ML12249A394

Issuance of Amendments to Revise the Technical Specification to Allow Use of Areva Advanced W17 High Thermal Performance Fuel (TS-SQN-2011-07)

ML12249A394
Person / Time
Site:	Sequoyah
Issue date:	09/26/2012
From:	Siva Lingam Plant Licensing Branch II
To:	James Shea Tennessee Valley Authority
Lingam S
Shared Package
ML12249A388	List: ML12249A394
References
TAC ME6538, TAC ME6539
Download: ML12249A394 (42)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Official Use Only Proprietary Information September 26,2012 Mr. Joseph W. Shea Manager, Corporate Nuclear Licensing Tennessee Valley Authority 3R Lookout Place 1101 Market Street Chattanooga, TN 37402-2801

SUBJECT:

SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 ISSUANCE OF AMENDMENTS TO REVISE THE TECHNICAL SPECIFICATION TO ALLOW USE OF AREVA ADVANCED W17 HIGH THERMAL PERFORMANCE FUEL (TS-SQN-2011-07)

(TAC NOS. ME6538 AND ME6539)

Dear Mr. Shea:

The Nuclear Regulatory Commission (NRC or the Commission) has issued the enclosed Amendment No. 331 to Facility Operating License No. DPR-77 and Amendment No. 324 to Facility Operating License No. DPR-79 for the Sequoyah Nuclear Plant, Units 1 and 2, respectively. These amendments are in response to your application dated June 17, 2011, as supplemented by letters dated July 27, 2011, November 14, 2011, March 23, April 26, May 15, May 24, and June 26,2012. The proposed Technical Specification changes allow the use of AREVA Advanced W17 High Thermal Performance Fuel to address fuel assembly distortion and its resultant fuel handling issues.

The NRC has determined that the related safety evaluation (SE) contains proprietary information pursuant to Title 10 of the Code of Federal Regulations, section 2.390, "Public inspections, exemptions, requests for withholding." Accordingly, the NRC staff has also prepared a redacted, publicly available, non-proprietary version of the SE.

Document transmitted herewith contains sensitive unclassified information. When separated from Enclosure 4, this document is decontrolled.

Official Use Only Propl'letary Information

OffiGial Use Only Proprietary Information

- 2 Copies of the proprietary and non-proprietary versions of the SE are enclosed. A notice of issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely,

~<f'~

Siva P. Lingam, Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-327 and 50-328

Enclosures:

1. Amendment No. 331 to DPR-77

2. Amendment No. 324 to DPR-79

3. Non-Proprietary Safety Evaluation

4. Proprietary Safety Evaluation cc w/o encl 4: Distribution via Listserv Offisial Use Only Proprietary Information

UNrrED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555"()001 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-327 SEQUOYAH NUCLEAR PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.331 License No. DPR-77

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

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated June 17, 2011, and supplemented by letters dated July 27, 2011, November 14, 2011, March 23, April 26, May 15, May 24, and June 26,2012, 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, 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 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. DPR-77 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 331

, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance, to be implemented prior to startup from Unit 1 fall 2013 refueling outage.

FOR THE NUCLEAR REGULATORY COMMISSION SleF.Q~~f Plant Licensing Branch 11-2 Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to the Facility Operating License and Technical Specifications Date of Issuance: September 26, 2012

ATTACHMENT TO LICENSE AMENDMENT NO.331 FACILITY OPERATING LICENSE NO. DPR-77 DOCKET NO. 50-327 Replace Page 3 of Operating License DPR-77 with the attached page 3.

Replace the following pages of the Appendix A Technical Specifications with the attached pages.

The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

REMOVE INSERT 2-1 2-1 2-2 2-2 2-5 2-5 3/42-17 3/42-17 6-13 6-13 6-13a 6-13a 6-14 6-14

C

- 3 (4)

Pursuant to the Act and 10 CFR Parts 3D, 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 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.

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 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.

331 are hereby incorporated into this 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;

c.

Performance of any test at power level different from there described; and Facility Operating License No. DPR-77 Amendment No~31

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETIINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (Tavg) shall not exceed the limits shown in Figure 2.1-1 and the following SLs shall not be exceeded:

2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained?::. 1.132 for the BHTP correlation, ?::. 1.21 for the BWU-N correlation, and?::. 1.21 for the BWCMV correlation.

2.1.1.2 The maximum local fuel pin centerline temperature shall be maintained ~ 4901 of, decreasing by 13.7°F per 10,000 MWD/MTU of bumup for COPERNIC applications, and

~ 4642°F, decreasing by 58°F per 10,000 MWD/MTU of burnup for TAC03 applications.

APPLICABILITY: MODES 1 and 2.

ACTION:

If SL 2.1.1 is violated, restore compliance and be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

APPLICABILITY: MODES 1, 2, 3, 4 and 5.

ACTION:

MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded 2735 pSig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

MODES 3, 4 and 5 Whenever the Reactor Coolant System pressure has exceeded 2735 pSig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes.

SEQUOYAH - UNIT 1 2-1 Amendment No. 41,331

I 620 600 500 +-----1 ACCEPTABLE OPERATION 5OO+----~--+--------+-------,_------_+-------r_----

540+-------~--------+_------_+--------+_------~------~

0,0 0.2 0.4 0,6 0,8 1.0 1.2 FRACTION OF RATED THERMAL POWER SEQUOYAH - UNIT 1 2-2 Amendment No. 19, 331 Figure 2.1-1 Reactor Core Safety Limit - Four Loops in Operation 600T-------~----------------~--------~------~------~

660~=_-----T--------,_----

UNACCEPTABLE OPERATION LL i!.

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U

~

TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT

1. Manual Reactor Trip

2. Power Range Neutron Flux

3. Power Range Neutron Flux High Positive Rate

4. Power Range Neutron Flux, High Negative Rate

5. Intermediate Range, Neutron Flux

6. Source Range Neutron Flux

7. Overtemperature.1.T

8. Overpower.1.T

9. Pressurizer Pressure--Low

10. Pressurizer Pressure--High

11. Pressurizer Water Level-High

12. Loss of Flow NOMINAL TRIP SETPOINT Not Applicable Low Setpoint - 25% of RATED THERMAL POWER High Setpoint - 109% of RATED THERMAL POWER 5% of RATED THERMAL POWER with a time constant 2 2 second 5% of RATED THERMAL POWER with a time constant 2 2 second 25% of RATED THERMAL POWER 105 counts per second See Note 1 See Note 2 1970 psig 2385 psig 92% of instrument span 90% of design flow per loop*

ALLOWABLE VALUES Not Applicable Low Setpoint - s: 27.4% of RATED THERMAL POWER High Setpoint - s: 111.4% of RATED THERMAL POWER s: 6.3% of RATED THERMAL POWER with a time constant 2 2 second s: 6.3% of RATED THERMAL POWER with a time constant :c. 2 second s: 45.20% of RATED THERMAL POWER

1.45 x 105 counts per second See Note 3 See Note 4

c. 1964.8 psig s: 2390.2 psig s: 92.7% of instrument span 2 89.6% of design flow per loop*

• Design flow is 94,600 (91 AOO X 1.035) gpm per loop.

SEQUOYAH - UNIT 1 2-5 Amendment No. 44,141,185,221,223,310, 331

Figure 3.2-1 Flow VS. Power for 4 Loops in Operation 38~lO ~------------------------------------------------------~

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(100,3784001 Acceptable Operiltlon Region unilcceptable Opl!'ratlon Region 98 100 t: 3/0000

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SEQUOYAH - UNIT 1 3/42-17 Amendment No. 223, 331

ADMINISTRATIVE CONTROLS MONTHLY REACTOR OPERATING REPORT 6.9.1.10 DELETED.

CORE OPERATING LIMITS REPORT 6.9.1.14 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

1.

f1(dl) limits for Overtemperature Delta T Trip Setpoints and Mdl) limits for Overpower Delta T Trip Setpoints for Specification 2.2.1.

2.

