ML12334A654

Official Exhibit - NYS00147B-00-BD01 - NUREG-1801, Rev. 2, Generic Aging Lessons Learned (GALL) Report, Final Report (December 2010) (GALL Rev 2)

ML12334A654
Person / Time
Site:	Indian Point
Issue date:	12/15/2011
From:	Office of Nuclear Reactor Regulation
To:	Atomic Safety and Licensing Board Panel
SECY RAS
References
RAS 21545, 50-247-LR, 50-286-LR, ASLBP 07-858-03-LR-BD01 NUREG-1801, Rev 2
Download: ML12334A654 (189)

Text

United States Nuclear Regulatory Commission Official Hearing Exhibit NYS00147B Entergy Nuclear Operations, Inc.

In the Matter of:

(Indian Point Nuclear Generating Units 2 and 3) Submitted: December 15, 2011 Ct"tp-f'REGU(.q" ASLBP #: 07-858-03-LR-BD01

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Docket #: 05000247 l 05000286

• 0 Exhibit #: NYS00147B-00-BD01 Identified: 10/15/2012

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Admitted: 10/15/2012 Withdrawn:

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"'1)..: 0' f Rejected: Stricken:

.. *** .. <I- Other:

oCD IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C')

CD A1 Reactor Vessel (BWR) 3 0-CD I\) Structure and/or Aging Effect/ Further a

...... Item Link Material Environment Aging Management Program (AMP) a Component Mechanism Evaluation IV.A1.RP-157 IV.A1- Reactor Vessel: Steel (with Reactor coolant Loss of material Chapter XI.M2, "Water Chemistry," No 8(RP-25) flanges; nozzles; stainless due to pitting and and penetrations; safe steel or crevice corrosion Chapter XI.M32, "One-Time ends; vessel nickel-alloy Inspection" shells, heads and cladding);

welds stainless steel; nickel alloy IV.A1.RP-50 IV.A1- Top head Steel Reactor coolant Loss of material Chapter XI.M2, "Water Chemistry," No 11 (R-59) enclosure due to general, and (without pitting, and Chapter XI.M32, "One-Time cladding): top crevice corrosion Inspection" head; nozzles (vent, top head spray or RCIC, and spare)

IV.A1.RP-51 IV.A1-9(R- Top head High- Air with reactor Cracking Chapter XI.M3, "Reactor Head No

60) enclosure: strength, coolant leakage due to stress Closure Stud Bolting" closure studs and low-alloy corrosion nuts steel cracking, intergranular stress corrosion cracking IV.A1.RP-201 Top head High- Air with reactor Cumulative Fatigue is a time-limited aging analysis Yes, TLAA enclosure: strength, coolant leakage fatigue damage (TLAA) to be evaluated for the period z closure studs and low-alloy due to fatigue of extended operation. See the SRP, c nuts steel Section 4.3 "Metal Fatigue," for

u acceptable methods for meeting the m

o...... requirements of 10 CFR 54.21 (c)(1) .

o 00 G) a a

o

u CD o <

--" i\J 0)

CD 0

1 o

o CD o

z IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM c

U A1 Reactor Vessel (BWR) m G)

I 00 a Structure and/or Aging Effect/ Further Item Link Material Environment Aging Management Program (AMP)

Component Mechanism Evaluation

u CD

< IV.A1.RP-165 Top head High- Air with reactor Loss of material Chapter XI.M3, "Reactor Head No I\)

enclosure: strength, coolant leakage due to general, Closure Stud Bolting" closure studs and low-alloy pitting, and nuts steel crevice corrosion, or wear IV.A1.R-61 IV.A1- Top head Stainless Air with reactor Cracking A plant-specific aging management Yes, plant-10(R-61) enclosure: vessel steel; coolant leakage due to stress program is to be evaluated because specific flange leak nickel alloy (Internal); or corrosion existing programs may not be capable detection line reactor coolant cracking, of mitigating or detecting crack intergranular initiation and growth due to SCC in the stress corrosion vessel flange leak detection line cracking

<: IV.A1.RP-227 IV.A1- Vessel shell Steel (with Reactor coolant Loss of fracture Chapter XI.M31, "Reactor Vessel Yes, plant

...... 14(R-63) (including or without and neutron flux toughness Surveillance" specific or I

(J) applicable cladding) due to neutron integrated beltline) irradiation surveillance components: embrittlement program shell; shell plates or forgings; shell welds; nozzle plates or forgings; nozzle welds IV.A1.R-64 IV.A1- Vessel shell: Stainless Reactor coolant Cracking Chapter XI.M4, "BWR VessellD No 12(R-64) attachment welds steel; due to stress Attachment Welds," and nickel alloy corrosion Chapter XI.M2, "Water Chemistry" cracking, intergranular 0 stress corrosion CD C') cracking CD 0

G) 3 0-CD 0 I\)

0 a 0

--" a 0)

CD 0

I 0

0 CD

oCD IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C')

CD A1 Reactor Vessel (BWR) 3 0-CD I\) Structure and!or Aging Effect! Further a Item Link Material Environment Aging Management Program (AMP) a Component Mechanism Evaluation IV.A1.R-62 IV.A1- Vessel shell: Steel (with Reactor coolant Loss of fracture Neutron irradiation embrittiement is a Yes, TLAA 13(R-62) intermediate or without and neutron flux toughness time-dependent aging mechanism beltline shell; stainless due to neutron evaluated for extended operation for beltline welds steel irradiation all ferritic materials that have a 2

cladding) embrittlement neutron fluence >1 E17 n!cm (E

>1 MeV) at the end of the period of extended operation. Aspects may involve a TLAA.

In accordance with approved BWRVIP-74, the TLAA evaluates the impact of neutron embrittlement on:

(a) adjusted reference temperature

<: values used for calculation of the

...... plant's pressure-temperature limits, (b)

-J I

need for inservice inspection of circumferential welds, and (c) Charpy upper shelf energy or the equivalent margins analyses performed in accordance with 10 CFR Part 50, Appendix G. Additionally, the applicant is to monitor axial beltline weld embrittiement. One acceptable method is to determine that the mean RTNDT of the axial beltline welds at the end of the extended period of operation is less than the value z specified by the staff in its March 7, c

u 2000 letter (ADAMS ML031430372).

m G) See the Standard Review Plan, I

Section 4.2 "Reactor Vessel Neutron 0 00 G) a Embrittlement" for acceptable methods

U for meeting the requirements of 10 0

0 CD CFR 54.21 (c).

0 <

--" i\J 0)

CD 0

I 0

0 CD N

A2. REACTOR VESSEL (PRESSURIZED WATER REACTOR)

Systems, Structures, and Components This section addresses the pressurized water reactor (PWR) vessel pressure boundary and consists of the vessel shell and flanges, the top closure head and bottom head, the control rod drive (CRD) mechanism housings, nozzles (including safe ends) for reactor coolant inlet and outlet lines and safety injection, and penetrations through either the closure head or bottom head domes for instrumentation and leakage monitoring tubes. Attachments to the vessel such as core support pads, as well as pressure vessel support and attachment welds, are also included in the table. Based on Regulatory Guide 1.26, "Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants," all systems, structures, and components that comprise the reactor coolant system are governed by Group A Quality Standards.

Common miscellaneous material/environment combinations where aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation are included in IV.E.

System Interfaces The systems that interface with the PWR reactor vessel include the reactor vessel internals (lV.B2, IV.B3, and IV.B4, respectively, for Westinghouse, Combustion Engineering, and Babcock and Wilcox designs), the reactor coolant system and connected lines (lV.C2), and the emergency core cooling system (V.D1).

December 201 0 IV A2-1 NUREG-1801, Rev. 2 OAGI0001390_00193

z IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM c

U A2 Reactor Vessel (PWR) m G)

I 00 Structure a Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.A2.RP-154 IV.A2- Bottom- Stainless Reactor coolant Cracking A plant-specific aging management Yes, plant-I\)

1 (RP-13) mounted steel due to stress program is to be evaluated specific instrument corrosion guide tube cracking (external to bottom head)

IV.A2.RP-52 IV.A2-2(R- Closure head: High- Air with reactor Cracking Chapter XI.M3, "Reactor Head No

71) stud assembly strength, coolant leakage due to stress Closure Stud Bolting" low-alloy corrosion steel cracking IV.A2.RP-54 IV.A2-4(R- Closure head: High- Air with reactor Cumulative Fatigue is a time-limited aging Yes, TLAA

73) stud assembly strength, coolant leakage fatigue damage analysis (TLAA) to be evaluated for

<: low-alloy due to fatigue the period of extended operation. See I\) steel the SRP, Section 4.3 "Metal Fatigue,"

I I\) for acceptable methods for meeting the requirements of 10 CFR 54.21 (c)(1).

IV.A2.RP-53 IV.A2-3(R- Closure head: High- Air with reactor Loss of material Chapter XI.M3, "Reactor Head No

72) stud assembly strength, coolant leakage due to general, Closure Stud Bolting" low-alloy pitting, and steel crevice corrosion, orwear IV.A2.R-74 IV.A2-5(R- Closure head: Stainless Air with reactor Cracking A plant-specific aging management Yes, plant-

74) vessel flange steel coolant leakage due to stress program is to be evaluated because specific leak detection (Internal); or corrosion existing programs may not be capable line reactor coolant cracking of mitigating or detecting crack initiation and growth due to SCC in 0 the vessel flange leak detection line CD C')

CD IV.A2.R-80 IV.A2-8(R- Control rod Stainless Air (with reactor Loss of preload Chapter XI.M18, "Bolting Integrity" No 0 80) drive head steel coolant due to thermal

>> 3 0-G) ...,

CD penetration: leakage) effects, gasket 0 I\) Flange bolting creep, and self-0 a 0 loosening

--" a 0)

CD 0

I 0

0 CD

.j::>.

oCD IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C')

CD A2 Reactor Vessel (PWR) 3 0-CD Structure I\) Aging Effect/ Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.A2.R-78 IV.A2-6(R- Control rod Stainless Air with reactor Cracking Chapter XI.M18, "Bolting Integrity" No

78) drive head steel coolant leakage due to stress penetration: corrosion flange bolting cracking IV.A2.R-79 IV.A2-7(R- Control rod Stainless Air with reactor Loss of material Chapter XI.M18, "Bolting Integrity" No

79) drive head steel coolant leakage due to wear penetration:

flange bolting IV.A2.RP-186 IV.A2-9(R- Control rod Nickel alloy Reactor coolant Cracking Chapter XI.M1, "ASME Section XI No

75) drive head due to primary Inservice Inspection, Subsections penetration: water stress IWB, IWC, and IWD" for Class 1 nozzle welds corrosion components, and

<: cracking Chapter XI.M2, "Water Chemistry,"

I\) and I

CJ.) Chapter XI.M11 B, "Cracking of Nickel-Alloy Components and Loss of Material Due to Boric Acid-Induced Corrosion in RCPB Components (PWRs Only)"

IV.A2.R-77 IV.A2- Control rod Cast Reactor coolant Loss of fracture Chapter XI.M12, "Thermal Aging No 10(R-77) drive head austenitic >250°C toughness Embrittlement of Cast Austenitic penetration: stainless (>482°F) due to thermal Stainless Steel (CASS)"

pressure steel aging housing embrittlement IV.A2.RP-55 IV.A2- Control rod Stainless Reactor coolant Cracking Chapter XI.M1, "ASME Section XI No z 11(R-76) drive head steel; nickel due to stress Inservice Inspection, Subsections c

u penetration

alloy corrosion IWB, IWC, and IWD" for Class 1 m pressure cracking, primary components, and G)

I housing water stress Chapter XI.M2, "Water Chemistry" 0 a 00 corrosion G) cracking

U 0 CD 0

0  :<

0)

CD 0

I 0

0 CD (J1

z IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM c

U A2 Reactor Vessel (PWR) m G)

I 00 Structure a Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.A2.RP-S7 IV.A2- Core support Nickel alloy Reactor coolant Cracking Chapter XI.M1, "ASME Section XI No I\)

12(R-88) pads; core due to primary Inservice Inspection, Subsections guide lugs water stress IWB, IWC, and IWD" for Class 1 corrosion components, and cracking Chapter XI.M2, "Water Chemistry" IV.A2.R-17 IV.A2- External Steel Air with borated Loss of material Chapter XI.M1 0, "Boric Acid No 13(R-17) surfaces water leakage due to boric acid Corrosion" corrosion IV.A2.RP-379 IV.A2- External Steel Air with borated Loss of material Chapter XI.M1 0, "Boric Acid No 13(R-17) surfaces: water leakage due to boric acid Corrosion," and reactor vessel corrosion Chapter XI.M11 B, "Cracking of top head and Nickel-Alloy Components and Loss of

<: bottom head Material Due to Boric Acid-Induced I\) Corrosion in RCPB Components I

.j:>.

(PWRs Only)"

IV.A2.RP-28 IV.A2- Flanges; Steel (with Reactor coolant Loss of material Chapter XI.M2, "Water Chemistry" No 14(RP-28) nozzles; stainless due to pitting and penetrations; steel or crevice corrosion pressure nickel-alloy housings; safe cladding);

ends; vessel stainless shells, heads steel; nickel welds alloy 0

CD C')

CD 0

G) 3 0-CD 0 I\)

0 a 0

--" a 0)

CD 0

I 0

0 CD (J)

oCD IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C')

CD A2 Reactor Vessel (PWR) 3 0-CD Structure I\) Aging Effect/ Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.A2.RP-234 IV.A2- Nozzle safe Stainless Reactor coolant Cracking Chapter XI.M1, "ASME Section XI No 1S(R-83) ends and steel; nickel due to stress Inservice Inspection, Subsections welds: inlet; alloy welds corrosion IWB, IWC, and IWD" for Class 1 outlet; safety and/or cracking, primary components, and injection buttering water stress Chapter XI.M2, "Water Chemistry,"

corrosion and cracking Chapter XI.M11 B, "Cracking of Nickel-Alloy Components and Loss of Material Due to Boric Acid-Induced Corrosion in RCPB Components (PWRs Only)" for nickel alloy components

<: IV.A2.RP-228 IV.A2- Nozzles: inlet; Steel (with Reactor coolant Loss of fracture Chapter XI.M31, "Reactor Vessel Yes, plant I\)

17(R-82) outlet; safety or without and neutron flux toughness Surveillance" specific or I

0"1 injection cladding) due to neutron integrated irradiation surveillance embrittlement program z

c

u m

G)

I 0 a 00 G)

U 0 CD 0

0  :<

0)

CD 0

I 0

0 CD

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z IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM c

U A2 Reactor Vessel (PWR) m G)

I 00 Structure a Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.A2.R-81 IV.A2- Nozzles: inlet; Steel (with Reactor coolant Loss of fracture Neutron irradiation embrittlement is a Yes, TLAA I\)

16(R-81) outlet; safety stainless and neutron flux toughness TLAA evaluated for extended injection steel or due to neutron operation for all ferritic materials with nickel-alloy irradiation a neutron fluence greater than 2

cladding) embrittlement 1E17 n!cm (E >1 MeV) at the end of the period of extended operation.

The TLAA is to evaluate the impact of neutron embrittlement on: (a) the RTPTS value based on the requirements in 10 CFR 50.61 , (b) the adjusted reference temperature values used for calculation of the

<: plant's pressure-temperature limits, I\)

and (c) the Charpy upper shelf energy I

(J) or the equivalent margins analyses performed in accordance with 10 CFR Part 50, Appendix G requirements.

The applicant may choose to demonstrate that the materials in the inlet, outlet, and safety injection nozzles are not controlling for the TLAA evaluations.

IV.A2.R-90 IV.A2- Penetrations: Nickel alloy Reactor coolant Cracking Chapter XI.M1, "ASME Section XI No 18(R-90) head vent pipe due to primary Inservice Inspection, Subsections (top head); water stress IWB, IWC, and IWD" for Class 1 instrument corrosion components, and tubes (top cracking Chapter XI.M2, "Water Chemistry,"

head) and 0

CD C')

Chapter XI.M11 B, "Cracking of Nickel-0 CD Alloy Components and Loss of G) 3 0- Material Due to Boric Acid-Induced CD Corrosion in RCPB Components 0 I\)

0 a (PWRs Only)"

0

--" a 0)

CD 0

I 0

0 CD CD

oCD IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C')

CD A2 Reactor Vessel (PWR) 3 0-CD Structure I\) Aging Effect/ Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.A2.RP-59 IV.A2- Penetrations: Nickel alloy Reactor coolant Cracking Chapter XI.M1, "ASME Section XI No 19(R-89) instrument due to primary Inservice Inspection, Subsections tubes (bottom water stress IWB, IWC, and IWD" for Class 1 head) corrosion components, and cracking Chapter XI.M2, "Water Chemistry,"

and Chapter XI.M11 B, "Cracking of Nickel-Alloy Components and Loss of Material Due to Boric Acid-Induced Corrosion in RCPB Components (PWRs Only)"

IV.A2.R-70 IV.A2- Pressure vessel Steel Air - indoor, Cumulative Fatigue is a time-limited aging Yes, TLAA

<: 20(R-70) support skirt uncontrolled fatigue damage analysis (TLAA) to be evaluated for I\)

and attachment due to fatigue the period of extended operation. See I

-J welds the SRP, Section 4.3 "Metal Fatigue,"

for acceptable methods for meeting the requirements of 10 CFR 54.21 (c)(1).

IV.A2.R-219 IV.A2- Reactor vessel Steel (with Reactor coolant Cumulative Fatigue is a TLAA evaluated for the Yes, TLAA 21 (R-219) components: or without fatigue damage period of extended operation, and for flanges; nickel-alloy due to fatigue Class 1 components environmental nozzles; or stainless effects on fatigue are to be penetrations; steel addressed. (See SRP, Sec 4.3 "Metal pressure cladding); Fatigue," for acceptable methods to housings; safe stainless comply with 10 CFR 54.21 (c)(1))

z ends; thermal steel; nickel c sleeves; vessel alloy

u m shells, heads G)

I and welds 0 a 00 G)

U 0 CD 0

0  :<

0)

CD 0

I 0

0 CD CD

z IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM c

U A2 Reactor Vessel (PWR) m G)

I 00 Structure a Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.A2.R-85 IV.A2- Vessel shell: SA508-C12 Reactor coolant Crack growth Growth of intergranular separations Yes, TLAA I\)

22(R-85) upper shell; forgings clad due to cyclic (underclad cracks) in low-alloy steel intermediate (with loading forging heat affected zone under shell; lower stainless austenitic stainless steel cladding is a shell (including steel) using time-limited aging analysis (TLAA) to beltline welds) a high-heat- be evaluated for the period of input extended operation for all the SA 508-welding CI 2 forgings where the cladding was process deposited with a high heat input welding process. The methodology for evaluating an underclad flaw is in accordance with the current well-

<: established flaw evaluation procedure I\)

and criterion in the ASME Section XI 00 I

Code. See the Standard Review Plan, Section 4.7, "Other Plant-Specific Time-Limited Aging Analysis," for generic guidance for meeting the requirements of 10 CFR 54.21 (c).

IV.A2.RP-229 IV.A2- Vessel shell: Steel (with Reactor coolant Loss of fracture Chapter XI.M31, "Reactor Vessel Yes, plant 24(R-86) upper shell; or without and neutron flux toughness Surveillance" specific or intermediate cladding) due to neutron integrated shell; lower irradiation surveillance shell (including embrittlement program beltline welds) 0 CD C')

CD 0

G) 3 0-CD 0 I\)

0 a 0

--" a 0)

CD 0

I 0

0 N

0 0

oCD IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C')

CD A2 Reactor Vessel (PWR) 3 0-CD Structure I\) Aging Effect! Further a Item Link and!or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.A2.R-84 IV.A2- Vessel shell: Steel (with Reactor coolant Loss of fracture Neutron irradiation embrittlement is a Yes, TLAA 23(R-84) upper shell; stainless and neutron flux toughness TLAA evaluated for extended intermediate steel or due to neutron operation for all ferritic materials with shell; lower nickel-alloy irradiation a neutron fluence greater than 2

shell (including cladding) embrittlement 1E17 n!cm (E >1 MeV) at the end of beltline welds) the period of extended operation. The TLAA is to evaluate the impact of neutron embrittlement on: (a) the RTPTS value based on the requirements in 10 CFR 50.61 , (b) the adjusted reference temperature values used for calculation of the

<: plant's pressure-temperature limits, I\)

and (c) the Charpy upper shelf energy I

CD or the equivalent margins analyses performed in accordance with 10 CFR Part 50, Appendix G requirements.

See the Standard Review Plan, Section 4.2 "Reactor Vessel Neutron Embrittlement" for acceptable methods for meeting the requirements of 10 CFR 54.21 (c).

IV.A2.R-87 IV.A2- Vessel shell: Steel Reactor coolant Loss of material Chapter XI.M1, "ASME Section XI No 25(R-87) vessel flange due to wear Inservice Inspection, Subsections IWB, IWC, and IWD" for Class 1 z components c

u m

G)

I 0 a 00 G)

U 0 CD 0

0  :<

0)

CD 0

I 0

0 N

0

B1. REACTOR VESSEL INTERNALS (BOILING WATER REACTOR)

Systems, Structures, and Components This section addresses the boiling water reactor (BWR) vessel internals and consists of the core shroud (including repairs) and core plate, the top guide, feedwater spargers, core spray lines and spargers, jet pump assemblies, fuel supports and control rod drive (CRO), and instrument housings, such as the intermediate range monitor (lRM) dry tubes, the low power range monitor (LPRM) dry tubes, and the source range monitor (SRM) dry tubes. Based on Regulatory Guide 1.26, "Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants," all structures and components that comprise the reactor vessel are governed by Group A or B Quality Standards.

Common miscellaneous material/environment combinations where aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation are included in IV.E.

System Interfaces The systems that interface with the reactor vessel internals include the reactor pressure vessel (lV.A 1) and the reactor coolant pressure boundary (lV.C1).

December 201 0 IV 81-1 NUREG-1801, Rev. 2 OAGI0001390_00202

z IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM c

U B1 Reactor Vessel Internals (BWR) m G)

I 00 Structure a Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.B1.R-92 IV.B1-1 (R- Core shroud Stainless Reactor coolant Cracking Chapter XI.M9, "BWR Vessel No I\)

92) (including steel and neutron flux due to stress Internals" for core shroud, and repairs) and corrosion Chapter XI.M2, "Water Chemistry" core plate: core cracking, shroud (upper, intergranular central, lower) stress corrosion cracking, irradiation-assisted stress corrosion cracking IV.B1.R-96 IV.B1-2(R- Core shroud Nickel alloy Reactor coolant Cracking Chapter XI.M9, "BWR Vessel No

96) (including and neutron flux due to stress Internals" for shroud support, and

<: repairs) and corrosion Chapter XI.M2, "Water Chemistry" III

...... core plate: cracking, I

I\) shroud support intergranular structure stress corrosion (shroud support cracking, cylinder, shroud irradiation-support plate, assisted stress shroud support corrosion cracking legs)

IV.B1.R-95 IV.B1-4(R- Core shroud Nickel alloy Reactor coolant Cracking Chapter XI.M1, "ASME Section XI No

95) and core plate: and neutron flux due to stress Inservice Inspection, Subsections access hole corrosion IWB, IWC, and IWD" for Class 1 cover cracking, components, and (mechanical) intergranular Chapter XI.M2, "Water Chemistry" stress corrosion 0 cracking, CD C') irradiation-CD assisted stress 0

G) 3 0- corrosion cracking CD 0 I\)

0 a 0

--" a 0)

CD 0

I 0

0 N

0 0)

oCD IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C')

CD B1 Reactor Vessel Internals (BWR) 3 0-CD Structure I\) Aging Effect/ Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.B1.R-94 IV.B1-5(R- Core shroud Nickel alloy Reactor coolant Cracking Chapter XI.M1, "ASME Section XI No

94) and core plate: and neutron flux due to stress Inservice Inspection, Subsections access hole corrosion IWB, IWC, and IWD" for Class 1 cover (welded) cracking, components, and intergranular Chapter XI.M2, "Water Chemistry" stress corrosion Because cracking initiated in crevice cracking, regions is not amenable to visual irradiation- inspection, for BWRs with a crevice in assisted stress the access hole covers, an corrosion cracking augmented inspection is to include ultrasonic testing (UT) or other demonstrated acceptable inspection

<: of cover welds.

