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Enclosure 3: Tennessee Valley Authority Browns Ferry Nuclear Plant, Units 1, 2, and 3 - License Renewal Application Examples
	ML023240328
Person / Time
Site:	Browns Ferry
Issue date:	10/31/2002
From:	; Tennessee Valley Authority
To:	; Office of Nuclear Reactor Regulation
	Hoffman ST, NRR/DRIP/RLEP, 415-3245
References
	Download: ML023240328 (29)
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Browns Ferry License Renewal Project Application Examples Tab I BFN Database Excerpt - Main Steam System Tab 2 Sections 2 & 3 - Residual Heat Removal & Recirculation Systems Tab 3 Sections 2 & 3 - Main Steam System Tab 4 Section 4 - RPV Integrity Tab 5 Appendix B Example - Flow Accelerated Corrosion Program Description

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Bolting CLASS 1 MISCELLANEOUS APPURTENANCES CLASS 1 PIPING CLASS 1 RESTRICTING ORIFICE CLASS 1 VALVES NON-CLASS 1 MISCELLANEOUS APPURTENANCES NON-CLASS 1 PIPING Sub-Material: ASTM A 106 GRADE B Sub-Material: ASTM A 155 GRADE KC-70 CLASS 1 Sub-Material: ASME SA 335 GRADE P11 External: Inside Air ACEffEval Internal: Reactor steam t-'_"5 Gall: VIlI-62 1-a E2 1 1 52 1 - AMPs: 3 1 2 Gall: VIII-52 1-b B2 11 8_2 1 2 E

P 82 1 5 52 1 6 7 AMPs: 3 14 Gall: VII-B2 1-c B2 1 1 52 1 2 62 1 3 52 1 4 B2 1 5 B2 1 6 Gall: I1.,1-B2 E,-: AMPs: 3 12 Internal: Reactor water A-f"E, -

Gall: VI',-B2 1-a K2 1 B2 1 2 AMPs: 3 1 2 Gall: VIII-62 1-b B2 1 1 62 1 2

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Sub-Material: ASME SA 312 GRADE TP316 Sub-Material:

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ASME SA 479 GRADE 316L ASTM A 376 GRADE TP 316 ASTM A 376 GRADE TYPE 304 316 ASME SA 312 GRADE TP 304 ASTM A 213 GRADE TP 304-OR 316 ASTM A 312 GRADE TP 304 ASTM A 312 GRADE TP 316 SS External: Inside Air AcEffEval Internal: Reactor steam AgEffEva: Gall: VIII-B2 BFN-7 AMPs: 3 1 2 Group: NON-CLASS 1 RESTRICTING ORIFICE Group: NON-CLASS 1 TUBING Group: NON-CLASS 1 VALVES Group: STRAINERS F

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Browns Ferry Nuclear Plant, Units 1, 2, and 3 Application for Renewed Operating Licenses Aaministrative ano Tecnnicai information 2.3.1.4 Reactor Recirculation System (System 68)

Description The Reactor Recirculation system functions to:

Provide sufficient subcooled water to the reactor core during normal operations to maintain normal core operating temperatures.

Control reactor power by varying recirculation flow during normal operations.

Provide a flow path for LPCI flow from the RHR system to the reactor vessel during design basis accidents.

Provide a flow path to and from the RHR system for decay heat removal at low temperatures.

The Reactor Recirculation system for each unit shares no components and has no interconnections with the other units. The Reactor Recirculation system for each unit consists of two piping loops connected to but external to the reactor vessel Each loop has a single variable-speed motor-driven pump with pump suction and discharge valves. Each pump takes a suction from the reactor vessel downcomer region and discharges into a manifold that supplies flow to five jet pumps internal to the reactor vessel. With the exception of instrumentation piping that penetrates the primary containment, the recirculation system is located inside the primary containment.

The recirculation loop piping is stainless steel that is either IGSCC resistant or has been treated to improve resistance to IGSCC.

UFSAR Reference Additional description of the Reactor Recirculation system is contained in UFSAR 3 7.6, 4.3, 7.9, 7.19, and Appendix G.

Intended Functions The Reactor Recirculation system is in the scope of 10 CFR 54 because it contains components that meet the following criteria of 10 CFR 54.4.

Unit No.

(a)(1)

(a)(2)

(a)(3) FP (a)(3) EQ (a)(3) ATWS (a)(3) SBO U1 Yes No No No No No U2 & U3 Yes No Yes Yes Yes No The intended functions of the Reactor Recirculation system are:

Reactor coolant pressure boundary - The entire system provides a fission product and pressure barrier.

Primary containment boundary - Instrument lines that penetrate the primary containment provide a closed boundary to limit the release of radioactivity post accident Core cooling - The Reactor Recirculation system piping provides an ECCS flow path for RHR flow for core flooding and log-term cooling.

Residual heat removal - The Reactor Recirculation system piping provides a flow path for the shutdown cooling mode of RHR. (FP function) iiTSmitigation - The recirculation pumps trip to limit the consequences of a failure to scram during T transient.

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Text enclosed in bold border indicates applicable to Units 2 and 3 only Page 2 3 1-1

Browns Ferry Nuclear Plant, Units 1, 2, and 3

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Signal input - Instrumentation signals are provided to RPS to support APRM trips, RHR for isolation, and RHR and CS for LPCI injection permissive.

License Renewal Boundary Essentially the entire Reactor Recirculation external to the reactor vessel is in license renewal scope. The scoping and screening results for Reactor Recirculation system components internal to the reactor vessel are presented in section 2.3.1.2. The Reactor Recirculation system boundary is denoted in red on the following drawings:

Unit I Unit 2 Unit 3 Shared 1-47E817-1-LR 2-47E817-1 -LR 3-47E817-1-LR None 1-47E822-1-LR 2-47E822-1 -LR 3-47E822-1 -LR System Components/Commodities Requiring Aging Management Review The list of Reactor Recirculation system components/commodities subject to an aging management review and their intended functions is shown on the following table. The AMR results for the Reactor Recirculation system are presented in Table 3 1.2.4.

Table 2.3.1.4 Component/Commodity Intended Function Bolting Closure Piping and Fittings Pressure Boundary Recirculation Pump Casing Pressure Boundary Valves Pressure Boundary

. Text enclosed in bold border indicates applicable to Units 2 and 3 only.

Page 2 3.1-2

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Application for Renewed Operating Licenses Administrative and Technical Information 2.3.2.4 Residual Heat Removal (System 074)

Description The Residual Heat Removal system functions are:

I Shutdown cooling (residual heat removal with the reactor at low temperature) during normal operations and regulated events - The system circulates water from the recirculation pump suction through the RHR heat exchangers back to the recirculation pump discharge into the reactor vessel and through the core for decay heat removal when the RCS is intact. This mode of operation is manually initiated when required.

