ML15344A321

Issuance of Amendment Regarding Modification of Technical Specification 3.4.9, RCS Pressure and Temperature (P/T) Limits

ML15344A321
Person / Time
Site:	Browns Ferry
Issue date:	01/07/2016
From:	Farideh Saba Plant Licensing Branch II
To:	James Shea Tennessee Valley Authority
Saba F
References
CAC MF5659
Download: ML15344A321 (23)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 7, 2016 Mr. Joseph W. Shea Vice President, Nuclear Licensing Tennessee Valley Authority 1101 Market Street, LP 3R-C Chattanooga, TN 37402-2801

SUBJECT:

BROWNS FERRY NUCLEAR PLANT, UNIT 3 - ISSUANCE OF AMENDMENT REGARDING MODIFICATION OF TECHNICAL SPECIFICATION 3.4.9, "RCS PRESSURE AND TEMPERATURE (PIT) LIMITS" (CAC NO. MF5659)

Dear Mr. Shea:

The U.S. Nuclear Regulatory Commission (NRC or the Commission) has issued the enclosed Amendment No. 278 to Renewed Facility Operating License No. DPR-68, for the Browns Ferry Nuclear Plant (BFN), Unit 3. This amendment is in response to Tennessee Valley Authority's (TVA, or the licensee) application dated January 27, 2015, as supplemented by letters dated August 13 and October 23, 2015.

The amendment revises BFN, Unit 3, Technical Specification (TS) 3.4.9, "RCS [reactor coolant system] Pressure and Temperature (PIT) Limits, Figures 3.4.9-1 and 3.4.9-2 that currently provide the PIT limits for up to 20 Effective Full Power Years (EFPYs), and greater than 20 EFPYs to less than or equal to 28 EFPYs, respectively. The proposed PIT limits for Figures 3.4.9-1 and 3.4.9-2 are applicable to 38 EFPYs, and greater than 38 EFPYs to less than or equal to 54 EFPYs, respectively. The amendment also revises Note 1 of TS Surveillance Requirement 3.4.9.1 to change the vessel pressure from less than 312 pounds per square inch gauge (psig) to less than 313 psig to conform to the modified PIT limit curves. The revision satisfies the requirements of NUREG-1843, "Safety Evaluation Report Related to the License Renewal of the Browns Ferry Nuclear Plant, Units 1, 2, and 3," dated April 2006, Commitment 39, that required the development and submittal of revised PIT limit curves for NRC approval prior to the period of extended operation.

The NRC staff has completed its review of the information provided by the licensee. The licensee's submittal contained proprietary information withheld from the public pursuant to Title 10 of the Code of Federal Regulations (1 O CFR), Section 2.390, "Public inspections, exemptions, requests for withholding." However, the enclosed NRC safety evaluation (SE) does not contain any proprietary information withheld under 10 CFR 2.390. The NRC will delay placing the enclosed (SE) in the public document room for a period of 10 working days from the date of this letter to provide TVA the opportunity to comment on any proprietary aspects of the SE. If you believe that Enclosure 2 contains proprietary information. please identify such information line by-line and define the basis for withholding pursuant to the criteria of 10 CFR 2.390. After 10 working days, the enclosed SE wiii be made publicly available, if the NRC is not notified of any existing proprietary information.

J. Shea A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely,

~ dJOvJi r Farideh E. Saba, Senior Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-296

Enclosures:

1. Amendment No. 278 to DPR-68

2. Safety Evaluation cc w/enclosures: Addressee cc w/enclosures 1O working days after issuance: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-296 BROWNS FERRY NUCLEAR PLANT. UNIT 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 278 Renewed License No. DPR-68

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

A. The application for amendment by Tennessee Valley Authority (the licensee) dated January 27, 2015, as supplemented by letters dated August 13, 2015, and October 23, 2015, 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 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 Renewed Operating License and Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-68 is hereby amended as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 278are hereby incorporated in the renewed operating 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 and shall be implemented within 60 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Benjamin G. Beasley, Chief 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: January 7, 201 6

ATIACHMENT TO LICENSE AMENDMENT NO. 278 RENEWED FACILITY OPERATING LICENSE NO. DPR-68 DOCKET NO. 50-296 Replace the following page of Renewed Facility Operating License No. DPR-68 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 3 Replace the following pages of Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

REMOVE INSERT 3.4-26 3.4-26 3.4-29 3.4-29 3.4-29a 3.4-29a 3.4-29b 3.4-29b 3.4-29c 3.4-29c

(3) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use at any time any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form for sample analysis or equipment and instrument calibration or associated with radioactive apparatus or components; (5) Pursuant to the Act and 10 CFR Parts 30 and 70, to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I:

Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 3458 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 278, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

For Surveillance Requirements (SRs) that are new in Amendment 212 to Facility Operating License DPR-68, the first performance is due at the end of the first surveillance interval that begins at implementation of the Amendment 212. For SRs that existed prior to Amendment 212, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first performance is due at the end of the first surveillance interval that begins on the date the surveillance was last performed prior to implementation of Amendment 212.