III

...... IV.B1.R-93 IV.B1-6(R- Core shroud Stainless Reactor coolant Cracking Chapter XI.M9, "BWR Vessel No I

CJ.) 93) and core plate: steel and neutron flux due to stress Internals" for core plate, and core plate and corrosion Chapter XI.M2, "Water Chemistry" plate bolts cracking, (used in early intergranular BWRs) stress corrosion cracking, irradiation-assisted stress corrosion cracking IV.B1.R-97 IV.B1-3(R- Core shroud Stainless Reactor coolant Cracking Chapter XI.M9, "BWR Vessel No

97) and core plate: steel and neutron flux due to stress Internals" for the LPCI coupling, and z LPCI coupling corrosion Chapter XI.M2, "Water Chemistry" c cracking,

u m intergranular G)

I stress corrosion 0 a 00 cracking, G) irradiation-0

U assisted stress CD 0 corrosion cracking 0

<

0)

CD 0

I 0

0 N

0

.j::>.

z IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM c

U B1 Reactor Vessel Internals (BWR) m G)

I 00 Structure a Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.B1.R-99 IV.B1-7(R- Core spray Stainless Reactor coolant Cracking Chapter XI.M9, "BWR Vessel No I\)

99) lines and steel and neutron flux due to stress Internals" for core spray internals, and spargers: core corrosion Chapter XI.M2, "Water Chemistry" spray lines cracking, (headers); intergranular spray rings; stress corrosion spray nozzles; cracking, thermal sleeves irradiation-assisted stress corrosion cracking IV.B1.R-104 IV.B1-8(R- Fuel supports Stainless Reactor coolant Cracking Chapter XI.M9, "BWR Vessel No 104) and control rod steel due to stress Internals" for lower plenum, and

<: drive corrosion Chapter XI.M2, "Water Chemistry" III

...... assemblies: cracking, I

.j:>. control rod drive intergranular housing stress corrosion cracking IV.B1.RP-220 IV.B1-9(R- Fuel supports Cast Reactor coolant Loss of fracture Chapter XI.M9, "BWR Vessel No 103) and control rod austenitic >2S0°C toughness Internals" drive stainless (>482°F) and due to thermal assemblies: steel neutron flux aging, neutron orificed fuel irradiation support embrittiement IV.B1.R-10S IV.B1- Instrumentation: Stainless Reactor coolant Cracking Chapter XI.M9, "BWR Vessel No 10(R-1 05) Intermediate steel and neutron flux due to stress Internals" for lower plenum, and range monitor corrosion Chapter XI.M2, "Water Chemistry" (lRM) dry tubes; cracking, 0 source range intergranular CD C') monitor (SRM) stress corrosion CD 0 dry tubes; cracking, G) 3 0- incore neutron irradiation-CD 0 I\) flux monitor assisted stress 0 a guide tubes corrosion cracking 0

--" a 0)

CD 0

I 0

0 N

0 (J1

oCD IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C')

CD B1 Reactor Vessel Internals (BWR) 3 0-CD Structure I\) Aging Effect/ Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.B1.RP-219 IV.B1- Jet pump Cast Reactor coolant Loss of fracture Chapter XI.M9, "BWR Vessel No 11 (R-1 01) assemblies: austenitic >250°C toughness Internals" castings stainless (>482°F) and due to thermal steel neutron flux aging, neutron irradiation embrittiement IV.B1.R-100 IV.B1- Jet pump Stainless Reactor coolant Cracking Chapter XI.M9, "BWR Vessel No 13(R-100) assemblies: steel; nickel and neutron flux due to stress Internals" for jet pump assembly, and thermal sleeve; alloy corrosion Chapter XI.M2, "Water Chemistry" inlet header; cracking, riser brace arm; intergranular holddown stress corrosion

<: beams; inlet cracking, III

...... elbow; mixing irradiation-I 0"1 assembly; assisted stress diffuser corrosion cracking castings IV.B1.R-53 IV.B1- Reactor vessel Stainless Reactor coolant Cumulative Fatigue is a time-limited aging Yes, TLAA 14(R-53) internal steel; nickel fatigue damage analysis (TLAA) to be evaluated for components alloy due to fatigue the period of extended operation. See the SRP, Section 4.3 "Metal Fatigue,"

for acceptable methods for meeting the requirements of 10 CFR 54.21 (c)(1).

z c

u m

G)

I 0 a 00 G)

U 0 CD 0

0  :<

0)

CD 0

I 0

0 N

0 (J)

z IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM c

U B1 Reactor Vessel Internals (BWR) m G)

I 00 Structure a Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.B1.RP-182 Reactor vessel PH Reactor coolant Loss of fracture Chapter XI.M9, "BWR Vessel No I\)

internals martensitic >2S0°C toughness Internals" components stainless (>482°F) and due to thermal steel (17- neutron flux aging, neutron 4PH and irradiation 1S-SPH); embrittiement martensitic stainless steel (SS 403,410, 431, etc.)

IV.B1.RP-26 IV.B1- Reactor vessel Stainless Reactor coolant Loss of material Chapter XI.M1, "ASME Section XI No

<: 1S(RP-26) internals steel; nickel due to pitting and Inservice Inspection, Subsections III

...... components alloy crevice corrosion IWB, IWC, and IWD" for Class 1 I

(J) components, and Chapter XI.M2, "Water Chemistry" IV.B1.RP-381 Reactor vessel X-7S0 alloy Reactor coolant Cracking Chapter XI.M9, "BWR Vessel No internals and neutron flux due to Internals" for core plate, and components intergranular Chapter XI.M2, "Water Chemistry" stress corrosion cracking IV.B1.RP-200 Reactor vessel X-7S0 alloy Reactor coolant Loss of fracture Chapter XI.M9, "BWR Vessel No internals and neutron flux toughness Internals" components due to neutron irradiation embrittiement IV.B1.RP-377 Reactor vessel Stainless Reactor coolant Loss of material Chapter XI.M9, "BWR Vessel No 0 internals steel due to wear Internals" CD C') components:

CD 0 Jet pump G) 3 0-wedge surface CD 0 I\)

0 a 0

--" a 0)

CD 0

I 0

0 N

0

-....J

oCD IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C')

CD B1 Reactor Vessel Internals (BWR) 3 0-CD Structure I\) Aging Effect/ Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.B1.RP-155 IV.B1- Steam dryers Stainless Reactor coolant Cracking Chapter XI.M9, "BWR Vessel No 16(RP-18) steel due to flow- Internals" for steam dryer induced vibration IV.B1.R-98 IV.B1- Top guide Stainless Reactor coolant Cracking Chapter XI.M9, "BWR Vessel No 17(R-98) steel and neutron flux due to stress Internals" for top guide, and corrosion Chapter XI.M2, "Water Chemistry" cracking, intergranular stress corrosion cracking, irradiation-assisted stress corrosion cracking z

c

u m

G)

I 0 a 00 G)

U 0 CD 0

0  :<

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CD 0

I 0

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0 CD

82. REACTOR VESSEL INTERNALS (PWR) - WESTINGHOUSE Systems, Structures, and Components This section addresses the Westinghouse pressurized water reactor (PWR) vessel internals and consists of the upper internals assembly, the control rod guide tube assemblies, the core barrel, the baffle/former assembly, the lower internal assembly, and the instrumentation support structures. Based on Regulatory Guide 1.26, "Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants," all structures and components that comprise the reactor vessel are governed by Group A or B Quality Standards.

Common miscellaneous material/environment combinations where aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation are included in IV.E.

System Interfaces The systems that interface with the reactor vessel internals include the reactor pressure vessel (lV.A2).

Inspection Plan An applicant will submit an inspection plan for reactor internals to the NRC for review and approval with the application for license renewal in accordance with Chapter XI.M16A, "PWR Vessel Internals."

December 201 0 IV 82-1 NUREG-1801, Rev. 2 OAGI0001390_00209

z IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM c

U B2 Reactor Vessel Internals (PWR) - Westinghouse m

G),

00 Structure a Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.B2.RP-300 IV.B2- Alignment and Stainless Reactor coolant Loss of preload Chapter XI.M16A, "PWR Vessel No I\)

33(R-108) interfacing steel and neutron flux due to thermal Internals" components: and irradiation Primary components (identified in the internals hold enhanced stress "Structure and Components" column) down spring relaxation; no Expansion components loss of material due to wear IV.B2.RP-301 IV.B2- Alignment and Stainless Reactor coolant Cracking 'Chapter XI.M2, "Water Chemistry," No 40(R-112) interfacing steel and neutron flux due to stress and components: corrosion cracking Chapter XI.M16A, "PWR Vessel upper core Internals" plate alignment Existing Program components

<: pins (identified in the "Structure and III Components" column)

I\)

I\) no Expansion components IV.B2.RP-299 IV.B2- Alignment and Stainless Reactor coolant Loss of material Chapter XI.M16A, "PWR Vessel No 34(R-11S) interfacing steel and neutron flux due to wear Internals" components: Existing Program components upper core (identified in the "Structure and plate alignment Components" column) pins no Expansion components IV.B2.RP-271 IV.B2- Baffle-to-former Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No 10(R-12S) assembly: steel and neutron flux due to irradiation- and accessible assisted stress Chapter XI.M16A, "PWR Vessel baffle-to-former corrosion cracking Internals" bolts and fatigue Primary components (identified in the "Structure and Components" column) 0 (for Expansion components see AMR CD C') Items IV.B2.RP-273 and IV.B2.RP-CD 0 286)

G) 3 0-CD 0 I\)

0 a 0

--" a 0)

CD 0

I 0

0 N

0

oCD IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C')

CD B2 Reactor Vessel Internals (PWR) - Westinghouse 3

0-CD Structure I\) Aging Effect/ Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.B2.RP-272 IV.B2-6(R- Baffle-to-former Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No 128) assembly: steel and neutron flux toughness Internals" accessible due to neutron Primary components (identified in the baffle-to-former irradiation "Structure and Components" column) bolts embrittiement; (for Expansion components see AMR change in Items IV.B2.RP-274 and IV.B2.RP-dimension 287) due to void swelling; loss of preload due to thermal and irradiation

<: enhanced stress III I\)

relaxation I

CJ.) IV.B2.RP-270 IV.B2-1 (R- Baffle-to-former Stainless Reactor coolant Change in Chapter XI.M16A, "PWR Vessel No 124) assembly: steel and neutron flux dimension Internals" baffle and due to void Primary components (identified in the former plates swelling "Structure and Components" column) no Expansion components IV.B2.RP-27S IV.B2-6(R- Baffle-to-former Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No 128) assembly: steel and neutron flux due to irradiation- and baffle-edge assisted stress Chapter XI.M16A, "PWR Vessel bolts (all plants corrosion cracking Internals" with baffle-edge and fatigue Primary components (identified in the bolts) "Structure and Components" column) z no Expansion components c

u m

G)

I 0 a 00 G)

U 0 CD 0

0  :<

0)

CD 0

I 0

0 N

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U B2 Reactor Vessel Internals (PWR) - Westinghouse m

G)

I 00 Structure a Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.B2.RP-3S4 Baffle-to-former Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No I\)

assembly: steel and neutron flux toughness Internals" baffle-edge due to neutron Primary components (identified in the bolts (all plants irradiation "Structure and Components" column) with baffle-edge embrittiement; no Expansion components bolts) change in dimension due to void swelling; loss of preload due to thermal and irradiation

<: enhanced stress III I\)

relaxation I

.j:>. IV.B2.RP-273 IV.B2- Baffle-to-former Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No 10(R-12S) assembly: steel and neutron flux due to irradiation- and barrel-to-former assisted stress Chapter XI.M16A, "PWR Vessel bolts corrosion cracking Internals" and fatigue Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Item IV.B2.RP-271) 0 CD C')

CD 0

G) 3 0-CD 0 I\)

0 a 0

--" a 0)

CD 0

I 0

0 N

N

oCD IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C')

CD B2 Reactor Vessel Internals (PWR) - Westinghouse 3

0-CD Structure I\) Aging Effect/ Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.B2.RP-274 IV.B2-6(R- Baffle-to-former Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No 128) assembly: steel and neutron flux toughness Internals" barrel-to-former due to neutron Expansion components (identified in bolts irradiation the "Structure and Components" embrittiement; column) change in (for Primary components see AMR dimension Item IV.B2.RP-272) due to void swelling; loss of preload due to thermal and irradiation

<: enhanced stress III I\)

relaxation I

0"1 IV.B2.RP-284 IV.B2- Bottom Stainless Reactor coolant Loss of material Chapter XI.M16A, "PWR Vessel No 12(R-143) mounted steel (with and neutron flux due to wear Internals" instrument or without Existing Program components system: flux chrome (identified in the "Structure and thimble tubes plating) Components" column)

No expansion components; and Chapter XI.M37, "Flux Thimble Tube Inspection" IV.B2.RP-293 IV.B2- Bottom- Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No 24(R-138) mounted steel and neutron flux due to fatigue and instrumentation Chapter XI.M16A, "PWR Vessel z system: bottom- Internals" c mounted Expansion components (identified in

u m instrumentation the "Structure and Components" G)

I (BMI) column column) 0 a 00 bodies (for Primary components see AMR G) Item IV.B2.RP-298)

U 0 CD 0

0  :<

0)

CD 0

I 0

0 N

0)

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U B2 Reactor Vessel Internals (PWR) - Westinghouse m

G)

I 00 Structure a Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.B2.RP-292 IV.B2- Bottom- Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No I\)

21(R-140) mounted steel and neutron flux toughness Internals" instrumentation due to neutron Expansion components (identified in system: bottom- irradiation the "Structure and Components" mounted embrittiement column) instrumentation (for Primary components see AMR (BMI) column Item IV.B2.RP-297) bodies IV.B2.RP-296 Control rod Stainless Reactor coolant Loss of material Chapter XI.M16A, "PWR Vessel No guide tube steel and neutron flux due to wear Internals" (CRGT) Primary Components (identified in the assemblies: "Structure and Components" column)

<: CRGT guide (for Expansion components see AMR III plates (cards) Line Item IV.B2.RP-386)

I\)

I (J) IV.B2.RP-298 IV.B2- Control rod Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No 28(R-118) guide tube steel and neutron flux due to stress and (CRGT) corrosion cracking Chapter XI.M16A, "PWR Vessel assemblies: and fatigue Internals" CRGT lower Primary components (identified in the flange welds "Structure and Components" column)

(accessible) (for Expansion components see AMR Items IV.B2.RP-291 and IV.B2.RP-293)

IV.B2.RP-297 Control rod Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No guide tube steel and neutron flux toughness Internals" (CRGT) due to thermal Primary components (identified in the assemblies: aging and neutron "Structure and Components" column) 0 CRGT lower irradiation (for Expansion components see AMR CD C') flange welds embrittiement Items IV.B2.RP-290 and IV.B2.RP-CD 0 (accessible) 292)

G) 3 0-CD 0 I\)

0 a 0

--" a 0)

CD 0

I 0

0 N

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oCD IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C')

CD B2 Reactor Vessel Internals (PWR) - Westinghouse 3

0-CD Structure I\) Aging Effect/ Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.B2.RP-386 Control rod Stainless Reactor coolant Loss of material Chapter XI.M16A, "PWR Vessel No guide tube steel and neutron flux due to wear Internals" (CRGT) Expansion components (identified in assemblies: C- the "Structure and Components" tubes and column) are only the components sheaths associated with a primary component that exceeded the acceptance limit.

(for Primary components see AMR Item IV.B2.RP-296)

IV.B2.RP-355 Control rod Nickel alloy Reactor coolant Cracking A plant-specific aging management Yes, plant-guide tube and neutron flux due to stress program is to be evaluated specific assemblies: corrosion cracking

<: guide tube and fatigue III support pins I\)

I

-J IV.B2.RP-356 Control rod Nickel alloy Reactor coolant Loss of material A plant-specific aging management Yes, plant-guide tube and neutron flux due to wear program is to be evaluated specific assemblies:

guide tube support pins IV.B2.RP-387 Core barrel Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No assembly: core steel and neutron flux due to stress and barrel axial corrosion Chapter XI.M16A, "PWR Vessel welds cracking, and Internals" irradiation- Expansion components (identified in assisted stress the "Structure and Components" z corrosion cracking column) c (for Primary components see AMR

u m Item IV.B2.RP-276)

G)

I 0 a 00 G)

U 0 CD 0

0  :<

0)

CD 0

I 0

0 N

(J1

z IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM c

U B2 Reactor Vessel Internals (PWR) - Westinghouse m

G)

I 00 Structure a Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.B2.RP-388 Core barrel Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No I\)

assembly: core steel and neutron flux toughness Internals" barrel axial due to neutron Expansion components (identified in welds irradiation the "Structure and Components" embrittiement column)

(for Primary components see AMR Item IV.B2.RP-276)

IV.B2.RP-282 IV.B2-8(R- Core barrel Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No 120) assembly: core steel and neutron flux due to stress and barrel flange corrosion cracking Chapter XI.M16A, "PWR Vessel and fatigue Internals" Expansion components (identified in

<: the "Structure and Components" III column)

I\)

I 00 (for Primary components see AMR Item IV.B2.RP-276)

IV.B2.RP-34S Core barrel Stainless Reactor coolant Loss of material Chapter XI.M16A, "PWR Vessel No assembly: core steel and neutron flux due to wear Internals" barrel flange Existing Program components (identified in the "Structure and Components" column) no Expansion components IV.B2.RP-278 IV.B2-8(R- Core barrel Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No 120) assembly: core steel and neutron flux due to stress and barrel outlet corrosion cracking Chapter XI.M16A, "PWR Vessel nozzle welds and fatigue Internals" Expansion component (identified in 0 the "Structure and Components" CD C') column)

CD 0 (for Primary components see AMR G) 3 0- Item IV.B2.RP-276)

CD 0 I\)

0 a 0

--" a 0)

CD 0

I 0

0 N

(J)

oCD IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C')

CD B2 Reactor Vessel Internals (PWR) - Westinghouse 3

0-CD Structure I\) Aging Effect/ Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.B2.RP-280 IV.B2-8(R- Core barrel Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No 120) assembly: steel and neutron flux due to stress and lower core corrosion cracking Chapter XI.M16A, "PWR Vessel barrel flange and irradiation- Internals" weld assisted stress Expansion component (identified in corrosion cracking the "Structure and Components" column)

(for Primary components see AMR Item IV.B2.RP-276)

IV.B2.RP-281 IV.B2-9(R- Core barrel Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No 122) assembly: steel and neutron flux toughness Internals" lower core due to neutron Expansion Components (identified in

<: barrel flange irradiation the "Structure and Components" III weld embrittiement column)

I\)

I CD (for Primary components see AMR Item IV.B2.RP-276)

IV.B2.RP-276 IV.B2-8(R- Core barrel Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No 120) assembly: steel and neutron flux due to stress and upper core corrosion cracking Chapter XI.M16A, "PWR Vessel barrel flange and irradiation- Internals" weld assisted stress Primary components (identified in the corrosion cracking "Structure and Components" column)

(for Expansion components see AMR Items IV.B2.RP-278, IV.B2.RP-280, IV.B2.RP-282, IV.B2.RP-294, z IV.B2.RP-295,IV.B2.RP-281, c IV.B2.RP-387, and IV.B2.RP-388)

u m IV.B2.RP-285 IV.B2- Lower internals Nickel alloy Reactor coolant Loss of material Chapter XI.M16A, "PWR Vessel No G)

I 14(R-137) assembly: and neutron flux due to wear Internals" 0 a 00 clevis insert Existing Program components G) bolts (identified in the "Structure and

U Components" column) 0 CD 0

0  :< no Expansion components 0)

CD 0

I 0

0 N

-....J

z IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM c

U B2 Reactor Vessel Internals (PWR) - Westinghouse m

G),

00 Structure a Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.B2.RP-289 IV.B2- Lower internals Stainless Reactor coolant Cracking 'Chapter XI.M2, "Water Chemistry," No I\)

20(R-130) assembly: steel and neutron flux due to irradiation- and lower core plate assisted stress Chapter XI.M16A, "PWR Vessel and extra-long corrosion Internals" (XL) lower core cracking, and Existing Program components plate fatigue (identified in the "Structure and Components" column) no Expansion components IV.B2.RP-288 IV.B2- Lower internals Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No 18(R-132) assembly: steel and neutron flux toughness Internals" lower core plate due to neutron Existing Program components and extra-long irradiation (identified in the "Structure and

<: (XL) lower core embrittiement; Components" column)

III I\) plate loss of material no Expansion components

...... due to wear a

IV.B2.RP-291 IV.B2- Lower support Cast Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No 24(R-138) assembly: austenitic and neutron flux due to irradiation- and lower support stainless assisted stress Chapter XI.M16A, "PWR Vessel column bodies steel corrosion cracking Internals" (cast) Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Item IV.B2.RP-298)

IV.B2.RP-290 IV.B2- Lower support Cast Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No 21(R-140) assembly: austenitic and neutron flux toughness Internals" lower support stainless due to thermal Expansion components (identified in 0 column bodies steel aging and neutron the "Structure and Components" CD C') (cast) irradiation column)

CD 0 embrittiement (for Primary components see AMR G) 3 0- Item IV.B2.RP-297)

CD 0 I\)

0 a 0

--" a 0)

CD 0

I 0

0 N

CD

oCD IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C')

CD B2 Reactor Vessel Internals (PWR) - Westinghouse 3

0-CD Structure I\) Aging Effect/ Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.B2.RP-294 IV.B2- Lower support Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No 24(R-138) assembly: steel and neutron flux due to irradiation- and lower support assisted stress Chapter XI.M16A, "PWR Vessel column bodies corrosion cracking Internals" (non-cast) Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Item IV.B2.RP-276)

IV.B2.RP-295 IV.B2- Lower support Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No 22(R-141) assembly: steel and neutron flux toughness Internals" lower support due to neutron Expansion Components (identified in

<: column bodies irradiation the "Structure and Components" III I\) (non-cast) embrittiement column)

I (for Primary components see AMR Item IV.B2.RP-276)

IV.B2.RP-286 IV.B2- Lower support Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No 16(R-133) assembly: steel; nickel and neutron flux due to irradiation- and lower support alloy assisted stress Chapter XI.M16A, "PWR Vessel column bolts corrosion cracking Internals" and fatigue Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Item IV.B2.RP-271) z c

u m

G)

I 0 a 00 G)

U 0 CD 0

0  :<

0)

CD 0

I 0

0 N

CD

z IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM c

U B2 Reactor Vessel Internals (PWR) - Westinghouse m

G)

I 00 Structure a Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.B2.RP-287 IV.B2- Lower support Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No I\)

17(R-135) assembly: steel; nickel and neutron flux toughness Internals" lower support alloy due to neutron Expansion component (identified in column bolts irradiation the "Structure and Components" embrittiement; column) loss of preload (for Primary components see AMR due to thermal Item IV.B2.RP-272) and irradiation enhanced stress relaxation IV.B2.RP-303 IV.B2- Reactor vessel Stainless Reactor coolant Cumulative Fatigue is a time-limited aging Yes, TLAA 31 (R-53) internal steel; nickel and neutron flux fatigue damage analysis (TLAA) to be evaluated for

<: components alloy due to fatigue the period of extended operation. See III I\) the SRP, Section 4.3 "Metal Fatigue,"

I for acceptable methods for meeting I\)

the requirements of 10 CFR 54.21 (c)(1).

IV.B2.RP-24 IV.B2- Reactor vessel Stainless Reactor coolant Loss of material Chapter XI.M2, "Water Chemistry" No 32(RP-24) internal steel; nickel and neutron flux due to pitting and components alloy crevice corrosion IV.B2.RP-268 Reactor vessel Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," Yes, if internal steel; nickel and neutron flux due to stress and accessible components alloy corrosion Chapter XI.M16A, "PWR Vessel Primary, (inaccessible cracking, and Internals" Expansion or locations) irradiation- Existing assisted stress program corrosion cracking components 0 indicate aging CD C')

CD effects that 0 need G) 3 0-CD management 0 I\)

0 a 0

--" a 0)

CD 0

I 0

0 N

N 0

oCD IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C')

CD B2 Reactor Vessel Internals (PWR) - Westinghouse 3

0-CD Structure I\) Aging Effect/ Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.B2.RP-269 Reactor vessel Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel Yes, if internal steel; nickel and neutron flux toughness Internals" accessible components alloy due to neutron Primary, (inaccessible irradiation Expansion or locations) embrittiement; Existing change in program dimension components due to void indicate aging swelling; effects that loss of preload need due to thermal management and irradiation

<: enhanced stress III relaxation; I\)

I loss of material CJ.)

due to wear IV.B2.RP-265 Reactor vessel Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No internal steel; nickel and neutron flux due to stress and components alloy corrosion Chapter XI.M16A, "PWR Vessel with no cracking, and Internals" additional irradiation- Note: Components with no additional measures assisted stress measures are not uniquely identified corrosion cracking in GALL tables - Components with no additional measures are defined in Section 3.3.1 of MRP-227, "Materials z Reliability Program: Pressurized c Water Reactor Internals Inspection

u m and Evaluation Guidelines" G)

I 0 a 00 G)

U 0 CD 0

0  :<

0)

CD 0

I 0

0 N

N

z IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM c

U B2 Reactor Vessel Internals (PWR) - Westinghouse m

G)

I 00 Structure a Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.B2.RP-267 Reactor vessel Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No I\)

internal steel; nickel and neutron flux toughness Internals" components alloy due to neutron Note: Components with no additional with no irradiation measures are not uniquely identified additional embrittiement; in GALL tables - Components with no measures change in additional measures are defined in dimension Section 3.3.1 of MRP-227, "Materials due to void Reliability Program: Pressurized swelling; Water Reactor Internals Inspection loss of preload and Evaluation Guidelines" due to thermal and irradiation

<: enhanced stress III relaxation; I\)

I loss of material

.j:>.

due to wear IV.B2.RP-382 IV.B2- Reactor vessel Stainless Reactor coolant Cracking, or Chapter XI.M1, "ASME Section XI No 26(R-142) internals: core steel; nickel and neutron flux Loss of material Inservice Inspection, Subsections support alloy; cast due to wear IWB, IWC, and IWD" structure austenitic stainless steel IV.B2.RP-302 Thermal shield Stainless Reactor coolant Cracking Chapter XI.M16A, "PWR Vessel No assembly: steel and neutron flux due to fatigue; Internals" thermal shield loss of material Primary components (identified in the flexures due to wear "Structure and Components" column) no Expansion components 0

CD C')

CD 0

G) 3 0-CD 0 I\)

0 a 0

--" a 0)

CD 0

I 0

0 N

N N

oCD IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C')

CD B2 Reactor Vessel Internals (PWR) - Westinghouse 3

0-CD Structure I\) Aging Effect/ Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.B2.RP-346 Upper internals Stainless Reactor coolant Cracking 'Chapter XI.M2, "Water Chemistry," No assembly: steel and neutron flux due to stress and upper support corrosion cracking Chapter XI.M16A, "PWR Vessel ring or skirt and fatigue Internals" Existing Program components (identified in the "Structure and Components" column) no Expansion components III I\)

0"1 z

c

u m

G) 0 a 00 G)

U 0 CD 0

0  :<

0)

CD 0

I 0

0 N

N 0)

B3. REACTOR VESSEL INTERNALS (PWR) - COMBUSTION ENGINEERING Systems, Structures, and Components This section addresses the Combustion Engineering pressurized water reactor (PWR) vessel internals and consists of the upper internals assembly, the control element assembly (CEA) shrouds, the core support barrel, the core shroud assembly, and the lower internal assembly.

Based on Regulatory Guide 1.26, "Quality Group Classifications and Standards for Water-,

Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants," all structures and components that comprise the reactor vessel are governed by Group A or B Quality Standards.

Common miscellaneous material/environment combinations where aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation are included in IV.E.

System Interfaces The systems that interface with the reactor vessel internals include the reactor pressure vessel (lV.A2).

Inspection Plan An applicant will submit an inspection plan for reactor internals to the NRC for review and approval with the application for license renewal in accordance with Chapter XI.M16A, "PWR Vessel Internals."

December 201 0 IV 83-1 NUREG-1801, Rev. 2 OAGI0001390_00224

z IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM c

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00 Structure a Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.B3.RP-312 IV.B3-2(R- Control Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No I\)

149) Element steel and neutron flux due to stress and Assembly corrosion cracking Chapter XI.M16A, "PWR Vessel (CEA): shroud and fatigue Internals" assemblies: Primary components (identified in the instrument "Structure and Components" column) guide tubes in (for Expansion components see AMR peripheral CEA Item IV.B3.RP-313) assemblies IV.B3.RP-313 Control Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No Element steel and neutron flux due to stress and Assembly corrosion cracking Chapter XI.M16A, "PWR Vessel

<: (CEA): shroud and fatigue Internals" III assemblies: Expansion components (identified in CJ.)