Core flooding to limit, in conjunction with the other ECCS systems during design basis accidents, the peak fuel clad temperature over the complete spectrum of possible break sizes in the reactor coolant pressure boundary - In LPCI mode, the system automatically starts to pump water from the suppression pool into the reactor vessel through the recirculation pumps discharge line for core flooding during a LOCA.

Long term cooling following loss of coolant accidents - Following LPCI during a LOCA, cooling is manually initiated using the LPCI flowpath for long term cooling.

0 Containment pressure and temperature control during loss of coolant accidents - In containment spray mode, the system pumps water from the suppression pool to spray headers in the containment to condense steam and control pressure in the containment. This mode of operation is manually initiated when required following a LOCA. In suppression pool cooling mode, the system pumps water from the suppression pool through the heat exchangers back to the suppression pool. This mode of operation is manually initiated when required during both normal and accident conditions.

Containment flooding post accident - Water can be pumped from non-accident unit suppression pools or from the RHRSW system to flood the containment. This mode of operation is manually initiated.

Suppression pool level control - Provisions are provided for both makeup and reject to maintain the suppression pool level within required limits.

0 Supplemental fuel pool cooling - Cross-connections with the FPC system allow the RHR heat exchangers to be used for spent fuel residual heat removal.

Each unit has two Residual Heat Removal system loops with each loop having two RHR pumps and two RHR heat exchangers The pump suction header and heat exchanger discharge header of one loop in Ul and one loop in U2 can be cross-connected. A similar cross-connection is provided between U2 and U3.

Provisions are provided to ensure the integrity of the reactor coolant boundary, primary containment, and secondary containment. Major components are located in the Reactor Building.

UFSAR Reference Additional descriptive information for the Residual Heat Removal system is found in UFSAR 4.8, 5 2 3, 6.4 4, 7.3, 7.4, 10.5, 10.9, and Appendices F and G.

Intended Function The Residual Heat Removal system is in the scope of 10 CFR 54 because it contains components that meet the following criteria of 10 CFR 54.4.

Unit No.

(a)(1)

(a)(2)

(a)(3) FP (a)(3) EQ (a)(3) ATWS (a)(3) SBO U1 Yes No No No No Yes U2 & U3 Yes No Yes Yes No Yes The intended functions of the Residual Heat Removal system are.

Reactor coolant pressure boundary - Portions of the RHR system provide a fission product and pressure barrier. When a loop is in standby, the pressure boundary extends to the inboard isolation Page 2 3 2-1

Browns Feriy Nuciear Plant, Unils i, 2, and 3 ApJ~

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Administrative and Technical Information valve on the SDC suction line and the check valve on the injection lines When a loop is operating, the boundary includes the entire loop Pnmary containment - Portions of the RHR system that penetrate the primary containment provide a closed boundary to limit the release of radioactivity post accident. When a loop is in standby, the pressure boundary extends to the outboard isolation valves on the SDC suction line and the injection lines When a loop is operating, the boundary includes the entire loop.

"* Primary containment isolation - Valves in lines that penetrate the primary containment are closed or can be closed during accidents.

"* Secondary containment boundary - Portions of the RHR system that penetrate the secondary containment provide a closed boundary to limit the release of radioactivity post accident.

"* Core cooling - The RHR system is an ECCS system that provides core flooding and long-term cooling capability. Included in this function is the capability to maintain readiness for low pressure injection while the RCS is being depressurized.

"* Residual heat removal - The RHR system transfers decay heat at low temperatures during regulated events (FP, SBO) from the RCS to the RHRSW system.

"* Containment pressure and temperature control - The RHR system removes heat from the suppression pool to the RHRSW system The containment spray capability provides for the condensation of steam in the drywell and torus air space.

"* Signal Input - RHR system instruments are used for RHR controls, ADS permissive, and for load shedding and unit prioritization logic License Renewal Boundary Essentially the entire Residual Heat Removal system is within the scope of license renewal requiring an AMR. The Residual Heat Removal system license renewal boundary is denoted in red on the following drawings*

Unit I

Unit 2 Unit 3 Shared 1-47E811-1-LR 2-47E81 1-1-LR 3-47E81 1-1-LR None ComponentslCommodities Subject to Aging Management Review The list of Residual Heat Removal system mechanical components/commodities subject to an AMR and their passive intended functions is shown on the following table. The AMR results for the RHR system are presented in Table 3 2.2.4.

Table 2.3 2.4 Component/Commodity Intended Function Bolting Closure Heat exchangers Heat transfer Pressure boundary Nozzles and Orifices Throttle Pressure boundary Piping and fittings Pressure boundary Pumps Pressure boundary Valves Pressure boundary Page 2.3 2-2

Rrowns Ferry Nucier Plnt, I Inits 1, 2, and 3 ADDlication for Renewed Operating I irPn :*

Administrative and Technical Information 2.3.4.1 Main Steam (MS - System 001)

System Description

The MS system consists of four main steam lines that transfer steam from reactor vessel to the various steam loads in the turbine building during normal plant operation. Two MSIVs are provided in each steam line to isolate the RCPB and the primary containment. Steam supply lines for the HPCI and the RCIC systems branch off the main steam lines between the reactor vessel and the MSIVs. A flow restrictor is provided in each main steam line. The flow restrictor allows for measurement of steam flow and limits the steam flow rate in the event of a downstream steam line break. Thirteen main steam relief valves are provided on the main steam lines upstream of the flow restrictors for automatic overpress ure protection and for automatic depressurization following Small Brenk I n,;q of C..ol;nt Arrf;idp,,ts* The MS -;vctpmn fr each unit shares no components with the other units lRadioactivity release from MSIV leakage to the 1Ftmosphere is minimized by, plateout on MS piping downstream of the MSIVs.

UFSAR Reference Additional descriptive information for the Main Steam system is found in UFSAR 4.4, 4.5, 4.6, 4.11, 5.2, 64 2, 7.3, 7.11, 10, and Appendix G.

Intended Functions The MS system is in the scope of 10 CFR 54 because it contains components that meet the criteria of 10 CFR 54.4 for the following paragraphs Unit No.

(a)(1)

(a)(2)

(a)(3) FP (a)(3) EQ (a)(3) ATWS (a)(3) SBO U1 Yes No No No No Yes U2&U3 Yes Yes Yes Yes No Yes The Main Steam (MS) system intended functions are:

"* Pressure retaining flow path - Steam from the reactor vessel is rovided to the HPCI and RCIC systems for design basis accidents, SBi and fire protection.