BFN-UNIT 3 Renewed License No. DPR-68 Amendment No. 278

RCS PIT Limits 3.4.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.9.1 --------------------------N()TES-----------------------

1. Only required to be performed during RCS heatup and cooldown operations or RCS inservice leak and hydrostatic testing when the vessel pressure is> 313 psig.

2. The limits of Figure 3.4.9-2 may be applied during nonnuclear heatup and ambient loss cooldown associated with inservice leak and hydrostatic testing provided that the heatup and cooldown rates are::::: 15°F/hour.

3. The limits of Figures 3.4.9-1 and 3.4.9-2 do not apply when the tension from the reactor head flange bolting studs is removed.

Verify: 30 minutes

a. RCS pressure and RCS temperature are within the limits specified by Curves No. 1 and No. 2 of Figures 3.4.9-1 and 3.4.9-2; and

b. RCS heatup and cooldown rates are

100°F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period.

SR 3.4.9.2 Verify RCS pressure and RCS temperature Once within 15 are within the criticality limits specified in minutes prior to Figure 3.4.9-1, Curve No. 3. control rod withdrawal for the purpose of achieving criticality (continued)

BFN-UNIT 3 3.4-26 Amendment No.~. 247,278

RCS Pff Limits 3.4.9 1400 I Curve No. 1 BRO\'\'NS FERRY UNIT 3 Ii Minimum temperature for bottom head 1300 CURVES I, 2, A~D 3 during mechanical ARE VALID FOR 38 EFPY If heatup or cooldown OF OPERATION ": 1 1200 2 3 I following nuclear shutdown.

I Ii Ii .'

1100 . I I Curve No. 2 Minimum temperature

II for upper RPV and

.21 ' beltline during Ill tl.

1000 '"

+' ,..

• • • • • • )

I mechanical heatup or cooldown following 0

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w .' I nuclear shutdown .

I:

a..

900 I

Curve No. 3

,J 0

I-

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(f)

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w ' I I/' I

> I I a:: 700 Notes 0

I-0 I These curves include t:i 600

~ I sufficient margin to a:: provide protection

~

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/ I I against feedwater nozzle degradation.

~ 500 ~/I 490 psig I II 490 psig I The curves allow for

...J I I shifts in RTNm of the w

a::

I 450 psig

/

II Reactor vessel

(/) 400 beltline materials, (f) w in accordance with er: 313'p~~] Reg. Guide 1.99, Rev.

a..

- F~

2, to compensate for I/

300 ,,,.,,,,,,,

'f '"'"'i'"'""

radiation embrittlement for 38 200 i/: ), EFPY.

BOTTOM HEAD

/1 '

iI I /I / I I The acceptable area for operation is to 100 r-- 68.F v v _..,.JBOLTUPl r ~I the right of the 0

1

• i ~

v* .

I 83.F I applicable curves.

0 25 50 75 WO 125 150 175 200 225 250 275 300 MINIMUM R,EACTOR VESSEL METAL TEMPERATURE (°F)

Figure 3.4.9-1 Pressure/Temperature Limits for Mechanical Heatup, Cooldown following Shutdown, and Reactor Critical Operations BFN-UNIT 3 3.4-29 Amendment No. 217, 233, 247 278

RCS PIT Limits 3.4.9 1400 Curve No. 1 BRO\VNS FERRY t:~l!T 3 Minimum temperature for 1300 Ct:RVES l AND 2 ARE bottom head during I*

in-service leak or VALID FOR 38 EFPY OF OPERATION f' 1 hydrostatic testing.

1200 . . :i Curve No. 2 Minimum temperature for

'i i! upper RPV and beltline 1100 f during in-service leak or hydrostatic testing .

.!?

~ 1000 .

• ---**+****~***-->--- ----~---..-1-------**--4---,~*----*--~-**-**--**

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w Notes J: 900 -********-**********-*-....,.':_! These curves include ll.

sufficient margin to

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0 beltline materials, in u accordance with Reg.

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600 D: Guide 1.99, Rev. 2, to compensate for radiation l yi .)

z embrittlement for 38

.... EFPY.

SE 500 ' ' '

....J w I ! The acceptable area for a:: '

operation is to the right

~ 400 BOTTOM _j __._______+*****-*-'"-*--*--*--"*****-----------+*--**-****--*-*-.. of the applicable curves.

(/)

HEAD  !

w 68°F '

a::

ll.