Moderator Temperature Coefficient BOL and EOL limits and 300 ppm surveillance limit for Specification 3/4.1.1.3,

3.

Shutdown Bank Insertion Limit for Specification 3/4.1.3.5,

4.

Control Bank Insertion Limits for Specification 3/4.1.3.6,

5.

AXIAL FLUX DIFFERENCE Limits for Specification 3/4.2.1,

6.

Heat Flux Hot Channel Factor and K(z) for Specification 3/4.2.2, and

7.

Nuclear Enthalpy Rise Hot Channel Factor for Specification 3/4.2.3.

6.9.1.14.a The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC, specifically those described in the following documents:

The COLR will contain the complete identification for each of the TS referenced topical reports used to prepare the COLR (Le., report number, title, revision, date, and any supplements).

1.

BAW-10180P-A, Revision 1, "NEMO - Nodal Expansion Method Optimized," March 1993

2.

BAW-10169P-A, Revision 0, "RSG Plant Safety Analysis - B&W Safety Analysis Methodology for Recirculating Steam Generator Plants," October 1989

3.

BAW-10163P-A, Revision 0, "Core Operating Limit Methodology for Westinghouse Designed PWRs," June 1989

4.

EMF-2328{P)(A), "PWR Small Break LOCA Evaluation Model," March 2001

5.

BAW-10227P-A, Revision 1, "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel," June 2003

6.

BAW-10186P-A, Revision 2, "Extended Burnup Evaluation," June 2003

7.

EMF-2103P-A, Revision 0, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," April 2003 SEQUOYAH - UNIT 1 6-13 Amendment No. 52, 58, 72, 74, 117, 152,155,156.171,216,223, 281,300.314, 331

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (continued)

8.

BAW-1 0241 P-A, Revision 1, "BHTP DNB Correlation Applied with L YNXT," July 2005

9.

BAW-10199P-A, Revision 0, "The BWU Critical Heat Flux Correlations,* August 1996

10.

BAW-10189P-A, "CHF Testing and Analysis of the Mark-BW Fuel Assembly Design,"

January 1996

11.

BAW-10159P-A, "BWCMV Correlation of Critical Heat Flux in Mixing Vane Grid Fuel Assemblies," August 1990

12.

BAW-10231(P)(A), Revision 1, "COPERNIC Fuel Rod Design Computer Code,"

January 2004 6.9.1.14.b The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

6.9.1.14.c THE CORE OPERATING LIMITS REPORT shall be provided within 30 days after cycle start up (Mode 2) for each reload cycle or within 30 days of issuance of any midcycle revision of the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS (PTLR) REPORT 6.9.1.15 RCS pressure and temperature limits for heatup, cooldown. low temperature operation, criticality, and hydrostatic testing, L TOP arming, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

Specification 3.4.9.1, "RCS Pressure and Temperature (PIT) Limits" Specification 3.4.12, "Low Temperature Over Pressure Protection (L TOP) System" 6.9.1.15.a The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1.

Westinghouse Topical Report WCAP-14040-NP-A, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves."

2.

Westinghouse Topical Report WCAP-15293, "Sequoyah Unit 1 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation."

3.

Westinghouse Topical Report WCAP-15984, "Reactor Vessel Closure HeadNessel Flange Requirements Evaluation for Sequoyah Units 1 and 2."

6.9.1.15.b The PTLR shall be provided to the NRC within 30 days of issuance of any revision or supplement thereto.

SEQUOYAH - UNIT 1 6-13a Amendment No. 52, 58, 72, 74, 117,155,223,241,258,294, 297,306,314.320,331

ADMINISTRATIVE CONTROLS STEAM GENERATOR TUBE INSPECTION REPORT 6.9.1.16 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 6.8.4.k, Steam Generator (SG) Program. The report shall include:

a.

The scope of inspections performed on each SG,

b. Active degradation mechanisms found,

c.

Nondestructive examination techniques utilized for each degradation mechanism,

d.

Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e.

Number of tubes plugged during the inspection outage for each active degradation mechanism,

f.

Total number and percentage of tubes plugged to date,

g.

The results of condition monitoring, including the results of tube pulls and in-situ testing, and

h.

The effective plugging percentage for all plugging in each SG.

SPECIAL REPORTS 6.9.2.1 Special reports shall be submitted within the time period specified for each report, in accordance with 10 CFR 50.4.

6.9.2.2 This speCification has been deleted.

6.10 RECORD RETENTION (DELETED)

SEQUOYAH - UNIT 1 6-14 Amendment No. 42, 52, 58, 72, 74, 117, 148, 155,163,174,178,223 233,241,258,294,297,306,331

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-328 SEQUOYAH NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.324 License No. DPR-79

1.

The Nuclear Regulatory Commission (the Commission) has found that A.

The application for amendment by Tennessee Valley Authority (the licensee) dated June 17, 2011, and supplemented by letters dated July 27,2011!

November 14, 2011, March 23, April 26, May 15, May 24, and June 26,2012, 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 (ii) 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.

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. DPR-79 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 324, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance, to be implemented prior to startup from Unit 2 fall 2012 refueling outage.

FOR THE NUCLEAR REGULATORY COMMISSION

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(... l: _ l Jessie F. Quichocho, Acting Cnlef Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Attachment Changes to the Facility Operating License and Technical Specifications Date of Issuance: September 26, 2012

ATTACHMENT TO LICENSE AMENDMENT NO.324 FACILITY OPERATING LICENSE NO. DPR-79 DOCKET NO. 50-328 Replace Page 3 of Operating License DPR-79 with the attached page.

Replace the following pages of the Appendix A Technical Specifications with the attached pages.

The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

REMOVE INSERT 2-1 2-1 2-2 2-2 2-5 2-5 3/42-15 3/42-15 6-13 6-13 6-14 6-14 6-14a 6-14a

-3*

(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 chemica! 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 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. 324 are hereby incorporated into this license. The licensee shall operate the facUtty In accordance with the Technical Specifications.

(3)

Initial Test Program The Tennessee Valley Authority shall conduct the post@fuei-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;

c.

Performance of any test at power level different from there described; and Facility Operating License No. DPR*79 Amendment No.324

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETIINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (T avg) shall not exceed the limits shown in Figure 2.1-1 and the following SLs shall not be exceeded:

2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained ~ 1.132 for the BHTP correlation, ~ 1.21 for the BWU-N correlation, and ~ 1.21 for the BWCMV correlation.

2.1.1.2 The maximum local fuel pin centerline temperature shall be maintained ~ 4901°F, decreasing by 13.7°F per 10,000 MWD/MTU of burn up for COPERNIC applications, and ~ 4642°F, decreasing by 58°F per 10,000 MWD/MTU of burnup for TAC03 applications.

APPLICABILITY: MODES 1 and 2.

ACTION:

If SL 2.1.1 is violated, restore compliance and be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

APPLICABILITY: MODES 1, 2, 3, 4 and 5.

ACTION:

MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded 2735 pSig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

MODES 3, 4 and 5 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes.