I\) remaining the "Structure and Components" instrument column) guide tubes in (for Primary components see AMR CEA Item IV.B3.RP-312) assemblies IV.B3.RP-320 IV.B3-9(R- Core shroud Stainless Reactor coolant Cracking 'Chapter XI.M2, "Water Chemistry," No 162) assemblies (all steel and neutron flux due to fatigue and plants): guide Chapter XI.M16A, "PWR Vessel lugs and guide Internals" lug insert bolts Existing Program components (identified in the "Structure and Components" column) no Expansion components 0

CD C')

CD 0

G) 3 0-CD 0 I\)

0 a 0

--" a 0)

CD 0

I 0

0 N

N (J1

oCD IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C')

CD B3 Reactor Vessel Internals (PWR) - Combustion Engineering 3

0-CD Structure I\) Aging Effect/ Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.B3.RP-319 IV.B3-9(R- Core shroud Stainless Reactor coolant Loss of material Chapter XI.M16A, "PWR Vessel No 162) assemblies (all steel and neutron flux due to wear; Internals" plants): guide Loss of preload Existing Program components lugs and guide due to thermal and (identified in the "Structure and lug insert bolts irradiation Components" column) enhanced stress no Expansion components relaxation IV.B3.RP-358 Core shroud Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No assemblies (for steel and neutron flux due to irradiation- and bolted core assisted stress Chapter XI.M16A, "PWR Vessel shroud corrosion cracking Internals" assemblies): Expansion components (identified in

<: (a) shroud the "Structure and Components" III plates and (b) column)

CJ.)

I CJ.) former plates (for Primary component see AMR Item IV.B3.RP-314)

IV.B3.RP-318 IV.B3-8(R- Core shroud Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No 163) assemblies (for steel and neutron flux toughness Internals" bolted core due to neutron Primary components (identified in the shroud irradiation "Structure and Components" column) assemblies): embrittlement; no Expansion components (a) shroud change in plates and (b) dimension former plates due to void swelling z

c

u m

G)

I 0 a 00 G)

U 0 CD 0

0  :<

--" I\)

0)

CD 0

I 0

0 N

N (J)

z IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM c

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G)

I 00 Structure a Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.B3.RP-316 IV.B3-9(R- Core shroud Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No I\)

162) assemblies (for steel and neutron flux due to irradiation- and bolted core assisted stress Chapter XI.M16A, "PWR Vessel shroud corrosion cracking Internals" assemblies): Expansion components (identified in barrel-shroud the "Structure and Components" bolts with column) neutron (for Primary components see AMR exposures Item IV.B3.RP-314) greater than 3 dpa IV.B3.RP-317 IV.B3-7(R- Core shroud Stainless Reactor coolant Loss of preload Chapter XI.M16A, "PWR Vessel No 165) assemblies (for steel; nickel and neutron flux due to thermal and Internals" bolted core alloy irradiation Expansion components (identified in shroud enhanced stress the "Structure and Components" assemblies): relaxation; column) barrel-shroud loss of fracture (for Primary components see AMR bolts with toughness Item IV.B3.RP-315) neutron due to neutron exposures irradiation greater than 3 embrittlement dpa IV.B3.RP-314 IV.B3-9(R- Core shroud Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No 162) assemblies (for steel and neutron flux due to irradiation- and bolted core assisted stress Chapter XI.M16A, "PWR Vessel shroud corrosion cracking Internals" assemblies): and fatigue Primary components (identified in the 0 core shroud "Structure and Components" column)

CD C') bolts (for Expansion components see AMR CD (accessible) Items IV.B3.RP-316, IV.B3.RP-330, 0

G) 3 0- and IV.B3.RP-358)

CD 0 I\)

0 a 0

--" a 0)

CD 0

I 0

0 N

N

-....J

oCD IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C')

CD 83 Reactor Vessel Internals (PWR) - Combustion Engineering 3

0-CD Structure I\) Aging Effect/ Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.83.RP-315 IV.83-7(R- Core shroud Stainless Reactor coolant Loss of preload Chapter XI.M16A, "PWR Vessel No 165) assemblies (for steel and neutron flux due to thermal and Internals," Primary components bolted core irradiation (identified in the "Structure and shroud enhanced stress Components" column) assemblies): relaxation; (for Expansion components see AMR core shroud loss of fracture Items IV.83.RP-317, and IV.83.RP-bolts toughness 331)

(accessible) due to neutron irradiation embrittlement; change in dimension

<: due to void III CJ.)

swelling I

0"1 IV.83.RP-359 Core shroud Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No assemblies steel and neutron flux toughness Internals," Primary components (welded): due to neutron (identified in the "Structure and (shroud plates irradiation Components" column) and (b) former embrittlement; no Expansion components plates change in dimension due to void swelling z

c

u m

G)

I 0 a 00 G)

U 0 CD 0

0  :<

--" I\)

0)

CD 0

I 0

0 N

N CD

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I 00 Structure a Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.B3.RP-322 Core shroud Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No I\)

assembly (for steel and neutron flux due to irradiation- and welded core assisted stress Chapter XI.M16A, "PWR Vessel shrouds in two corrosion cracking Internals" vertical Primary components (identified in the sections): Core "Structure and Components" column) shroud plate- (for Expansion components see AMR former plate Item IV.B3.RP-323) weld (a) The axial and horizontal weld seams at the

<: core shroud re-III CJ.)

entrant corners I

(J) as visible from the core side of the shroud, within six inches of the central flange and horizontal stiffeners, and (b) the horizontal stiffeners in shroud plate-to-former plate weld 0

CD C')

CD 0

G) 3 0-CD 0 I\)

0 a 0

--" a 0)

CD 0

I 0

0 N

N CD

oCD IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C')

CD B3 Reactor Vessel Internals (PWR) - Combustion Engineering 3

0-CD Structure I\) Aging Effect/ Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.B3.RP-326 Core shroud Stainless Reactor coolant Change in Chapter XI.M16A, "PWR Vessel No assembly (for steel and neutron flux dimension Internals" welded core due to void Primary components (identified in the shrouds in two swelling "Structure and Components" column) vertical no Expansion components sections): gap between the upper and lower plates IV.B3.RP-323 Core shroud Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No assembly (for steel and neutron flux due to irradiation- and welded core assisted stress Chapter XI.M16A, "PWR Vessel shrouds in two corrosion cracking Internals" vertical Expansion components (identified in sections): the "Structure and Components" remaining axial column) welds in (for Primary components see AMR shroud plate- Item IV.B3.RP-322) to-former plate z

c

u m

G)

I 0 a 00 G)

U 0 CD 0

0  :<

--" I\)

0)

CD 0

I 0

0 N

0) 0

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I 00 Structure a Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.B3.RP-324 Core shroud Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No I\)

assembly (for steel and neutron flux due to irradiation- and welded core assisted stress Chapter XI.M16A, "PWR Vessel shrouds with corrosion cracking Internals" full-height Primary components (identified in the shroud plates): "Structure and Components" column) axial weld (for Expansion components see AMR seams at the Item IV.B3.RP-32S) core shroud re-entrant corners, at the core mid-plane

<: (+3 feet in III CJ.)

height) as 00 I

visible from the core side of the shroud 0

CD C')

CD 0

G) 3 0-CD 0 I\)

0 a 0

--" a 0)

CD 0

I 0

0 N

0)

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CD 83 Reactor Vessel Internals (PWR) - Combustion Engineering 3

0-CD Structure I\) Aging Effect/ Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.83.RP-360 Core shroud Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No assembly (for steel and neutron flux toughness Internals" welded core due to neutron Primary components (identified in the shrouds with irradiation "Structure and Components" column) full-height embrittlement (for Expansion components see AMR shroud plates): Item IV.83.RP-361) axial weld seams at the core shroud re-entrant corners, at the core mid-plane

<: (+3 feet in III CJ.)

height) as I

CD visible from the core side of the shroud IV.83.RP-32S Core shroud Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No assembly (for steel and neutron flux due to irradiation- and welded core assisted stress Chapter XI.M16A, "PWR Vessel shrouds with corrosion cracking Internals" full-height Expansion components (identified in shroud plates): the "Structure and Components" remaining axial column) welds, ribs, (for Primary components see AMR z and rings Item IV.83.RP-324) c

u m

G)

I 0 a 00 G)

U 0 CD 0

0  :<

--" I\)

0)

CD 0

I 0

0 N

0)

N

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G)

I 00 Structure a Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.B3.RP-361 Core shroud Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No I\)

assembly (for steel and neutron flux toughness Internals" welded core due to neutron Expansion components (identified in shrouds with irradiation the "Structure and Components" full-height embrittlement column) shroud plates): (for Primary components see AMR remaining axial Item IV.B3.RP-360) welds, ribs, and rings IV.B3.RP-362 Core support Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No barrel steel and neutron flux toughness Internals" assembly: due to neutron Expansion components (identified in

<: lower cylinder irradiation the "Structure and Components" III CJ.) welds embrittlement column)

I (for Primary components see AMR a

Item IV.B3.RP-327)

IV.B3.RP-329 IV.B3- Core support Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No 15(R-155) barrel steel and neutron flux due to stress and assembly: corrosion cracking Chapter XI.M16A, "PWR Vessel lower cylinder Internals" welds and Expansion components (identified in remaining core the "Structure and Components" barrel column) assembly (for Primary components see AMR welds Item IV.B3.RP-327)

IV.B3.RP-333 Core support Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," Yes, evaluate barrel steel and neutron flux due to fatigue and to determine 0 assembly: Chapter XI.M16A, "PWR Vessel the potential CD C') lower flange Internals" locations and CD 0 weld, if fatigue Primary components (identified in the extent of G) 3 0- life cannot be "Structure and Components" column) fatigue CD 0 I\) demonstrated no Expansion components cracking 0 a by TLAA 0

--" a 0)

CD 0

I 0

0 N

0) 0)

oCD IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C')

CD B3 Reactor Vessel Internals (PWR) - Combustion Engineering 3

0-CD Structure I\) Aging Effect/ Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.B3.RP-389 Core support Stainless Reactor coolant Cumulative fatigue Fatigue is a time-limited aging Yes, TLAA barrel steel and neutron flux damage analysis (TLAA) to be evaluated for assembly: due to fatigue the period of extended operation. See lower flange the SRP, Section 4.3 "Metal Fatigue,"

weld (if fatigue for acceptable methods for meeting analysis exists) the requirements of 10 CFR 54.21 (c)(1).

IV.B3.RP-328 IV.B3- Core support Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No 15(R-155) barrel steel and neutron flux due to stress and assembly: corrosion cracking Chapter XI.M16A, "PWR Vessel surfaces of the and fatigue Internals" lower core Primary components (identified in the

<: barrel flange "Structure and Components" column)

III CJ.) weld no Expansion components I

(accessible surfaces)

IV.B3.RP-332 IV.B3- Core support Stainless Reactor coolant Loss of material Chapter XI.M16A, "PWR Vessel No 17(R-156) barrel steel and neutron flux due to wear Internals" assembly: Existing Program components upper core (identified in the "Structure and barrel flange Components" column) no Expansion components IV.B3.RP-327 IV.B3- Core support Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No 15(R-155) barrel steel and neutron flux due to stress and assembly: corrosion cracking Chapter XI.M16A, "PWR Vessel z upper core Internals" c support barrel Primary components (identified in the

u m flange weld "Structure and Components" column)

G)

I (accessible (for Expansion components see AMR 0 a 00 surfaces) Items IV.B3.RP-329, IV.B3.RP-335, G) IV.B3.RP-362, IV.B3.RP-363,

U IV. B3. RP-364) 0 CD 0

0  :<

--" I\)

0)

CD 0

I 0

0 N

0)

.j::>.

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00 Structure a Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.B3.RP-3S7 Incore Zircaloy-4 Reactor coolant Loss of material A plant-specific aging management Yes, plant-I\)

instrumentation and neutron flux due to wear program is to be evaluated specific (ICI): ICI thimble tubes -

lower IV.B3.RP-336 IV.B3- Lower support Stainless Reactor coolant Loss of material Chapter XI.M16A, "PWR Vessel No 22(R-170) structure: A286 steel and neutron flux due to wear; Internals" fuel alignment loss of fracture Existing Program components pins (all plants toughness (identified in the "Structure and with core due to neutron Components" column) shroud irradiation no Expansion components assembled in embrittlement;

<: two vertical loss of preload III CJ.) sections) due to thermal and

...... irradiation I\)

enhanced stress relaxation IV.B3.RP-334 IV.B3- Lower support Stainless Reactor coolant Cracking 'Chapter XI.M2, "Water Chemistry," No 23(R-167) structure: A286 steel and neutron flux due to irradiation- and fuel alignment assisted stress Chapter XI.M16A, "PWR Vessel pins (all plants corrosion cracking Internals" with core and fatigue Existing Program components shroud (identified in the "Structure and assembled Components" column) with full-height no Expansion components shroud plates)

IV.B3.RP-364 Lower support Cast Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No 0 structure: core austenitic and neutron flux toughness Internals" CD C') support column stainless due to neutron Expansion components (identified in CD 0 steel irradiation and the "Structure and Components" G) 3 0- thermal column)

CD 0 I\) embrittlement (for Primary components see AMR 0 a Item IV.B3RP-327) 0

--" a 0)

CD 0

I 0

0 N

0)

(J1

oCD IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C')

CD B3 Reactor Vessel Internals (PWR) - Combustion Engineering 3

0-CD Structure I\) Aging Effect/ Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.B3.RP-363 Lower support Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No structure: core steel and neutron flux toughness Internals" support column due to neutron Expansion components (identified in irradiation the "Structure and Components" embrittlement column)

(for Primary components see AMR Item IV.B3RP-327)

IV.B3.RP-330 IV.B3- Lower support Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No 23(R-167) structure: core steel and neutron flux due to irradiation- and support column assisted stress Chapter XI.M16A, "PWR Vessel bolts corrosion cracking Internals" and fatigue Expansion components (identified in

<: the "Structure and Components" III CJ.) column)

...... (for Primary components see AMR CJ.)

Item'IV.B3.RP-314)

IV.B3.RP-331 Lower support Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No structure: core steel and neutron flux toughness Internals" support column due to neutron Expansion components (identified in bolts irradiation the "Structure and Components" embrittlement column)

(for Primary components see AMR Item'IV.B3.RP-315)

IV.B3.RP-335 IV.B3- Lower support Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No 23(R-167) structure: core steel and neutron flux due to stress and z support column corrosion Chapter XI.M16A, "PWR Vessel c welds, cracking, Internals"

u m applicable to irradiation- Expansion components (identified in G)

, all plants assisted stress the "Structure and Components" 0 a 00 except those corrosion column)

G) assembled cracking, and (for Primary components see AMR

U with full-height fatigue Item IV.B3.RP-327) 0 CD 0

0  :< shroud plates

--" I\)

0)

CD 0

I 0

0 N

0)

(J)

z IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM c

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G)

I 00 Structure a Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.83.RP-365 Lower support Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No I\)

structure: core steel and neutron flux toughness Internals" support plate due to neutron Primary component (identified in the irradiation "Structure and Components" column) embrittlement no Expansion components IV.83.RP-343 Lower support Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry", Yes, evaluate structure: core steel and neutron flux due to fatigue and to determine support plate Chapter XI.M16A, "PWR Vessel the potential (applicable to Internals" locations and plants with a Primary components (identified in the extent of core support "Structure and Components" column) fatigue plate), if fatigue no Expansion components cracking life cannot be demonstrated by TLAA IV.83.RP-390 Lower support Stainless Reactor coolant Cumulative fatigue Fatigue is a time-limited aging Yes, TLAA structure: core steel and neutron flux damage analysis (TLAA) to be evaluated for support plate due to fatigue the period of extended operation. See (applicable to the SRP, Section 4.3 "Metal Fatigue,"

plants with a for acceptable methods for meeting core support the requirements of 10 CFR plate), if fatigue 54.21 (c)(1).

analysis exists IV.83.RP-342 Lower support Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No structure: deep steel and neutron flux due to stress and beams corrosion Chapter XI.M16A, "PWR Vessel (applicable cracking, Internals" 0 assemblies irradiation- Primary components (identified in the CD C') with full height assisted stress "Structure and Components" column)

CD 0 shroud plates) corrosion no Expansion components G) 3 0- cracking, and CD 0 I\)

fatigue 0 a 0

--" a 0)

CD 0

I 0

0 N

0)

-....J

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CD B3 Reactor Vessel Internals (PWR) - Combustion Engineering 3

0-CD Structure I\) Aging Effect/ Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.B3.RP-366 Lower support Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No structure: deep steel and neutron flux toughness Internals" beams due to neutron Primary components (identified in the (applicable irradiation "Structure and Components" column) assemblies embrittlement no Expansion components with full height shroud plates)

IV.B3.RP-339 IV.B3- Reactor vessel Stainless Reactor coolant Cumulative fatigue Fatigue is a time-limited aging Yes, TLAA 24(R-53) internal steel; nickel and neutron flux damage analysis (TLAA) to be evaluated for components alloy due to fatigue the period of extended operation. See the SRP, Section 4.3 "Metal Fatigue,"

for acceptable methods for meeting

<: the requirements of 10 CFR III CJ.) 54.21 (c)(1).

I IV.B3.RP-24 IV.B3- Reactor vessel Stainless Reactor coolant Loss of material Chapter XI.M2, "Water Chemistry" No 0"1 25(RP-24) internal steel; nickel and neutron flux due to pitting and components alloy crevice corrosion IV.B3.RP-309 Reactor vessel Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," Yes, if internal steel; nickel and neutron flux due to stress and accessible components alloy corrosion Chapter XI.M16A, "PWR Vessel Primary, (inaccessible cracking, and Internals" Expansion or locations) irradiation- Existing assisted stress program corrosion cracking components indicate aging z effects that c

u need m

G) management I

0 a 00 G)

U 0 CD 0

0  :<

--" I\)

0)

CD 0

I 0

0 N

0)

CD

z IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM c

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G)

I 00 Structure a Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.B3.RP-311 Reactor vessel Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel Yes, if I\)

internal steel; nickel and neutron flux toughness Internals" accessible components alloy due to neutron Primary, (inaccessible irradiation Expansion or locations) embrittlement; Existing change in program dimension components due to void indicate aging swelling; effects that loss of preload need due to thermal and management irradiation

<: enhanced stress III relaxation; CJ.)

I loss of material (J) due to wear IV.B3.RP-306 Reactor vessel Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No internal steel; nickel and neutron flux due to stress and components alloy corrosion Chapter XI.M16A, "PWR Vessel with no cracking, and Internals" additional irradiation- Note: Components with no additional measures assisted stress measures are not uniquely identified corrosion cracking in GALL tables - Components with no additional measures are defined in Section 3.3.1 of MRP-227, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines" 0

CD C')

CD 0

G) 3 0-CD 0 I\)

0 a 0

--" a 0)

CD 0

I 0

0 N

0)

CD

oCD IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C')

CD B3 Reactor Vessel Internals (PWR) - Combustion Engineering 3

0-CD Structure I\) Aging Effect/ Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.B3.RP-307 Reactor vessel Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No internal steel; nickel and neutron flux toughness Internals" components alloy due to neutron Note: Components with no additional with no irradiation measures are not uniquely identified additional embrittlement; in GALL tables - Components with no measures change in additional measures are defined in dimension Section 3.3.1 of MRP-227, "Materials due to void Reliability Program: Pressurized swelling; Water Reactor Internals Inspection loss of preload and Evaluation Guidelines" due to thermal and irradiation enhanced stress relaxation; loss of material due to wear IV.B3.RP-382 IV.B3- Reactor vessel Stainless Reactor coolant Cracking, or Chapter XI.M1, "ASME Section XI No 22(R-170) internals: core steel; nickel and neutron flux Loss of material Inservice Inspection, Subsections support alloy; cast due to wear IWB, IWC, and IWD" structure austenitic stainless steel z

c

u m

G)

I 0 a 00 G)

U 0 CD 0

0  :<

--" I\)

0)

CD 0

I 0

0 N

.j::>.

0

z IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM c

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G),

00 Structure a Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.B3.RP-338 Upper internals Stainless Reactor coolant Cracking 'Chapter XI.M2, "Water Chemistry," Yes, evaluate I\)

assembly: fuel steel and neutron flux due to fatigue and to determine alignment plate Chapter XI.M16A, "PWR Vessel the potential (applicable to Internals" locations and plants with Primary components (identified in the extent of core shrouds "Structure and Components" column) fatigue assembled no Expansion components cracking with full height shroud plates),

if fatigue life cannot be demonstrated

<: by TLAA III CJ.) IV.B3.RP-391 Upper internals Stainless Reactor coolant Cumulative fatigue Fatigue is a time-limited aging Yes, TLAA

...... assembly: fuel steel and neutron flux damage analysis (TLAA) to be evaluated for 00 alignment plate due to fatigue the period of extended operation. See (applicable to the SRP, Section 4.3 "Metal Fatigue,"

plants with for acceptable methods for meeting core shrouds the requirements of 10 CFR assembled 54.21 (c)(1).

with full height shroud plates),

if fatigue analysis exists 0

CD C')

CD 0

G) 3 0-CD 0 I\)

0 a 0

--" a 0)

CD 0

I 0

0 N

.j::>.

B4. REACTOR VESSEL INTERNALS (PWR) - BABCOCK AND WILCOX Systems, Structures, and Components This section addresses the Babcock and Wilcox pressurized water reactor (PWR) vessel internals and consists of the plenum cover and plenum cylinder, the upper grid assembly, the control rod guide tube (CRGT) assembly, the core support shield assembly, the core barrel assembly, the lower grid assembly, and the flow distributor assembly. Based on Regulatory Guide 1.26, "Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants," all structures and components that comprise the reactor vessel are governed by Group A or B Quality Standards.

Common miscellaneous material/environment combinations where aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation are included in IV.E.

System Interfaces The systems that interface with the reactor vessel internals include the reactor pressure vessel (lV.A2).

Inspection Plan An applicant will submit an inspection plan for reactor internals to the NRC for review and approval with the application for license renewal in accordance with Chapter XI.M16A, "PWR Vessel Internals."

December 201 0 IV 84-1 NUREG-1801, Rev. 2 OAGI0001390_00242

z IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM c

U B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox m

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I 00 Structure a Aging Effect! Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.B4.RP-242 IV.B4-4(R- Control rod Cast Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No I\)

183) guide tube austenitic and neutron flux toughness Internals" (CRGT) stainless due to thermal Expansion components (identified in assembly: steel aging the "Structure and Components" accessible embrittiement column) surfaces at four (for Primary components see AMR screw locations Items IV.B4.RP-253 and IV.B4.RP-(every 90 258) degrees) for CRGT spacer castings IV.B4.RP-245 IV.B4- Core barrel Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No

<: 13(R-194) assembly: (a) steel; nickel and neutron flux due to stress and III upper thermal alloy corrosion cracking Chapter XI.M16A, "PWR Vessel

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I I\) shield bolts; (b) Internals" surveillance Expansion components (identified in specimen the "Structure and Components" holder tube column) bolts (Davis- (for Primary components see AMR Besse, only); Items IV.B4.RP-247 and IV.B4.RP-(c) surveillance 248) specimen tube holder studs, and nuts (Crystal River Unit 3, only) 0 CD C')

CD 0 3 G) 0-CD 0 I\)

0 a 0

--" a 0)

CD 0

I 0

0 N

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0-CD Structure I\) Aging Effect! Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.B4.RP-247 IV.B4- Core barrel Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No 13(R-194) assembly: steel; nickel and neutron flux due to stress and Chapter XI.M16A, "PWR Vessel accessible alloy corrosion cracking Internals" lower core Primary components (identified in the barrel (LCB) "Structure and Components" column) bolts and (for Expansion components see AMR locking devices Items IV.B4.RP-245, IV.B4.RP-246, IV.B4.RP-254, and IV.B4.RP-256)

IV.B4.RP-249 IV.B4- Core barrel Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No 12(R-196) assembly: steel and neutron flux toughness Internals" baffle plate due to neutron Primary components (identified in the accessible irradiation "Structure and Components" column) surfaces within embrittiement (for Expansion components see AMR one inch Item IV.B4.RP-250) around each baffle plate flow and bolt hole IV.B4.RP-241 IV.B4-7(R- Core barrel Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No 125) assembly: steel and neutron flux due to stress and baffle/former corrosion Chapter XI.M16A, "PWR Vessel assembly: (a) cracking, Internals" accessible irradiation- Primary Components (identified in the baffle-to-former assisted stress "Structure and Components" column) bolts and corrosion cracking (for Expansion components see AMR screws; (b) Items IV.B4.RP-244 and IV.B4.RP-z accessible 375) c locking devices

u m (including G)

I welds) of 0 a 00 baffle-to-former G) bolts

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CD 0

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Mechanism Evaluation

u Component CD

< IV.B4.RP-240 IV.B4-1 (R- Core barrel Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No I\)

128) assembly: steel and neutron flux toughness Internals."

baffle/former due to neutron Primary components (identified in the assembly: (a) irradiation "Structure and Components" column) accessible embrittiement; (for Expansion components see AMR baffle-to-former loss of preload Item IV.B4.RP-243.)

bolts and due to thermal screws; (b) and irradiation accessible enhanced stress locking devices relaxation; (including loss of material welds) of due to wear baffle-to-former bolts IV.B4.RP-2S0 IV.B4- Core barrel Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No 12(R-196) assembly: core steel and neutron flux toughness Internals" barrel cylinder due to neutron Expansion components (identified in (including irradiation the "Structure and Components" vertical and embrittiement column) circu mferential (for Primary components see AMR seam welds); Item IV.B4.RP-249) former plates IV.B4.RP-37S Core barrel Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No assembly: steel and neutron flux due to fatigue and internal baffle- Chapter XI.M16A, "PWR Vessel to-baffle bolts Internals" Expansion components (identified in 0 the "Structure and Components" CD column)

C')

CD 0 3 (for Primary components see AMR G) 0- Item IV.B4.RP-241)

CD 0 I\)

0 a 0

--" a 0)

CD 0

I 0

0 N

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oCD IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C')

CD B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox 3

0-CD Structure I\) Aging Effect! Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.B4.RP-244 IV.B4-7(R- Core barrel Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No 125) assembly; (a) steel and neutron flux due to irradiation- and external baffle- assisted stress Chapter XI.M16A, "PWR Vessel to-baffle bolts; corrosion cracking Internals" (b) core barrel- Expansion components (identified in to-former bolts; the "Structure and Components" (c) locking column) devices (for Primary components see AMR (including Item IV.B4.RP-241) welds) of external baffle-to-baffle bolts