"* Over-pressure protection for the reactor vessel - MSRVs function automatically or may be operated manually to limit reactor vessel pressure transients

"* Depressurization during a SBLOCA (ADS function) - MSRVs open automatically to depressurize the reactor vessel to allow LPCI and CS injection flow.

Limitation of release of radioactive material to the environment - Radioactivity in MSIV leakage plates out on MS system piping.*

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Reactor coolant pressure boundary - The piping from the reactor vessel to the outboard primary containment isolation valves (MSIVs, MSL drains) provides a fission product and pressure barrier.

Primary containment - Portions of the Main Steam system that penetrate the primary containment provide a closed boundary to limit the release of radioactivity post accident Primary containment isolation - Valves in lines that penetrate the primary containment are closed or automatically close during accidents Secondary containment - Portions of the MS system that penetrate the secondary containment provide a closed boundary to limit the release of radioactivity post accident.

Signal input - Initiation and control signals to the RPS, PCIS, and ADS are provided.

Text enclosed in bold border indicates applicable to Units 2 and 3 only Page 2.3 4-1

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Application for Renewed Operating Licenses Aiii;n;it ative ai j T -chnI i

ri, rii orni ation License Renewal Boundary The license renewal boundary extends from the reactor vessel to the outboard MSIVs and then to the main turbine stop valves and the stop or isolation valves for other loads Also included are the main steam line drain lines to the outboard containment isolation valves and then to the condenser.*

I The following drawings depict the MS system license renewal boundary in red (or blue if only intended function is Seismic Il/I)

Unit I Unit 2 Unit 3 Shared 1-47E801-1-LR 2-47E2847-9-LR 3-47E3847-9-LR None 1-47E817-1-LR 2-47E801-1-LR 3-47E801-1-LR 1-47E801-2-LR 2-47E801-2-LR 3-47E801-2-LR 1-47E807-1-LR 2-47E807-1-LR 3-47E807-1-LR 1-47E807-2-LR 2-47E807-2-LR 3-47E807-2-LR System Components/Commodities Requiring Aging Management Review The MS system components/commodities subject to an aging management review and their passive intended functions are listed in following table. The AMR results for the MS system are presented in Table 3.4.2.1 Table 2.3.4.1 Component/Commodity Intended Function Bolting Closure Miscellaneous Appurtenances - Class 1 Pressure Boundary (Instrument fittings at flow restrictors)

Miscellaneous Appurtenances - Class 1 Throttle (Venturi inserts for MS Flow Restnctors Piping and Fittings - Class 1 Pressure Boundary Throttle Piping and Fittings - Non Class 1 Pressure Boundary Throttle Restricting Orifice -Class 1 Pressure Boundary Throttle Restricting Orifice -Non-Class 1 Pressure Boundary Throttle Strainers - Non-Class 1 Pressure Boundary Valves - Class 1 Pressure Boundary Valves - Non-Class 1 Pressure Boundary Text enclosed in bold border indicates applicable to Units 2 and 3 only.

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A po-mm" "M a" Browns Ferry Nuclear Plant, Units 1, 2, and 3 Application for Renewed Operating Liceinses Aging Management Review Re!;ulls Table 3 1 2.4:

Reactor Recirculation System (System 068)

Summary of Aging Management Component Function Material Environment Aging Effect NUREG -1801 AMP Notes Vol 2 Item Vol 1 item Bolting Closure Carbon &

Internal. None Loss of Material/

IV C1.3-e 3.1.1 26 Bolting Integrity Note A.

Low Alloy External: Drywell Wear IV C1 2-d (XI.M18)

Steel atmosphere with App. B, B.2.11 metal temperature Loss of pre-loadl IV C1.3-f 3.1.1.26 Bolting Integrity Note A.

up to 550'F stress relaxation IV C1.2.e (XI.M18)

App. B B.2.11 Fatigue IV C1.3-g 3.1.1.26 TLAA Ch. 4.3.11 Note A.

IV C1.2-f Piping and Fittings Pressure Stainless Treated water Crack initiation and IV C1.1-f 3.1.1.29 BWR stress Note A.

boundary Steel (e.g.,

growth due to SCC corrosion cracking type 304, and IGSCC (MI.M7) 316, or App. B. B.1.32 316NG)

Chemistry Control (XI.M2)

App. B. B.1.4 Crack initiation and IV C1.1-i 3.1.1.7 ASME Section XI Note A growth due to SCC Inservice Plant specific Note 1 and IGSCC in small Inspection, (XI.M1) bore piping App B, B.1.7 Chemistry Control Program (XI.M2)

App. B, B.1.4 One time inspection (XI.M32)

App B, B 2.11 Recirculation Pump Pressure CASS Treated water Cumulative fatigue 3.1.1.1 TLAA Note A Casing boundary damage Ch 4.3 86 Crack initiation and IV C1.2-b 3.1.1.29 BWR stress Note A.

growth due to SCC corrosion cracking and IGSCC (Ml.M7)

App. B, B.1.32 Chemistry Control Program (XI.M2)

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Browns Ferry Nuclear Plant, Units 1, 2, and.3 Application for Renewed Operating Licenses; Aging Management Review Results; Table 3.1.2.4:

Reactor Recirculation System (System 068)

Summary of Aging Management Component Function Material Environment Aging Effect NUREG -1801 AMP Notes Vol 2 Item Vol 1 Item Recirculation Pump Pressure CASS Treated water Loss of fracture IV C1 2-c 3.1.1.23 ASME Section XI Note A.

Casing (continued) boundary toughness from Inservice thermal aging Inspection, (XI.M1) embrittleness App B, B.1.7 Valves -

Pressure CASS Treated water Wall thinning due to IV.C1.3-a 3.1.1.25 Flow Accelerated Note A.

boundary FAC Corrosion (Xl M17)

App. B. B.1.6 Loss of fracture IV Cl.3-b 3.1.1.23 ASME Section XI Note A.

toughness from Inservice thermal aging Inspection, (XI.M1) embrittleness App B, B.1.7 Crack initiation and IV C1.3-c 3.1.1.29 BWR stress Note A.

growth due to SCC corrosion cracking and IGSCC (MI.M7)

App. B, B.1.32 Chemistry Control Program (XI M2)

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IApp. B. B.1.4 Cumulative fatigue IV C1.3-d 3.1.1.1 TLAA Note A.

damage I

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1 Page 3.1-4

Table 3.1.2.4 Notes:

General:

A. Consistent with NUREG-1801 item for component, material, environment, aging effect and aging management program.

B. Component is different, but consistent with NUREG-1801 item for material, environment, aging effect and aging management program.