SEQUOYAH - UNIT 2 2-1 Amendment No. 33, 324

660 620 600 500 ACCEPTABLE OPERATION 500+-~~---r------~--------+--------+--------+-----~-~

5~+-------~--------~------~--------+-------~------~

0.0 0.2 0.4 0.6 0.8 1.0 1.2 FRACTION OF RATED THERMAL POWER SEQUOYAH - UNIT 2 2-2 Amendment No. 21, 324 Figure 2.1-1 Reactor Core Safety Limit - Four Loops in Operation 600~------~--------~-------T--------~------~------~

UNACCEPTABLE OPERATION E

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TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT

1. Manual Reactor Trip

2. Power Range, Neutron Flux

3. Power Range, Neutron Flux, High Positive Rate

4. Power Range, Neutron Flux, High Negative Rate

5. Intermediate Range, Neutron Flux

6. Source Range, Neutron Flux

7. Overtemperature !:;.T

8. Overpower!:;. T

9. Pressurizer Pressure--Low

10. Pressurizer Pressure--High

11. Pressurizer Water Level-High

12. Loss of Flow NOMINAL TRIP SETPOINT Not Applicable Low Setpoint - 25% of RATED THERMAL POWER High Setpoint - 109% of RATED THERMAL POWER 5% of RATED THERMAL POWER with a time constant

~2 seconds 5% of RATED THERMAL POWER with a time constant

~2 seconds 25% of RATED THERMAL POWER 105 counts per second See Note 1 See Note 2 1970 psig 2385 psig 92% of instrument span 90% of design flow per loop" ALLOWABLE VALUES Not Applicable Low Setpoint - ~ 27.4% of RATED THERMAL POWER High Setpoint - ~ 111.4% of RATED THERMAL POWER 5, 6.3% of RATED THERMAL POWER with a time constant

~2 seconds 5, 6.3% of RATED THERMAL POWER with a time constant

~2 seconds 5, 45.20% of RATED THERMAL POWER 5, 1.45 x 105 counts per second See Note 3 See Note 4

~ 1964.8 psig

~ 2390.2 psig

~ 92.7% of instrument span

~ 89.6% of design flow per loop"

.. Design flow is 94,600 (91,400 x 1.035) gpm per loop.

SEQUOYAH - UNIT 2 2-5 Amendment No. 36, 132, 177,203,212, 214,299, 324

Figure 3.2-1 Flow VS. Power for 4 Loops in Operation 380000*~---------------------------------------------------,

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Unac(eptabie Operation Region Y6 100 SEQUOYAH - UNIT 2 3/42-15 Amendment No. 214,324

ADMIN ISTRATIVE CONTROLS MONTHLY REACTOR OPERATING REPQRT 6.9.1.10 DELETED CORE OPERATING LIMITS REPORT 6.9.1.14 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

1.

f1(.:1I) limits for Overtemperature Delta T Trip Setpoints and f2(.:1I) limits for Overpower Delta T Trip Setpoints for Specification 2.2.1.

2.

Moderator Temperature Coefficient BOL and EOL limits and 300 ppm surveillance limit for Specification 3/4.1.1.3,

3.

Shutdown Bank Insertion Limit for Specification 3/4.1.3.5,

4.

Control Bank Insertion Limits for Specification 3/4.1.3.6,

5.

AXIAL FLUX DIFFERENCE Limits for Specification 3/4.2.1,

6.

Heat Flux Hot Channel Factor and K(z) for Specification 3/4.2.2, and

7.

Nuclear Enthalpy Rise Hot Channel Factor for Specification 3/4.2.3.

6.9.1.14.a The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC, specifically those described in the following documents:

The COLR will contain the complete identification for each of the TS referenced topical reports used to prepare the COLR (I.e., report number, title, revision, date, and any supplements).

1.

BAW-10180P-A, Revision 1, "NEMO - Nodal Expansion Method Optimized," March 1993

2.

BAW-10169P-A, Revision 0, "RSG Plant Safety Analysis - B&W Safety Analysis Methodology for Recirculating Steam Generator Plants," October 1989

3.

BAW-10163P-A, Revision 0, "Core Operating Limit Methodology for Westinghouse Designed PWRs," June 1989

4.

EMF-2328(P)(A), "PWR Small Break LOCA Evaluation Model," March 2001

5.

BAW-10227P-A, Revision 1, "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel," June 2003

6.

BAW-10186P-A, Revision 2, "Extended Burnup Evaluation," June 2003

7.

EMF-2103P-A, Revision 0, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," April 2003 SEQUOYAH - UNIT 2 6-13 Amendment No. 44, 50, 64, 66, 107, 134,142,146,161,206,214,223, 272,289,303,324

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (continued)

8.

BAW-10241P-A, Revision 1, uBHTP DNB Correlation Applied with LYNXT," July 2005

9.

BAW-10199P-A, Revision 0, "The BWU Critical Heat Flux Correlations," August 1996

10.

BAW-10189P-A, "CHF Testing and Analysis of the Mark-BW Fuel Assembly Design,"

January 1996

11.

BAW-10159P-A, "BWCMV Correlation of Critical Heat Flux in Mixing Vane Grid Fuel Assemblies," August 1990

12.

BAW-10231 (P)(A), Revision 1, "COPERNIC Fuel Rod Design Computer Code," January 2004 6.9.1.14.b The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

6.9.1.14.c THE CORE OPERATING LIMITS REPORT shall be provided within 30 days after cycle start up (Mode 2) for each reload cycle or within 30 days of issuance of any midcycle revision of the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS (PTLR) REPORT 6.9.1.15 RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing, LTOP arming. and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

Specification 3.4.9.1. "RCS Pressure and Temperature (PIT) Limits" Specification 3.4.12, "Low Temperature Over Pressure Protection (L TOP) System" 6.9.1.15.a The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1.

Westinghouse Topical Report WCAP-14040-NP-A, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves."

2.

Westinghouse Topical Report WCAP-15321, "Sequoyah Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation."

3.

Westinghouse Topical Report WCAP-15984. "Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Sequoyah Units 1 and 2."

6.9.1.15.b The PTLR shall be provided to the NRC within 30 days of issuance of any revision or supplement thereto.

SEQUOYAH - UNIT 2 6-14 Amendment No. 44. 50, 64. 66.107, 134,146,206,214,231.

249,284,303,305.311,324

ADMINISTRATIVE CONTROLS STEAM GENERATOR (SG) TUBE INSPECTION REPORT 6.9.1.16.1 A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 6.8.4.k, "Steam Generator (SG) Program." The report shall include:

a. The scope of inspections performed on each SG,

b. Active degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism,

d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e. Number of tubes plugged during the inspection outage for each active degradation mechanism,

f.

Total number and percentage of tubes plugged to date,

g. The results of condition monitoring, including the results of tube pulls and in-situ testing, and

h.

The effective plugging percentage for all plugging in each SG.

SEQUOYAH - UNIT 2 6-14a Amendment No. 305, 323324

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 331 TO FACILITY OPERATING LICENSE NO. DPR-77 AND AMENDMENT NO. 324 TO FACILITY OPERATING LICENSE NO. DPR-79 TENNESSEE VALLEY AUTHORITY SEOUOYAH NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328

1.0 INTRODUCTION

By letter dated June 17, 2011 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML111720780), as supplemented by letters dated July 27,2011, November 14,2011, March 23, April 26, May 15, May 24, and June 26,2012 (ADAMS Accession Nos. ML112101798, ML113200023, ML120880317, ML121180215, ML12137A297, ML12153A377, and ML121850008, respectively), Tennessee Valley Authority (TVA or the licensee) requested a license amendment for Sequoyah (SON), Units 1 and 2, Technical Specifications (TSs), The proposed change would allow the use of AREVA Advanced W17 High Thermal Performance (HTP) fuel in the SON, Units 1 and 2 reactors.

The supplemental letters dated July 27,2011, November 14, 2011, March 23, April 26, May 15, May 24, and June 26, 2012, provided additional information that clarified the application, and did not change the Nuclear Regulatory Commission (NRC or the Commission) staffs original proposed no significant hazards consideration determination as published in the Federal Register on August 23, 2011 (76 FR 52703),

2.0 REGULATORY EVALUATION

The regulatory requirements and guidance documents on which the NRC staff bases its acceptance are:

• Title 10 of the Code of Federal Regulation (10 CFR), Part 50, Section 50.36, "Technical specifications. "

10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors."

- 2 10 CFR 50.59, "Changes, tests, and experiments."

10 CFR 50.67, "Accident source term."

10 CFR 50.92, "Issuance of amendment."

10 CFR Part 50 Appendix A, General Design Criterion 19 (GDC-19), "Control room."

10 CFR Part 100, "Reactor Site Criteria."

Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000.

RG 1.4, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors [PWRs]," Rev. 2, June 1974.

RG 1.24, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Pressurized Water Reactor Radioactive Gas Storage Tank Failure," March 1972.