300 200 FLANGE f------+---+----1 REGla'll 83°F 100 0

0 25 50 75 100 125 150 175 200 2.25 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F)

Figure 3.4.9-2 Pressure/Temperature Limits for Reactor In-Service Leak and Hydrostatic Testing BFN-UNIT 3 3.4-29a Amendment No. ~ 2 78

RCS P!T Limits 3.4.9 1400

...*..: I

' I BRO\\"N'S FERRY UNIT 3 Curve No. 1 Minimum temperature CURVES I. ) AND.~ ~.

I 2 3 1300 ARE VAUD FOR 54 for bottom head during I

EFPY 01: OPERATION mechanical heatup or cooldown following 1200 .............. **************** .......... nuclear shutdown.

---~**~***

I Curve No. 2 1100 ,, """"~ "l'" . l Minimum temperature for upper RPV and

' I I

beltline during Cl , i

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.e, 1000 ,.'

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I mechanical heatup or cooldown following a '

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x: 900

Curve No. 3

a. '

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...J w 800 * (criticality).

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(/')

(/')

, Notes

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ct: 700 ~*-*-***** --*-:--*--****-** *---**-****- .... ...... '***** ---. ****-*- These curves include 0 sufficient margin to I-(.)

w 600 J j i

provide protection against feedwater I ',.'

ct: ' nozzle degradation.

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I z The curves allow for

~

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...J w I~ * - -* *-* *1 f.rI I f I i Reactor vessel beltline materials, in ct:

50 psig 420 ps1g

-t*+-.. -.. *------t-*- - I 420 psig I accordance with Reg .

~-

(/') 400 ............. ............. c- ... --**- I

• • ~*-*w***-**-**

Guide 1.99, Rev. 2, to

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300 .,'

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,,, : .* j  ! The acceptable area 200 BOTTOM

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100 ""-'-"

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...' v14---t:isl~~~~f>]

' I I I

j 0 25 50 75 100 125 150 175 200 225 250 275 300 MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F)

Figure 3.4.9-1 Pressure/Temperature Limits for Mechanical Heatup, Cooldown following Shutdown, and Reactor Critical Operations BFN-UNIT 3 3.4-29b Amendment No. ~' 278

RCS PfT Limits 3.4.9 1400 I

..'. I l

I Curve No. 1 I

Minimum 1300 BROWNS FERRY UNIT 3 ** temperature CURVES l AND 2 ARf ' for bottom head during 1200 ~

VAi.ID FOR 54 EFPY

.l 2J in-service leak or hydrostatic testing .

I OF OPERATION I I I I

Curve No. 2

'1100 I

I Minimum I I I Cl 1000 I. .

I temperature for upper RPV and beltline during in-I I

c<( '

I service leak or w hydrostatic testing.

J: 900 '---*

0 0...

I-

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(/)

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w 800 I

I I

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0 I-(.)

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w 0:::

z 700 600 630 pstg L.. .'*

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t *:I *--

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provide protection against feedwater nozzle degradation.

The curves allow for 1510 psig I shifts in RTNoT of the I-

~ 500 *- . l/j ...... .. ................. "'"'""'"""' ----

Reactor vessel beltline materials,

...J I .

I I

in accordance with UJ t I

0::: I I Reg. Guide 1.99, Rev.

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---

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embrittlement for 54 EFPY.

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t The acceptable area 200 FLANGE t

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83'F the right of the applicable curves.

mo t

I I

' I 0

0 25 50 75 100 125 150 175 200 225 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE ("F)

Figure 3.4.9-2 Pressure/Temperature Limits for Reactor In-Service Leak and Hydrostatic Testing BFN-UNIT 3 3.4-29c Amendment No. ~ , 278

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 278 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-68 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT. UNIT 3 DOCKET NO. 50-296

1.0 INTRODUCTION

By application dated January 27, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML15040A698), as supplemented by letters dated August 13 and October 23, 2015 (ADAMS Accession Nos. ML15226A324 and ML15296A527, respectively), the Tennessee Valley Authority (TVA, or the licensee), requested changes to the Technical Specifications (TSs) for Browns Ferry Nuclear Plant (BFN), Unit 3. The supplements dated August 13 and October 23, 2015, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register on May 5, 2015 (80 FR 25720).

The amendment revises BFN, Unit 3, Technical Specification (TS) 3.4.9, "RCS [Reactor Coolant System] Pressure and Temperature (PIT) Limits," Figures 3.4.9-1 and 3.4.9-2 that currently provide the PIT limits for up to 20 Effective Full Power Years (EFPYs), and greater than 20 EFPYs to less than or equal to 28 EFPYs, respectively. The amendment also revises Note 1 of TS Surveillance Requirement 3.4.9.1 to change the vessel pressure from less than 312 pounds per square inch gauge (psig) to less than 313 psig to conform to the modified PIT limit curves. The proposed PIT limits for Figures 3.4.9-1 and 3.4.9-2 are applicable to 38 EFPYs, and greater than 38 EFPYs to less than or equal to 54 EFPYs, respectively. The revision satisfies the requirements of NUREG-1843, "Safety Evaluation Report Related to the License Renewal of the Browns Ferry Nuclear Plant, Units 1, 2, and 3," dated April 2006 (ADAMS Accession No. ML061030032), Commitment 39, that required the development and submittal of revised PIT limit curves for U.S. Nuclear Regulatory Commission (NRC) approval prior to the period of extended operation.