<: and core III

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barrel-to-former I

0"1 bolts IV.B4.RP-243 IV.B4-1 (R- Core barrel Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No 128) assembly; (a) steel and neutron flux toughness Internals" external baffle- due to neutron Expansion components (identified in to-baffle bolts; irradiation the "Structure and Components" (b) core barrel- embrittiement; column) to-former bolts; loss of preload (for Primary components see AMR (c) locking due to thermal Item IV.B4.RP-240) devices and irradiation (including enhanced stress welds) of relaxation; z external baffle- loss of material c to-baffle bolts due to wear

u m and core G) barrel-to-former I

0 00 bolts; (d)

>> a internal baffle-G) 0

U to-baffle bolts CD 0

0  :<

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CD 0

I 0

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Mechanism Evaluation

u Component CD

< IV.B4.RP-248 IV.B4- Core support Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No I\)

12(R-196) shield (CSS) steel; nickel and neutron flux due to stress and Chapter XI.M16A, "PWR Vessel assembly: alloy corrosion cracking Internals" accessible Primary components (identified in the upper core "Structure and Components" column) barrel (UCB) (for Expansion components see AMR bolts and Items IV.B4.RP-24S, IV.B4.RP-246, locking devices IV.B4.RP-2S4, IV.B4.RP-247, and IV.B4.RP-2S6)

IV.B4.RP-2S3 IV.B4- Core support Cast Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No 21 (R-191) shield (CSS) austenitic and neutron flux toughness Internals" assembly: (a) stainless due to thermal Primary components (identified in the

<: CSS cast outlet steel aging "Structure and Components" column)

III nozzles embrittiement (for Expansion components see AMR

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I (J) (Oconee Unit 3 Item IV.B4.RP-242) and Davis-Besse, only);

(b) CSS vent valve discs IV.B4.RP-2S2 IV.B4- Core support Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No 16(R-188) shield (CSS) steel and neutron flux toughness Internals" assembly: (a) due to thermal Primary components (identified in the CSS vent valve aging "Structure and Components" column) disc shaft or embrittiement No Expansion components hinge pin (b)

CSS vent valve top retaining 0 ring (c) CSS CD vent valve C')

CD 0 3 bottom G) 0- retaining ring CD 0 I\)

0 a 0

--" a 0)

CD 0

I 0

0 N

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CD B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox 3

0-CD Structure I\) Aging Effect! Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.B4.RP-2S1 IV.B4- Core support Stainless Reactor coolant Loss of material Chapter XI.M16A, "PWR Vessel No 1S(R-190) shield (CSS) steel and neutron flux due to wear Internals" assembly: CSS Primary component (identified in the top flange; "Structure and Components" column) plenum cover No Expansion components assembly:

plenum cover weldment rib pads and plenum cover support flange IV.B4.RP-2S6 IV.B4- Flow distributor Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No 2S(R-210) assembly: flow steel; nickel and neutron flux due to stress and distributor bolts alloy corrosion cracking Chapter XI.M16A, "PWR Vessel and locking Internals," Expansion components devices (identified in the "Structure and Components" column)

(for Primary components see AMR Items IV.B4.RP-247 and IV.B4.RP-248)

IV.B4.RP-2S9 IV.B4- Incore Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No 31 (R-20S) Monitoring steel; nickel and neutron flux toughness Internals" Instrumentation alloy due to thermal Primary components (identified in the (IMI) guide aging, neutron "Structure and Components" column) tube assembly: irradiation (for Expansion components see Item z accessible top embrittiement IV.B4.RP-260) c surfaces of IMI

u m guide tube G)

I spider-to-Iower 0 a 00 grid rib G) sections welds

U 0 CD 0

0  :<

0)

CD 0

I 0

0 N

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G),

00 Structure a Aging Effect! Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.B4.RP-2S8 IV.B4-4(R- Incore Cast Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No I\)

183) Monitoring austenitic and neutron flux toughness Internals" Instrumentation stainless due to thermal Primary components (identified in the (IMI) guide steel aging, neutron "Structure and Components" column) tube assembly: irradiation (for Expansion components see Item accessible top embrittiement IV.B4.RP-242) surfaces of IMI Incore guide tube spider castings IV.B4.RP-2S4 IV.B4- Lower grid Nickel alloy Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No 2S(R-210) assembly: and neutron flux due to stress and

<: alloy X-7S0 corrosion cracking Chapter XI.M16A, "PWR Vessel III lower grid Internals," Expansion components

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00 shock pad bolts (identified in the "Structure and and locking Components" column) devices (TMI-1, (for Primary components see AMR only) Items IV.B4.RP-247 and IV.B4.RP-248)

IV.B4.RP-246 IV.B4- Lower grid Stainless Reactor coolant Cracking 'Chapter XI.M2, "Water Chemistry," No 12(R-196) assembly: steel; nickel and neutron flux due to stress and lower thermal alloy corrosion cracking Chapter XI.M16A, "PWR Vessel shield (L TS) Internals" bolts Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR 0 Items IV.B4.RP-247 and IV.B4.RP-CD 248)

C')

CD 0 3 G) 0-CD 0 I\)

0 a 0

--" a 0)

CD 0

I 0

0 N

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oCD IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C')

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0-CD Structure I\) Aging Effect! Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.B4.RP-260 IV.B4- Lower grid Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No 31 (R-205) assembly: (a) steel; nickel and neutron flux toughness Internals" accessible alloy due to neutron Expansion components (identified in pads; (b) irradiation the "Structure and Components" accessible pad- embrittiement column) to-rib section (for Primary components see AMR welds; (c) Item IV.B4.RP-259) accessible alloy X-750 dowels, cap screws and locking devices

<: IV.B4.RP-262 IV.B4- Lower grid Nickel alloy Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No III 32(R-203) assembly: and neutron flux due to stress and

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I CD accessible corrosion cracking Chapter XI.M16A, "PWR Vessel alloy X-750 Internals" dowel-to-Iower Expansion components (identified in fuel assembly the "Structure and Components" support pad column) welds (for Primary components see AMR Item IV.B4.RP-261)

IV.B4.RP-261 IV.B4- Lower grid Nickel alloy Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No 32(R-203) assembly: alloy and neutron flux due to stress and X-750 dowel- corrosion cracking Chapter XI.M16A, "PWR Vessel to-guide block Internals" z welds Primary components (identified in the c "Structure and Components" column)

u m (for Expansion components see AMR G)

I Items IV.B4.RP-262 and IV.B4.RP-0 a 00 352)

G)

U 0 CD 0

0  :<

0)

CD 0

I 0

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I 00 Structure a Aging Effect! Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.B4.R-53 IV.B4- Reactor vessel Stainless Reactor coolant Cumulative Fatigue is a time-limited aging Yes, TLAA I\)

37(R-53) internal steel; nickel and neutron flux fatigue damage analysis (TLAA) to be evaluated for components alloy due to fatigue the period of extended operation. See the SRP, Section 4.3 "Metal Fatigue,"

for acceptable methods for meeting the requirements of 10 CFR 54.21 (c)(1).

IV.B4.RP-24 IV.B4- Reactor vessel Stainless Reactor coolant Loss of material Chapter XI.M2, "Water Chemistry" No 38(RP-24) internal steel; nickel and neutron flux due to pitting and components alloy crevice corrosion IV.B4.RP-376 Reactor vessel Stainless Reactor coolant Reduction in Ductility - Reduction in Fracture Yes, TLAA

<: internal steel; nickel and neutron flux ductility and Toughness is a TLAA (BAW-2248A)

III components alloy fracture to be evaluated for the period of

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I toughness extended operation. See the SRP, a due to neutron Section 4.7, "Other Plant-Specific irradiation TLAAs," for acceptable methods for meeting the requirements of 10 CFR 54.21 (c)(1).

IV.B4.RP-238 Reactor vessel Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," Yes, if internal steel; nickel and neutron flux due to stress and accessible components alloy corrosion Chapter XI.M16A, "PWR Vessel Primary, (inaccessible cracking, and Internals" Expansion or locations) irradiation- Existing assisted stress program corrosion cracking components indicate aging effects that 0 need CD C')

CD management 0 3 G) 0-CD 0 I\)

0 a 0

--" a 0)

CD 0

I 0

0 N

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oCD IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C')

CD B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox 3

0-CD Structure I\) Aging Effect! Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.B4.RP-239 Reactor vessel Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel Yes, if internal steel; nickel and neutron flux toughness Internals" accessible components alloy due to neutron Primary, (inaccessible irradiation Expansion or locations) embrittiement; Existing change in program dimension components due to void indicate aging swelling; effects that loss of preload need due to thermal management and irradiation enhanced stress relaxation; loss of material due to wear IV.B4.RP-236 Reactor vessel Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry" No internal steel; nickel and neutron flux due to stress and components alloy corrosion Chapter XI.M16A, "PWR Vessel with no cracking, and Internals" additional irradiation- Note: Components with no additional measures assisted stress measures are not uniquely identifies corrosion cracking in GALL tables - Components with no additional measures are defined in Section 3.3.1 of MRP-227, "Materials z Reliability Program: Pressurized c Water Reactor Internals Inspection

u m and Evaluation Guidelines" G)

I 0 a 00 G)

U 0 CD 0

0  :<

0)

CD 0

I 0

0 N

(J1 N

z IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM c

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G)

I 00 Structure a Aging Effect! Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.B4.RP-237 Reactor vessel Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No I\)

internal steel; nickel and neutron flux toughness Internals" components alloy due to neutron Note: Components with no additional with no irradiation measures are not uniquely identified additional embrittiement; in GALL tables - Components with no measures change in additional measures are defined in dimension Section 3.3.1 of MRP-227, "Materials due to void Reliability Program: Pressurized swelling; Water Reactor Internals Inspection loss of preload and Evaluation Guidelines" due to thermal and irradiation

<: enhanced stress III relaxation;

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I loss of material I\)

due to wear IV.B4.RP-382 IV.B4- Reactor vessel Stainless Reactor coolant Cracking, or Chapter XI.M1, "ASME Section XI No 42(R-179) internals: core steel; nickel and neutron flux Loss of material Inservice Inspection, Subsections support alloy; cast due to wear IWB, IWC, and IWD" structure austenitic stainless steel IV.B4.RP-3S2 Upper grid Nickel alloy Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No assembly: alloy and neutron flux due to stress and X-7S0 dowel- corrosion cracking Chapter XI.M16A, "PWR Vessel to-upper fuel Internals" assembly Expansion components (identified in 0 support pad the "Structure and Components" CD welds (all column)

C')

CD 0 plants except (for Primary components see AMR

>> 3 G) 0- Davis-Besse) Item IV.B4.RP-261)

CD 0 I\)

0 a 0

--" a 0)

CD 0

I 0

0 N

(J1 0)

C1. REACTOR COOLANT PRESSURE BOUNDARY (BOILING WATER REACTOR)

Systems, Structures, and Components This section addresses the boiling water reactor (BWR) primary coolant pressure boundary and consists of the reactor coolant recirculation system and portions of other systems connected to the pressure vessel extending to the second containment isolation valve or to the first anchor point outside containment. The connected systems include the residual heat removal (RHR),

low-pressure core spray (LPCS), high-pressure core spray (HPCS), low-pressure coolant injection (LPCI), high-pressure coolant injection (HPCI), reactor core isolation cooling (RCIC),

isolation condenser (lC), reactor water cleanup (RWC), standby liquid control (SLC), feedwater (FW), and main steam (MS) systems; and the steam line to the HPCI and RCIC pump turbines.

Based on Regulatory Guide 1.26, "Quality Group Classifications and Standards for Water-,

Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants," all systems, structures, and components that comprise the reactor coolant pressure boundary are governed by Group A Quality Standards.

Pump and valve internals perform their intended functions with moving parts or with a change in configuration, or are subject to replacement based on qualified life or specified time period.

Pursuant to 10 CFR 54.21 (a)(1), therefore, they are not subject to an aging management review.

Common miscellaneous material/environment combinations where aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation are included in IV.E.

System Interfaces The systems that interface with the reactor coolant pressure boundary include the reactor pressure vessel (lV.A 1), the emergency core cooling system (V.02), the standby liquid control system (VII.E2), the reactor water cleanup system (VII.E3), the shutdown cooling system (older plants) (VII.E4), the main steam system (VII I. B2), and the feedwater system (VII I. 02).

December 201 0 IV C1-1 NUREG-1801, Rev. 2 OAGI0001390_00254

z IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM c

U C1 Reactor Coolant Pressure Boundary (BWR) m G)

I 00 Structure a Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.C1.RP-230 IV.C1-1 (R- Class 1 Steel; Reactor coolant Cracking Chapter XI.M1, "ASME Section XI No I\)

03) piping, fittings stainless due to stress Inservice Inspection, Subsections and branch steel corrosion IWB, IWC, and IWD" for Class 1 connections < cracking, components, NPS 4 intergranular Chapter XI.M2, "Water Chemistry,"

stress corrosion and cracking (for XI.M35, "One-Time Inspection of stainless steel ASME Code Class 1 Small-bore only), and Piping" thermal, mechanical, and vibratory loading

<: IV.C1.R-52 IV.C1-2(R- Class 1 Cast Reactor coolant Loss of fracture Chapter XI.M12, "Thermal Aging No

() 52) piping, piping austenitic >250°C toughness Embrittiement of Cast Austenitic I

I\) components, stainless (>482°F) due to thermal Stainless Steel (CASS)"

and piping steel aging elements embrittiement IV.C1.R-08 IV.C1-3(R- Class 1 pump Cast Reactor coolant Loss of fracture Chapter XI.M1, "ASME Section XI No

08) casings; valve austenitic >250°C toughness Inservice Inspection, Subsections bodies and stainless (>482°F) due to thermal IWB, IWC, and IWD" for Class 1 bonnets steel aging components embrittiement For pump casings and valve bodies, screening for susceptibility to thermal aging is not necessary. The ASME Section XI inspection requirements are sufficient for managing the effects of loss of fracture toughness due to 0 thermal aging embrittlement of CASS CD C')

pump casings and valve bodies.

CD 0

G) 3 0-CD 0 I\)

0 a 0

--" a 0)

CD 0

I 0

0 N

(J1 (J1

oCD IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C')

CD C1 Reactor Coolant Pressure Boundary (BWR) 3 0-CD Structure I\) Aging Effect/ Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.C1.RP-43 IV.C1- Closure Steel; Air Loss of preload Chapter XI.M18, "Bolting Integrity" No 10(R-27) bolting stainless due to thermal steel effects, gasket creep, and self-loosening IV.C1.RP-42 IV.C1- Closure Steel; Air with reactor Loss of material Chapter XI.M18, "Bolting Integrity" No 12(R-26) bolting stainless coolant leakage due to general steel (steel only),

pitting, and crevice corrosion or wear IV.C1.R-15 IV.C1-4(R- Isolation Stainless Reactor coolant Cracking Chapter XI.M1, "ASME Section XI Yes, detection

<: 15) condenser steel due to stress Inservice Inspection, Subsections of aging

()

, components corrosion IWB, IWC, and IWD" for Class 1 effects is to be CJ.) cracking, components, and evaluated intergranular Chapter XI.M2, "Water Chemistry" stress corrosion The AMP in Chapter XI.M1 is to be cracking augmented to detect cracking due to stress corrosion cracking and verification of the program's effectiveness is necessary to ensure that significant degradation is not occurring and the component intended function will be maintained during the extended period of operation. An z acceptable verification program c includes temperature and radioactivity

u m monitoring of the shell side water, and G)

, eddy current testing of tubes.

0 a 00 G)

U 0 CD 0

0  :<

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0)

CD 0

I 0

0 N

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U C1 Reactor Coolant Pressure Boundary (BWR) m G),

00 Structure a Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.C1.R-225 IV.C1-5(R- Isolation Steel; Reactor coolant Cracking Chapter XI.M1, "ASME Section XI Yes, detection I\)

225) condenser stainless due to cyclic Inservice Inspection, Subsections of aging components steel loading IWB, IWC, and IWD" for Class 1 effects is to be components evaluated The AMP in Chapter XI.M1 is to be augmented to detect cracking due to cyclic loading and verification of the program's effectiveness is necessary to ensure that significant degradation is not occurring and the component intended function will be maintained during the extended period of

<: operation. An acceptable verification

() program includes temperature and

, radioactivity monitoring of the shell

.j:>.

side water, and eddy current testing of tubes.

IV.C1.RP-39 IV.C1-6(R- Isolation Steel; Reactor coolant Loss of material Chapter XI.M1, "ASME Section XI No

16) condenser stainless due to general Inservice Inspection, Subsections components steel (steel only), IWB, IWC, and IWD," and pitting, and Chapter XI.M2, "Water Chemistry" crevice corrosion IV.C1.R-23 IV.C1-7(R- Piping, piping Steel Reactor coolant Wall thinning Chapter XI.M17, "Flow-Accelerated No

23) components, due to flow- Corrosion" and piping accelerated elements corrosion IV.C1.R-21 IV.C1-8(R- Piping, piping Nickel alloy Reactor coolant Cracking Chapter XI.M7, "BWR Stress No 0 21) components, due to stress Corrosion Cracking," and CD C') and piping corrosion Chapter XI.M2, "Water Chemistry" CD 0 elements cracking, G) 3 0- greater than intergranular CD 0 I\) or equal to 4 stress corrosion 0 a NPS cracking 0

--" a 0)

CD 0

I 0

0 N

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CD C1 Reactor Coolant Pressure Boundary (BWR) 3 0-CD Structure I\) Aging Effect/ Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.C1.R-20 IV.C1-9(R- Piping, piping Stainless Reactor coolant Cracking Chapter XI.M7, "BWR Stress No

20) components, steel due to stress Corrosion Cracking," and and piping corrosion Chapter XI.M2, "Water Chemistry" elements cracking, greater than intergranular or equal to 4 stress corrosion NPS cracking IV.C1.RP-44 IV.C1- Pump and Steel; System Cumulative Fatigue is a TLAA evaluated for the Yes, TLAA 11 (R-28) valve closure stainless temperature up fatigue damage period of extended operation; check bolting steel to 288°C due to fatigue ASME Code limits for allowable cycles (550°F) (less than 7000 cycles) of thermal stress range. (SRP Sec 4.3 "Metal

<: Fatigue," for acceptable methods to

() comply with 10 CFR 54.21 (c)(1))

I 0"1 IV.C1.RP-158 IV.C1- Reactor Steel (with Reactor coolant Loss of material Chapter XI.M2, "Water Chemistry," No 14(RP-27) coolant stainless due to pitting and and pressure steel or crevice corrosion Chapter XI.M32, "One-Time boundary nickel-alloy Inspection" components cladding);

stainless steel; nickel alloy IV.C1.R-220 IV.C1- Reactor Steel (with Reactor coolant Cumulative Fatigue is a TLAA evaluated for the Yes, TLAA 15(R-220) coolant or without fatigue damage period of extended operation, and for pressure nickel-alloy due to fatigue Class 1 components environmental z boundary or stainless effects on fatigue are to be addressed.

c components: steel (See SRP, Sec 4.3 "Metal Fatigue," for

u m piping, piping cladding); acceptable methods to comply with 10 G)

I components, stainless CFR 54.21 (c)(1))

0 a 00 and piping steel; nickel G) elements alloy

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C2. REACTOR COOLANT SYSTEM AND CONNECTED LINES (PRESSURIZED WATER REACTOR)

Systems, Structures, and Components This section addresses the pressurized water reactor (PWR) primary coolant pressure boundary and consists of the reactor coolant system and portions of other connected systems generally extending up to and including the second containment isolation valve or to the first anchor point and including the containment isolation valves, the reactor coolant pump, valves, pressurizer, and the pressurizer relief tank. The connected systems include the residual heat removal (RH R) or low pressure injection system, high pressure injection system, sampling system, and the small-bore piping. With respect to other systems such as the core flood system (CFS) or the safety injection tank (SIT) and the chemical and volume control system (CVCS), the isolation valves associated with the boundary between ASME Code class 1 and 2 are located inside the containment. Based on Regulatory Guide 1.26, "Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants," and with the exception of the pressurizer relief tank, which is governed by Group B Quality Standards, all systems, structures, and components that comprise the reactor coolant system are governed by Group A Quality Standards. The recirculating pump seal water heat exchanger is discussed in V.D1.

Pump and valve internals perform their intended functions with moving parts or with a change in configuration, or are subject to replacement based on qualified life or specified time period.

Pursuant to 10 CFR 54.21 (a)(1), therefore, they are not subject to an aging management review.

Common miscellaneous material/environment combinations where aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation are included in IV.E.

System Interfaces The systems that interface with the reactor coolant pressure boundary include the reactor pressure vessel (lV.A2), the steam generators (lV.D1 and IV.D2), the emergency core cooling system (V.D1), and the chemical and volume control system (VII.E1).

December 201 0 IV C2-1 NUREG-1801, Rev. 2 OAGI0001390_00259

z IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM c

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Mechanism Evaluation

u Component CD

< IV.C2.RP-235 IV.C2-1 (R- Class 1 piping, Stainless Reactor coolant Cracking Chapter XI.M1, "ASME Section XI No I\)

02) fittings and steel; steel due to stress Inservice Inspection, Subsections branch with corrosion IWB, IWC, and IWD" for Class 1 connections stainless cracking, components,

< NPS4 steel intergranular Chapter XI.M2, "Water Chemistry,"

cladding stress corrosion and cracking (for XI.M35, "One-Time Inspection of stainless steel ASME Code Class 1 Small-bore only), and Piping" thermal, mechanical, and vibratory loading

<: IV.C2.R-05 IV.C2-3(R- Class 1 piping, Cast Reactor coolant Cracking Monitoring and control of primary Yes, plant-() 05) piping austenitic due to stress water chemistry in accordance with specific I\)

I I\) components, stainless corrosion cracking EPRI 1014986 minimize the potential and piping steel for SCC. Material selection according elements to NUREG-0313, Rev. 2, guidelines of

0.035% C and 2

7.5% ferrite reduces susceptibility to SCC.

For CASS components that do not meet either one of the above, a plant-specific aging management program is evaluated The program is to include (a) adequate inspection methods to ensure detection of cracks, and (b) flaw evaluation methodology for CASS components that are 0

CD C')

susceptible to thermal aging 0 CD embrittiement.

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CD C2 Reactor Coolant System and Connected Lines (PWR) 3 0-CD Structure I\) Aging Effect/ Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.C2.R-52 IV.C2-4(R- Class 1 piping, Cast Reactor coolant Loss of fracture Chapter XI.M12, "Thermal Aging No

52) piping austenitic >250°C toughness Embrittlement of Cast Austenitic components, stainless (>482°F) due to thermal Stainless Steel (CASS)"

and piping steel aging elements embrittlement IV.C2.RP-344 IV.C2-2(R- Class 1 piping, Stainless Reactor coolant Cracking Chapter XI.M1, "ASME Section XI No

07) piping steel; steel due to stress Inservice Inspection, Subsections components, with corrosion cracking IWB, IWC, and IWD" for Class 1 and piping stainless components, and elements steel Chapter XI.M2, "Water Chemistry" cladding IV.C2.R-09 IV.C2-5(R- Class 1 pump Steel (with Reactor coolant Cracking Chapter XI.M1, "ASME Section XI No

<: 09) casings; valve stainless due to stress Inservice Inspection, Subsections

()

I\) bodies steel corrosion cracking IWB, IWC, and IWD" for Class 1 I

CJ.) cladding); components, and stainless Chapter XI.M2, "Water Chemistry" steel IV.C2.R-08 IV.C2-6(R- Class 1 pump Cast Reactor coolant Loss of fracture Chapter XI.M1, "ASME Section XI No

08) casings; valve austenitic >250°C toughness Inservice Inspection, Subsections bodies and stainless (>482°F) due to thermal IWB, IWC, and IWD" for Class 1 bonnets steel aging components embrittlement For pump casings and valve bodies, screening for susceptibility to thermal aging is not necessary. The ASME Section XI inspection requirements z are sufficient for managing the effects c of loss of fracture toughness due to

u m thermal aging embrittiement of CASS G)

I pump casings and valve bodies.

0 a 00 G)

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Mechanism Evaluation

u Component CD

< IV.C2.R-11 IV.C2-7(R- Closure bolting High- Air with reactor Cracking Chapter XI.M18, "Bolting Integrity" No I\)

11 ) strength, coolant leakage due to stress low-alloy corrosion cracking steel; stainless steel IV.C2.R-12 IV.C2-8(R- Closure bolting Low-alloy Air (with reactor Loss of preload Chapter XI.M18, "Bolting Integrity" No

12) steel, coolant due to thermal stainless leakage) effects, gasket steel creep, and self-loosening IV.C2.RP-166 Closure bolting Steel Air - indoor, Loss of material Chapter XI.M18, "Bolting Integrity" No

<: uncontrolled due to general,

()

I\) pitting, and I

.j:>. crevice corrosion IV.C2.RP-167 Closure bolting Steel Air with borated Loss of material Chapter XI.M1 0, "Boric Acid No water leakage due to boric acid Corrosion" corrosion IV.C2.R-17 IV.C2-9(R- External Steel Air with borated Loss of material Chapter XI.M1 0, "Boric Acid No

17) surfaces water leakage due to boric acid Corrosion" corrosion IV.C2.RP-380 IV.C2-9(R- External Steel Air with borated Loss of material Chapter XI.M1 0, "Boric Acid No

17) surfaces: water leakage due to boric acid Corrosion," and reactor coolant corrosion Chapter XI.M11 B, "Cracking of pressure Nickel-Alloy Components and Loss of boundary piping Material Due to Boric Acid-Induced 0 or components Corrosion in RCPB Components CD adjacent to (PWRs Only)"

C')

CD 0 dissimilar metal G) 3 0- (Alloy 82/1 82)

CD welds 0 I\)

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CD C2 Reactor Coolant System and Connected Lines (PWR) 3 0-CD Structure I\) Aging Effect/ Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.C2.R-18 IV.C2- Piping and Steel; System Cumulative Fatigue is a time-limited aging Yes, TLAA 10(R-18) components stainless temperature up fatigue damage analysis (TLAA) to be evaluated for (External steel to 340°C due to fatigue the period of extended operation. See surfaces); (644 OF) the SRP, Section 4.3 "Metal Fatigue,"

bolting for acceptable methods for meeting the requirements of 10 CFR 54.21 (c)(1).