RG 1.157, "Best-Estimate Calculations of Emergency Core Cooling System Performance,"

May 1989.

NUREG-0800, "Standard Review Plan," (SRP) Section 4.2, "Fuel System Design," Rev. 3, March 2007.

NUREG-0800, SRP Section 4.3, "Nuclear Design," Rev. 3, March 2007.

NUREG-0800, SRP Section 4.4, "Thermal and Hydraulic Design," Rev. 2, March 2007.

NUREG-0800, SRP Section 6.3, "Emergency Core Cooling Systems," Rev. 3.

March 2007.

NUREG-0800, SRP Section 6.4, "Control Room Habitability System," Rev. 3, March 2007.

NUREG-0800, SRP Section 15.0, "Introduction - Transient and Accident Analyses,"

Rev. 3, March 2007.

NUREG-0800, SRP, Section 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms," Rev. 0, July 2000.

NUREG-0800, SRP, Section 15.1.5, Appendix A, "Radiological Consequences of Main Steam Line Failures Outside Containment of a PWR," Rev. 2, July 1981.

NUREG-0800, SRP, Section 15.4.8, Appendix A, "Radiological Consequences of a Control Rod Ejection Accident (PWR)," Rev. 1, July 1981.

- 3 NUREG-0800, SRP, Section 15.6.3, "Radiological Consequences of Steam Generator Tube Failure," Rev. 2, July 1981.

NUREG-0800, SRP, Section 15.6.5, Appendix A, "Radiological Consequences of a Design Basis Loss-of-Coolant Accident Including Containment Leakage Contribution,"

Rev. 1, July 1981.

NUREG-0800, SRP, Section 15.6.5, Appendix B, "Radiological Consequences of a Design Basis Loss-of-Coolant Accident: Leakage From Engineered Safety Feature Components Outside Containment," Rev. 1, July 1981.

NUREG-0800, SRP, Section 15.7.3, "Postulated Radioactive Releases Due to Liquid-Containing Tank Failures," Rev. 2, July 1981.

NUREG-0800, SRP, Section 15.7.4, "Radiological Consequences of Fuel Handling Accidents," Rev. 1, July 1981.

For the fuel-handling accident (FHA), the NRC staff evaluated the licensee's proposed change against the requirements specified in 10 CFR 50.67(b)(2). Section 50.67(b)(2) of 10 CFR requires that the licensee's analysis demonstrates with reasonable assurance that:

• An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 0.25 Sv [sievert] (25 rem [roentgen equivalent man]) total effective dose equivalent (TEDE).

• An individual located at any pOint on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).

• Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) total effective dose equivalent (TEDE) for the duration of the accident.

The regulatory requirements from which the NRC staff based its acceptance for the FHA are the reference values in 10 CFR 50.67, and the accident specific guideline values in Regulatory Position 4.4 of RG 1.183 and Table 1 of SRP Section 15.0.1. RG 1.183 provides guidance to licensees on acceptable application of alternate source term (AST) submittals, including acceptable radiological analysis assumptions for use in conjunction with the accepted AST.

For all other design-basis accidents (DBAs) discussed in Section 3.1 below, the NRC staff evaluated the licensee's proposed change against the requirements specified in 10 CFR Part 100 and 10 CFR Part 50, Appendix A, GDC-19. It is required by 10 CFR Part 100, and 10 CFR Part 50, Appendix A, GDC-19 that the licensee's analysis demonstrates with reasonable assurance that:

-4

• An individual located at any point on the boundary of the exclusion area for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> immediately following onset of the postulated fission product release, would not receive a radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.

• An individual located at any pOint on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release during the entire period of its passage, would not receive a radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.

• Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident.

The regulatory requirements from which the NRC staff based its acceptance, for DBAs discussed in Section 3.1 below, other than the FHA, are the reference values in 10 CFR Part 100, GDC-19, and the accident specific guideline values in the SRPs and RGs.

Section 50.59 of 10 CFR Part 50 requires: (1) under 50.59(a)(1) a modification or addition to, or removal from, the facility or procedures that affects a design function, method of performing or controlling the function, or an evaluation that demonstrates that intended function will be accomplished; and (2) under 50.59(a)(2) departure from a method of evaluation described in the Updated Final Safety Analysis Report (UFSAR) unless the results of the analysis are conservative or essentially the same Section 50.92 of 10 CFR Part 50 requires involvement of the material alteration of a licensed facility.

Section 50.36 of 10 CFR Part 50: 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 Commission's 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 systems settings, and limiting control settings (10 CFR 50.36(c)(1>>; (2) Limiting conditions for operation (10 CFR 50.36(c)(2>>; (3) Surveillance requirements (10 CFR 50.36(c)(3>>; (4) Design features (10 CFR 50.34(c)(4)}; and (5) Administrative controls (10 CFR 50.36(c)(5>>.

In general, there are two classes of changes to TSs: (1) Changes needed to reflect modifications to the design basis (TSs are derived from the design basis), and (2) voluntary changes to take advantage of the evolution in policy and guidance as to the required content and preferred format of TSs over time. This license amendment deals with the first class of changes. In determining the acceptability of SON, Units 1 and 2 proposed safety limit and administrative controls TS changes, the NRC staff used the accumulation of generically approved guidance in NUREG-1431, Revision 3, "Standard Technical Specifications, Westinghouse Plants," dated June 2004.

- 5 The requirements pertaining to emergency core cooling system (ECCS) performance are contained in 10 CFR 50.46(a)(1)(i). Pertinent requirements contained in 10 CFR 50.46(a)(1)(i) include the following:

The ECCS performance must be calculated using an acceptable evaluation model. The analytical technique realistically describes the behavior of the reactor system performance during a loss-of-coolant accident (LOCA). A number of postulated LOCAs of different sizes, locations, and other properties must be analyzed, to provide assurance that the most severe postulated LOCAs have been calculated. Uncertainty in the analysis method and inputs must be identified and assessed. The uncertainty must be accounted for so that, when the results are compared to the criteria contained in 10 CFR 50.46(b), there is a high level of probability that the acceptance criteria would not be exceeded.

The acceptance criteria for the results of the ECCS include the following from 10 CFR 50.46, to which the AREVA realistic evaluation model is intended to show conformance:

(b)(1) The calculated maximum fuel element cladding temperature [peak cladding temperature, (PCT)] shall not exceed 2200 of [degrees Fahrenheit].

(b)(2) The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation. If cladding rupture is calculated to occur, the inside surfaces of the cladding shall be included in the oxidation, beginning at the calculated time of rupture.

(b)(3) The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.

The licensee proposes to incorporate some features that are additional to the generically approved, or accepted, evaluation model described in EMF-21 031. These include corrections to incorporate more realistic modeling of the fuel mechanical effects expected at the peak temperatures predicted for the SON application.

The NRC staff reviewed the additional model features in consideration of the guidance contained in NRC RG 1.157.

3.0 TECHNICAL EVALUATION

3.1 DBA Radiological Consequences On July 27, 2011, TVA submitted to the NRC a report prepared by its contractor, AREVA NP Inc.,

entitled "Sequoyah HTP Fuel Transition (NP)," ANP-2986(NP), Revision 3 (ADAMS Accession No. ML11210B532). This report presents technical information related to the transition from the use 1 The words "approved" and "accepted" may be used interchangeably. Accepted is used because it is consistent with the language in the regulation, referring to the use of an acceptable evaluation model. The word approved is also used because one way to identify an acceptable evaluation model is by its description in a licensing topical report that has received NRC review and approval.

- 6 of Mark-BW fuel to the Advanced W17 HTP fuel design. The NRC staff reviewed the AREVA report and docketed supplemental information.

This section addresses the impact of the proposed change on previously analyzed DBA radiological consequences. Two parameters in the DBA radiological consequences analyses potentially impacted by the change in fuel design are the DBA source term and the mass and energy released after a DBA. The impact of the fuel change on the DBA source term and mass and energy releases is discussed below in Sections 3.1.1 and 3.1.2. The NRC staff safety evaluation of the proposed change with regard to specific design basis analyses is provided from Sections 3.1.3 to 3.1.10.