C. Material not in NUREG-1 801 for this component.

D. Environment not in NUREG-1801 for this component and material.

E. Aging effect not in NUREG-1801 for this component, material, and environment combination.

F. AMP not consistent with NUREG-1 801 for this component, material, environment and aging effect combination Plant Specific*

1 Provide required further evaluation.

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Browns Ferry Nuclear Plant, Units 1, 2, and 3 Application for Renewed Operating Licenses Aging Management Review Re3uls Table 3.2.2.4:

Residual Heat Removal System (System 074)

Summar! of Agin anagement Component Function Material Environment Aging Effect NUREG -1801 AMP Notes Vol 2 Item Vol 1 item Bolting Closure Carbon &

Internal: None Loss of Material/

IV.C1.3-e 3.1.1.26 Bolting Integrity Note A.

Low Alloy External: Air with Wear V.E.2-a 3.2.1.18 (XI M18)

Steel metal temperature App. B, B.2.11 up to 550°F Loss of pre-load/

IV.C1.3-f 3.1.1.26 Bolting Integrity Note A.

stress relaxation V.E.2-b 32.1.18 (XI.M18)

App. B, B.2.11 Fatigue IV.Cl.3-g 3.1.1.26 App. B, B.2.11 Note A.

TLAA Ch. 4 3.22 Heat Exchangers Pressure Carbon &

Primary side:

Loss of material due V.D.2.4-a 3 2.2.12 Open Cycle Note A.

boundary Low Alloy treated water to general corrosion, Closed Cooling Steel Secondary side:

pitting, crevice and Water System Raw water MIC corrosion (X1.M12)

App B, B 2 33 Heat Carbon &

Primary side:

Buildup of deposits V.D 2.4-b 32.2.12 Open Cycle Note A.

transfer Low Alloy treated water Biofouling Closed Cooling Steel Secondary side:

Water System Raw water (X1.M12)

App B, B.2.33 Pressure Carbon &

Primary side:

Loss of material due V.D 2.4-b 3.2.2.13 Closed Cycle Note A.

boundary Low Alloy treated water to general corrosion, Closed Cooling Steel Secondary side:

pitting, and crevice Water System Closed cycle corrosion (X1.M21) cooling treated App B, B.2.34 water Nozzles and orifices Pressure Carbon &

Drywell Air Loss of material due V.D.2.5-a 3.2 2.3 Spray Nozzle Note A.

(Drywell and suppression boundary Low Alloy to general corrosion Surveillance chamber spray)

Steel Program App B, B.2.40 Throttle Carbon &

Drywell Air Plugging due to V.D.2.5-b 3.2.2.9 Spray Nozzle Note A.

Low Alloy general corrosion Surveillance Steel Program I-_ I App B, B.2.40 Page 3.2-3

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Residual Heat Removal System (System 074)

Browns Ferry Nuclear Plant, Units 1, 2, arid 3 Application for Renewed Operating Licenses; Aging Management Review Results; Summa or Pginq ianagemen Component Function Material Environment Aging Effect NUREG -1801 AMP Notes Vol 2 Item Vol I Item Piping and Fittings Pressure CASS Treated water Crack initiation and V D.2.1-c 3 2.2.16 BWR stress Note A.

boundary growth due to SCC corrosion cracking and IGSCC (MI.M7)

App. B, B.1.32 Chemistry Control Program (XI.M2)

App. B, B.1.4 Loss of fracture IV C1.1-g 3.1.1.24 ASME Section Xl Note A.

toughness from V.D2.1-d 3.2.2.11 Inservice thermal aging Inspection, (XI.M1) embrittleness App B, B.1.7 Thermal Aging embrittlement of CASS (X1.M12)

App B, B.2.51 Crack initiation and IV.C1.1.i 3.1.1.7 ASME Section Xl Note A.

growth due to SCC Inservice Plant specific Note 1 and IGSCC (Small Inspection, (XI.M1)

Bore)

App B, B.1.7 Chemistry Control Program (XI M2)

App. B, B.1.4 One time inspection (XI.M32)

App B, B.2.11 Piping and Fittings Pressure Carbon steel Internal*

Loss of material due V.C.1-b 3.2.2.6 Chemistry Control Note A.

boundary Treated water to general, pitting, V.D2.1-a 3 2.1 2 Program (XI.M2)

Plant specific Note 2 and crevice V.D2.1-a 3.2.1.4 App. B, B.1.4 Plant specific Note 2 corrosion One time inspection (XI.M32)

App B B.2.41 External: Air Loss of material due V.E.1-b 3.2.2.10 Inspection of Note A.

to general corrosion External Surfaces Program App B, B.2.13 Page 3.2-4

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Residual Heat Removal System (System 074) 5Rimmnrv of Anini IVMananment

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Browns Ferry Nuclear Plant, Units 1, 2, and 3 Application for Renewed Operating Licen,;es Aging Management Review Results Component Function Material Environment Aging Effect NUREG -1801 AMP Notes Vol 2 Item Vol 1 item Pump Casing Pressure Carbon steel Internal Loss of material due V.D2.2-a 3.2.1.2 Chemistry Control Note A.

boundary Treated water to general, pitting, 32.1.4 Program (XI.M2).

Plant specific Note 2 and crevice App. B, B.1.4 corrosion One time inspection (Xl M32)

App B, B.2.11 Pump Casing Pressure Carbon steel External: Air Loss of material due V.E.1-b 3.2.2.10 Inspection of Note A.

boundary to general corrosion External Surfaces Program App B, B 2.61 Valves Pressure CASS Treated water Wall thinning due to IV.C1.3-a 3.1.1.25 Flow Accelerated Note A.

boundary FAC Corrosion (XI.M17)

App. B, B.1.6 Loss of fracture IV.C1.3-b 3.1.1.23 ASME Section XI Note A.

toughness from Inservice thermal aging Inspection, (XI.M1) embrittleness App B, B.1.7 Thermal Aging embrittlement of CASS (X1.M12)

App B, B 2.51 Crack initiation and IV.C1.3-c 3.1.1.29 BWR stress Note A.

growth due to SCC V.D2.3-c 3.2.2.16 corrosion cracking and IGSCC (Ml.M7)

App. B, B.1.32 Chemistry Control Program (XI.M2)

App. B, 6.1.4 Cumulative fatigue IV.C1.3-d 3.1.1.1 TLAA Note A.

damage Ch. 4.6.17.