2.0 REGULATORY EVALUATION

2.1 System Description The RCS is designed to withstand the effects of cyclic loads from system pressure and temperature changes due to startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. The PIT changes are limited during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.

The reactor pressure vessel (RPV) contains the reactor core and all associated support and alignment devices. The RPV is a part of the RCS pressure boundary, the second barrier to the release of fission products to the environment.

2.2 Applicable Regulatory Requirements The NRC established requirements in Part 50 of Title 1O of the Code of Federal Regulations (10 CFR) Part 50 to protect the integrity of the reactor coolant pressure boundary in nuclear power plants. The NRC staff evaluates the acceptability of a facility's proposed PIT limits based on the following NRC regulations and guidance:

Section 50.36 of 10 CFR requires that the TS include items in specific categories, including:

(1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements (SRs); (4) design features; and (5) administrative controls. This license amendment request is regarding the limiting conditions for operation and SR of the TS.

Section 50.60 of 10 CFR, "Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for normal operation," imposes fracture toughness and material surveillance program requirements, which are set forth in 10 CFR Part 50, Appendices G and H.

General Design Criterion (GDC) 14 in Appendix A to 10 CFR Part 50, "Reactor Coolant Pressure Boundary," requires the design, fabrication, erection, and testing of the reactor coolant pressure boundary (RCPB) so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.

GDC 30, "Quality of Reactor Coolant Pressure Boundary," requires, in part, that components comprising the RCPB be designed, fabricated, erected, and tested to the highest quality standards practical.

GDC 31, "Fracture Prevention of Reactor Coolant Pressure Boundary," requires that RCPB be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions, (1) the boundary behaves in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized.

Appendix G to 10 CFR Part 50, "Fracture Toughness Requirements," specifies fracture toughness requirement for ferritic materials of pressure-retaining components of the RCS boundary to provide adequate margins of safety during any condition of normal operation.

Regulatory Guide (RG) 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence" (ADAMS Accession No. ML010890301), describes methods and assumptions acceptable to the NRC staff for determining the pressure vessel neutron fluence with respect to the GDC contained in Appendix A of 10 CFR Part 50. Specifically GDC 14, 30, and 31 are applicable.

RG 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials" (ADAMS Accession No. ML003740284), contains guidance on methodologies the NRC considers acceptable for determining the increase in transition temperature and the decrease in upper-shelf energy resulting from neutron radiation.

Generic Letter (GL) 92-01, Revision 1, "Reactor Vessel Structural Integrity;" requested that licensees submit the RPV data for their plants to the NRC for review. Supplement 1 to GL 92-01, Revision 1, requested that licensees provide and assess data from other licensees that could affect their RPV integrity evaluations.

NUREG-800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition," Section 5.3.2, "Pressure-Temperature Limits, Upper-Shelf Energy, and Pressurized Thermal Shock," (ADAMS Accession No. ML070380185) describes acceptance criteria for determining the PIT limits for ferritic materials in the beltline of the RPV based on the American Society of Mechanical Engineers (ASME) Code Appendix G methodology.

Topical Report NEDC-32983P-A, Revision 2, "General Electric [GE] Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluations," dated January 2006 (ADAMS Accession No. ML072480121 ), provides a methodology intended for the determination of the fast neutron fluence accumulated by the pressure vessel and internal components of U.S. boiling-water reactor (BWR) plants. As noted by the safety evaluation enclosed in the NRG-approved topical report, the NRC determined that this methodology is generically acceptable for reference in licensing actions.

Topical Report NEDC-33178P-A, Revision 1, "GE Hitachi Nuclear Energy Methodology for Development of Reactor Pressure Vessel Pressure-Temperature Curves" (ADAMS Accession No. ML092370487) provides generic upper vessel and bottom head PIT limit curves along with beltline curves that are shifted by the plant-specific adjusted reference temperature, as described in RG 1.99, Revision 2, as well as guidance on the application of the 1998 edition, 2000 addenda of the ASME Boiler and Pressure Vessel (BPV) Code,Section XI, Appendix G and 10 CFR Part 50, Appendix G. Additionally, data from the Integrated Surveillance Program BWRVIP-135, "BWR Vessel and Internals Project Integrated Surveillance Program (BWRVIP ISP)

Data Source Book and Plant Evaluations," is requested to be submitted by NEDC-33178P-A. The BWRVIP-135 source book is used by the industry in compliance with BWRVIP-86, Revision 1, "BWR Vessel and Internals Project - Updated BWR Integrated Surveillance Program Implementation Plan" (ADAMS Accession No. ML090300556).