3.1.1 DBA Source Term TVA stated that the Advanced W 17 HTP fuel design is similar both physically and neutronically, to the Mark-BW fuel pin design. Operation design characteristics (power and burnup) are unchanged or more restrictive than previous fuel cycles. Because several DBAs do not expect fuel cladding damage, the source term is set by TS limits that will not change. In addition, TVA stated that no new failure mechanisms are introduced by the use of the Advanced W17 fuel.

Therefore, the source terms used in the DBA analyses remain unaffected by implementation of the Advanced W17 HTP fuel.

3.1.2 Mass and Energy Releases TVA stated that the Advanced W17 HTP fuel is thermally similar and hydraulically compatible with the Mark-BW assemblies. Therefore, TVA stated that the mass and energy releases used in environmental consequences (DBA radiological consequences) analyses remain unaffected by the Advanced W17 HTP fuel.

3.1.3 Specific Design Basis Accidents This section provides the NRC staff's evaluation of the DBA radiological consequences analysis results reported in the amendment submittal. The NRC staff evaluated the analysis inputs and assumptions against analysis descriptions in the Sequoyah UFSAR, regulatory guidance, and NRC staff experience in performing similar reviews.

TVA considered the impact of the Advanced W17 HTP fuel design on the previously analyzed DBAs. The DBAs considered included:

Loss of Alternating Current Power Waste Gas Decay Tank Failure Loss-of-Coolant Accident Main Steam Line Break Steam Generator Tube Rupture (SGTR)

Fuel Handling Accident Rod Ejection Accident (REA)

- 7 3.1.4 Loss of Alternating Current (AC) Power In the event of a loss of offsite AC power, emergency diesel generators will start and power vital loads. The main condenser circulating water pumps are not vital loads and will be unavailable.

Without circulating water, the main condenser will not be available to receive and condense steam generated during the plant cool down. As such, this steam is released to the environment by steam safety valves and atmospheric dump valves. This steam may be contaminated due to leakage of reactor coolant into the steam generators (SGs) via small tube leaks (i.e.,

primary-to-secondary leakage). It is assumed that the cool down, and the release, will continue for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The release is based on the transfer of reactor coolant containing radionuclides (corresponding to the TS limits on primary coolant activity) at a rate equivalent to primary-to-secondary leak rate technical specification. The release also includes the radionuclides initially in the SGs at the event onset.

TVA stated and the NRC staff agrees that all the input parameters for this analysis are not affected by the change in fuel design. These parameters include: primary-to-secondary leakage rate, primary coolant activity, iodine activity in the secondary side liquid, the SG partition factor, and the steam release to cool the plant. The primary-to-secondary leakage rate, primary and secondary side liquid activities are all set by TS limits and, therefore, are not affected by the fuel design. The partition factor is also independent of fuel design. The NRC staff also finds that it is reasonable that the secondary side steam needed to cool the plant after the accident would not change since the total reactor power would not change. Since all the parameters in this analysis are not changed by the Advanced W17 HTP fuel, the results of the existing analysis remain valid.

Therefore, the NRC staff has reasonable assurance that the radiological consequences continue to meet the criterion of GDC-19 and are a small fraction (10 percent) of the 10 CFR Part 100 requirements.

3.1.5 Waste Gas Decay Tank Failure Waste gas decay tanks contain radioactive gases generated during plant operation until the gases have decayed sufficiently to allow release to the environment. This DBA postulates the failure of a waste gas decay tank that releases its contents. The content of the tank is released non-mechanistically to the environment.

The tank activity is based upon the assumption of 1 percent defective fuel and the reactor coolant system (RCS) volume. Both these parameters are not affected by a change in fuel design and, therefore, the tank activity will remain unchanged. Since all the parameters in this analysis are not changed by the Advanced W17 HTP fuel, the results of the existing analysis remain valid.

3.1.6 Loss-of-Coolant Accident A LOCA is a failure of the RCS that results in the loss of reactor coolant and could possibly result in reactor core damage. The containment building is designed to hold up the majority of the radioactivity released from the core. Evaluation of the effectiveness of plant safety features, such as ECCS, has shown that core melt is unlikely. The objective of this DBA is to evaluate the ability of the plant design to mitigate the release of radionuclides to the environment in the unlikely event that ECCS is not effective. Two release pathways are considered: (1) leakage of containment

- 8 atmosphere, and (2) leakage from systems that recirculate contaminated sump water outside of containment (e.g., certain ECCS).

TVA states and the NRC staff agrees that the parameters used in the LOCA analysis remain unaffected by the change in fuel design. The one key parameter in the LOCA analysis that could change as the result of new fuel would be the accident source term. TVA states that the burnup limit, the maximum fuel enrichment (5 percent Uranium 235) and the power do not change. In addition the fuel rod materials, fuel pellet materials, and fuel design are similar to the current Mark-BW fuel and the impact on this change on the neutron spectra and resulting fission product inventory is insignificant. Based on the information provided by TVA and NRC staffs experience in reviews for similar plants, the NRC staff accepts the TVA conclusion that the change in fission product inventory is insignificant. Since all the parameters in this analysis are not changed by the Advanced W17 HTP fuel, the results of the existing analysis remain valid. Therefore, the NRC staff has reasonable assurance that the radiological consequences continue to meet 10 CFR Part 100 requirements and the criterion of GDC-19 for this accident.

3.1.7 Main Steam Line Break This DBA postulates an unisolable failure in one of the four main steam lines at a location outside of containment, resulting in the release of steam from the affected steam line. Since a loss of offsite power is assumed to occur, the main condenser is not available as a heat sink and the unaffected SGs are used to cool down the plant by dumping steam to the environment. The released steam may be contaminated due to leakage of reactor coolant into the SGs via small tube leaks (Le., primary-to-secondary leakage).

TVA stated and the NRC staff agrees that all the input parameters for this analysis are not affected by the change in fuel design. These parameters include: primary-to-secondary leakage rate, primary coolant activity, iodine activity in the secondary side liquid, the SG partition factor, and the steam release to cool the plant. Since the primary-to-secondary leakage rate and the primary and secondary side liquid activities are all set by TS limits, they are not affected by the fuel design. The partition factor is also independent of fuel design. The NRC staff also finds that it is reasonable that the secondary side steam needed to cool the plant after the accident would not change since the total reactor power would not change. Since all the parameters in this analysis are not changed by the Advanced W17 HTP fuel, the results of the existing analysis remain valid.

Therefore, the NRC staff has reasonable assurance that the radiological consequences continue to meet the criterion of GDC-19 and are a small fraction (10 percent) of the 10 CFR Part 100 requirements.

3.1.8 Steam Generator Tube Rupture This DBA postulates a rupture in a tube in one of the SGs resulting in the transfer of reactor coolant water to the ruptured SG. Since a loss of offsite power is assumed to occur, the main condenser is not available as a heat sink and the unaffected SGs are used to cool down the plant by dumping steam to the environment. The released steam may be contaminated due to leakage of reactor coolant into the SGs.

TVA stated and the NRC staff agrees that all the input parameters for this analysis are not affected by the change in fuel design. These parameters include: primary-to-secondary leakage rate,

- 9 primary coolant activity, iodine activity in the secondary side liquid, the SG partition factor, and the steam release to cool the plant. Since the primary-to-secondary leakage rate and the primary and secondary side liquid activities are all set by TS limits, they are not affected by the fuel design. The partition factor is also independent of fuel design. The NRC staff also finds that it is reasonable that the secondary side steam needed to cool the plant after the accident would not change since the total reactor power would not change. Since all the parameters in this analysis are not changed by the Advanced W17 HTP fuel, the results of the existing analysis remain valid.

Therefore, the NRC staff has reasonable assurance that the radiological consequences continue to (1) meet the 10 CFR Part 100 requirements for the SGTR with an assumed pre-accident spike, (2) remain within 10 percent of 10 CFR Part 100 requirements for a SGTR with an equilibrium iodine concentration in combination with an assumed accident-generated spike, and (3) remain within the criterion of GDC-19 for this accident.

