposed change does not [significantly] increase the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation.

Therefore, this proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Based on the Operability of the required containment ESF [engineered safety feature] systems for containment heat removal, the proposed change ensures that the accident analysis assumptions continue to be met. The design and operation of these systems are not affected by the proposed change. The safety analysis acceptance criteria are not altered by the proposed change.

Therefore, the proposed change does not involve a significant reduction in the margin of safety.

Based on the above determination evaluation, the NRC staff concludes that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff has made a final determination that no significant hazards considerations are involved for the proposed amendment.

7.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Tennessee State official was notified of the proposed issuance of the amendment. The State official had no comments.

8.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in10 CFR Part 20.

The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (78 FR 33117). The Commission has made a final determination that the amendment involves no significant hazards consideration.

The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite,

-8 and that there is no significant increase in individual or cumulative occupational radiation exposure.

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

9.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, there is reasonable assurance that: (1) the health and safety of the public will not be endangered by operation in the proposed manner, (2) that such activities will be conducted in compliance with the Commission's 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.

Contributors: J. Bettie, NRR M. Hamm, NRR A. Hon, NRR J. Parillo, NRR Date: June 22,2013

J. Shea -2 A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice. If you have any questions regarding this letter, please contact me at (301 )415-8480.

Sincerely, IRA!

Andrew Hon, Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-390

Enclosures:

1. Amendment No. 93 to NPF-90

2. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:

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DATE 06/20/2013 06/21/2013 06/20/2013 06/20/2013 06/20/2013 OFFICE NRR/APLAlBC OGC/NLO NRRlLPB2-/BC NRR/LP2-2/PM JQuichocho NAME HHamzehee* DRoth AHon (SLingam for)

DATE 06/19/2013 06/21/2013 06/22/2013 06/22/2013 OFFICIAL AGENCY RECORD