3.0 TECHNICAL EVALUATION

The operating limits for Prr are required for three categories of operation by 10 CFR Part 50 Appendix G:

• Hydrostatic pressure tests and leak tests, referred to as Curve A;

• Non-nuclear heatup/cooldown (core not critical), referred to as Curve B; and

• Core critical operation, referred to as Curve C.

The current Prr operating limits are set forth as TS Figures 3.4.9-1 and 3.4.9-2, valid for up to 20 and 28 EFPYs. The licensee proposed to replace these existing Figures 3.4.9-1 (Curves 2 and 3 are equivalent to Curves 8 and C) and 3.4.9-2 (Curve 2 is equivalent to Curve A) with corresponding new ones for up to 38 and 54 EFPYs.

The licensee proposes to revise Note 1 of TS SR 3.4.9.1 to change the vessel pressure from

> 312 psig to> 313 psig to conform to the modified Prr limit curves.

In addition, an associated note for each figure is changed to reflect the new operational applicability limit with respect to EFPY.

The NRC staff's review of the licensee's submittal was to determine if both the neutron fluence experienced by the reactor vessel and the Prr limit curves were appropriately developed using NRG-approved guidance. The NRC staff's review is below.

3.1 Fluence Calculation 3.1.1 Regulatory Guide 1.190 Criteria The guidance provided in RG 1.190 states that an acceptable fluence calculation has the following attributes:

• Performed using an acceptable methodology

• Contains an analytic uncertainty analysis identifying possible sources of uncertainty

• Contains a benchmark comparison to approved results of a test facility

• Demonstrates plant-specific qualification by comparison to measured fluence values The NRC staff reviewed the licensee's submittal against the guidance provided in RG 1.190 to determine the acceptability of the fluence calculation.

3.1.2 Acceptability of the Methodology For input to the Prr Limits, TVA performed the fluence calculations in accordance with GE Topical Report NEDC-32983P-A, Revision 2, and RG 1.190. A solution to the Boltzmann transport equation is approximated using the two-dimensional discrete ordinates code. The licensee used a cross-section library that the NRC staff has found generically acceptable (refer to Section 3.1 of the safety evaluation contained within NEDC-32983-A, Revision 2). Numeric

approximations include a P3 Legendre expansion for anisotropic scattering and the modeling uses Ss order of angular quadrature. These cross-section data and approximations are in accordance with the modeling guidance contained in RG 1.190. Additionally, the spatial distribution of neutron source density is assumed to be proportional to the relative cycle-averaged energy production at each fuel node and bundle location. The neutron source and transport calculations, as described above, were performed in accordance with the guidance set forth in RG 1.190.

In the supplement dated August 13, 2015, the licensee provided an explanation for how the neutron fluence calculation was updated to reflect the ATRIUM-10 specific fuel bundles located on the periphery of the reactor core and the NRC staff reviewed this supplemental information.

The evaluation was representative of an extended power uprate (EPU) core design utilizing a full core of ATRIUM-10/ Blended Low Enriched Uranium (BLEU) fuel, and the AREVA calculations were performed in accordance with RG 1.190. A comparison of the calculated neutron flux values from General Electric - Hitachi (GEH) and AREVA at the limiting locations on both the shroud and reactor vessel inner surfaces showed that the ATRIUM-10/BLEU flux values are bounded by those calculated by GEH and used in development of the BFN, Unit 3, PIT limit curves. The shroud and reactor vessel flux profiles reflect the same shape for both the GEH and AREVA calculations, but the flux values predicted by GEH are significantly higher.

Based on the review of the supplemental information, NRC staff finds that the ATRIUM-10 specific fuel bundles were updated with the bounding NEDC-32983P-A, Revision 2, calculation.

In the supplement dated October 23, 2015, the licensee provided information on the validation of neutron fluence calculations outside of the beltline and above the core, and the NRC staff reviewed this information. The water density used in the region outside the beltline is compared to the results for Browns Ferry from the GEH best-estimate thermal hydraulic code, TRACG, for nominal EPU/Maximum Extended Load Line Limit Analysis Plus conditions. The water density used in the region outside the beltline is from a reactor that is (1) at the end of cycle, (2) at an uprated power of 120 percent of original licensed thermal power, and (3) at 85 percent of rated RCS flow. The water density used in the neutron fluence analysis is more conservative than the most bounding power/flow water density in the upper plenum during reactor operation. The variation in water density associated with the most important region, the downcomer, is already characterized and accounted for in the determination of the overall uncertainty, as described in NEDC-32983P-A, Revision 2. The volume averaged water density at either axial level above the core is bounded by the water density used in the neutron fluence analysis. Although the analytical uncertainty is not specifically quantified, the following analysis decisions conservatively account for water density variation in the fluence: (1) thermal hydraulic analysis results chosen for comparison are from the most limiting operational characteristics (i.e., 120 percent of original licensed thermal power and 85 percent of rated flow), (2) the impact of the conservatism in water density is integrated over the entire operating period, (3) the water density in the neutron fluence analysis bounds the region of highest importance amongst the rings in the upper plenum (peripheral ring), and (4) the water density in the neutron fluence analysis bounds the volume averaged water density from the best-estimate thermal hydraulic analysis. Based on the conservative analysis decisions, the NRC staff finds that validation of the neutron fluence calculations outside the beltline and above the core to be adequate.