3.1.9 Fuel-Handling Accident This DBA postulates the drop of an irradiated fuel assembly during refueling operations. All of the fuel rods in the assembly, including the 24 Tritium Producing Burnable Absorber Rods, are assumed to rupture, releasing the radionuclides within the fuel clad gap to the fuel pool or reactor cavity water, Two cases are considered: (1) an FHA within the containment, and (2) an FHA in the auxiliary building spent fuel pool area, TVA stated that all the input parameters for this analysis are not affected by the change in fuel design, TVA stated the fuel rod and fuel pellet materials, and design are similar to the current Mark-BW fuel and the impact of this change on the neutron spectra and resulting fission product inventory is inSignificant. The fuel burnup limits, power, and number of fuel rods failed are also bounded by the current Mark-BW fuel. Based on the information provided by TVA and the NRC staffs experience in reviews for similar plants, the NRC staff accepts the TVA conclusion that the change in fission product inventory is insignificant.

The NRC staff agrees that all the input parameters for this analysis are not affected by the change in fuel design and, therefore, the NRC staff has reasonable assurance that the licensee's estimates of the TEDE will continue to comply with the requirements of 10 CFR 50,67 and the guidance of RG 1,183, 3,1.10 Rod Ejection Accident TVA stated that the radiological consequences of this event are bounded by the evaluation of the DBA LOCA The Sequoyah licensing basis (UFSAR) does not include a radiological consequence analysis specific to the REA The NRC staff, based on its experience in reviews at similar plants, accepts the TVA conclusion, 3,2 Fuel Mechanical DeSign The AREVA designed Advanced W17 HTP fuel features M5 cladding, zircaloy-4 MONOBLOC guide tubes, intermediate flow mixers, HTP intermediate spacers, debris resistance lower tie plate, and reconstitutable top nozzle, The M5 cladding was approved for fuel rod design (Reference 7.1),

- 10 The licensee will use the approved fuel performance code COPERNIC (Reference 7.2) to evaluate the thermal and mechanical analyses for the Advanced W17 HTP fuel design. The COPERNIC code is a contemporary code with new features including fuel thermal conductivity degradation (TCD) with burnup. The TCD issue was described in References 7.3 and 7.4. TCD is a physical phenomenon due to irradiation damage and the progressive buildup of fission products in fuel pellets resulting in reduced thermal conductivity of the pellets. The COPERNIC code was approved to a peak rod average burnup of 62 gigawatt-days per metric ton of uranium (GWD/MTU).

3.2.1 Fuel Design Methodologies The generic advanced 17x17 fuel design was described in the approved EMF-93-074(P)(A),

Revision 0, "Generic Mechanical Licensing Report for Advanced 17x17 Fuel Design." The Advanced W17 HTP fuel design was described in ANP-2986(P/NP), Revision 003, "Sequoyah HTP Fuel Transition (Attachments 1 and 2 of the licensee's supplemental letter dated July 27,2011; ADAMS Accession Nos. ML11210B533 and ML11210B532}." In responding to the NRC staff requests for additional information (RAls) (Enclosures 1 and 2 of the licensee's letter dated May 24,2012; ADAMS Accession NOs. ML12153A379 and ML12153A378), the licensee stated that the Advanced W17 HTP fuel design was evolved from the approved Advanced Mark-BW fuel design (Reference 7.5) using the approved generic fuel rod thermal-mechanical design methodology (Reference 7.6). The approved generic fuel rod thermal-mechanical design methodology allows AREVA to make changes to the existing Advanced 17x17 fuel adopting advanced features from the Advanced Mark-BW fuel without prior NRC staff review and approval.

The generic fuel rod thermal-mechanical design methodology has been used to acquire new fuel designs including the Advanced W17 HTP fuel design.

Based on the approved fuel design topical reports and methodologies, the NRC staff concludes that the fuel design methodologies used in acquiring the Advanced W17 HTP fuel design are acceptable for SON, Units 1 and 2.

3.2.2 Cladding Strain The design criterion for cladding strain is that the total cladding strain shall not exceed 1-percent strain during normal operation and transients. The licensee will adhere to this criterion using the approved COPERNIC code by demonstrating that cycle specific calculations of linear heat generation rate (LHGR) limits will meet the 1-percent strain limit with adequate margin. In addition, AREVA performed cladding strain analysis using COPERNIC to confirm the 1-percent strain limit for uranium dioxide (U02) and U02-Gd20 3 [gadolinium oxide or gadolinia] fuel pellets in the Advanced W17 HTP fuel design.

Based on the approved COPERNIC code, the NRC staff concludes that the Advanced W17 HTP fuel design criterion meets the cladding strain limit for SON, Units 1 and 2.

3.2.3 Fuel Melting The design criterion for fuel centerline melt is that the fuel pellet centerline temperature shall not exceed the melting temperature during normal operation and transients. The licensee will adhere to this criterion using the approved COPERNIC code by demonstrating that cycle specific

calculations of LHGR limits will not result in fuel centerline melting with adequate margin. In addition, AREVA performed fuel pellet centerline temperature analysis using COPERNIC to confirm no fuel centerline melting for U02 and U02-Gd 20 3 fuel pellets in the Advanced W17 HTP fuel design.

Based on the approved COPERNIC code, the NRC staff concludes that the Advanced W17 HTP fuel design criterion meets the fuel centerline melt criterion for SON, Units 1 and 2.

3.2.4 Strain Fatigue The design criterion for cladding strain fatigue is that the maximum fuel rod fatigue usage factor shall not exceed (( )) with a minimum safety factor of 2 on the stress amplitude or a minimum safety factor of 20 on the number of cycles. The design criterion is based on the O'Donnell and Langer curve for the use of fatigue usage factor.

In responding to the NRC staff RAls (Enclosures 1 and 2 of the licensee's letter dated May 24, 2012), the licensee stated that the approved COPERNIC code will be used for the strain fatigue analysis. In addition, previous analyses of this criterion had shown that the cladding strain fatigue was not limiting, and significant margins were existed for typical operation conditions.

Based on the approved COPERNIC code, the NRC staff concludes that the Advanced W17 HTP fuel design criterion meets the cladding strain fatigue criterion for SON, Units 1 and 2.

3.2.5 Gap Inventory In response to an RAI regarding the applicability of the RG 1.183, Table 3 non-LOCA source terms to the Sequoyah AST (Enclosures 1 and 2 of the licensee's letter dated November 14, 2011), the licensee stated the following:

If fuel rod burn ups were to exceed 54 GWD/MTU and any pins exceed the LHGR of 6.3 kilowatt per foot (kW/ft), then the gap release fraction for the non-LOCA events would be conservatively doubled or evaluated using the [American National Standards Institute/American Nuclear Society] ANSI/ANS-5.4 methodology based on the maximum burnup. Alternatively, the core design would be modified to ensure that the LHGR criterion is met.

This response prompted concerns regarding the details of the analytical techniques used to develop alternate gap release fractions using the ANSI/ANS-5.4 methodology. In response to a subsequent RAI requesting further information (Enclosures 1 and 2 of the licensee's letter dated May 24, 2012), the licensee detailed two analytical paths for addressing fuel rods that exceed the footnote 11 applicability limit. The first approach involved multiplicative adjustment factors to the gap fractions. These adjustments factors are provided in Table 7 of ANP-3053(P). As indicated in the licensee's response, this type of approach has been accepted by the NRC staff in previous AST license amendment requests (LARs) including Calvert Cliffs and Fort Calhoun. In these previous LARs, the licensees for Calvert Cliffs and Fort Calhoun calculated gap fractions using the ANSIIANS-5.4-1982 methodology and showed that doubling the gap release fractions in RG 1.183 Table 3 was bounding. A brief search of ADAMS also revealed NRC staff acceptance of similar gap fraction adjustments for Byron Station Units 1 and 2, Braidwood Station Units 1 and 2, and Palisades Nuclear Plant.