IV.C2.RP-222 IV.C2- Piping, piping Copper Closed-cycle Loss of material Chapter XI.M21A, "Closed Treated No 11 (RP-11) components, alloy cooling water due to pitting, Water Systems" and piping crevice, and elements galvanic corrosion IV.C2.RP-12 IV.C2- Piping, piping Copper Closed-cycle Loss of material Chapter XI.M33, "Selective Leaching" No

<: 12(RP-12) components, alloy (>15% cooling water due to selective

()

I\) and piping Zn or >8% leaching I

0"1 elements AI)

IV.C2.RP-159 IV.C2- Piping, piping Nickel alloy Reactor coolant Cracking Chapter XI.M1, "ASME Section XI No 13(RP-31) components, or steam due to primary Inservice Inspection, Subsections and piping water stress IWB, IWC, and IWD" for Class 1 elements corrosion cracking components, and Chapter XI.M2, "Water Chemistry,"

and Chapter XI.M11 B, "Cracking of Nickel-Alloy Components and Loss of Material Due to Boric Acid-Induced Corrosion in RCPB Components z (PWRs Only)"

c

u IV.C2.RP-221 IV.C2- Piping, piping Steel Closed-cycle Loss of material Chapter XI.M21A, "Closed Treated No m 14(RP-10) components, cooling water due to general, Water Systems" G)

I and piping pitting, and 0 a 00 elements crevice corrosion G)

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Mechanism Evaluation

u Component CD

< IV.C2.RP-23 IV.C2- Piping, piping Steel (with Reactor coolant Loss of material Chapter XI.M2, "Water Chemistry" No I\)

15(RP-23) components, stainless due to pitting and and piping steel or crevice corrosion elements; nickel-alloy flanges; heater cladding);

sheaths and stainless sleeves; steel; nickel penetrations; alloy thermal sleeves; vessel shell heads and welds

<: IV.C2.R-58 IV.C2- Pressurizer Steel (with Reactor coolant Cracking Chapter XI.M1, "ASME Section XI No

() 18(R-58) components stainless due to cyclic Inservice Inspection, Subsections I\)

I (J) steel or loading IWB, IWC, and IWD" for Class 1 nickel-alloy components, and cladding); Chapter XI.M2, "Water Chemistry" stainless Cracks in the pressurizer cladding steel could propagate from cyclic loading into the ferrite base metal and weld metal. However, because the weld metal between the surge nozzle and the vessel lower head is subjected to the maximum stress cycles and the area is periodically inspected as part of the lSI program, the existing AMP is adequate for managing the effect of pressurizer clad cracking.

0 CD C')

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CD C2 Reactor Coolant System and Connected Lines (PWR) 3 0-CD Structure I\) Aging Effect/ Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.C2.R-2S IV.C2- Pressurizer Steel (with Reactor coolant Cracking Chapter XI.M1, "ASME Section XI No 19(R-2S) components stainless due to stress Inservice Inspection, Subsections steel or corrosion IWB, IWC, and IWD" for Class 1 nickel-alloy cracking, primary components, and cladding); water stress Chapter XI.M2, "Water Chemistry" stainless corrosion cracking steel IV.C2.R-217 IV.C2- Pressurizer Stainless Reactor coolant Cracking Chapter XI.M1, "ASME Section XI No 20(R-217) heater sheaths steel due to stress Inservice Inspection, Subsections and sleeves; corrosion cracking IWB, IWC, and IWD" for Class 1 heater bundle components, and diaphragm plate Chapter XI.M2, "Water Chemistry"

<: IV.C2.RP-37 IV.C2- Pressurizer Nickel alloy; Reactor coolant Cracking Chapter XI.M1, "ASME Section XI No

()

I\) 21 (R-06) instrumentation nickel-alloy due to primary Inservice Inspection, Subsections I

-J penetrations; cladding water stress IWB, IWC, and IWD" for Class 1 heater sheaths corrosion cracking components, and and sleeves; Chapter XI.M2, "Water Chemistry,"

heater bundle and diaphragm Chapter XI.M11 B, "Cracking of plate; manways Nickel-Alloy Components and Loss of and flanges Material Due to Boric Acid-Induced Corrosion in RCPB Components (PWRs Only)"

IV.C2.RP-231 IV.C2- Pressurizer Stainless Treated borated Cracking Chapter XI.M1, "ASME Section XI No 22(R-14) relief tank: tank steel; steel water >60°C due to stress Inservice Inspection, Subsections z shell and with (>140°F) corrosion cracking IWB, IWC, and IWD" for ASME Code c heads; flanges; stainless components, and

u m nozzles steel Chapter XI.M2, "Water Chemistry" G)

I cladding 0 a 00 G)

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I 00 Structure a Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.C2.R-13 IV.C2- Pressurizer Steel (with Treated borated Cumulative Fatigue is a time-limited aging Yes, TLAA I\)

23(R-13) relief tank: tank stainless water fatigue damage analysis (TLAA) to be evaluated for shell and steel or due to fatigue the period of extended operation. See heads; flanges; nickel-alloy the SRP, Section 4.3 "Metal Fatigue,"

nozzles cladding) for acceptable methods for meeting the requirements of 10 CFR 54.21 (c)(1).

IV.C2.RP-383 Pressurizer Stainless Treated borated Cracking Chapter XI.M2, "Water Chemistry," No relief tank: tank steel; steel water >60°C due to stress and shell and with (>140°F) corrosion cracking Chapter XI.M32, "One-Time heads; flanges; stainless Inspection" nozzles (non- steel

<: ASME Section cladding

() XI components)

I\)

I 00 IV.C2.RP-156 IV.C2- Pressurizer Nickel alloy Reactor coolant Cracking Chapter XI.M1, "ASME Section XI No 24(RP-22) surge and or steam due to primary Inservice Inspection, Subsections steam space water stress IWB, IWC, and IWD" for Class 1 nozzles; welds corrosion cracking components, and Chapter XI.M2, "Water Chemistry,"

and Chapter XI.M11 B, "Cracking of Nickel-Alloy Components and Loss of Material Due to Boric Acid-Induced Corrosion in RCPB Components (PWRs Only)"

IV.C2.R-19 IV.C2- Pressurizer: Steel; Air with metal Cracking Chapter XI.M1, "ASME Section XI No 16(R-19) integral support stainless temperature up due to cyclic Inservice Inspection, Subsections 0 steel to 288°C loading IWB, IWC, and IWD" for Class 1 CD C') (550°F) components CD 0

G) 3 0-CD 0 I\)

0 a 0

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CD 0

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(J)

(J)

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CD C2 Reactor Coolant System and Connected Lines (PWR) 3 0-CD Structure I\) Aging Effect/ Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.C2.RP-40 IV.C2- Pressurizer: Nickel alloy Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No 17(R-24) spray head due to stress and corrosion Chapter XI.M32, "One-Time cracking, primary Inspection" water stress corrosion cracking IV.C2.RP-41 IV.C2- Pressurizer: Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," No 17(R-24) spray head steel due to stress and corrosion cracking Chapter XI.M32, "One-Time Inspection" IV.C2.R-223 IV.C2- Reactor coolant Steel (with Reactor coolant Cumulative Fatigue is a TLAA evaluated for the Yes, TLAA 25(R-223) pressure or without fatigue damage period of extended operation, and for

<: boundary nickel-alloy due to fatigue Class 1 components environmental

()

I\) components: or stainless effects on fatigue are to be I

CD piping, piping steel addressed. (See SRP, Sec 4.3 "Metal components, cladding); Fatigue," for acceptable methods to and piping stainless comply with 10 CFR 54.21 (c)(1))

elements; steel; nickel flanges; nozzles alloy and safe ends; pressurizer vessel shell heads and welds; heater sheaths and z sleeves; c penetrations;

u m thermal sleeves G)

I 0 a 00 G)

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I 00 Structure a Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.C2.R-56 IV.C2- Reactor coolant Steel (with Reactor coolant Cracking Chapter XI.M1, "ASME Section XI No I\)

26(R-56) system piping stainless due to cyclic Inservice Inspection, Subsections and fittings: cold steel loading IWB, IWC, and IWD" for Class 1 leg; hot leg; cladding); components surge line; stainless spray line steel IV.C2.R-30 IV.C2- Reactor coolant Steel (with Reactor coolant Cracking Chapter XI.M1, "ASME Section XI No 27(R-30) system piping stainless due to stress Inservice Inspection, Subsections and fittings: cold steel corrosion cracking IWB, IWC, and IWD" for Class 1 leg; hot leg; cladding); components, and surge line; stainless Chapter XI.M2, "Water Chemistry" spray line steel

()

I\)

I a

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01. STEAM GENERATOR (RECIRCULATING)

Systems, Structures, and Components This section addresses the recirculating-type steam generators, as found in Westinghouse and Combustion Engineering pressurized water reactors (PWRs), including all internal components and water/steam nozzles and safe ends. Based on Regulatory Guide 1.26, "Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants," the primary water side (tube side) of the steam generator is governed by Group A Quality Standards, and the secondary water side is governed by Group B Quality Standards.

Common miscellaneous material/environment combinations where aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation are included in IV.E.

System Interfaces The systems that interface with the steam generators include the reactor coolant system and connected lines (lV.C2), the containment isolation components (V.C), the main steam system (VIII.B1), the feedwater system (VIII.D1), the steam generator blowdown system (VIII.F), and the auxiliary feedwater system (VIII.G).

December 201 0 IV 01-1 NUREG-1801, Rev. 2 OAGI0001390_00269

z IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM c D1 Steam Generator (Recirculating)

U m

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I Structure 00 a Aging Effect! Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.D1.R-10 IV.D1-2(R- Closure bolting Steel Air with reactor Cracking Chapter XI.M18, "Bolting Integrity" No I\) 10) coolant leakage due to stress corrosion cracking IV.D1.RP-46 IV.D1- Closure bolting Steel; Air - indoor, Loss of preload Chapter XI.M18, "Bolting Integrity" No 10(R-32) stainless uncontrolled due to thermal steel (External) effects, gasket creep, and self-loosening IV.D1.R-17 IV.D1-3(R- External Steel Air with borated Loss of material Chapter XI.M1 0, "Boric Acid No

17) surfaces water leakage due to boric acid Corrosion" corrosion

<: IV.D1.RP-36 IV.D1-4(R- Instrument Steel (with Reactor coolant Cracking Chapter XI.M1 , "ASME Section XI No o 01) penetrations nickel-alloy due to primary Inservice Inspection, Subsections I

I\)

and primary cladding); water stress IWB, IWC, and IWD" for Class 1 side nozzles; nickel alloy corrosion components, and safe ends; cracking Chapter XI.M2, "Water Chemistry,"

welds and Chapter XI.M11 B, "Cracking of Nickel-Alloy Components and Loss of Material Due to Boric Acid-Induced Corrosion in RCPB Components (PWRs Only)"

IV.D1.R-37 IV.D1-5(R- Pressure Steel Secondary Wall thinning Chapter XI.M17, "Flow-Accelerated No

37) boundary and feedwater or due to flow- Corrosion" structural: steam accelerated steam nozzle corrosion 0 and safe end; CD C') feedwater 0 CD G) 3 0-nozzle and safe CD end 0

0 I\)

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CD D1 Steam Generator (Recirculating) 3 0-CD Structure I\) Aging Effect! Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.D1.RP-17 IV.D1- Primary side Stainless Reactor coolant Cracking Chapter XI.M2, "Water Chemistry" No 7(RP-17) components: steel due to stress divider plate corrosion cracking IV.D1.RP-367 IV.D1- Primary side Steel (with Reactor coolant Cracking Chapter XI.M2, "Water Chemistry" Yes, detection 6(RP-21) components: nickel-alloy due to primary For nickel alloy divider plate of aging divider plate cladding); water stress assemblies and associated welds effects is to be nickel alloy corrosion made of Alloy 600, effectiveness of evaluated cracking the chemistry control program should be verified to ensure that cracking due to PWSCC is not occurring.

IV.D1.R-221 IV.D1-8(R- Recirculating Steel (with Reactor coolant Cumulative Fatigue is a TLAA evaluated for the Yes, TLAA

<: 221) steam or without fatigue damage period of extended operation, and for o generator nickel-alloy due to fatigue Class 1 components environmental I

CJ.) components: or stainless effects on fatigue are to be flanges; steel addressed. (See SRP, Sec 4.3 "Metal penetrations; cladding); Fatigue," for acceptable methods to nozzles; safe stainless comply with 10 CFR 54.21 (c)(1))

ends; lower steel; nickel heads and alloy welds IV.D1.RP-372 Steam Steel Secondary Loss of material Chapter XI.M2, "Water Chemistry" No generator feedwater or due to general, and components: steam pitting, and Chapter XI.M32, "One-Time shell assembly crevice corrosion Inspection" z

c

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U m

G)

I Structure 00 a Aging Effect! Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.D1.R-33 IV.D1- Steam Steel Secondary Cumulative Fatigue is a time-limited aging Yes, TLAA I\) 11 (R-33) generator feedwater or fatigue damage analysis (TLAA) to be evaluated for components: steam due to fatigue the period of extended operation. See top head; the SRP, Section 4.3 "Metal Fatigue,"

steam nozzle for acceptable methods for meeting and safe end; the requirements of 10 CFR upper and 54.21 (c)(1).

lower shell; feedwater (FW) and auxiliary FW nozzle and safe end; FW impingement

<: plate and o

I support

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0 CD G) 3 0-0 CD 0 I\)

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CD D1 Steam Generator (Recirculating) 3 0-CD Structure I\) Aging Effect! Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.D1.RP-368 IV.D1- Steam Steel Secondary Loss of material Chapter XI.M1 , "ASME Section XI Yes, detection 12(R-34) generator feedwater or due to general, Inservice Inspection, Subsections of aging components: steam pitting, and IW8, IWC, and IWD" for Class 2 effects is to be upper and crevice corrosion components, and evaluated lower shell; Chapter XI.M2, "Water Chemistry" transition cone; As noted in NRC IN 90-04, if general new transition and pitting corrosion of the shell cone closure exists, Chapter XI.M1 methods may weld not be sufficient to detect general and pitting corrosion (and the resulting corrosion-fatigue cracking), and additional inspection procedures are

<: to be developed. This issue is limited o to Westinghouse Model 44 and 51 I

0"1 Steam Generators where a high stress region exists at the shell to transition cone weld. The new transition is only applicable to replacement recirculating steam generators.

IV.D1.R-39 IV.D1- Steam Steel Secondary Loss of material A plant-specific aging management Yes, plant-13(R-39) generator feedwater due to erosion program is to be evaluated specific feedwater impingement plate and z support c IV.D1.RP-48 IV.D1- Steam Steel Secondary Wall thinning Chapter XI.M19, "Steam Generators," No

u m 16(R-41) generator feedwater or due to flow- and G)

I structural: tube steam accelerated Chapter XI.M2, "Water Chemistry" 0 a 00 support lattice corrosion and G) bars general corrosion

U 0 CD 0

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U m

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I Structure 00 a Aging Effect! Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.D1.R-42 IV.D1- Steam Steel Secondary Ligament Chapter XI.M19, "Steam Generators," No I\) 17(R-42) generator feedwater or cracking and structural: tube steam due to corrosion Chapter XI.M2, "Water Chemistry" support plates IV.D1.RP-384 IV.D1- Steam Steel; Secondary Cracking Chapter XI.M19, "Steam Generators," No 14(RP-14) generator chrome feedwater or due to stress and structural: U- plated steel; steam corrosion Chapter XI.M2, "Water Chemistry" bend supports stainless cracking or other including anti- steel; nickel mechanism(s) vibration bars alloy IV.D1.RP-225 IV.D1- Steam Steel; Secondary Loss of material Chapter XI.M19, "Steam Generators" No 15(RP-15) generator chrome feedwater or due to fretting

<: structural: U- plated steel; steam o bend supports stainless I

(J) including anti- steel; nickel vibration bars alloy IV.D1.RP-226 IV.D1- Steam Steel; Secondary Loss of material Chapter XI.M19, "Steam Generators," No 15(RP-15) generator chrome feedwater or due to general and structural: U- plated steel; steam (steel only), Chapter XI.M2, "Water Chemistry" bend supports stainless pitting, and including anti- steel; nickel crevice corrosion vibration bars alloy IV.D1.RP-232 IV.D1-1 (R- Steam Stainless Reactor coolant Cracking Chapter XI.M1 , "ASME Section XI No

07) generator: steel; steel due to stress Inservice Inspection, Subsections primary with corrosion IW8, IWC, and IWD" for Class 1 nozzles; nozzle stainless cracking components, and to safe end steel Chapter XI.M2, "Water Chemistry" 0 welds; cladding CD C') manways; 0 CD flanges G) 3 0-0 CD 0 I\)

0 a

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CD 0

I 0

0 N

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oCD IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C')

CD D1 Steam Generator (Recirculating) 3 0-CD Structure I\) Aging Effect! Further a Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component IV.D1.RP-161 IV.D1- Steam Steel Secondary Loss of material Chapter XI.M19, "Steam Generators," No 9(RP-16) generator: feedwater or due to erosion, and Tube bundle steam general, pitting, Chapter XI.M2, "Water Chemistry" wrapper and and crevice associated corrosion supports and mounting hardware IV.D1.R-40 IV.D1- Tube plugs Nickel alloy Reactor coolant Cracking Chapter XI.M19, "Steam Generators," No 18(R-40) due to primary and water stress Chapter XI.M2, "Water Chemistry" corrosion

<: cracking o IV.D1.R-43 IV.D1- Tubes Nickel alloy Secondary Changes in Chapter XI.M19, "Steam Generators," No I

-J 19(R-43) feedwater or dimension and steam ("denting") Chapter XI.M2, "Water Chemistry" due to corrosion of carbon steel tube support plate IV.D1.R-44 IV.D1- Tubes and Nickel alloy Reactor coolant Cracking Chapter XI.M19, "Steam Generators," No 20(R-44) sleeves due to primary and water stress Chapter XI.M2, "Water Chemistry" corrosion cracking z IV.D1.R-46 IV.D1- Tubes and Nickel alloy Reactor coolant Cumulative Fatigue is a time-limited aging Yes, TLAA c

u 21 (R-46) sleeves and secondary fatigue damage analysis (TLAA) to be evaluated for m feedwater/steam due to fatigue the period of extended operation. See G)

I the SRP, Section 4.3 "Metal Fatigue,"

0 a 00 for acceptable methods for meeting G) the requirements of 10 CFR

U 0 CD 54.21 (c)(1).

0 0  :<

--" I\)

0)

CD 0

I 0

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z IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM c D1 Steam Generator (Recirculating)

U m

G)

I Structure 00 a Aging Effect! Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.D1.R-48 IV.D1- Tubes and Nickel alloy Secondary Cracking Chapter XI.M19, "Steam Generators," No I\) 22(R-48) sleeves feedwater or due to and steam intergranular Chapter XI.M2, "Water Chemistry" attack IV.D1.R-47 IV.D1- Tubes and Nickel alloy Secondary Cracking Chapter XI.M19, "Steam Generators," No 23(R-47) sleeves feedwater or due to outer and steam diameter stress Chapter XI.M2, "Water Chemistry" corrosion cracking IV.D1.RP-233 IV.D1- Tubes and Nickel alloy Secondary Loss of material Chapter XI.M19, "Steam Generators" No 24(R-49) sleeves feedwater or due to fretting steam and wear

<: IV.D1.R-50 IV.D1- Tubes and Nickel alloy Secondary Loss of material Chapter XI.M19, "Steam Generators," No o 25(R-50) sleeves feedwater or due to wastage and I

00 (exposed to steam and pitting Chapter XI.M2, "Water Chemistry" phosphate corrosion chemistry)

IV.D1.RP-385 Tube-to-tube Nickel alloy Reactor coolant Cracking Chapter XI.M2, "Water Chemistry" Yes, plant-sheet welds due to primary A plant-specific program is to be specific.

water stress evaluated; the effectiveness of the corrosion water chemistry program should be cracking verified to ensure cracking is not occurring.

IV.D1.RP-49 IV.D1- Upper Steel Secondary Wall thinning Chapter XI.M19, "Steam Generators," No 26(R-51) assembly and feedwater or due to flow- and separators steam accelerated Chapter XI.M2, "Water Chemistry" including: corrosion 0 feedwater inlet CD 0

C')

CD ring and G) 3 0-support 0

CD 0 I\)

0 a

--" a 0)

CD 0

I 0

0 N

-....J (J)

D2.STEAM GENERATOR (ONCE-THROUGH)

Systems, Structures, and Components This section addresses the once-through type steam generators, as found in Babcock & Wilcox pressurized water reactors (PWRs), including all internal components and water/steam nozzles and safe ends. Based on Regulatory Guide 1.26, "Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants,"

the primary water side (tube side) of the steam generator is governed by Group A Quality Standards, and the secondary water side is governed by Group B Quality Standards.

Common miscellaneous material/environment combinations where aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation are included in IV.E.

System Interfaces The systems that interface with the steam generators include the reactor coolant system and connected lines (lV.C2), the main steam system (VIII.B1), the feedwater system (VIII.D1), the steam generator blowdown system (VII I. F), and the auxiliary feedwater system (VIII.G).

December 201 0 IV 02-1 NUREG-1801, Rev. 2 OAGI0001390_00277

z IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM c

U D2 Steam Generator (Once-Through) m G)

I 00 a Structure and/or Aging Effect/ Aging Management Program Further Item Link Material Environment Component Mechanism (AMP) Evaluation

u CD

< IV.D2.RP-46 IV.D2- Closure bolting Steel; Air - indoor, Loss of preload Chapter XI.M18, "Bolting Integrity" No I\)

6(R-32) stainless uncontrolled due to thermal steel (External) effects, gasket creep, and self-loosening IV.D2.R-17 IV.D2- External surfaces Steel Air with borated Loss of material Chapter XI.M1 0, "Boric Acid No 1(R-17) water leakage due to boric acid Corrosion" corrosion IV.D2.RP-36 IV.D2- Instrument Steel (with Reactor coolant Cracking Chapter XI.M1, "ASME Section XI No 2(R-01) penetrations and nickel-alloy due to primary Inservice Inspection, Subsections primary side cladding); water stress IWB, IWC, and IWD" for Class 1 nozzles; safe nickel alloy corrosion components, and

<: ends; welds cracking Chapter XI.M2, "Water Chemistry,"

o I\)

I and I\)

Chapter XI.M11 B, "Cracking of Nickel-Alloy Components and Loss of Material Due to Boric Acid-Induced Corrosion in RCPB Components (PWRs Only)"

IV.D2.R-222 IV.D2- Once-through Steel (with Reactor coolant Cumulative Fatigue is a TLAA evaluated for the Yes, TLAA 3(R-222) steam generator or without fatigue damage period of extended operation, and for components: nickel-alloy due to fatigue Class 1 components environmental primary side or stainless effects on fatigue are to be nozzles, safe steel addressed. (See SRP, Sec 4.3 ends, and welds cladding); "Metal Fatigue," for acceptable stainless methods to comply with 10 CFR steel; nickel 54.21 (c) (1 ))

0 alloy CD C')

CD 0

G) 3 0-CD 0 I\)

0 a 0

--" a 0)

CD 0

I 0

0 N

-....J CD

oCD IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C')

CD D2 Steam Generator (Once-Through) 3 0-CD I\) Structure and/or Aging Effect/ Aging Management Program Further a Item Link Material Environment a Component Mechanism (AMP) Evaluation IV.D2.RP-47 IV.D2- Primary side Steel (with Reactor coolant Cracking Chapter XI.M1, "ASME Section XI No 4(R-35) components: stainless due to stress Inservice Inspection, Subsections upper and lower steel or corrosion IWB, IWC, and IWO" for Class 1 heads, and tube nickel-alloy cracking, primary components, and sheet welds cladding) water stress Chapter XI.M2, "Water Chemistry" exposed to corrosion reactor coolant cracking IV.D2.R-31 IV.D2- Secondary Steel Air with leaking Loss of material Chapter XI.M1, "ASME Section XI No 5(R-31) manway covers; secondary-side due to erosion Inservice Inspection, Subsections hand hole covers water and/or IWB, IWC, and IWO" for Class 2 steam components IV.D2.R-36 IV.D2- Steam generator Nickel alloy Secondary Cracking Chapter XI.M2, "Water Chemistry," No

<: 9(R-36) components: feedwater or due to stress and o

I\) secondary side steam corrosion Chapter XI.M32, "One-Time I

CJ.) nozzles (vent, cracking Inspection," or drain, and Chapter XI.M1, "ASME Section XI instrumentation) Inservice Inspection, Subsections IWB, IWC, and IWO."

IV.D2.R-38 IV.D2- Steam generator Steel Secondary Wall thinning Chapter XI.M17, "Flow-Accelerated No 7(R-38) components: feedwater or due to flow- Corrosion" feedwater (FW) steam accelerated and auxiliary FW corrosion nozzles and safe ends; steam nozzles and safe z ends c IV.D2.RP-153 IV.D2- Steam generator Steel Secondary Loss of material Chapter XI.M2, "Water Chemistry," No

u m 8(R-224) components

shell feedwater or due to general, and G)

I assembly steam pitting, and Chapter XI.M32, "One-Time 0 a 00 crevice corrosion Inspection" G)

U 0 CD 0

0  :<

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0)

CD 0

I 0

0 N

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z IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM c

U D2 Steam Generator (Once-Through) m G)

I 00 a Structure and/or Aging Effect/ Aging Management Program Further Item Link Material Environment Component Mechanism (AMP) Evaluation

u CD

< IV.D2.R-33 IV.D2- Steam generator Steel Secondary Cumulative Fatigue is a time-limited aging Yes, TLAA I\)

10(R-33) components: top feedwater or fatigue damage analysis (TLAA) to be evaluated for head; steam steam due to fatigue the period of extended operation.

nozzle and safe See the SRP, Section 4.3 "Metal end; upper and Fatigue," for acceptable methods for lower shell; meeting the requirements of 10 CFR feedwater (FW) 54.21 (c)(1).

and auxiliary FW nozzle and safe end; FW impingement plate and support

<: IV.D2.R-42 IV.D2- Steam generator Steel Secondary Ligament Chapter XI.M19, "Steam No o

I\)

11 (R-42) structural: tube feedwater or cracking Generators," and I

.j:>. support plates steam due to corrosion Chapter XI.M2, "Water Chemistry" IV.D2.RP-162 Steam generator: Steel Secondary Loss of material Chapter XI.M19, "Steam No tube bundle feedwater or due to erosion, Generators," and wrapper and steam general, pitting, Chapter XI.M2, "Water Chemistry" associated and crevice supports and corrosion mounting hardware IV.D2.R-40 IV.D2- Tube plugs Nickel alloy Reactor coolant Cracking Chapter XI.M19, "Steam No 12(R-40) due to primary Generators," and water stress Chapter XI.M2, "Water Chemistry" corrosion cracking 0

CD C')

CD 0

G) 3 0-CD 0 I\)

0 a 0

--" a 0)

CD 0

I 0

0 N

CD 0

oCD IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C')

CD D2 Steam Generator (Once-Through) 3 0-CD I\) Structure and/or Aging Effect/ Aging Management Program Further a Item Link Material Environment a Component Mechanism (AMP) Evaluation IV.D2.R-226 IV.D2- Tubes Nickel alloy Secondary Changes in Chapter XI.M19, "Steam No 13(R-226) feedwater or dimension Generators," and steam ("denting") Chapter XI.M2, "Water Chemistry" due to corrosion of carbon steel tube support plate IV.D2.R-44 IV.D2- Tubes and Nickel alloy Reactor coolant Cracking Chapter XI.M19, "Steam No 14(R-44) sleeves due to primary Generators," and water stress Chapter XI.M2, "Water Chemistry" corrosion cracking

<: IV.D2.R-46 IV.D2- Tubes and Nickel alloy Reactor coolant Cumulative Fatigue is a time-limited aging Yes, TLAA o

I\) 15(R-46) sleeves and secondary fatigue damage analysis (TLAA) to be evaluated for I

0"1 feedwater/steam due to fatigue the period of extended operation.