Based on the fact that (1) the neutron source and transport calculations were performed in accordance with the guidance set forth in RG 1.190, (2) the ATRIUM-10 specific fuel bundles

were updated with the bounding NEDC-32983P-A, Revision 2, calculation, and (3) the validation of the neutron fluence calculations outside the beltline and above the core are adequate, the NRC staff finds that the fluence calculation was performed using an acceptable methodology.

3.1.3 Uncertainty Analysis The NRG-approved methods in NEDC-32983P-A, Revision 2, are supported by an analytic uncertainty analysis. The estimated uncertainty analysis is less than 20 percent, which is in accordance with RG 1.190. Thus, the NRC staff determined that the neutron fluence calculation contains an acceptable analytic uncertainty analysis.

3.1.4 Benchmark Comparison NEDC-32983P-A, Revision 2, describes the methods qualification using the standard benchmark problems found in RG 1.190. The calculations were compared with the benchmark measurements from the vessel fluence benchmark problems provided in NUREG/CR-6115, "PWR [Pressurized-Water Reactor] and BWR Pressure Vessel Fluence Calculation Benchmark Problems and Solutions," (ADAMS Accession No. ML012900043) referenced in RG 1.190 and were found to have very similar results. The NRC staff finds that these calculations constitute acceptable test facilities because they are specifically referenced in RG 1.190.

3.1.5 Plant Specific Qualification NEDC-32983P-A, Revision 2, contains acceptable plant-specific benchmarking for BFN, Unit 3, as it contains a database of BWR dosimetry benchmarking, and BFN, Unit 3, unit geometry (BWR/4 Reactor Vessel) is well represented within the database. The GE-specific benchmarking documented in NEDC-32983P-A, Revision 2, indicates the surveillance capsule fluence can be calculated within 20 percent of measured values, which is in accordance with RG 1.190. The NRC staff finds that the fluence calculation demonstrates plant-specific qualification by comparison to measured fluence values.

3.1.6 Summary The NRC staff reviewed the fluence calculation and found that it meets the four criteria in RG 1.190; therefore, the NRC staff concludes that the fluence calculation is acceptable for use in the development of the PIT limit curves.

3.2 Consistency Between BFN, Unit 1, and BFN. Unit 3 The NRC staff has reviewed the applications to update the PIT Limits Curves TS for all three BFN reactors. Amendment Number 287 for BFN, Unit 1, was issued on February 2, 2014 (ADAMS Accession No. ML14325A501 ). During the review of the application for BFN, Unit 3, the NRC staff identified a possible discrepancy between the BFN, Unit 1, and the BFN, Unit 3, applications, despite significant similarity in plant designs. The NRC staff discovered that the axial weld peak neutron fluence values at 38 EFPYs for BFN, Unit 1, were similar to those same welds for BFN, Unit 3, at 54 EFPYs. The NRC staff approved 5.00 percent power uprates for BFN, Unit 3, and BFN, Unit 1, in 1998 and 2007, respectively.

In the supplement dated August 13, 2015, the licensee provided information regarding the comparison of fluence calculation inputs and results for BFN, Unit 1, and BFN, Unit 3. The NRC staff reviewed the supplementary information regarding the operational histories, neutron flux, and the similarity between the axial weld peak neutron fluence values of 54 EFPYs and 38 EFPYs for BFN, Unit 3, and BFN, Unit 1, respectively, to determine if the discrepancy indicated errors in the applications. For the operational histories, the original licensed thermal power, current licensed thermal power, and EPU power levels are consistent between the units and the same peak flux was used for both fluence calculations. For the neutron flux, the licensee applied the bounding EPU flux distribution for all fluence calculations. Regarding the differences between the axial weld peak neutron fluence values, they are due to the differences in the selection of peak or specific locations on the reactor vessel. BFN, Unit 1, applied the peak vessel fluence value, which is the peak over all azimuths for 38 EFPYs, to the axial welds rather than calculate the fluence at the specific weld locations. BFN, Unit 3, calculated the axial weld fluence at the limiting axial weld location for 54 EFPYs. Based on the review of the supplementary information, the NRC staff finds that the similarity of the axial weld peak neutron fluence values of 54 EFPYs and 38 EFPYs for BFN, Unit 3, and BFN, Unit 1, respectively, has a valid technical basis and no errors existed in the applications regarding the axial weld peak neutron fluence values.