- 12 The NRC has endorsed the ANSI/ANS-S.4-2011 standard for these applications. The ANSIIANS-S.4-2011 standard predicts significantly lower iodine release fractions relative to the ANSI/ANS-S.4-1982 standard. As a result of the lower iodine release fractions, application of the 2011 standard would promote lower calculated doses for a given fuel rod design and power history.

With respect to previous applications, increasing the 1-131 gap fraction from 8 percent to 16 percent for fuel rods that exceed the applicability footnote is more conservative than previously recognized. Based upon the NRC staffs prior acceptance of the doubling factor method and the inherent margin which exists relative to the ANSI/ANS-S.4-2011 standard, the NRC staff finds the adjusted gap fractions in Table 7 of RAI No. 12 response acceptable.

Part (b) of RAI No. 12 response details an alternate approach for calculating revised gap fractions.

During subsequent discussions, the licensee stated that this alternate approach was not necessary based upon fuel management studies. As a result, the NRC staff did not review this detailed analytical procedure. Part (b) of RAI No. 12 response is not acceptable due to this lack of review.

3.3 TS Changes and Large Break Loss-of-Coolant Accident Evaluation 3.3.1 SL 2.1.1, Reactor Core Safety Limit The proposed TS changes include the departure from nucleate boiling ratio (DNBR) limits in SL 2.1.1.1 and the limitations for local fuel pin centerline temperature in SL 2.1.1.2.

The DNBR limits will be reflected in SL 2.1.1.1 as follows.

SL 2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained

1.132 for BHTP correlation,

;:: 1.21 for the BWU-N correlation, and:;:: 1.21 for the BWCMV correlation.

Limitation for local fuel pin centerline temperature will be added as follows.

SL 2.1.1.2 The maximum local fuel pin centerline temperature shall be maintained s 4901 of, decreasing by 13.7 of per 10,000 MWD/MTU [megawatt days per metric ton of uranium] of burnup for COPERNIC applications, and s 4642 of, decreasing by S8 OF per 10,000 MWD/MTU of burnup for TAC03 applications.

The NRC staff reviewed the proposed TS changes, and found them acceptable, because:

(1) approved methodologies (I.e., TAC03 for Zircaloy-4 cladding and COPERNIC for MS advanced alloy cladding) are used for the fuel transition applications; (2) the restrictions of the SL prevent overheating of the fuel and possible cladding perforation that would result in the release of fission products to the reactor coolant; (3) the automatic reactor protection system trip function ensures that the local linear heat rate will not exceed the centerline fuel melt limits; and (4) the change to the SL will add limits for the DNBR for each of the resident fuel types during fuel transition (I.e., Advanced W17 HTP fuel and Mark-BW fuel).

- 13 3.3.2 Figure 2.1-1, Reactor Core Safety Limit-Four Loops in Operation The proposed change to Figure 2.1-1 is to reflect the correct statepoints for the departure from nucleate boiling (DNB) safety limit maximum allowable peaking limits based on the protected Over Temperature Delta T (OT!:iT) limit lines, instead of the OT!:iT trip setpoint lines. The NRC staff has reviewed the proposal and found it acceptable because the proposed change due to the transition to the Advanced W17 HTP design and the implementation of the BHTP DNB correlation is to modify the existing core safety limit lines to be consistent in statepoint basis with the protected OT!:iT limit lines for each specified system pressure and to re-establish compliance with the approved methodology for SON, Units 1 and 2 safety analysis applications.

3.3.3 Table 2.2-1, Reactor Trip System Instrumentation Trip Setpoints The proposed increase in design flow from 90,045 to 94,600 gallons per minute (gpm) per loop in item 12 of Table 2.2-1, "Reactor Trip System Instrumentation Trip Setpoints," is acceptable because the thermal hydraulic analysis indicates that the transition from a full core of Mark-BW fuel to a full core of Advanced W17 HTP fuel will result in a small increase in bypass flow and a small decrease in the RCS loop flow due to the higher pressure drop of the Advanced W17 HTP fuel, the new replacement of SGs with minimal tube plugging, and favorable historical measured flow (the licensee's letter dated June 26,2012).

3.3.4 Figure 3.2-1, Flow versus Power for 4 Loops in Operation The proposed changes to the total RCS flow rate: from 360,100 gpm to 378,400 gpm at 100 percent Thermal Power Fraction (TPF); and from 342,095 gpm to 359,480 gpm at 90 percent TPF.

The NRC staff reviewed the proposed changes and found them acceptable because a small net increase in RCS loop flow due to the effect of the fuel transition and the SG replacement was confirmed by the measurement (the licensee's letter dated June 26,2012).

3.3.5 TS 6.9.1.14.a, Core Operating Limits Report The proposed revision to TS 6.9.1.14.a includes: (1) deletion of TS 6.9.1.14.a.5 and TS 6.9.1.14.a.6; (2) updating information for TS 6.9.1.14.a.1 through TS 6.9.1.14.a.3; (3) replacing TS 6.9.1.14.a.4 by EMF-2328(P)(A), "PWR Small Break LOCA Evaluation Model,"

March 2001; (4) renumbering the TS; and (5) addition of new TS 6.9.14.a.8 through TS 6.9.14.a.12.

The NRC staff reviewed the proposed revision to TS 6.9.1.14.a relating to approved methodologies used to support cycle-specific parameters listed in TS 6.9.1.14 and found them acceptable because: (1) all the methodologies listed in TS 6.9.1.14.a are reviewed and approved by NRC; (2) deletion of TS 6.9.1.14.a.5 and TS 6.9.1.14.a.6, and updating revision number and approved date for TS 6.9.1.14.a.1 through TS 6.9.1.14.a.3 are administrative in nature; and (3) addition of new TS 6.9.1.14.a.8 through TS 6.9.1.14.a.12 is to reflect methodologies applied to the proposed fuel transition.

- 14 3.3.6 Large Break LOCA Analysis Evaluation The license amendment request included a re-evaluation of ECCS performance that explicitly incorporated the features of the HTP fuel. Since the licensee had previously received NRC approval to implement EMF-2103(P)(A), "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," the NRC staff's review in this case was limited to issues with EMF-2103(P)(A) that have been identified by the NRC staff and addressed through the plant-specific RAI process.

Because the licensee has already received NRC staff approval to implement EMF-2103(P)(A), the NRC staff's review was limited to the following areas:

• Analytic treatment of clad swelling and rupture and fuel relocation for high-PCT applications.

Correction of the Sieicher-Rouse heat transfer correlation coding error.

Expansion of the run set from 59 to 93 cases.

For this LAR, the licensee omitted a clad ballooning and rupture model, citing qualitative considerations of the heating and cooling effects of cladding rupture. This modeling approach is discussed in response to recent NRC staff RAI No.9, beginning on page 6-24 of ANP-2970(NP)

(ADAMS Accession No. ML11172A064). The NRC staff position is that, when rupture is evaluated in consideration of the limited, but widely ranging data concerning the amount of fuel relocation possible, the heating effects can outweigh the cooling effects when fuel relocation is considered using a bounding assessment. Therefore, it is necessary to consider high fuel relocation packing fractions, either in an explicit uncertainty treatment, or in a bounding sense. If fuel relocation to an 80-percent packing fraction is considered, the cladding surface will heat more if the fuel clad ruptures.

In response to this concern, TVA re-analyzed its case set using the same statistical seed as for the cases presented in ANP-2970(NP). The licensee stated in ANP-2970Q1 (NP) that it is widely recognized that there would be some cooling as a result of steam de-superheating via droplet shattering against the intruding rupture and a related heat transfer enhancement; however, the licensee chose to conservatively neglect these cooling phenomena (ADAMS Accession No. ML12118A166). The results of the licensee's study indicated that the predicted limiting PCT would increase by 44 of to 1985 of.