Carbon Steel Internal:

Loss of material due V.C.1-b 3.2.2.6 Chemistry Control Note A.

treated water to general, pitting, V.D2.1-a 3.2.1.2 Program (XI.M2)

Plant specific Note 2 and crevice V.D2.1-a 32.1.4 App. B B.1.4 corrosion One time inspection (XI M32)

I App B, B.2.11 Page 3.2-5

Browns Ferry Nuclear Plant, Units 1, 2, an J 3 Application for Renewed Operating Licenmes Aging Management Review Results Table 3 2 2.4:

Residual Heat Removal System (System 074)

Summar g

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Management Component Function Material Environment Aging Effect NUREG -1801 AMP Notes Vol 2 Item Vol 1 item Valves Pressure Carbon Steel External:

Loss of material due V E.1-b 3.2.2.10 Inspection of Note A (cont) boundary (cont) air to general corrosion External Surfaces (cont)

Program I

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Table 3.2.2.4 Notes:

General:

A. Consistent with NUREG-1801 item for component, material, environment, aging effect and aging management program.

B.

Component is different, but consistent with NUREG-1 801 item for material, environment, aging effect and aging management program.

C. Material not in NUREG-1 801 for this component.

D. Environment not in NUREG-1 801 for this component and material.

E. Aging effect not in NUREG-1801 for this component, material, and environment combination.

F. AMP not consistent with NUREG-1801 for this component, material, environment and aging effect combination Plant Specific:

1 Discuss one time inspection.

2 For NUREG-1801 item V.C.1.b, the only inside surface environment applicable to the RHR system is treated water.

Page 3 2-7

Table 3.4.2.1:

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Application for Renewed Operating Licenses Aging Management Review Re3ults Main Steam (System 001)

Summary of Aging Management Page 3.4-3 Component Function Material Environment Aging Effect NUREG -1801 AMP Notes Vol 2 Item Vol 1 Item Bolting Closure Carbon & Low Internal: None Loss of Material/

IV C1.3-e 3.1.1 26 Bolting Integrity Note A.

Alloy Steel External: Air Wear VIII H2-a 3.4.1 8 (XI.M18) with metal App. B B 2.11 temperature up Loss of pre-load/

IV C1.3-f 3.1.1.26 Bolting Integrity Note A.

to 550°F stress relaxation VIII H2-b 3.4.1.8 (XI M18)

App. B B 2.11 Fatigue IV C1.3-g 3.1.1.26 TLAA See Ch 4 Note A.

Miscellaneous Pressure Carbon steel Internal:

Loss of material -

Chemistry Control Note C.

Appurtenances - Class 1 Boundary Treated water galvanic corrosion Program (XI M2)

Plant specific nc te I (Instrument fittings at App. B B.1.4 flow restrictors)

Miscellaneous Throttle CASS

Internal, Change in material 3.1.1.24 Thermal Aging Note B Appurtenances - Class 1 Treated water properties/ reduction embrittlement of Plant specific nc te :2 (Venturi inserts for MS in fracture toughness CASS (X1.M12)

Flow Restrictors due to thermal aging:

App B B 2.51 Piping and Fittings -

Pressure Carbon & Low Internal:

Wall Thinning - FAC IV Cl.1-a 3.1.1.xx Flow Accelerated Note A.

Class 1 Boundary Alloy Steel Treated water Corrosion (XI M17)

App. B B.1 6 Loss of material due Chemistry Control Note B.

to general, pitting and Program (XI.M2)

Plant specific Note 3 crevice corrosion App. B B.1.4 Fatigue IV C1.1-b 3.1.1.1 TLAA See Note A.

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Main Steam (System 001)

Summary of Aging Management Component Function Material Environment Aging Effect NUREG -1801 AMP Notes Vol 2 Item Vol 1 item Piping and Fittings -

Pressure Small bore

Internal, Crack initiation and IV C1.1-i 3.1.1.7 ASME Section XI Note A.

Class 1 Boundary piping less Treated water growth due to SCC, Inservice Plant specific note 4 than NPS 4 IASCC, thermal and Inspection, (XI.M1)

CASS and mechanical loading App B B.1.7 Wrought Chemistry Control Stainless steel Program (XI.M2)

App. B B.1.4 One time inspection (XI.M32)

App B B.2.11

CASS, Internal:

Loss of material due Chemistry Control Note B Wrought Treated water to pitting and crevice Program (XI M2)

Note E Stainless steel corrosion App B B.1.4 Plant specific nole 5 Piping and Fittings' -

Pressure Carbon & Low Internal:

Loss of material due VIII B2.1-a 3.4.1.7 Chemistry Control Note A Non Class 1 Boundary Alloy steel Treated water to pitting and crevice Program (XI.M2) corrosion

'App. B B.1.4 Wall thinning due to VIII B2.1-b 34.1.6 Flow Accelerated Note A flow-accelerated Corrosion (XI M17) corrosion App. B B 1 6 Cumulative fatigue VIII B2.1-c 3.4.1.1 TLAA Note A damage due to fatigue Chapter 4 3 22 Loss of material due Chemistry Control Note B 7

to' general corrosion Program (Xl M2)

Plant specific No e E; App. B B.1.4 Stainless Steel Internal:

Chemistry Control Note C (CASS and Treated Water Program (XI.M2)

Plant specific No e 7 Wrought)

App. B B.1.4 Loss of material due to pitting corrosion and crevice corrosion:

Crack initiation and Chemistry Control Note C growth due to SCC Program (XI.M2)

Plant specific Noe 8 App. B B.1.4 mil-

. Text enclosed in bold border indicates applicable to Units 2 and 3 only.

Page 3.4-4 Pý Oft....a P.-M awmma

Table 3 4.2 1:

Browns Ferry Nuclear Plant, Units 1, 2, *ind 3 Application for Renewed Operating Licenses Aging Management Review Results Main Steam (System 001)

Summary of Aging Management Component Function Material Environment Aging Effect NUREG -1801 AMP Notes Vol 2 Item Vol 1 item Restricting Orifice -

Throttle Stainless Steel Internal.

Loss of material due Chemistry Control Note B Class 1 Pressure (CASS and Treated Water to crevice corrosion Program (XI.M2)

Plant specific Note 9 boundary Wrought) and pitting corrosion:

App. B B.1.4 Crack initiation and 3.1.1.7 ASME Section Xl Note B growth due to SCC ISI, (XI.MI)

Plant specific Note App B B.1.7 10 Restricting Orifice - Non-Throttle Carbon and Internal:

Loss of material due Chemistry Control Note B Class 1 Pressure Low Alloy Steel Treated Water to general corrosion, Program (XI.M2)

Plant specific Note Boundary crevice corrosion, and App. B, B.1.4 11 pitting corrosion:

Strainers - Non Class 1 Pressure Carbon and Treated Water Loss of material due Chemistry Control Note B boundary Low Alloy Steel to general, pitting, and Program (XI.M2)

Plant specific Note crevice corrosion App. B, B.1.4 12 Carbon and Treated Water Loss of material due Chemistry Control Note B Low Alloy Steel to galvanic corrosion Program (XI.M2)

Plant specific Note (Strainer App. B, B.1.4 13 elements are stainless steel)

Valves - Class 1 Pressure Carbon & Low Internal:

Wall thinning due to IV C1.3-a 3.1.1.25 Flow Accelerated Note A.