See the SRP, Section 4.3 "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR 54.21 (c)(1).

IV.D2.R-48 IV.D2- Tubes and Nickel alloy Secondary Cracking Chapter XI.M19, "Steam No 16(R-48) sleeves feedwater or due to Generators," and steam intergranular Chapter XI.M2, "Water Chemistry" attack IV.D2.R-47 IV.D2- Tubes and Nickel alloy Secondary Cracking Chapter XI.M19, "Steam No 17(R-47) sleeves feedwater or due to outer Generators," and z steam diameter stress Chapter XI.M2, "Water Chemistry" c corrosion

u m cracking G)

I IV.D2.RP-233 IV.D2- Tubes and Nickel alloy Secondary Loss of material Chapter XI.M19, "Steam Generators" No 0 a 00 18(R-49) sleeves feedwater or due to fretting G) steam and wear

U 0 CD 0

0  :<

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0)

CD 0

I 0

0 N

CD

z IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM c

U D2 Steam Generator (Once-Through) m G)

I 00 a Structure and/or Aging Effect/ Aging Management Program Further Item Link Material Environment Component Mechanism (AMP) Evaluation

u CD

< IV.D2.RP-185 IV.D2- Tube-to-tube Nickel alloy Reactor coolant Cracking due to Chapter XI.M2, "Water Chemistry" Yes, plant-I\)

4(R-35) sheet welds primary water A plant-specific program is to be specific stress corrosion evaluated; the effectiveness of the cracking water chemistry program should be verified to ensure cracking is not occurring o

I\)

I (J) 0 CD C')

CD 0

G) 3 0-CD 0 I\)

0 a 0

--" a 0)

CD 0

I 0

0 N

CD N

E. COMMON MISCELLANEOUS MATERIAL/ENVIRONMENT COMBINATIONS Systems, Structures, and Components This section addresses the aging management programs for miscellaneous material/environment combinations which may be found throughout the reactor vessel, internals and reactor coolant system's structures and components. For the material/environment combinations in this part, aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation, therefore, no resulting aging management programs for these structures and components are required.

System Interfaces The structures and components covered in this section belong to the engineered safety features in PWRs and BWRs. (For example, see System Interfaces in V.A to V.D2 for details.)

December 201 0 IV E-1 NUREG-1801, Rev. 2 OAGI0001390_00283

z IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM c

U m E Common Miscellaneous Material Environment Combinations G)

I 00 Structure a Aging Effect! Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

u Component CD

< IV.E.RP-03 IV.E-1 (RP- Piping, piping Nickel alloy Air - indoor, None None No I\)

03) components, uncontrolled and piping (External) elements IV.E.RP-378 Piping, piping Nickel alloy Air with borated None None No components, water leakage and piping elements IV.E.RP-04 IV.E-2(RP- Piping, piping Stainless Air - indoor, None None No

04) components, steel uncontrolled and piping (External) elements

<: IV.E.RP-05 IV.E-3(RP- Piping, piping Stainless Air with borated None None No m I 05) components, steel water leakage I\)

and piping elements IV.E.RP-06 IV.E-4(RP- Piping, piping Stainless Concrete None None No

06) components, steel and piping elements IV.E.RP-07 IV.E-5(RP- Piping, piping Stainless Gas None None No

07) components, steel and piping elements IV.E.RP-353 IV.E-6(RP- Piping, piping Steel Concrete None None, provided: No, if

01) components, 1) attributes of the concrete are conditions are and piping consistent with ACI 318 or ACI 349 (low met.

0 water-to-cement ratio, low permeability, CD elements C')

CD and adequate air entrainment) as cited 0 3 G) 0-in NUREG-1557, and CD 2) plant OE indicates no degradation of 0 I\)

the concrete 0 a 0

--" a 0)

CD 0

I 0

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CD

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CHAPTER V ENGINEERED SAFETY FEATURES December 201 0 V-i NUREG-1801, Rev. 2 OAGI0001390_00285

MAJOR PLANT SECTIONS A. Containment Spray System (Pressurized Water Reactors)

B. Standby Gas Treatment System (Boiling Water Reactors)

C. Containment Isolation Components

01. Emergency Core Cooling System (Pressurized Water Reactors)

02. Emergency Core Cooling System (Boiling Water Reactors)

E. External Surfaces of Components and Miscellaneous Bolting F. Common Miscellaneous Material/Environment Combinations December 201 0 V-iii NUREG-1801, Rev. 2 OAGI0001390_00286

A. CONTAINMENT SPRAY SYSTEM (PRESSURIZED WATER REACTORS)

Systems, Structures, and Components This section addresses the containment spray system for pressurized water reactors (PWRs) designed to lower the pressure, temperature, and gaseous radioactivity (iodine) content of the containment atmosphere following a design basis event. Spray systems using chemically treated borated water are reviewed. The system consists of piping and valves, including containment isolation valves, flow elements, orifices, pumps, spray nozzles, eductors, and the containment spray system heat exchanger (for some plants).

Based on Regulatory Guide 1.26, "Quality Group Classifications and Standards for Water-,

Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants," all components that comprise the containment spray system outside or inside the containment are governed by Group B Quality Standards.

Pumps and valve internals perform their intended functions with moving parts or with a change in configuration. They are also subject to replacement based on qualified life or a specified time period. Pursuant to 10 CFR 54.21(a)(1), therefore, they are not subject to an aging management review.

Aging management programs for the degradation of external surfaces of components and miscellaneous bolting are included in V.E. Common miscellaneous material/environment combinations, where aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation, are included in V.F.

The system piping includes all pipe sizes, including instrument piping.

System Interfaces The systems that interface with the containment spray system are the PWR emergency core cooling (V.D1), and open- or closed-cycle cooling water systems (VII.C1 or VII.C2).

December 201 0 V A-1 NUREG-1801, Rev. 2 OAG10001390_00287

z V ENGINEERED SAFETY FEATURES c

U A Containment Spray System (PWR) m G)

I 00 a Structure and/or Aging Effect! Aging Management Further Item Link Material Environment Component Mechanism Program (AMP) Evaluation

u CD

< VAE-26 V.A-1 (E- Ducting, piping, and Steel Air - indoor, Loss of material Chapter XI.M36, "External No I\)

26) components (External uncontrolled due to general corrosion Surfaces Monitoring of surfaces) (External) Mechanical Components" VAEP-42 V.A- Encapsu lation Steel Air - indoor, Loss of material Chapter XI.M38, No 2(EP- components uncontrolled due to general, pitting, "Inspection of Internal

42) (Internal) and crevice corrosion Surfaces in Miscellaneous Piping and Ducting Components" VAEP-43 V.A- Encapsu lation Steel Air with Loss of material Chapter XI.M38, No 3(EP- components borated water due to general, pitting, "Inspection of Internal

43) leakage crevice, and boric acid Surfaces in Miscellaneous (Internal) corrosion Piping and Ducting

< Components"

>> I VAE-28 V.A-4(E- External surfaces Steel Air with Loss of material Chapter XI.M1 0, "Boric No I\)

28) borated water due to boric acid Acid Corrosion" leakage corrosion VAEP-94 V.A- Heat exchanger Copper alloy Closed-cycle Loss of material Chapter XI.M21A, "Closed No 5(EP- components cooling water due to pitting, crevice, Treated Water Systems"

13) and galvanic corrosion VAEP-37 V.A- Heat exchanger Copper alloy Closed-cycle Loss of material Chapter XI.M33, "Selective No 6(EP- components (>15% Zn or cooling water due to selective leaching Leaching"

37) >8% AI)

VAEP-93 V.A-7(E- Heat exchanger Stainless Closed-cycle Loss of material Chapter XI.M21A, "Closed No

19) components steel cooling water due to pitting and crevice Treated Water Systems" corrosion 0 VAEP-91 V.A-8(E- Heat exchanger Stainless Raw water Loss of material Chapter XI.M20, "Open- No CD C') 20) components steel due to pitting, crevice, Cycle Cooling Water CD 0 and microbiologically- System" G) 3 0-influenced corrosion; CD 0 I\) fouling that leads to 0 a corrosion 0

--" a 0)

CD 0

I 0

0 N

CD CD

oCD V ENGINEERED SAFETY FEATURES C')

CD A Containment Spray System (PWR) 3 0-CD I\) Structure and/or Aging Effect! Aging Management Further a Item Link Material Environment a Component Mechanism Program (AMP) Evaluation VAEP-92 V.A-9(E- Heat exchanger Steel Closed-cycle Loss of material Chapter XI.M21A, "Closed No

17) components cooling water due to general, pitting, Treated Water Systems" crevice, and galvanic corrosion VAEP-90 V.A- Heat exchanger Steel Raw water Loss of material Chapter XI.M20, "Open- No 10(E-18) components due to general, pitting, Cycle Cooling Water crevice, and System" microbiolog ically-influenced corrosion; fouling that leads to corrosion VAEP-100 V.A- Heat exchanger tubes Copper alloy Closed-cycle Reduction of heat Chapter XI.M21A, "Closed No

< 11 (EP- cooling water transfer Treated Water Systems"

>> I 39) due to fouling CJ.)

VAEP-78 V.A- Heat exchanger tubes Copper alloy Lubricating oil Reduction of heat Chapter XI.M39, No 12(EP- transfer "Lubricating Oil Analysis,"

47) due to fouling and Chapter XI.M32, "One-Time Inspection" VAEP-96 V.A- Heat exchanger tubes Stainless Closed-cycle Reduction of heat Chapter XI.M21A, "Closed No 13(EP- steel cooling water transfer Treated Water Systems"

35) due to fouling VAEP-79 V.A- Heat exchanger tubes Stainless Lubricating oil Reduction of heat Chapter XI.M39, No 14(EP- steel transfer "Lubricating Oil Analysis,"

z 50) due to fouling and c Chapter XI.M32, "One-

u m Time Inspection" G)

I VAE-21 V.A- Heat exchanger tubes Stainless Raw water Reduction of heat Chapter XI.M20, "Open- No 0 a 00 15(E-21) steel transfer Cycle Cooling Water G) due to fouling System"

U 0 CD 0

0  :<

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0)

CD 0

I 0

0 N

CD CD

z V ENGINEERED SAFETY FEATURES c

U A Containment Spray System (PWR) m G)

I 00 a Structure and/or Aging Effect! Aging Management Further Item Link Material Environment Component Mechanism Program (AMP) Evaluation

u CD

< VAEP-74 V.A- Heat exchanger tubes Stainless Treated water Reduction of heat Chapter XI.M2, "Water No I\)

16(EP- steel transfer Chemistry," and

34) due to fouling Chapter XI.M32, "One-Time Inspection" VAEP-75 V.A- Heat exchanger tubes Steel Lubricating oil Reduction of heat Chapter XI.M39, No 17(EP- transfer "Lubricating Oil Analysis,"

40) due to fouling and Chapter XI.M32, "One-Time Inspection" VAE-43 V.A- Motor cooler Gray cast iron Treated water Loss of material Chapter XI.M33, "Selective No 18(E-43) due to selective leaching Leaching"

< VAE-29 V.A- Piping and components Steel Air - indoor, Loss of material Chapter XI.M38, No

>> I 19(E-29) (Internal surfaces) uncontrolled due to general corrosion "Inspection of Internal

.j:>.

(Internal) Surfaces in Miscellaneous Piping and Ducting Components" VAEP-97 V.A- Piping, piping Copper alloy Closed-cycle Loss of material Chapter XI.M21A, "Closed No 20(EP- components, and piping cooling water due to pitting, crevice, Treated Water Systems"

36) elements and galvanic corrosion VAEP-76 V.A- Piping, piping Copper alloy Lubricating oil Loss of material Chapter XI.M39, No 21 (EP- components, and piping due to pitting and crevice "Lubricating Oil Analysis,"

45) elements corrosion and Chapter XI.M32, "One-Time Inspection" VAEP-27 V.A- Piping, piping Copper alloy Closed-cycle Loss of material Chapter XI.M33, "Selective No 0 22(EP- components, and piping (>15% Zn or cooling water due to selective leaching Leaching" CD C') 27) elements >8% AI)

CD 0

G) 3 0-VAEP-95 V.A- Piping, piping Stainless Closed-cycle Loss of material Chapter XI.M21A, "Closed No CD 23(EP- components, and piping steel cooling water due to pitting and crevice Treated Water Systems" 0 I\)

0 a 33) elements corrosion 0

--" a 0)

CD 0

I 0

0 N

CD 0

oCD V ENGINEERED SAFETY FEATURES C')

CD A Containment Spray System (PWR) 3 0-CD I\) Structure and/or Aging Effect! Aging Management Further a Item Link Material Environment a Component Mechanism Program (AMP) Evaluation VAEP-98 V.A- Piping, piping Stainless Closed-cycle Cracking Chapter XI.M21A, "Closed No 24(EP- components, and piping steel cooling water due to stress corrosion Treated Water Systems"

44) elements >60°C cracking

(>140°F)

VAEP-77 V.A- Piping, piping Steel Lubricating oil Loss of material Chapter XI.M39, No 25(EP- components, and piping due to general, pitting, "Lubricating Oil Analysis,"

46) elements and crevice corrosion and Chapter XI.M32, "One-Time Inspection" VAEP-81 V.A- Piping, piping Stainless Condensation Loss of material Chapter XI.M38, No 26(EP- components, and piping steel (Internal) due to pitting and crevice "Inspection of Internal

53) elements (Internal corrosion Surfaces in Miscellaneous

< surfaces); tanks Piping and Ducting

>> I Components" 0"1 VAEP-41 V.A- Piping, piping Stainless Treated water Loss of material Chapter XI.M2, "Water No 27(EP- components, and piping steel (borated) due to pitting and crevice Chemistry"

41) elements; tanks corrosion VAE-12 V.A- Piping, piping Stainless Treated water Cracking Chapter XI.M2, "Water No 28(E-12) components, and piping steel (borated) due to stress corrosion Chemistry" elements; tanks >60°C cracking

(>140°F) z c

u m

G)

I 0 a 00 G)

U 0 CD 0

0  :<

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CD 0

I 0

0 N

CD

B. STANDBY GAS TREATMENT SYSTEM (BOILING WATER REACTORS)

Systems, Structures, and Components This section addresses the standby gas treatment system found in boiling water reactors (BWRs) and consists of ductwork, filters, and fans. Based on Regulatory Guide 1.26, "Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants," all components that comprise the standby gas treatment system are governed by Group B Quality Standards.

Specifically, charcoal absorber filters are to be addressed consistent with the NRC position on consumables, provided in the NRC letter from Christopher I. Grimes to Douglas J. Walters of NEI, dated March 10, 2000. Components that function as system filters are typically replaced based on performance or condition monitoring that identifies whether these components are at the end of their qualified lives and may be excluded, from an aging management review (on a plant-specific basis), under 10 CFR 54.21(a)(1)(ii). As part of the methodology description, the application should identify the standards that are relied on for replacement, for example, National Fire Protection Association (NFPA) standards for fire protection equipment.

Aging management programs for the degradation of external surfaces of components and miscellaneous bolting are included in V.E. Common miscellaneous material/environment combinations, where aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation, are included in V.F.

System Interfaces There are no system interfaces with the standby gas treatment system addressed in this section.

December 201 0 V 8-1 NUREG-1801, Rev. 2 OAG10001390_00292

z V ENGINEERED SAFETY FEATURES c

U B Standby Gas Treatment System (BWR) m G)

I 00 Structure a Aging Effect! Aging Management Program Further Item Link and/or Material Environment Mechanism (AMP) Evaluation

u Component CD

< V.B.E-25 V.B-1 (E- Ducting and Steel Air - indoor, Loss of material Chapter XI.M38, "Inspection of No I\)

25) components uncontrolled due to general Internal Surfaces in Miscellaneous (Internal (Internal) corrosion Piping and Ducting Components" surfaces)

V.B.E-40 V.B-2(E- Ducting, closure Steel Air - indoor, Loss of material Chapter XI.M36, "External Surfaces No

40) bolting uncontrolled due to general Monitoring of Mechanical (External) corrosion Components" V.B.E-26 V.B-3(E- Ducting, piping, Steel Air - indoor, Loss of material Chapter XI.M36, "External Surfaces No

26) and components uncontrolled due to general Monitoring of Mechanical (External (External) corrosion Components" surfaces)

V.B.EP-59 V.B-4(E- Elastomer seals Elastomers Air - indoor, Hardening and Chapter XI.M36, "External Surfaces No

< 06) and components uncontrolled loss of strength Monitoring of Mechanical III I

I\) (External) due to elastomer Components" degradation V.B.EP-58 V.B-4(E- Elastomer seals Elastomers Air - indoor, Hardening and Chapter XI.M38, "Inspection of No

06) and components uncontrolled loss of strength Internal Surfaces in Miscellaneous (Internal) due to elastomer Piping and Ducting Components" degradation V.B.EP-37 V.B-5(EP- Heat exchanger Copper alloy Closed-cycle Loss of material Chapter XI.M33, "Selective No

37) components (>15% Zn or cooling water due to selective Leaching"

>8% AI) leaching V.B.EP-97 V.B-6(EP- Piping, piping Copper alloy Closed-cycle Loss of material Chapter XI.M21A, "Closed Treated No

36) components, cooling water due to pitting, Water Systems" and piping crevice, and elements galvanic corrosion 0 V.B.EP-27 V.B-7(EP- Piping, piping Copper alloy Closed-cycle Loss of material Chapter XI.M33, "Selective No CD C') 27) components, (>15% Zn or cooling water due to selective Leaching" CD 0 and piping >8% AI) leaching G) 3 0-elements CD 0 I\)

0 a 0

--" a 0)

CD 0

I 0

0 N

CD 0)

oCD V ENGINEERED SAFETY FEATURES C')

CD B Standby Gas Treatment System (BWR) 3 0-CD Structure I\) Aging Effect! Aging Management Program Further a Item Link and/or Material Environment a Mechanism (AMP) Evaluation Component V.B.EP-54 V.B-8(EP- Piping, piping Gray cast Soil Loss of material Chapter XI.M33, "Selective No

54) components, iron due to selective Leaching" and piping leaching elements V.B.EP-111 V.B-9(E- Piping, piping Steel (with Soil or concrete Loss of material Chapter XI.M41, "Buried and No

42) components, coating or due to general, Underground Piping and Tanks" and piping wrapping) pitting, crevice, elements and microbiologically-influenced corrosion V.B.EP-103 Piping, piping Stainless Air - outdoor Cracking Chapter XI.M36, "External Surfaces Yes,

< components, steel due to stress Monitoring of Mechanical environmental III I and piping corrosion cracking Components" conditions CJ.)

elements; tanks need to be evaluated V.B.EP-107 Piping, piping Stainless Air - outdoor Loss of material Chapter XI.M36, "External Surfaces Yes, components, steel due to pitting and Monitoring of Mechanical environmental and piping crevice corrosion Components" conditions elements; tanks need to be evaluated z

c

u m

G)

I 0 a 00 G)

U 0 CD 0

0  :<

--" I\)

0)

CD 0

I 0

0 N

CD

.j::>.

C. CONTAINMENT ISOLATION COMPONENTS Systems, Structures, and Components This section addresses the containment isolation components found in all designs of boiling water reactors (BWR) and pressurized water reactors (PWR) in the United States. The system consists of isolation barriers in lines for BWR and PWR nonsafety systems, such as the plant heating, waste gas, plant drain, liquid waste, and cooling water systems. Based on Regulatory Guide 1.26, "Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants," all components that comprise the containment isolation components are governed by Group A or B Quality Standards.

The aging management programs for hatchways, hatch doors, penetration sleeves, penetration bellows, seals, gaskets, and anchors are addressed in II.A and II. B. The containment isolation valves for in-scope systems are addressed in the appropriate sections in IV, VII, and VIII.

Aging management programs for the degradation of external surfaces of components and miscellaneous bolting are included in V.E. Common miscellaneous material/environment combinations, where aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation, are included in V.F.

System Interfaces There are no system interfaces with the containment isolation components addressed in this section.

December 201 0 V C-1 NUREG-1801, Rev. 2 OAGI0001390_00295

z V ENGINEERED SAFETY FEATURES c

U C Containment Isolation Components m

G)

I 00 Structure a Aging Effect/ Aging Management Program Further Item Link and/or Material Environment Mechanism (AMP) Evaluation

u Component CD

< V.C.E-35 V.C-1 (E- Containment Steel Air - indoor, Loss of material Chapter XI.M36, "External Surfaces No I\)

35) isolation piping uncontrolled due to general Monitoring of Mechanical and components (External) corrosion Components" (External surfaces)

V.C.E-30 V.C-2(E- Containment Steel Condensation Loss of material Chapter XI.M36, "External Surfaces No

30) isolation piping (External) due to general Monitoring of Mechanical and components corrosion Components" (External surfaces)

V.C.E-34 V.C-3(E- Containment Stainless Raw water Loss of material Chapter XI.M20, "Open-Cycle No

34) isolation piping steel due to pitting, Cooling Water System"

< and components crevice, and

() (Internal microbiolog ically-I I\)

surfaces) influenced corrosion; fouling that leads to corrosion V.C.EP-63 V.C-4(E- Containment Stainless Treated water Loss of material Chapter XI.M2, "Water Chemistry," No

33) isolation piping steel due to pitting and and and components crevice corrosion Chapter XI.M32, "One-Time (Internal Inspection" surfaces)

V.C.E-22 V.C-5(E- Containment Steel Raw water Loss of material Chapter XI.M20, "Open-Cycle No

22) isolation piping due to general, Cooling Water System" and components pitting, crevice, (Internal and 0 surfaces) microbiologically-CD C') influenced CD 0 corrosion; fouling G) 3 0-that leads to CD 0 I\) corrosion 0 a 0

--" a 0)

CD 0

I 0

0 N

CD (J)

oCD V ENGINEERED SAFETY FEATURES C')

CD C Containment Isolation Components 3

0-CD Structure I\) Aging Effect/ Aging Management Program Further a Item Link and/or Material Environment a Mechanism (AMP) Evaluation Component V.C.EP-62 V.C-6(E- Containment Steel Treated water Loss of material Chapter XI.M2, "Water Chemistry," No

31) isolation piping due to general, and and components pitting, and crevice Chapter XI.M32, "One-Time (Internal corrosion Inspection" surfaces)

V.C.EP-95 V.C- Piping, piping Stainless Closed-cycle Loss of material Chapter XI.M21A, "Closed Treated No 7(EP-33) components, steel cooling water due to pitting and Water Systems" and piping crevice corrosion elements V.C.EP-98 V.C- Piping, piping Stainless Closed-cycle Cracking Chapter XI.M21A, "Closed Treated No 8(EP-44) components, steel cooling water due to stress Water Systems" and piping >60°C (>140°F) corrosion cracking

< elements

()

I CJ.)

V.C.EP-99 V.C- Piping, piping Steel Closed-cycle Loss of material Chapter XI.M21A, "Closed Treated No 9(EP-48) components, cooling water due to general, Water Systems" and piping pitting, and crevice elements corrosion V.C.EP-103 Piping, piping Stainless Air - outdoor Cracking Chapter XI.M36, "External Surfaces Yes, components, steel due to stress Monitoring of Mechanical environmental and piping corrosion cracking Components" conditions elements; tanks need to be evaluated V.C.EP-107 Piping, piping Stainless Air - outdoor Loss of material Chapter XI.M36, "External Surfaces Yes, components, steel due to pitting and Monitoring of Mechanical environmental and piping crevice corrosion Components" conditions z elements; tanks need to be c

u evaluated m

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CD

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D1. EMERGENCY CORE COOLING SYSTEM (PRESSURIZED WATER REACTORS)

Systems, Structures, and Components This section addresses the emergency core cooling systems for pressurized water reactors (PWRs) designed to cool the reactor core and provide safe shutdown following a design basis accident. The core cooling systems consist of the core flood system (CFS), residual heat removal (RHR) (or shutdown cooling (SOC)), high-pressure safety injection (HPSI), low-pressure safety injection (LPSI), and spent fuel pool (SFP) cooling systems, the lines to the chemical and volume control system (CVCS), the emergency sump, the HPSI and LPSI pumps, the pump seal coolers, the RHR heat exchanger, and the refueling water tank (RWT).

Based on Regulatory Guide 1.26, "Quality Group Classifications and Standards for Water-,

Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants," all components that comprise the emergency core cooling system are governed by Group B Quality Standards. Portions of the RHR, HPSI, and LPSI systems and the CVCS extending from the reactor coolant system up to and including the second containment isolation valve are governed by Group A Quality Standards and covered in IV.C2.

Pumps and valve internals perform their intended functions with moving parts or with a change in configuration. They are also subject to replacement based on qualified life or a specified time period. Pursuant to 10 CFR 54.21(a)(1), therefore, they are not subject to an aging management review.

Aging management programs for the degradation of external surfaces of components and miscellaneous bolting are included in V.E. Common miscellaneous material/environment combinations where aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation are included in VI.F.

The system piping includes all pipe sizes, including instrument piping.

System Interfaces The systems that interface with the emergency core cooling system include the reactor coolant system and connected lines (lV.C2), the containment spray system (V.A), the spent fuel pool cooling and cleanup system (VII.A3), the closed-cycle cooling water system (VII.C2), the ultimate heat sink (VII.C3), the chemical and volume control system (VII.E1), and the open-cycle cooling water system (service water system) (VII.C1).