Because the licensee modeled fuel cladding swelling and rupture using a model that conservatively assumes a bounding 80-percent value of post-rupture fuel relocation, the NRC staff finds the licensee's revised treatment of fuel cladding swelling and rupture acceptable. This approach is consistent with Regulatory Position 1 of RG 1.157, which states in part that "a best-estimate model should provide a realistic calculation of the important parameters associated with a particular phenomenon to the degree practical with the currently available data and knowledge of the phenomenon." Although the RG encourages the use of the mean of the data, Regulatory Position 4.3.2 also suggests that uncertainties in fuel behavior, which may not be incorporated into the code results, should be included in the determination of the overall calculational uncertainty. The NRC staff finds that the modeling bias to include a maximum fuel relocation fraction for ruptured fuel provides a reasonable bound for this phenomenon and ensures that the uncertainty estimate (Le., the 95/95 upper tolerance limit result), incorporates the effects of uncertainties associated with the fuel relocation phenomenon.

- 15 The licensee provided information concerning a correction for an error that was identified in the implementation of the Sieicher-Rouse heat transfer correlation. The NRC staff notes that TVA's correction is similar in nature and magnitude of PCT effect to recently reported error corrections from other licensees using the AREVA realistic large break LOCA analysis method. However, because evaluation model error reporting does not require NRC staff review and approval, the NRC staff merely acknowledges that the SON large break LOCA analysis includes this correction. The correction reduces the upper tolerance limit PCT value by 35 of.

In consideration of the swelling rupture and relocation study and the Sieicher-Rouse error correction, the updated licensing basis PCT value is 1950 of. Because this value remains less than the 10 CFR 50.46(b)(1) regulatory acceptance criterion of 2200 of, the NRC staff finds the updated result acceptable.

Figure 3-19 of ANP-2970(NP) indicated a slight reduction in down comer liquid level from 700-800 seconds. At the NRC staffs request, the licensee provided figures in ANP-297001 (NP),

which showed that, for a period following 800 seconds, the downcomer liquid level remained stable and increasing. Because the plots confirmed that the plant was trending to a stable post-LOCA recovery with respect to vessel liquid mass, the NRC staff finds the supplemental information acceptable.

The run set included the analysis of 93 cases as opposed to a more conventional 59 cases, as set forth in Section 5.2.1 of EMF-21 03(P)(A) (non-proprietary report ADAMS Accession No. ML032691424). Although the generic topical report states that this practice may be employed, it is not addressed in the NRC staffs safety evaluation approving EMF-2103(P)(A). Therefore, the NRC staff requested additional information concerning the implementation of this sampling practice.

The NRC staff requested that the licensee indicate whether the decision to execute additional cases prior to or after completing an analysis using the standard number of cases. In response the licensee stated that ((

))

The licensee also stated that the decision was made because there was a reasonable expectation concerning the results based on prior analyses completed for the same plant. and that the eliminated case was significantly different from the expected results, while the remaining cases aligned more closely with expectations. To justify this assertion, the licensee also provided tables of run matrix data. ((

- 16

)) The information provided by the licensee showed that the randomly sampled parameters combined to produce an unusually pessimistic result, which confirmed that the case was likely an outlier, and its elimination was justified. Based on the information provided by the licensee, the NRC staff accepts the licensee's justification.

The supplemental information provided by the licensee also showed that, ((

))

Because of the small difference between the maximum predicted PCT of the sample set and the upper tolerance limit, the NRC staff also finds that the alternative sampling practice does not jeopardize the requisite, per 10 CFR 50.46(a)(1 )(i), high level of probability that the regulatory limit of 2200 of is not exceeded. Based on this consideration, the NRC staff finds the licensee's alternative sampling practice acceptable. Because the NRC staffs basis for approving this approach pertains to Sequoyah-specific results, the NRC staff also finds that broadening this practice in additional applications would warrant additional NRC staff review.

3.4 Summary The NRC staff reviewed the analysis used by the licensee to assess the radiological impacts of the transition to AREVA Advanced W17 HTP fuel at SON, Units 1 and 2. The NRC staff finds that the licensee used methods consistent with regulatory requirements and guidance identified in Section 2.0 above. The NRC staff finds, with reasonable assurance that the licensee's estimates of the exclusion area boundary, low-population zone, and control room doses will continue to comply with these criteria. Therefore, the proposed change is acceptable with regard to the radiological consequences of postulated DBAs.

The NRC staff has reviewed the TVA submittal of the Advanced W17 HTP fuel design as described in ANP-2986(NP), Revision 3. Based on the evaluation, the NRC staff concludes that the Advanced W17 HTP fuel mechanical design is acceptable for SON, Units 1 and 2. The Advanced W17 HTP fuel design is approved to the peak rod average burnup of 62 GWD/MTU.

The NRC staff finds the licensee's re-evaluation of ECCS performance acceptable. The licensee has used an acceptable evaluation model that had been previously approved for implementation at SON. The model incorporates the effects of the high thermal performance fuel, and accounts for issues recently identified with the EMF-2103(P)(A) calculational framework, that have been addressed through the plant-specific review process. Based on the NRC staffs review, the NRC staff concludes that the results of the evaluation demonstrate that there is a high level of probability that the 10 CFR 50.46(b) acceptance criteria would not be exceeded for SON, when using AREVA High Thermal Performance Fuel.

4.0 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.

- 17

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to installation or use of a facility component 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 that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (76 FR 52703, August 23, 2011). Accordingly, the amendments meet 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 amendments.

6.0 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) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.

7.0 REFERENCES

7.1 BAW-10227P-A, "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel," February 2000 (ADAMS Accession No. ML003686134).

7.2 BAW-10231PA, "COPERNIC Fuel Rod Design Computer Code," April 2002.

7.3 Information Notice 2009-23, "Nuclear Fuel Thermal Conductivity Degradation,"

October 8, 2009 (ADAMS Accession No. ML091550527).

7.4 Information Notice 2011-21, "Realistic Emergency Core Cooling System Evaluation Model Effects Resulting from Nuclear Fuel Thermal Conductivity Degradation,"

December 13, 2011 (ADAMS Accession No. ML113430785).

7.5 BAW-10239P-A, Revision 0, "Advanced Mark-BW Fuel Assembly Mechanical Design Topical Report," July 2004.

7.6 EMF-92-116(P)(A), "Generic Mechanical Design Criteria for PWR Fuel Designs,"

February 1999.

Principal Contributors: M. Blumberg T.Huang S. Wu L. Ward B. Parks P. Clifford Dated: September 26. 2012

- 2 Copies of the proprietary and non-proprietary versions of the SE are enclosed. A notice of issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, Ira!

Siva P. Lingam, Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-327 and 50-328

Enclosures:

1. Amendment No. 331 to DPR-77

2. Amendment No. 324 to DPR-79

3. Non-Proprietary Safety Evaluation

4. Proprietary Safety Evaluation cc wlo encl 4: Distribution via Listserv Distribution:

Public RidsNrrDorlDpr RidsNrrLABClayton RidsAcrsAcnw _MaiICTR LPL2-2 RlF RidsNrrDorlLpl2-2 RidsOgcRp RidsRgn2MailCenter RidsNrrDraAadb RidsNrrDssSnpb RidsNrrDssSrxb RidsNrrDssStsb S. Wu, NRR

1. Huang, NRR M. Blumberg, NRR RidsNrrPMSequoyah L. Ward, NRR B. Parks, NRR C. Schulten, NRR P. Clifford, NRR ADAMS ACCESSION NOS.:

Package: ML12249A388, Letter and Non-Proprietary SE: ML12249A394, Proprietary SE: ML12249A415

• by memo NRR-058 OFFICE LPL2-2/PM LPL2-2/LA MDB/BC SRXB/BC SNPB/BC STSB/BC OGC LPLlI-2/BC (A)

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RElliott w/comments LSubin "NLO" w/comments JQuichocho DATE 9/20/12 9/20/12 4/10/12 8/24/12 8/10/12 9/14/12 9/19/12 9/26/12 OFFICIAL RECORD COpy