Boundary Alloy steel Treated water FAC Corrosion (XI M17)

App. B B.1.6 Stainless Steel

Internal, Crack IV C1 3-c 3.1.1.29 Chemistry Control Note A (CASS and Treated water Initiation/Growth due Program (XI.M2)

Wrought) to stress corrosion App. B B.1.4 cracking (SCC):

Carbon & Low Internal:

Cumulative fatigue IV C1.3-d 3.1.1.1 TLAA Note A.

Alloy steel, Treated water damage due to fatigue Chapter 4.3.14 Stainless Steel (CASS and Wrought)

Carbon & Low Internal:

Loss of material due Chemistry Control Note D.

Alloy steel Treated Water:

to pitting corrosion, Program (XI.M2)

Plant specific note general corrosion, and App. B B.1.4

14.

crevice corrosion Stainless steel Internal:

Loss of material due Chemistry Control Note E (CASS and Treated Water:

to pitting corrosion Program (XI.M2)

Plant specific Note wrought) and crevice App. B B.1.4 15 corrosion::

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Table 3.4 2.1:

Browns Ferry Nuclear Plant, Units 1, 2, ancd 3 Application for Renewed Operating Lice nses Aging Management Review Re suIts Main Steam (System 001)

Summary of Aging Management Component Functior Material Environment Aging Effect NUREG -1801 AMP Notes Vol 2 Item Vol 1 item Valves - Non Class 1 Pressure Carbon and Internal:

Wall thinning due to VIII A2-a 3.4.1.6 Flow Accelerated Note A.

Boundary Low Alloy Steel Treated water FAC Corrosion (XI.M17)

Plant specific Note App. B, B.1.6 16 Loss of material due VIII A2-b 34.1.6 Chemistry Control Note A.

to general, pitting, and Program (XI.M2)

Plant specific Note crevice corrosion App. B, B.1.4 14 Loss of material due Chemistry Control Note D to galvanic corrosion Program (XI.M2)

Note E and general corrosion App. B, B.1'.4 Plant specific Note 17 Stainless Steel Internal:

Loss of material due Chemistry Control Note C (CASS and Treated Water to crevice corrosion, Program (XI.M2)

Plant specific Note Wrought) pitting corrosion:

App. B, B.1.4 18 crack initiation and growth due SCC Page 3.4-6 am -- m N

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T db;t: 3.4.2.1 Noi~es:

General:

A. Consistent with NUREG-1 801 item for component, material, environment, aging effect and aging management program.

B. Component is different, but consistent with NUREG-1801 item for material, environment, aging effect and aging management program.

C. Material not in NUREG-1801 for this component.

D. Environment not in NUREG-1 801 for this component and material.

E. Aging effect not in NUREG-1801 for this component, material, and environment combination.

F. AMP not consistent with NUREG-1801 for this component, material, environment and aging effect combination Plant Specific Notes:

1 This grouping applies to carbon steel instrument piping where it contacts the stainless steel steam lines and flow restrictors The chemistry control program is adequate to manage the effects of galvanic corrosion of the carbon steel instrument piping where it contacts the stainless steel steam lines and flow restrictors because later 2

This grouping applies to the venturi inserts for the main steam line flow restrictors Reduction in fracture toughness due thermal aging is adequately managed by the thermal aging embrittlement of CASS AMP.

3 This grouping includes flow elements, thermowells, and piping/tubing fittings. The chemistry control program for this grouping adequately manages the effects of general, pitting and crevice corrosion because in both carbon and stainless steel flow elements, thermowells, and piping/tubing fittings because...

4 Class I piping with a diameter less than nominal pipe size 4 inch does not receive a volumetric examination during inservice inspection. The one time inspection is intended to confirm that crack initiation and growth due to SCC or cyclic loading is not occurring. The one time inspection program for small bore stainless steel piping on the RCPB is described in Appendix B section B.2.33.

5 This grouping includes thermowells, flow elements and pipe/tubing fittings.

6 This grouping applies to carbon steel valve bodies on the reactor coolant pressure boundary in the MS system.

The chemistry control program is adequate to manage the effects of general, pitting, and crevice corrosion because [later]

7 later 8

later 9

This grouping applies to stainless steel restricting orifices on the RCPB. The chemistry control program is adequate to manage the effects of pitting and crevice corrosion because these components similar to those already evaluated.

10 This grouping applies to stainless steel restricting orifices on the RCPB. The ISI program is adequate to manage the effects of SCC because the material and environments are similar to those already evaluated.

11 This grouping includes flow restrictors that [later] The chemistry control program is adequate to manage the effects of general, pitting, and crevice corrosion because [later]

12 This grouping applies to carbon steel strainers. The chemistry control program is adequate to manage the effects of general, pitting and crevice corrosion in carbon steel strainer bodies because [later]

13 This grouping applies to carbon steel strainers where the strainer elements are stainless steel The chemistry control program is adequate to manage the effects of galvanic corrosion on the carbon steel valve bodies because [later]

14 The chemistry control program is adequate to manage the effects of general, pitting and crevice corrosion in carbon steel valve bodies because [later]

15 The chemistry control program is adequate to manage the effects of general, pitting and crevice corrosion in stainless steel valve bodies because [later]

Page 3 4-7

16 The only turbine control valves included in the scope of license renewal at BFN are the turbine stop valves. The main steam safety/relief valves listed in this item are included as Class 1 valves in GALL Item IV.Cl.3.

17 This grouping applies to carbon steel valves and carbon steel instrument root valves in contact with stainless steel. The internal environment is stagnant treated water at low temperature. The chemistry control program is adequate to manage the effects of general and galvanic corrosion because [later]

18 This grouping applies to non-Class 1 stainless steel valves in the MS system downstream of the MSIVs. The chemistry control program is adequate to manage the effects of general, pitting, crevice, corrosion and stress corrosion cracking because [later]

Page 3.4-8

Browns Ferry Nuclear Plant, Units 1, 2, and 3 Application for Renewed Operating Licenses Time Limited AMina Analyses 4.2 Neutron Irradiation Embrittlement Appendix G to 10 CFR Part 50 specifies fracture toughness requirements for ferritic materials of the pressure-retaining components of the reactor coolant pressure boundary of light water nuclear power reactors to ensure adequate margins of safety during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. For the RPV, Appendix G to 10 CFR Part 50 requires an evaluation of the Charpy Upper Shelf Energy (USE) and an evaluation of the Adjusted Reference Temperature (ART) to determine pressure-temperature (P-T) limits for the RPV.