December 201 0 V 01-1 NUREG-1801, Rev. 2 OAG10001390_00298

z V ENGINEERED SAFETY FEATURES c

U D1 Emergency Core Cooling System (PWR) m G)

I 00 a Structure and/or Aging Effect! Aging Management Further Item Link Material Environment Component Mechanism Program (AMP) Evaluation

u CD

< V.D1.E-28 V.D1- External surfaces Steel Air with borated Loss of material Chapter XI.M1 0, "Boric No I\)

1(E-28) water leakage due to boric acid Acid Corrosion" corrosion V.D1.EP- V.D1- Heat exchanger Copper alloy Closed-cycle Loss of material Chapter XI.M21A, No 94 2(EP- components cooling water due to pitting, crevice, "Closed Treated Water

13) and galvanic corrosion Systems" V.D1.EP- V.D1- Heat exchanger Copper alloy Closed-cycle Loss of material Chapter XI.M33, No 37 3(EP- components (>15% Zn or >8% cooling water due to selective leaching "Selective Leaching"

37) AI)

V.D1.EP- V.D1- Heat exchanger Stainless steel Closed-cycle Loss of material Chapter XI.M21A, No 93 4(E-19) components cooling water due to pitting and crevice "Closed Treated Water

< corrosion Systems" o

I I\)

V.D1.EP- V.D1- Heat exchanger Stainless steel Raw water Loss of material Chapter XI.M20, "Open- No 91 5(E-20) components due to pitting, crevice, Cycle Cooling Water and microbiologically- System" influenced corrosion; fouling that leads to corrosion V.D1.EP- V.D1- Heat exchanger Steel Closed-cycle Loss of material Chapter XI.M21A, No 92 6(E-17) components cooling water due to general, pitting, "Closed Treated Water crevice, and galvanic Systems" corrosion V.D1.EP- V.D1- Heat exchanger Steel Raw water Loss of material Chapter XI.M20, "Open- No 90 7(E-18) components due to general, pitting, Cycle Cooling Water crevice, and System" microbiolog ically-0 CD C')

influenced corrosion; 0 CD fouling that leads to G) 3 0- corrosion CD 0 I\)

0 a 0

--" a 0)

CD 0

I 0

0 N

CD CD

oCD V ENGINEERED SAFETY FEATURES C')

CD D1 Emergency Core Cooling System (PWR) 3 0-CD I\) Structure and/or Aging Effect! Aging Management Further a Item Link Material Environment a Component Mechanism Program (AMP) Evaluation V.D1.EP- V.D1- Heat exchanger Copper alloy Lubricating oil Reduction of heat Chapter XI.M39, No 78 8(EP- tubes transfer "Lubricating Oil

47) due to fouling Analysis," and Chapter XI.M32, "One-Time Inspection" V.D1.EP- V.D1- Heat exchanger Stainless steel Closed-cycle Reduction of heat Chapter XI.M21A, No 96 19(EP- tubes cooling water transfer "Closed Treated Water

35) due to fouling Systems" V.D1.EP- V.D1- Heat exchanger Stainless steel Lubricating oil Reduction of heat Chapter XI.M39, No 79 10(EP- tubes transfer "Lubricating Oil

50) due to fouling Analysis," and Chapter XI.M32, "One-

< Time Inspection" o

I CJ.)

V.D1.E-21 V.D1- Heat exchanger Stainless steel Raw water Reduction of heat Chapter XI.M20, "Open- No 11 (E- tubes transfer Cycle Cooling Water

21) due to fouling System" V.D1.EP- V.D1- Heat exchanger Steel Lubricating oil Reduction of heat Chapter XI.M39, No 75 12(EP- tubes transfer "Lubricating Oil

40) due to fouling Analysis," and Chapter XI.M32, "One-Time Inspection" V.D1.E-43 V.D1- Motor cooler Gray cast iron Treated water Loss of material Chapter XI.M33, No 13(E- due to selective leaching "Selective Leaching" 43) z c

u m

G)

I 0 a 00 G)

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CD 0

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0 0) 0 0

z V ENGINEERED SAFETY FEATURES c

U D1 Emergency Core Cooling System (PWR) m G)

I 00 a Structure and/or Aging Effect! Aging Management Further Item Link Material Environment Component Mechanism Program (AMP) Evaluation

u CD

< V.D1.E-24 V.D1- Orifice (miniflow Stainless steel Treated water Loss of material A plant-specific aging Yes, plant-I\)

14(E- recirculation) (borated) due to erosion management program is specific

24) to be evaluated for erosion of the orifice due to extended use of the centrifugal HPSI pump for normal charging. See LER 50-275/94-023 for evidence of erosion.

V.D1.E-01 V.D1- Partially-encased Stainless steel Raw water Loss of material A plant-specific aging Yes, plant-15(E- tanks with breached due to pitting and crevice management program is specific

01) moisture barrier corrosion to be evaluated for pitting

< and crevice corrosion of o tank bottom because I

.j:>. moisture and water can egress under the tank due to cracking of the perimeter seal from weathering.

V.D1.EP- V.D2- Piping, piping Aluminum Air with borated Loss of material Chapter XI.M1 0, "Boric No 101 18(EP- components, and water leakage due to boric acid Acid Corrosion"

2) piping elements corrosion V.D1.E-47 V.D1- Piping, piping Cast austenitic Treated water Loss of fracture Chapter XI.M12, No 16(E- components, and stainless steel (borated) toughness "Thermal Aging

47) piping elements >250°C due to thermal aging Embrittlement of Cast

(>482°F) embrittlement Austenitic Stainless Steel (CASS)"

0 V.D1.EP- V.D1- Piping, piping Copper alloy Closed-cycle Loss of material Chapter XI.M21A, No CD C')

CD 97 17(EP- components, and cooling water due to pitting, crevice, "Closed Treated Water 0

G) 3 0-

36) piping elements and galvanic corrosion Systems" CD 0 I\)

0 a 0

--" a 0)

CD 0

I 0

0 0) 0

oCD V ENGINEERED SAFETY FEATURES C')

CD D1 Emergency Core Cooling System (PWR) 3 0-CD I\) Structure and/or Aging Effect! Aging Management Further a Item Link Material Environment a Component Mechanism Program (AMP) Evaluation V.D1.EP- V.D1- Piping, piping Copper alloy Lubricating oil Loss of material Chapter XI.M39, No 76 19(EP- components, and due to pitting and crevice "Lubricating Oil

45) piping elements corrosion Analysis," and Chapter XI.M32, "One-Time Inspection" V.D1.EP- V.D1- Piping, piping Copper alloy Closed-cycle Loss of material Chapter XI.M33, No 27 19(EP- components, and (>15% Zn or >8% cooling water due to selective leaching "Selective Leaching"

27) piping elements AI)

V.D1.EP- V.D1- Piping, piping Gray cast iron Closed-cycle Loss of material Chapter XI.M33, No 52 20(EP- components, and cooling water due to selective leaching "Selective Leaching"

52) piping elements

< V.D1.EP- V.D1- Piping, piping Gray cast iron Soil Loss of material Chapter XI.M33, No o 54 21 (EP- components, and due to selective leaching "Selective Leaching" I

0"1 54) piping elements V.D1.EP- V.D1- Piping, piping Stainless steel Closed-cycle Loss of material Chapter XI.M21A, No 95 22(EP- components, and cooling water due to pitting and crevice "Closed Treated Water

33) piping elements corrosion Systems" V.D1.EP- V.D1- Piping, piping Stainless steel Closed-cycle Cracking Chapter XI.M21A, No 98 23(EP- components, and cooling water due to stress corrosion "Closed Treated Water

44) piping elements >60°C (>140°F) cracking Systems" V.D1.EP- V.D1- Piping, piping Stainless steel Lubricating oil Loss of material Chapter XI.M39, No 80 24(EP- components, and due to pitting and crevice "Lubricating Oil

51) piping elements corrosion Analysis," and z Chapter XI.M32, "One-c Time Inspection"

u m V.D1.EP- V.D1- Piping, piping Stainless steel Raw water Loss of material Chapter XI.M20, "Open- No G)

I 55 25(EP- components, and due to pitting, crevice, Cycle Cooling Water 0 a 00

55) piping elements and microbiologically- System" G) influenced corrosion

U 0 CD 0

0  :<

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0)

CD 0

I 0

0 0) 0 N

z V ENGINEERED SAFETY FEATURES c

U D1 Emergency Core Cooling System (PWR) m G)

I 00 a Structure and/or Aging Effect! Aging Management Further Item Link Material Environment Component Mechanism Program (AMP) Evaluation

u CD

< V.D1.EP- V.D1- Piping, piping Stainless steel Soil or concrete Loss of material Chapter XI.M41, "Buried No I\)

72 26(EP- components, and due to pitting and crevice and Underground Piping

31) piping elements corrosion and Tanks" V.D1.E-13 V.D1- Piping, piping Stainless steel Treated water Cumulative fatigue Fatigue is a time-limited Yes, TLAA 27(E- components, and (borated) damage aging analysis (TLAA) to

13) piping elements due to fatigue be evaluated for the period of extended operation. See the SRP, Section 4.3 "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR

< 54.21 (c)(1).

o I V.D1.EP- V.D1- Piping, piping Steel Lubricating oil Loss of material Chapter XI.M39, No (J) 77 28(EP- components, and due to general, pitting, "Lubricating Oil

46) piping elements and crevice corrosion Analysis," and Chapter XI.M32, "One-Time Inspection" V.D1.EP- V.D1- Piping, piping Stainless steel Condensation Loss of material Chapter XI.M38, No 81 29(EP- components, and (Internal) due to pitting and crevice "Inspection of Internal

53) piping elements corrosion Surfaces in (Internal surfaces); Miscellaneous Piping tanks and Ducting Components" V.D1.EP- Piping, piping Stainless steel Air - outdoor Cracking Chapter XI.M36, Yes, 103 components, and due to stress corrosion "External Surfaces environmental piping elements; cracking Monitoring of Mechanical conditions 0 tanks Components" need to be CD C')

CD evaluated 0

G) 3 0-CD 0 I\)

0 a 0

--" a 0)

CD 0

I 0

0 0) 0 0)

oCD V ENGINEERED SAFETY FEATURES C')

CD D1 Emergency Core Cooling System (PWR) 3 0-CD I\) Structure and/or Aging Effect! Aging Management Further a Item Link Material Environment a Component Mechanism Program (AMP) Evaluation V.D1.EP- Piping, piping Stainless steel Air - outdoor Loss of material Chapter XI.M36, Yes, 107 components, and due to pitting and crevice "External Surfaces environmental piping elements; corrosion Monitoring of Mechanical conditions tanks Components" need to be evaluated V.D1.EP- V.D1- Piping, piping Stainless steel Treated water Loss of material Chapter XI.M2, "Water No 41 30(EP- components, and (borated) due to pitting and crevice Chemistry"

41) piping elements; corrosion tanks V.D1.E-12 V.D1- Piping, piping Stainless steel Treated water Cracking Chapter XI.M2, "Water No 31 (E- components, and (borated) >60°C due to stress corrosion Chemistry"

12) piping elements; (>140°F) cracking

< tanks o V.D1.EP- V.D1- Pump casings Steel (with Treated water Loss of material A plant-specific aging Yes, verify I

-J 49 32(EP- stainless steel (borated) due to cladding breach management program is that plant-

49) cladding) to be evaluated specific Reference NRC program Information Notice 94-63, addresses "Boric Acid Corrosion of clad breach Charging Pump Casings Caused by Cladding Cracks."

V.D1.E-38 V.D1- Safety injection tank Steel (with Treated water Cracking Chapter XI.M2, "Water No 33(E- (accumulator) stainless steel or (borated) >60°C due to stress corrosion Chemistry"

38) nickel-alloy (>140°F) cracking z cladding) c

u m

G)

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0  :<

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CD 0

I 0

0 0) 0

.j::>.

02. EMERGENCY CORE COOLING SYSTEM (BOILING WATER REACTORS)

Systems, Structures, and Components This section addresses the emergency core cooling systems for boiling water reactors (BWRs) designed to cool the reactor core and provide safe shutdown following a design basis accident.

The cooling systems consist of the high-pressure coolant injection (HPCI), reactor core isolation cooling (RCIC), high-pressure core spray (HPCS), automatic depressurization (ADS), low-pressure core spray (LPCS), low-pressure coolant injection (LPCI), and residual heat removal (RHR) systems, including various pumps and valves, the RHR heat exchangers, and the drywell and suppression chamber spray system (DSCSS). The auxiliary area ventilation system includes RCIC, HPCI, RHR, and core spray pump room cooling.

Based on Regulatory Guide 1.26, "Quality Group Classifications and Standards for Water-,

Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants," all components that comprise the emergency core cooling system outside the containment are governed by Group B Quality Standards and the portion of the DSCSS inside the containment up to the isolation valve is governed by Group A Quality Standards. Portions of the HPCI, RCIC, HPCS, LPCS, and LPCI (or RHR) systems extending from the reactor vessel up to and including the second containment isolation valve are governed by Group A Quality Standards and covered in IV.C1.

Pumps and valve internals perform their intended functions with moving parts or with a change in configuration. They are also subject to replacement based on qualified life or a specified time period. Pursuant to 10 CFR 54.21(a)(1), therefore, they are not subject to an aging management review.

The system piping includes all pipe sizes, including instrument piping.

Aging management programs for the degradation of external surfaces of components and miscellaneous bolting are included in V.E. Common miscellaneous material/environment combinations where aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation are included in VI.F.

System Interfaces The systems that interface with the emergency core cooling system include the reactor vessel (lV.A 1), the reactor coolant pressure boundary (lV.C1), the feedwater system (VII I. D2), the condensate system (VIII.E), the closed-cycle cooling water system (VII.C2), the open-cycle cooling water system (VII.C1), and the ultimate heat sink (VII.C3).

December 201 0 V 02-1 NUREG-1801, Rev. 2 OAG10001390_00305

z V ENGINEERED SAFETY FEATURES c

U D2 Emergency Core Cooling System (8WR) m G)

I 00 Structure a Aging Effect! Aging Management Further Item Link and/or Material Environment Mechanism Program (AMP) Evaluation

u Component CD

< V.D2.EP-113 V.D2- Drywell and Steel Air - indoor, Loss of material A plant-specific aging Yes, plant-I\)

1(E-04) suppression uncontrolled due to general management program is to be specific chamber spray (Internal) corrosion; fouling that evaluated system (internal leads to corrosion surfaces): flow orifice; spray nozzles V.D2.E-26 V.D2- Ducting, piping, Steel Air - indoor, Loss of material Chapter XI.M36, "External No 2(E-26) and components uncontrolled due to general Surfaces Monitoring of (External (External) corrosion Mechanical Components" surfaces)

V.D2.EP-94 V.D2- Heat exchanger Copper alloy Closed-cycle Loss of material Chapter XI.M21A, "Closed No

< 3(EP-13) components cooling water due to pitting, Treated Water Systems" o

I\) crevice, and galvanic I

I\) corrosion V.D2.EP-37 V.D2- Heat exchanger Copper alloy Closed-cycle Loss of material Chapter XI.M33, "Selective No 4(EP-37) components (>15% Zn or cooling water due to selective Leaching"

>8% AI) leaching V.D2.EP-93 V.D2- Heat exchanger Stainless steel Closed-cycle Loss of material Chapter XI.M21A, "Closed No 5(E-19) components cooling water due to pitting and Treated Water Systems" crevice corrosion V.D2.EP-91 V.D2- Heat exchanger Stainless steel Raw water Loss of material Chapter XI.M20, "Open-Cycle No 6(E-20) components due to pitting, Cooling Water System" crevice, and microbiolog ically-influenced corrosion; 0 fouling that leads to CD corrosion C')

CD 0 V.D2.EP-92 V.D2- Heat exchanger Steel Closed-cycle Loss of material Chapter XI.M21A, "Closed No G) 3 0- 7(E-17) components cooling water due to general, Treated Water Systems" CD 0 I\) pitting, crevice, and 0 a galvanic corrosion 0

--" a 0)

CD 0

I 0

0 0) 0 (J)

oCD V ENGINEERED SAFETY FEATURES C')

CD D2 Emergency Core Cooling System (8WR) 3 0-CD Structure I\) Aging Effect! Aging Management Further a Item Link and/or Material Environment a Mechanism Program (AMP) Evaluation Component V.D2.EP-90 V.D2- Heat exchanger Steel Raw water Loss of material Chapter XI.M20, "Open-Cycle No 8(E-18) components due to general, Cooling Water System" pitting, crevice, and microbiolog ically-influenced corrosion; fouling that leads to corrosion V.D2.EP-78 V.D2- Heat exchanger Copper alloy Lubricating oil Reduction of heat Chapter XI.M39, "Lubricating No 9(EP-47) tubes transfer Oil Analysis," and due to fouling Chapter XI.M32, "One-Time Inspection" V.D2.EP-96 V.D2- Heat exchanger Stainless steel Closed-cycle Reduction of heat Chapter XI.M21A, "Closed No

< 10(EP- tubes cooling water transfer Treated Water Systems" o

I\) 35) due to fouling I

CJ.)

V.D2.EP-79 V.D2- Heat exchanger Stainless steel Lubricating oil Reduction of heat Chapter XI.M39, "Lubricating No 11 (EP- tubes transfer Oil Analysis," and

50) due to fouling Chapter XI.M32, "One-Time Inspection" V.D2.E-21 V.D2- Heat exchanger Stainless steel Raw water Reduction of heat Chapter XI.M20, "Open-Cycle No 12(E-21) tubes transfer Cooling Water System" due to fouling V.D2.EP-74 V.D2- Heat exchanger Stainless steel Treated water Reduction of heat Chapter XI.M2, "Water No 13(EP- tubes transfer Chemistry," and

34) due to fouling Chapter XI.M32, "One-Time z Inspection" c V.D2.EP-75 V.D2- Heat exchanger Steel Lubricating oil Reduction of heat Chapter XI.M39, "Lubricating No

u m 14(EP- tubes transfer Oil Analysis," and G)

I

40) due to fouling Chapter XI.M32, "One-Time 0 a 00 Inspection" G) V.D2.E-23 V.D2- Heat exchanger Steel Raw water Reduction of heat Chapter XI.M20, "Open-Cycle No

U 0 CD 15(E-23) tubes transfer Cooling Water System" 0

0  :< due to fouling

--" I\)

0)

CD 0

I 0

0 0) 0

-....J

z V ENGINEERED SAFETY FEATURES c

U D2 Emergency Core Cooling System (8WR) m G)

I 00 Structure a Aging Effect! Aging Management Further Item Link and/or Material Environment Mechanism Program (AMP) Evaluation

u Component CD

< V.D2.E-29 V.D2- Piping and Steel Air - indoor, Loss of material Chapter XI.M38, "Inspection of No I\)

16(E-29) components uncontrolled due to general Internal Surfaces in (Internal (Internal) corrosion Miscellaneous Piping and surfaces) Ducting Components" V.D2.E-27 V.D2- Piping and Steel Condensation Loss of material Chapter XI.M38, "Inspection of No 17(E-27) components (Internal) due to general, Internal Surfaces in (Internal pitting, and crevice Miscellaneous Piping and surfaces) corrosion Ducting Components" V.D2.EP-71 V.D2- Piping, piping Aluminum Treated water Loss of material Chapter XI.M2, "Water No 19(EP- components, due to pitting and Chemistry," and

26) and piping crevice corrosion Chapter XI.M32, "One-Time elements Inspection"

< V.D2.E-11 V.D2- Piping, piping Cast austenitic Treated water Loss of fracture Chapter XI.M12, "Thermal No o

I\) 20(E-11) components, stainless steel >250°C (>482°F) toughness Aging Embrittlement of Cast I

.j:>.

and piping due to thermal aging Austenitic Stainless Steel elements embrittiement (CASS)"

V.D2.EP-97 V.D2- Piping, piping Copper alloy Closed-cycle Loss of material Chapter XI.M21A, "Closed No 21 (EP- components, cooling water due to pitting, Treated Water Systems"

36) and piping crevice, and galvanic elements corrosion V.D2.EP-76 V.D2- Piping, piping Copper alloy Lubricating oil Loss of material Chapter XI.M39, "Lubricating No 22(EP- components, due to pitting and Oil Analysis," and

45) and piping crevice corrosion Chapter XI.M32, "One-Time elements Inspection" V.D2.EP-27 V.D2- Piping, piping Copper alloy Closed-cycle Loss of material Chapter XI.M33, "Selective No 23(EP- components, (>15% Zn or cooling water due to selective Leaching"

27) and piping >8% AI) leaching 0 elements CD C')

CD V.D2.EP-54 V.D2- Piping, piping Gray cast iron Soil Loss of material Chapter XI.M33, "Selective No 0

G) 3 0-24(EP- components, due to selective Leaching" CD 54) and piping leaching 0 I\)

0 a elements 0

--" a 0)

CD 0

I 0

0 0) 0 CD

oCD V ENGINEERED SAFETY FEATURES C')

CD D2 Emergency Core Cooling System (BWR) 3 0-CD Structure I\) Aging Effect! Aging Management Further a Item Link and/or Material Environment a Mechanism Program (AMP) Evaluation Component V.D2.EP-95 V.D2- Piping, piping Stainless steel Closed-cycle Loss of material Chapter XI.M21A, "Closed No 25(EP- components, cooling water due to pitting and Treated Water Systems"

33) and piping crevice corrosion elements V.D2.EP-98 V.D2- Piping, piping Stainless steel Closed-cycle Cracking Chapter XI.M21A, "Closed No 26(EP- components, cooling water due to stress Treated Water Systems"

44) and piping >60°C (>140°F) corrosion cracking elements V.D2.EP-72 V.D2- Piping, piping Stainless steel Soil or concrete Loss of material Chapter XI.M41, "Buried and No 27(EP- components, due to pitting and Underground Piping and

31) and piping crevice corrosion Tanks" elements

< V.D2.EP-73 V.D2- Piping, piping Stainless steel Treated water Loss of material Chapter XI.M2, "Water No o

I\) 28(EP- components, due to pitting and Chemistry," and I

0"1

32) and piping crevice corrosion Chapter XI.M32, "One-Time elements Inspection" V.D2.E-37 V.D2- Piping, piping Stainless steel Treated water Cracking ChapterXI.M7, "BWR Stress No 29(E-37) components, >60°C (>140°F) due to stress Corrosion Cracking," and and piping corrosion cracking, Chapter XI.M2, "Water elements intergranular stress Chemistry" corrosion cracking V.D2.EP-77 V.D2- Piping, piping Steel Lubricating oil Loss of material Chapter XI.M39, "Lubricating No 30(EP- components, due to general, Oil Analysis," and

46) and piping pitting, and crevice Chapter XI.M32, "One-Time elements corrosion Inspection" z V.D2.E-07 V.D2- Piping, piping Steel Steam Wall thinning Chapter XI.M17, "Flow- No c

u 31 (E-07) components, due to flow- Accelerated Corrosion" m

G) and piping accelerated corrosion I

elements 0 a 00 G)

U 0 CD 0

0  :<

--" I\)

0)

CD 0

I 0

0 0) 0 CD

z V ENGINEERED SAFETY FEATURES c

U D2 Emergency Core Cooling System (8WR) m G)

I 00 Structure a Aging Effect! Aging Management Further Item Link and/or Material Environment Mechanism Program (AMP) Evaluation

u Component CD

< V.D2.E-10 V.D2- Piping, piping Steel Treated water Cumulative fatigue Fatigue is a time-limited aging Yes, TLAA I\)

32(E-10) components, damage analysis (TLAA) to be and piping due to fatigue evaluated for the period of elements extended operation. See the SRP, Section 4.3 "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR 54.21 (c)(1).

V.D2.EP-60 V.D2- Piping, piping Steel Treated water Loss of material Chapter XI.M2, "Water No 33(E-08) components, due to general, Chemistry," and and piping pitting, and crevice Chapter XI.M32, "One-Time

< elements corrosion Inspection" o

I\) V.D2.E-09 V.D2- Piping, piping Steel Treated water Wall thinning Chapter XI.M17, "Flow- No I

(J) 34(E-09) components, due to flow- Accelerated Corrosion" and piping accelerated corrosion elements V.D2.EP-61 V.D2- Piping, piping Stainless steel Condensation Loss of material Chapter XI.M38, "Inspection of No 35(E-14) components, (Internal) due to pitting and Internal Surfaces in and piping crevice corrosion Miscellaneous Piping and elements Ducting Components" (Internal surfaces)

V.D2.EP-103 Piping, piping Stainless steel Air - outdoor Cracking Chapter XI.M36, "External Yes, components, due to stress Surfaces Monitoring of environmental and piping corrosion cracking Mechanical Components" conditions elements; tanks need to be 0 evaluated CD C')

CD V.D2.EP-107 Piping, piping Stainless steel Air - outdoor Loss of material Chapter XI.M36, "External Yes, 0 components, due to pitting and Surfaces Monitoring of environmental

>> 3 0-G) ...,

CD and piping crevice corrosion Mechanical Components" conditions 0 I\) elements; tanks need to be 0 a 0 evaluated

--" a 0)

CD 0

I 0

0 0) 0

E. EXTERNAL SURFACES OF COMPONENTS AND MISCELLANEOUS BOLTING Systems, Structures, and Components This section addresses the aging management programs for the degradation of external surfaces of all steel structures and components, including closure boltings in the engineered safety features in pressurized water reactors (PWRs) and boiling water reactors (BWRs). For the steel components in PWRs, this section addresses only boric acid corrosion of external surfaces as a result of dripping borated water leaking from an adjacent PWR component. Boric acid corrosion can also occur for steel components containing borated water due to leakage, such components and the related aging management program are covered in the appropriate major plant sections in V.

System Interfaces The structures and components covered in this section belong to the engineered safety features in PWRs and BWRs. (For example, see System Interfaces in V.A to V.D2 for details.)