3.3 PIT Limit Curves To evaluate the licensee's input material property values for calculating the PIT limit curves, the NRC staff first examined the licensee's selection of limiting materials. Specifically, for beltline materials, the NRC staff found that the initial RT NDT, copper, and nickel values had not changed since the NRC approval of the TS change associated with updated PIT limits, dated January 15, 1999 (ADAMS Accession No. ML020100017).

The licensee reported that there are no best-estimate chemistries for the BFN-3 beltline materials described in BWRVIP-135 and, therefore, the information from BWRVIP-135 does not change the limiting beltline material previously identified by the NRC staff. The licensee only calculated adjusted reference temperature values for the RPV 1/4T location, following NEDC-33178P-A, Revision 1. The NRC staff finds this calculation acceptable because using the maximum tensile stress for either heatup or cooldown, and applying it at the 1/4T location, is equivalent to using the maximum thermal stress intensity factor and the minimum fracture toughness in the heatup and cooldown analysis. The staff notes that the approach is part of the NRG-approved methodology in NEDC-33178P-A, Revision 1.

The NRC staff reviewed the limiting PIT limit Curves A, B, and C. The composite curves are consistent with those generated independently by the NRC staff applying the GE methodology, shifting the approved generic GE curves by the adjusted reference temperature, as defined in RG 1.99, Revision 2, for the limiting material. For all conditions, the Appendix G to 10 CFR Part 50 requirements for the minimum metal temperature of the closure head flange and vessel flange regions produce limiting "notches," serving to explain the distinct vertical lines at constant temperature in the licensee's proposed PIT limits. The NRC staff notes that the licensee had updated the hydro test pressure from 312 to 313 psig for Unit 3, which is identical to the change made for Unit 2 that was approved in Amendment No. 314 on June 2, 2015 (ADAMS Accession No. ML15065A049). The "notch" pressure changed to account for the actual preservice hydrostatic test pressure. Thus, the licensee's request to revise Note 1 of TS SR 3.4.9.1 to

change the vessel pressure from less than 312 pounds per square inch gauge (psig) to less than 313 psig is acceptable because it applies a plant-specific value in place of the generic value used in NEDC-33178P-A, Revision 1.

The NRC staff reviewed how the nozzles are accounted for in the heatup and cooldown analysis and confirmed that only the N16 water level instrument nozzles were in the extended beltline region where the fluence is estimated to exceed 1 x 1017 n/cm 2 (E > 1 MeV). The licensee used Appendix J of the NEDC-33178P-A, Revision 1, methodology to calculate the PIT limits for the water level instrument nozzle. The NRC staff previously determined in the safety evaluation for BWR Owners Group Topical Report BWROG-TP-11-023, Revision 0, "Linear Elastic Fracture Mechanics Evaluation of General Electric Boiling Water Reactor Water Level Instrument Nozzles for Pressure-Temperature Curve Evaluations" (ADAMS Accession No. ML13183A017),

that the thermal stress value from Appendix J of the NEDC-33178P-A, Revision 1, methodology is acceptable because it is derived from a bounding transient. The bounding transient is a more severe transient than is required to be addressed by the ASME BPV Code,Section XI, Appendix G. Using the emergency transient, the resulting applied stress intensity factor for the water level instrument nozzle in the PIT limit curves would be higher than would be calculated according to the ASME BPV Code; therefore, the NRC staff finds the analysis acceptable.

The NRC staff also evaluated the analysis of non-beltline components and materials by comparing the application and the BFN, Unit 3, Updated Final Safety Analysis Report (UFSAR).

The NRC staff confirmed that the material properties conveyed in the UFSAR are consistent with the material properties stated in the application. to Enclosure 1 of the application provides proposed changes to Section 4.2.4 of the UFSAR to reflect implementation of the new PIT limit curves. The NRC staff notes that the proposed changes to the UFSAR are consistent with the new PIT limit curves and supporting technical information. The NRC staff did not review the UFSAR changes for approval.

Based on the evaluation described above, the NRC staff finds that the proposed PIT limits follow the same approved methodology as the current PIT limits, and that the UFSAR material properties are consistent with the application.

3.4 Summarv Based on the review of the application, the NRC staff has determined that the licensee appropriately followed NRG-approved guidance for calculating neutron fluence and developing new PIT limits. The NRC staff finds that the proposed BFN, Unit 3, PIT limits satisfy the requirements of Appendix G to Section XI of the ASME BPV Code and Appendix G of 10 CFR Part 50, and concludes that they are acceptable for use in TS 3.4.9, Figures 3.4.9-1 and 3.4.9-2.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Alabama State official was notified of the proposed issuance of the amendment on December 10, 2015. The State official had no comments.