Neutron irradiation causes a decrease in the Charpy USE and an increase in the ART of the RPV beltline materials.

This section presents TVA's evaluation of the impact of irradiation during the period of extended operation on the Charpy USE for the BFNP reactor vessels In addition, this section presents TVA's evaluation of the impact of irradiation during the period of extended operation on the BFNP RPV-temperature limits A calculation( Ref. 1) to determine the RPV ID neutron flux using the approved GE flux calculation methodology ( Ref. 2) has been performed for the extended power uprate thermal power limit of 3952 MWt The results of the calculation are that the peak RPV ID neutron flux is 1.4e9 n/sec-cm 2. The axial and azimuthal (at core midplane) flux distributions were also obtained.

The flux calculation results were used as the input to the revised ART, Charpy USE, and the P-T limits In addition, the calculations will assume that, at the end of the period of extended operation (60 calendar years), the end of life effective full power years (EFPY) will be 52. This determination is based on assumptions that the capacity factor will be 90% and the rated power level will be 3952 MWt for Unit 1 after [later] EFPPY, Unit 2 after 18.1 EFPY, and Unit 3 after 13 EFPY. The rated power level assumed for the units until implementing the power uprate is the currently licensed power level of 3458 MWt 4.2.1 RTNDT The ART is defined as the sum of the initial (unirradiated) reference temperature (initial RTNDT),

the mean value of the adjustment in reference temperature caused by irradiation (delta RTNDT),

and a margin (M) term The delta RTNDT is a product of a chemistry factor and a fluence factor.

The chemistry factor is dependent upon the amount of copper and nickel in the material and may be determined from tables in RG 1.99, Rev. 2, or from surveillance data. The fluence factor is dependent upon the neutron fluence at the maximum postulated flaw depth. The margin term is dependent upon whether the initial RTNDT is a plant-specific or a generic value and whether the chemistry factor (CF) was determined using the tables in RG 1.99, Rev. 2, or surveillance data.

The margin term is used to account for uncertainties in the values of the initial RTNDT, the copper and nickel contents, the fluence, and the calculation methods. RG 1.99, Rev. 2, describes the methodology to be used in calculating the margin term.

The 52 EFPYs ART for the limiting beltline material for Unit 1 [later] The 52 EFPYs ART for the limiting beltline material for Unit 2 (Heat C2463-1) at 1/4T has been determined to be 157.4 °F.

The 52 EFPYs ART for the limiting beltline material for Unit 3(Shell # 2, Heat C3222-2) at 1/4T is 157.4 OF. These values for ARTs are based on a neutron fluence value of 1.6E18 n/cm 2, the initial RTNDT values of 23.1°F for the units, the limiting Cu content of 0.24%, and the limiting Ni content of 0.37% for the units 4.2.2 Charpy Upper Shelf Energy Section IV.A.la of Appendix G to 10 CFR Part 50 requires, in part, that the RPV beltline materials have Charpy USE in the transverse direction for base metal and along the weld for weld material of no less than 50 ft-lb (68J), unless it is demonstrated in a manner approved by the Director, Page 4 2-1

Browns Ferry Nuclear Plant, Units 1, 2, and 3

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Time Limlted Aging Analyses Office of Nuclear Reactor Regulation, that lower values of Charpy USE will ensure margins of safety against fracture equivalent to those required by Appendix G of Section Xl of the ASME Code.

By letter dated April 30, 1993, the Boiling Water Reactor Owners Group (BWROG) submitted a topical report entitled "10 CFR Part 50 Appendix G Equivalent Margins Analysis for Low Upper Shelf Energy in BWR/2 Through BWR/6 Vessels," to demonstrate that BWR RPVs could meet margins of safety against fracture equivalent to those required by Appendix G of the ASME Code Section XI for Charpy USE values less than 50 ft-lb. In a letter dated December 8, 1993, the staff concluded that the topical report demonstrated that the evaluated materials have the margins of safety against fracture equivalent to Appendix G of ASME Code Section Xl, in accordance with Appendix G of 10 CFR Part 50. In this report, the BWROG derived through statistical analysis to derive the unirradiated USE values for materials that originally did not have documented unirradiated Charpy USE values. Using these statistically derived Charpy USE values, the BWROG predicted the end-of life (40 years of operation) USE values in accordance with RG 1.99, Rev. 2. According to this RG, the decrease in USE is dependent upon the amount of copper in the material and the neutron fluence predicted for the material. The BWROG analysis determined that the minimum allowable Charpy USE in the transverse direction for base metal and along the weld for weld metal was 35 ft-lb.

General Electric (GE) performed an update to the USE equivalent margins analysis, which is documented in EPRI TR-113596, "BWR Vessel and Internals Project BWR Reactor Pressure Vessel Inspection and Flaw Evaluation Guidelines,* BWRVIP-74, September 1999. The staff review and approval of EPRI TR-113596 is documented in a letter from C I. Grimes to C. Terry dated October 18, 2001. The analysis in EPRI TR-113596 determined the reduction in the unirradiated Charpy USE resulting from neutron radiation using the methodology in RG 1.99, Revision 2. Using this methodology and a correction factor of 65% for conversion of the longitudinal properties to transverse properties, the lowest irradiated Charpy USE at 54 EFPYs for all BWR/3-6 plates is projected to be 45 ft-lb. The correction factor for specimen orientation in plates is based on NRC Branch Technical position MTEB 5-2. Using the RG methodology, the lowest irradiated Charpy USE at 54 EFPY for BWR non-Linde 80 submerged arc welds is projected to be 43 ft-lb. EPRI TR-1 13596 indicates that the percent reduction in Charpy USE for the limiting BWR/3-6 beltline plates and BWR non-Linde 80 submerged arc welds are 23.5% and 39%, respectively. Since this is a generic analysis, TVA has performed a plant-specific analysis to demonstrate that the beltline materials of the BFNP Units 1, 2, and 3 RPVs meet the criteria in the report at the end of the license renewal period.

The BFNP analysis determined the predicted percent decrease of the beltline material USE values at 1/4T and 54 EFPYs was estimated using BWRVIP-74 and RG 1.99, Revision 2. The equivalent margin analysis was performed using information presented in Tables B-4 and B-5 of EPRI TR-1 13596. RG 1.99, Revision 2, predicted percent decrease in USE for the limiting beltline plate material at the end of the license renewal period is [later] for Unit 1, 17% for Unit 2, and 16%

for Unit 3 Both predicted values of USE are less than the generic value of 23.5% reported in EPRI TR-113596. Similarly, the RG 1.99, Revision 2, predicted percent decrease in USE for limiting weld material (Electroslag Weld at both units) at the end of license renewal period is [later]

for Unit 1 and 25.5% for both Unit 2 and Unit 3, which is less than the generic value of 39%

reported in EPRI TR-1 13596. The 52 EFPYs neutron fluence at 1/4T for the limiting beltline plate and weld materials of the units is 1.6E18 n/cm 2. The Cu contents for the limiting beltline materials are 0.24 wt% for the units.

While the margin has decreased, the BFNP analysis results are acceptable because the percent decrease in USE for limiting plate and weld materials at BFNP Units 1, 2, and 3 is bounded by the corresponding generic results obtained by the equivalent margin analysis presented in EPRI TR 113596 as mentioned above. Therefore, the Charpy USE values at 52 EFPYs for the limiting plate and weld materials at BFNP Units 1, 2, and 3 are greater than the minimum allowable value of 35 ft-lb, which demonstrates that the evaluated materials have the margins of safety against Page 4 2-2

Browns Ferry Nuclear Plant, Units 1, 2, and 3 ADolication for Rpnp~wpd Opertinn Licenses Time Limited Agina Analyses fracture equivalent to Appendix G of Section XI of the ASME Code, in accordance with Appendix G of 10 CFR Part 50, throughout the license renewal period.

4.2.3 Pressure/Temperature Limits BFNP Units 1, 2, and 3 Technical Specifications, Section 3.4.9, contain P/T limit curves for heatup, cooldown, criticality, and inservice leakage and hydrostatic testing. The curves are currently calculated for the operating periods ending with 12 EFPY, 17.2 EFPY, and 13.1 EFPY for Units 1, 2, and 3 respectively.

The P/T Limit curves are calculated using Code Case N-640.

Analysis The P-T limit curves are based on the following NRC regulations and guidance:

10 CFR Part 50, Appendix G; Generic Letter (GL) 88-11, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and Its Impact on Plant Operations";

G GL 92-01, "Reactor Vessel Structural Integrity, *Revision 1; GL 92-01, Revision 1, Supplement 1; RG 1.99, Revision 2 Standard Review Plan (SRP) Section 5.3.2, "Pressure-Temperature Limits and Pressurized Thermal Shock."

Appendix G to 10 CFR Part 50 requires that P-T limit curves for the RPV be at least as conservative as those obtained by the methodology of Appendix G Section Xl of the ASME Code.

The ASME Code Appendix G methodology requires that applicants determine the ART at the end of the operating period.

TVA has performed P/T limit curve calculations (Ref 4.2.1) for the period of extended operation using methodologies base on RG 1.99 Revision 2. Use of RG 1.99 requires that an allowance for margin be included in the bounding ART value. This ensures that adequate safety margins are maintained. The calculations show that the RPVs will be in compliance with regulatory requirements and adequate safety margins can be maintained during the period of extended operation. The results of the calculation are provided in Figures [later].

Therefore, operation of the BFNP RPVs to 52 EFPY (60 calendar years) will not have an adverse affect on reactor vessel fracture toughness Disposition: Revision, 10 CFR 54.21(c)(1)(ii)

Amendments to the Technical Specifications to revise the reactor vessel P/T limit curves will be requested and implemented as current P/T curves reach their operational limits.

Text enclosed in bold border indicates applicable to Units 2 and 3 only.

Page 4 2-3

Browns Ferry Nuclear Plant, Units 1,2, and 3 Appl;cation....

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J0t*i.u iI B.1.17 XI.M17 FLOW-ACCELERATED CORROSION (FAC)

Description The FAC aging management program is an existing program implemented by an administrative procedure that is applicable to all three units. The program is based on the EPRI guidelines in NSAC-202L-R2. The program predicts, detects, and monitors wall thinning in piping, fittings, and valve bodies due to FAC in the following systems: Main Steam (System 001), Condensate (System 002), Reactor Feedwater (System 003 ), Extraction Steam (System 05), Heater Drains and Vents (System 06), Aux. Boiler (System 12).

Program activities include analyses using the predictive CHECWORKS computer code to determine critical locations, baseline inspections to determine the extent of thinning at these critical locations, and follow-up inspections to confirm the predictions. Repairs and replacements are performed as necessary.

NUREG-1801 Consistency With enhancements the FAC AMP is consistent with the ten elements of aging management program XI.M17, "Flow-Accelerated Corrosion," specified in NUREG-1801.

Enhancements Enhancements include.

1. Expanding the scope of the program to include the main steam lines to the HPCI and RCIC turbines and the turbines' exhaust lines to the Torus'

2.

Implementing the program on Unit 1.

Operating Experience Industry Wall-thinning problems in single-phase systems have occurred in feedwater and condensate systems (NRC IE Bulletin No. 87-01 and NRC Information Notices (INs) 81-28, 92-35, and 95-11), in two-phase piping in extraction steam lines (NRC INs 89-53 and 97-84), and in moisture separation reheater and feedwater heater drains (NRC INs 89-53, 91-18, 93-21, and 97-84)

Browns Ferry The TVA experience with its flow-accelerated corrosion aging management program activities has shown that the program can determine susceptible locations for flow-accelerated corrosion, predict the component degradation, and detect the wall thinning in piping and valves due to flow accelerated corrosion In addition, the program provides for reevaluation, repair or replacement for locations where calculations indicate an area will reach minimum allowable thickness before the next inspection. When FAC problems have been identified, corrective actions have been taken to prevent recurrence. For example.

I Extraction steam, heater drain, and heater vent lines have experienced wall thinning due to FAC and this piping has been replaced.

Conclusion Based on the use of industry guidelines, NRC requirements, and BFN operating experience, there is reasonable assurance that, with the following commitments, the BFN FAC program will continue to adequately manage the aging effects (loss of material) due to flow accelerated corrosion such that the piping and components within the scope of license renewal will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation Text enclosed in bold border indicates applicable to Units 2 and 3 only.

Page B-1

Browns Ferry Nuclear Plant, Units 1,2, and 3 f-%

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Commitment 1 Prior to entering the period of extended operation, the scope of the FAC program will be expanded to include the main steam lines to the HPCI and RCIC turbines and the turbines' exhaust lines to the Torus.

Commitment 2 Prior to restart of Unit 1 from its current outage, the FAC program will be implemented on Unit 1.

Page B-2

ML023240328

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System Description

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