5.0 PUBLIC COMMENTS On May 5, 2015, the NRC staff published a "Notice of Consideration of Issuance of Amendments to Facility Operating Licenses, Proposed Significant Hazards Consideration Determination, and Opportunity for a Hearing," in the Federal Register associated with the proposed amendment request (80 FR 25720). In accordance with the requirements in 10 CFR 50.91, "Notice for public comment: State consultation," the notice provided a 30-day period for public comment on the proposed no significant hazards consideration (NSHC) determination. One public comment was received regarding the proposed amendment (ADAMS Accession No. ML15163A027). Some of the issues discussed in the public comment do not specifically pertain to the proposed NSHC determination. However, the NRC staff has addressed both the issues within the scope of the proposed NSHC and those that are not within the scope. A summary of the comment and the NRC staff response is provided below.

The comment states, in part:

It is criminal that you hide something as important as Pressure Temperature curves and changes under "no significant hazards". You can't get more significant than reactor pressure vessel failure.

NRC Response:

RPV failure is significant, and the NRC staff views RPV failure as a possible accident if the licensee exceeds P!T curves during operation. However, per Section 50.92(c) of 10 CFR, a "no significant hazards" determination is found for a license amendment if the proposed changes do 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. In this instance, none of the 10 CFR 50.92(c) qualifiers were met because the proposed changes to the P!T curves are based on NRC approved methods and guidance. Therefore, the NRC staff properly made a "no significant hazards" determination.

Additionally, the proposed P!T limit curves are publicly available. Please see the original submittal dated January 27, 2015 at ADAMS Accession No. ML15226A324. The licensee submitted proprietary information regarding the computation of those curves that is considered sensitive unclassified non-safeguards information and NRC staff performed reviews for withholding from the public under 10 CFR 2.390, "Public inspections, exemptions, request for withholding." Those reviews are documented in letters dated November 9, 2015 (ADAMS Accession Nos. ML15289A539 and ML15289A582).

The comment states, in part:

Embrittlement in nuclear reactors hasn't gotten better, it has only gotten worse over the course of the last 17 years. Reactor Pressure Vessel [RPV]

embrittlement will eventually lead to catastrophic failure and soon.

NRC Response:

Embrittlement increases over the service life of every RPV. The PIT limit curves and other NRC regulations set conservative limits to prevent RPV failure.

The comment states, in part:

I think that the reactors have already had license extension and are being allowed to operate past their shut-down date. And, you are aiming to justify it with a Pressure Temp curve which matches what you want to prove.

The only thing you have succeeded in proving is that nuclear "sciences" have special rules which do not match reality nor the scientific method. You are not supposed to extend operations of these old reactors without proving they are safe. You cannot prove what is false. It is not science to approve the extended operation and then manipulate statistically insignificant data to "prove" that a new Pressure Temp curve is ok.

NRC Response:

In a license extension request, each applicant must include (1) an Integrated Plant Assessment and (2) revised TS. An Integrated Plant Assessment identifies and lists structures and components subject to an aging management review. These include "passive" structures and components that perform their intended function without moving parts or without a change in configuration or properties. The licensee also submits TS changes or additions, with justification, necessary to manage the effects of aging during the period of extended operation.

The NRC staff reviews this information and if found to maintain plant safety throughout the period of extended operation, the NRC approves the license extension, but with specified commitments or conditions relating to the approval. This submission of the revised PIT curves before the period of extended operation is a commitment made by the licensee for the approval of their license extension. The NRC review of this submittal found the revised curves to provide a reasonable assurance of safety based on NRC regulations and guidance.

For more information, please visit http://www.nrc.gov/reactors/operating/licensing/renewal.html

6.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. 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, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there was one public comment on such finding published in the Federal Register on May 5, 2015 (80 FR 25720) that was responded to in Section 5.0 of this document. 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.

7.0 CONCLUSION

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

Principal Contributor: M. Hardgrove P. Purtscher Date:January 7, 2016

ML15344A321 *b1y memo OFFICE NRR/DORL/LPLll-1/PM NRR/DORL/LPLll-2/LA NRR/DSS/SRXB/BC* NRR/DE/EVIB/BC*

NAME MOrenak BClayton CJackson JMcHale DATE 12/16/2016 1/06/2016 11/04/2015 9/03/2015 OFFICE NRR/DSS/STSB/BC OGC-NLO w/comment NRR/DORL/LPLll-2/BC NRR/DORL/LPLll-2/PM NAME (MHamm for) RE Hiott Jlindell BBeasley (MOrenak for) FSaba DATE 12/18/2015 1/04/2016 1/06/2016 1/07/2016