December 201 0 V E-1 NUREG-1801, Rev. 2 OAGI0001390_00311

V ENGINEERED SAFETY FEATURES z E External Surfaces of Components and Miscellaneous Bolting c

U m

G) Structure I

Aging Effect/ Further 00 Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component

u V.E.EP-116 Bolting Copper Any Loss of preload Chapter XI.M18, "Bolting Integrity" No CD

< alloy environment due to thermal I\)

effects, gasket creep, and self-loosening V.E.EP-117 Bolting Nickel alloy Any Loss of preload Chapter XI.M18, "Bolting Integrity" No environment due to thermal effects, gasket creep, and self-loosening V.E.EP-120 Bolting Stainless Treated borated Loss of preload Chapter XI.M18, "Bolting Integrity" No steel water due to thermal effects, gasket

< creep, and self-m I loosening I\)

V.E.E-41 V.E-2(E- Bolting Steel Air with borated Loss of material Chapter XI.M1 0, "Boric Acid No

41) water leakage due to boric acid Corrosion" corrosion V.E.EP-64 V.E-1 (EP- Bolting Steel; Air - outdoor Loss of material Chapter XI.M18, "Bolting Integrity" No

1) stainless (External) due to general steel (steel only),

pitting, and crevice corrosion V.E.EP-118 Bolting Steel; Air - outdoor Loss of preload Chapter XI.M18, "Bolting Integrity" No stainless (External) due to thermal steel effects, gasket creep, and self-loosening 0

0 CD G)

C')

CD 3

0 0-0 ...,

CD 0 I\)

0) a CD a 0

I 0

0 0)

N

oCD V ENGINEERED SAFETY FEATURES C')

CD E External Surfaces of Components and Miscellaneous Bolting 3

0-CD Structure I\) Aging Effect/ Further a Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation a Component V.E.EP-121 Bolting Steel; Fuel oil Loss of preload Chapter XI.M18, "Bolting Integrity" No stainless due to thermal steel effects, gasket creep, and self-loosening V.E.EP-119 Bolting Steel; Raw water Loss of preload Chapter XI.M18, "Bolting Integrity" No stainless due to thermal steel effects, gasket creep, and self-loosening V.E.EP-122 Bolting Steel; Treated water Loss of preload Chapter XI.M18, "Bolting Integrity" No stainless due to thermal

< steel effects, gasket m I creep, and self-CJ.)

loosening V.E.E-02 V.E-6(E- Closure Steel Air with steam Loss of material Chapter XI.M18, "Bolting Integrity" No

02) bolting or water due to general leakage corrosion V.E.E-03 V.E-3(E- Closure Steel, high- Air with steam Cracking Chapter XI.M18, "Bolting Integrity" No

03) bolting strength or water due to cyclic leakage loading, stress corrosion cracking V.E.EP-70 V.E-4(EP- Closure Steel; Air - indoor, Loss of material Chapter XI.M18, "Bolting Integrity" No z 25) bolting stainless uncontrolled due to general c

u steel (External) (steel only),

m pitting, and G)

I crevice corrosion 00 0 a G) ;U CD 0

0 0 I\)

0)

CD 0

I 0

0 0) 0)

V ENGINEERED SAFETY FEATURES z E External Surfaces of Components and Miscellaneous Bolting c

U m

G) Structure I

Aging Effect/ Further 00 Item Link and/or Material Environment Aging Management Program (AMP) a Mechanism Evaluation Component

u V.E.EP-69 V.E-5(EP- Closure Steel; Air - indoor, Loss of preload Chapter XI.M18, "Bolting Integrity" No CD

< 24) bolting stainless uncontrolled due to thermal I\)

steel (External) effects, gasket creep, and self-loosening V.E.E-44 V.E-7(E- External Steel Air - indoor, Loss of material Chapter XI.M36, "External Surfaces No

44) surfaces uncontrolled due to general Monitoring of Mechanical (External) corrosion Components" V.E.E-45 V.E-8(E- External Steel Air - outdoor Loss of material Chapter XI.M36, "External Surfaces No

45) surfaces (External) due to general Monitoring of Mechanical corrosion Components" V.E.E-28 V.E-9(E- External Steel Air with borated Loss of material Chapter XI.M1 0, "Boric Acid No

28) surfaces water leakage due to boric acid Corrosion"

< corrosion m I

.j:>.

V.E.E-46 V.E-10(E- External Steel Condensation Loss of material Chapter XI.M36, "External Surfaces No

46) surfaces (External) due to general Monitoring of Mechanical corrosion Components" V.E.EP-114 Piping, piping Aluminum Air - outdoor Loss of material Chapter XI.M36, "External Surfaces No components, due to pitting and Monitoring of Mechanical and piping crevice corrosion Components" elements V.E.EP-38 V.E- Piping, piping Copper Air with borated Loss of material Chapter XI.M1 0, "Boric Acid No 11 (EP-38) components, alloy (>15% water leakage due to boric acid Corrosion" and piping Zn or >8% corrosion elements AI)

V.E.EP-123 Underground Steel; Air-indoor, Loss of material Chapter XI.M41, "Buried and No piping, piping stainless uncontrolled due to general Underground Piping and Tanks" 0

0 CD components, steel (External) or (steel only),

G)

C')

CD and piping condensation pitting and 3 elements (External) crevice corrosion 0 0-0 ...,

CD 0 I\)

0) a CD a 0

I 0

0 0)

.j::>.

F. COMMON MISCELLANEOUS MATERIAL/ENVIRONMENT COMBINATIONS Systems, Structures, and Components This section addresses the aging management programs for miscellaneous material/environment combinations which may be found throughout the emergency safety feature system's structures and components. For the material/environment combinations in this part, aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation and, therefore, no resulting aging management programs for these structures and components are required.

System Interfaces The structures and components covered in this section belong to the engineered safety features in pressurized water reactors (PWRs) and boiling water reactors (8WRs). (For example, see System Interfaces in V.A to V.D2 for details.)

December 201 0 V F-1 NUREG-1801, Rev. 2 OAGI0001390_00315

V ENGINEERED SAFETY FEATURES z F Common Miscellaneous Material/Environment Combinations c

U m

G) Structure I

Aging Effect/ Aging Management Program Further 00 Item Link and/or Material Environment a Mechanism (AMP) Evaluation Component

u V.F.EP-14 V.F-1 (EP- Ducting, Galvanized Air - indoor, None None No CD

< 14) piping, and steel controlled I\)

components (External)

V.F.EP-15 V.F-6(EP- Piping Glass Air - indoor, None None No

15) elements uncontrolled (External)

V.F.EP-87 Piping Glass Air - outdoor None None No elements V.F.EP-65 Piping Glass Air with borated None None No elements water leakage

< V.F.EP-68 Piping Glass Closed-cycle None None No

""T1 I

I\) elements cooling water V.F.EP-66 Piping Glass Condensation None None No elements (Internal/External)

V.F.EP-67 Piping Glass Gas None None No elements V.F.EP-16 V.F-7(EP- Piping Glass Lubricating oil None None No

16) elements V.F.EP-28 V.F-8(EP- Piping Glass Raw water None None No 0 28) elements 0 CD G)

C')

CD 3

0 0-0 ...,

CD 0 I\)

0) a CD a 0

I 0

0 0)

(J)

oCD V ENGINEERED SAFETY FEATURES C')

CD F Common Miscellaneous Material/Environment Combinations 3

0-CD I\) Structure a Aging Effect/ Aging Management Program Further Item Link and/or Material Environment a Mechanism (AMP) Evaluation Component V.F.EP-29 V.F- Piping Glass Treated water None None No 10(EP-29) elements V.F.EP-30 V.F-9(EP- Piping Glass Treated water None None No

30) elements (borated)

V.F.EP-3 V.F-2(EP- Piping, piping Aluminum Air - indoor, None None No

3) components, uncontrolled and piping (Internal/External) elements V.F.EP-10 V.F-3(EP- Piping, piping Copper alloy Air - indoor, None None No

< 10) components, uncontrolled

""T1 I

and piping (External)

CJ.)

elements V.F.EP-9 V.F-4(EP- Piping, piping Copper alloy Gas None None No

9) components, and piping elements V.F.EP-12 V.F-5(EP- Piping, piping Copper alloy Air with borated None None No

12) components, (::::15% Zn and water leakage and piping  ::::8% AI) elements V.F.EP-17 V.F- Piping, piping Nickel alloy Air - indoor, None None No 11 (EP-17) components, uncontrolled z and piping (External) c

u elements m

G) V.F.EP-115 Piping, piping Nickel alloy Air with borated None None No I

components, water leakage 00 0 a and piping G)

- elements

U 0 CD 0

0 I\)

0)

CD 0

I 0

0 0)

-....J

V ENGINEERED SAFETY FEATURES z F Common Miscellaneous Material/Environment Combinations c

U m

G) Structure I

Aging Effect/ Aging Management Program Further 00 Item Link and/or Material Environment a Mechanism (AMP) Evaluation Component

u V.F.EP-18 V.F- Piping, piping Stainless steel Air - indoor, None None No CD

< 12(EP-18) components, uncontrolled I\)

and piping (External) elements V.F.EP-82 Piping, piping Stainless steel Air - indoor, None None No components, uncontrolled and piping (Internal) elements V.F.EP-19 V.F- Piping, piping Stainless steel Air with borated None None No 13(EP-19) components, water leakage and piping elements V.F.EP-20 V.F- Piping, piping Stainless steel Concrete None None No

< 14(EP-20) components,

""T1 I and piping

.j:>.

elements V.F.EP-22 V.F- Piping, piping Stainless steel Gas None None No 15(EP-22) components, and piping elements V.F.EP-4 V.F- Piping, piping Steel Air - indoor, None None No 16(EP-4) components, controlled and piping (External) elements V.F.EP-112 V.F- Piping, piping Steel Concrete None None, provided No, if 17(EP-5) components, 1) attributes of the concrete are conditions are and piping consistent with ACI 318 or ACI 349 met.

elements (low water-to-cement ratio, low 0 permeability, and adequate air 0 CD G)

C')

CD entrainment) as cited in NUREG-3 1557, and 0 0-0 ...,

CD 2) plant OE indicates no 0 I\) degradation of the concrete 0) a CD a 0

I 0

0 0)

CD

oCD V ENGINEERED SAFETY FEATURES C')

CD F Common Miscellaneous Material/Environment Combinations 3

0-CD Structure I\)

a Aging Effect/ Aging Management Program Further Item Link and/or Material Environment a Mechanism (AMP) Evaluation Component V.F.EP-7 V.F- Piping, piping Steel Gas None None No 18(EP-7) components, and piping elements

""T1 I

0"1 z

c

u m

G)

I 00 0 a G)

U 0 CD 0

0 I\)

0)

CD 0

I 0

0 0)

CD

CHAPTER VI ELECTRICAL COMPONENTS December 201 0 VI-i NUREG-1801, Rev. 2 OAGI0001390_00320

ELECTRICAL COMPONENTS A. Equipment Not Subject to 10 CFR 50.49 Environmental Qualification Requirements B. Equipment Subject to 10 CFR 50.49 Environmental Qualification Requirements December 201 0 VI-iii NUREG-1801, Rev. 2 OAG10001390_00321

A. EQUIPMENT NOT SUBJECT TO 10 CFR 50.49 ENVIRONMENTAL QUALIFICATION REQUIREMENTS Systems, Structures, and Components This section addresses electrical cables and connections that are not subject to the environmental qualification requirements of 10 CFR 50.49 and that are installed in power and instrumentation and control (I&C) applications. The power cables and connections addressed are low-voltage <<1000 volts) and medium-voltage (2 kilovolts [kV] to 35 kV). High voltage (> 35 kV) power cables and connections have unique, specialized constructions and must be evaluated on a plant-specific basis.

This section also addresses components that are relied upon to meet the station blackout (SBO) requirements for restoration of offsite power. The offsite power system relied upon in the plant-specific current licensing basis for compliance with 10 CFR 50.63, that is used to connect the plant to the offsite power source, is included in the SBO restoration equipment scope. The electrical distribution equipment out to the first circuit breaker with the offsite distribution system (i.e., equipment in the switchyard) should be included within the SBO restoration equipment scope. This path typically includes the circuit breakers that connect to the offsite system power transformers (startup transformers), the transformers themselves, the intervening overhead or underground circuits between circuit breaker and transformer and transformer and onsite electrical distribution system, and associated control circuits and structures. However, the staff's review is based on the plant-specific current licensing basis, regulatory requirements, and offsite power design configurations.

Electrical cables and their required terminations (i.e., connections) are typically reviewed as a single commodity. The types of connections included in this review are splices, mechanical connectors, fuse holders, and terminal blocks. This common review is translated into program actions, which treat cables and connections in the same manner.

Electrical cables and connections that are in the plant's environmental qualification (EQ) program are addressed in VI. B.

System Interfaces Electrical cables and connections functionally interface with all plant systems that rely on electric power or instrumentation and control. Electrical cables and connections also interface with and are supported by structural commodities (e.g., cable trays, conduit, cable trenches, cable troughs, duct banks, cable vaults, and manholes) that are reviewed, as appropriate, in the Systems, Structures, and Components section.

December 201 0 VI A-1 NUREG-1801, Rev. 2 OAG10001390_00322

z VI ELECTRICAL COMPONENTS c

U fA. Equipment Not Subject to 10 CFR 50.49 Environmental Qualification Requirements m

G)

I 00 a Structure and/or Aging Effect! Aging Management Further Item Link Material Environment Component Mechanism Program (AMP) Evaluation

u CD

< VI.A.LP- VI.A- Cable connections Various Air - indoor, Increased resistance of Chapter XI.E6, "Electrical No I\) 30 1(LP-12) (metallic parts) metals used controlled or connection Cable Connections Not for electrical uncontrolled or due to thermal cycling, Subject to 10 CFR 50.49 contacts Air - outdoor ohmic heating, electrical Environmental Qualification ransients, vibration, Requirements" chemical contamination, corrosion, and oxidation VI.A.LP- VI.A-4(L- Conductor insulation Various Adverse localized Reduced insulation Chapter XI.E3, "Inaccessible No 35 03) for inaccessible organic environment resistance Power Cables Not Subject to power cables greater polymers caused by due to moisture 10 CFR 50.49 Environmental than or equal to 400 (e.g., EPR, significant Qualification Requirements" volts (e.g., installed in SR, EPDM, moisture conduit or direct P<LPE) buried)

VI.A.LP- VI.A-5(L- Connector contacts Various Air with borated Increased resistance of Chapter XI.M1 0, "Boric Acid No 36 04) for electrical metals used water leakage connection Corrosion" connectors exposed for electrical due to corrosion of to borated water contacts connector contact leakage surfaces caused by intrusion of borated water VI.A.LP- VI.A- Fuse holders (not Insulation Air - indoor, None None No 24 7(LP-02) part of active material: controlled or equipment): bakelite; uncontrolled insulation material phenolic melamine or ceramic; molded polycarbonate; 0

CD other C')

0 CD G) 3 0-CD 0 I\)

0 a 0

0) a CD 0

I 0

0 0)

N 0)

oCD VI ELECTRICAL COMPONENTS C')

CD fA. Equipment Not Subject to 10 CFR 50.49 Environmental Qualification Requirements 3

0-CD I\) Structure and/or Aging Effect! Aging Management Further a Item Link Material Environment Component Mechanism Program (AMP) Evaluation a

VI.A.LP- VI.A- Fuse holders (not Various Air - indoor, Increased resistance of Chapter XI.E5, "Fuse No 31 8(LP-01) part of active metals used controlled or connection Holders" equipment): metallic for electrical uncontrolled due to fatigue caused by No aging management clamps connections frequent manipulation or program is required for those vibration applicants who can demonstrate these fuse holders are located in an environment that does not subject them to environmental aging mechanisms or fatigue caused by frequent manipulation or vibration VI.A.LP- VI.A- Fuse holders (not Various Air - indoor, Increased resistance of Chapter XI.E5, "Fuse No 23 8(LP-01) part of active metals used uncontrolled connection Holders" equipment): metallic for electrical due to chemical clamps connections contamination, corrosion, and oxidation (in an air, indoor controlled environment, increased resistance of connection due to chemical contamination, corrosion and oxidation do not apply);

fatigue due to ohmic heating, z hermal cycling, electrical c

u ransients m

G) VI.A.LP- VI.A- High-voltage Porcelain; Air - outdoor Loss of material A plant-specific aging Yes, plant-I 32 10(LP- insulators malleable iron; due to mechanical wear management program is to be specific 0 00 G) a 11 ) aluminum; caused by wind blowing evaluated

U galvanized on transmission 0

0 CD steel; cement conductors 0  :<

--" I\)

0)

CD 0

I 0

0 0)

N

.j::>.

z VI ELECTRICAL COMPONENTS c

U fA. Equipment Not Subject to 10 CFR 50.49 Environmental Qualification Requirements m

G)

I 00 a Structure and/or Aging Effect! Aging Management Further Item Link Material Environment Component Mechanism Program (AMP) Evaluation

u CD

< VI.A.LP- VI.A- High-voltage Porcelain; Air - outdoor Reduced insulation A plant-specific aging Yes, plant-I\) 28 9(LP-07) insulators malleable iron; resistance management program is to be specific aluminum; due to presence of salt evaluated for plants located galvanized deposits or surface such that the potential exists steel; cement contamination for salt deposits or surface contamination (e.g., in the vicinity of salt water bodies or industrial pollution)

VI.A.LP- VI.A-2(L- Insulation material for Various Adverse localized Reduced insulation Chapter XI.E1, "Insulation No 33 01) electrical cables and organic environment resistance Material for Electrical Cables connections polymers caused by heat, due to and Connections Not Subject (including terminal (e.g., EPR, radiation, or hermal/thermoxidative to 10 CFR 50.49 blocks, fuse holders, SR, EPDM, moisture degradation of organics, Environmental Qualification etc.) P<LPE) radiolysis, and photolysis Requirements" (UV sensitive materials only) of organics; radiation-induced oxidation; moisture intrusion VI.A.LP- VI.A-3(L- Insulation material for Various Adverse localized Reduced insulation Chapter XI.E2, "Insulation No 34 02) electrical cables and organic environment resistance Material for Electrical Cables connections used in polymers caused by heat, due to and Connections Not Subject instrumentation (e.g., EPR, radiation, or hermal/thermoxidative to 10 CFR 50.49 circuits that are SR, EPDM, moisture degradation of organics, Environmental Qualification sensitive to reduction P<LPE) radiolysis, and photolysis Requirements Used in in conductor (UV sensitive materials Instrumentation Circuits" insulation resistance only) of organics; (lR) radiation-induced 0 oxidation; CD C')

0 CD moisture intrusion G) 3 0-CD 0 I\)

0 a 0

0) a CD 0

I 0

0 0)

N (J1

oCD VI ELECTRICAL COMPONENTS C')

CD fA. Equipment Not Subject to 10 CFR 50.49 Environmental Qualification Requirements 3

0-CD I\) Structure and/or Aging Effect! Aging Management Further a Item Link Material Environment Component Mechanism Program (AMP) Evaluation a

VI.A.LP- VI.A- Metal enclosed bus: Various Air - indoor, Increased resistance of Chapter XI.E4, "Metal No 25 11 (LP- bus/connections metals used controlled or connection Enclosed Bus"

04) for electrical uncontrolled or due to the loosening of bus and Air - outdoor bolts caused by thermal connections cycling and ohmic heating VI.A.LP- VI.A- Metal enclosed bus: Elastomers Air - indoor, Surface cracking, crazing, Chapter XI.E4, "Metal No 29 12(LP- enclosure assemblies controlled or scuffing, dimensional Enclosed Bus," or

10) uncontrolled or change (e.g. "ballooning" Chapter XI.M38, "Inspection Air - outdoor and "necking"), shrinkage, of Internal Surfaces in discoloration, hardening Miscellaneous Piping and and loss of strength Ducting Components" due to elastomer degradation VI.A.LP- VI.A- Metal enclosed bus: Galvanized Air - indoor, None None No 41 13(LP- external surface of steel; controlled or

06) enclosure assemblies aluminum uncontrolled VI.A.LP- VI.A- Metal enclosed bus: Galvanized Air - outdoor Loss of material Chapter XI.E4, "Metal No 42 13(LP- external surface of steel; due to pitting and crevice Enclosed Bus," or

06) enclosure assemblies aluminum corrosion Chapter XI.S6, "Structures Monitoring" VI.A.LP- VI.A- Metal enclosed bus: Steel Air - indoor, None None No 44 13(LP- external surface of controlled

06) enclosure assemblies VI.A.LP- VI.A- Metal enclosed bus: Steel Air - indoor, Loss of material Chapter XI.E4, "Metal No z 43 13(LP- external surface of uncontrolled or due to general, pitting, Enclosed Bus," or c

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organic Air - outdoor hermoxidative polymers degradation of organics/

hermoplastics, radiation-induced oxidation, moisture/debris intrusion, and ohmic heating VI.A.LP- VI.A- Switchyard bus and fA,luminum; Air - outdoor Loss of material A plant-specific aging Yes, plant-39 15(LP- connections copper; due to wind-induced management program is to be specific

09) bronze; abrasion; evaluated stainless steel; Increased resistance of galvanized connection steel due to oxidation or loss of pre-load VI.A.LP- VI.A- Transmission fA,luminum Air - outdoor Loss of conductor strength None - for Aluminum None 46 16(LP- conductors due to corrosion Conductor Aluminum Alloy

08) Reinforced (ACAR)

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08) evaluated for Aluminum Conductor Steel Reinforced (ACSR)

VI.A.LP- VI.A- Transmission fA,luminum; Air - outdoor Loss of material A plant-specific aging Yes, plant-47 16(LP- conductors Steel due to wind-induced management program is to be specific

08) abrasion evaluated for ACAR and 0 ACSR CD C') VI.A.LP- VI.A- Transmission fA,luminum; Air - outdoor Increased resistance of A plant-specific aging Yes, plant-0 CD 48 16(LP- connectors steel connection management program is to be specific

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B. EQUIPMENT SUBJECT TO 10 CFR 50.49 ENVIRONMENTAL QUALIFICATION REQUIREMENTS Systems, Structures, and Components The Nuclear Regulatory Commission (NRC) has established nuclear station environmental qualification (EQ) requirements in 10 CFR Part 50 Appendix A, Criterion 4, and in 10 CFR 50.49. 10 CFR 50.49 specifically requires that an EQ program be established to demonstrate that certain electrical components located in harsh plant environments (i.e., those areas of the plant that could be subject to the harsh environmental effects of a loss of coolant accident [LOCA], high energy line breaks [HELBs] or post-LOCA radiation) are qualified to perform their safety function in those harsh environments after the effects of inservice aging.

10 CFR 50.49 requires that the effects of significant aging mechanisms be addressed as part of environmental qualification. Components in the EQ program have a qualified life, and the components are replaced at the end of that qualified life if it is shorter than the current operating term. The qualified life may be extended by methods such as refurbishment or reanalysis, but the licensee is required by the EQ regulation (10 CFR 50.49) to replace the component when its qualified life has expired.

Similarly, some nuclear power plants have mechanical equipment that was qualified in accordance with the provisions of Criterion 4 of Appendix A to 10 CFR Part 50.

System Interfaces Equipment subject to 10 CFR 50.49 environmental qualification requirements functionally interfaces with all plant systems that rely on electric power or instrumentation and control.

December 201 0 VI 8-1 NUREG-1801, Rev. 2 OAG10001390_00328

z VI ELECTRICAL COMPONENTS c

U B Equipment Subject to 10 CFR 50.49 Environmental Qualification Requirements m

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See Chapter X.E1, "Environmental Qualification (EQ) of Electric Components," of this report for meeting

the requirements of 10 CFR III 54.21 (c)(1)(iii).

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CHAPTER VII AUXILIARY SYSTEMS December 201 0 VII-i NUREG-1801, Rev. 2 OAGI0001390_00330

MAJOR PLANT SECTIONS A 1. New Fuel Storage A2. Spent Fuel Storage A3. Spent Fuel Pool Cooling and Cleanup (PWR)

A4. Spent Fuel Pool Cooling and Cleanup (BWR)

AS. Suppression Pool Cleanup System (BWR)

B. Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems C1. Open-Cycle Cooling Water System (Service Water System)

C2. Closed-Cycle Cooling Water System C3. Ultimate Heat Sink D. Compressed Air System E1. Chemical and Volume Control System (PWR)

E2. Standby Liquid Control System (BWR)

E3. Reactor Water Cleanup System (BWR)

E4. Shutdown Cooling System (Older BWR)

ES. Waste Water Systems F1. Control Room Area Ventilation System F2. Auxiliary and Radwaste Area Ventilation System F3. Primary Containment Heating and Ventilation System F4. Diesel Generator Building Ventilation System G. Fire Protection H1. Diesel Fuel Oil System H2. Emergency Diesel Generator System I. External Surfaces of Components and Miscellaneous Bolting J. Common Miscellaneous Material/Environment Combinations December 201 0 VII-iii NUREG-1801, Rev. 2 OAGI0001390_00331

A1. NEW FUEL STORAGE Systems, Structures, and Components This section discusses those structures and components used for new fuel storage which include carbon steel new fuel storage racks located in the auxiliary building or the fuel handling building. The racks are exposed to the temperature and humidity in the auxiliary building. The racks are generally painted with a protective coating. Based on Regulatory Guide 1.26, "Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants," all components used for new fuel storage are governed by Group C Quality Standards.

Aging management programs for the degradation of external surfaces of components and miscellaneous bolting are included in VII.I. Common miscellaneous material/environment combinations where aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation are included in VII.J.

System Interfaces No other systems discussed in this report interface with those used for new fuel storage.

December 201 0 VII A1-1 NUREG-1801, Rev. 2 OAGI0001390_00332

z VII AUXILIARY SYSTEMS c

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A2. SPENT FUEL STORAGE Systems, Structures, and Components This section discusses those structures and components used for spent fuel storage and includes stainless steel spent fuel storage racks and neutron-absorbing materials (e.g.,

Boraflex, Boral, or boron-steel sheets, if used) submerged in chemically treated oxygenated boiling water reactor (BWR) or borated pressurized water reactor (PWR) water. The intended function of a spent fuel rack is to separate spent fuel assemblies. Boraflex sheets fastened to the storage cells provide for neutron absorption and help maintain subcriticality of spent fuel assemblies in the spent fuel pool.

Based on Regulatory Guide 1.26, "Quality Group Classifications and Standards for Water-,

Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants," all components used for spent fuel storage are governed by Group C Quality Standards. In some plants, the Boraflex has been replaced by Boral or boron steel.

Aging management programs for the degradation of external surfaces of components and miscellaneous bolting are included in VI 1.1. Common miscellaneous material/environment combinations where aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation are included in VII.J.

The system piping includes all pipe sizes, including instrument piping.

System Interfaces No other systems discussed in this report interface with those used for spent fuel storage.

December 201 0 VII A2-1 NUREG-1801, Rev. 2 OAGI0001390_00334

z VII AUXILIARY SYSTEMS c

U A2 Spent Fuel Storage m

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1(AP-79) components, stainless steel water due to pitting and and piping cladding); crevice corrosion elements stainless steel VII.A2.A-96 VII.A2-6(A- Spent fuel Stainless steel Treated water Cracking Chapter XI.M2, "Water Chemistry" No

96) storage racks >60°C (>140°F) due to stress (BWR) corrosion cracking VII.A2.A-97 VII.A2-7(A- Spent fuel Stainless steel Treated borated Cracking Chapter XI.M2, "Water Chemistry" No

97) storage racks water >60°C due to stress (PWR) (>140°F) corrosion cracking

VII.A2.A-87 VII.A2-2(A- Spent fuel Boraflex Treated water Reduction of Chapter XI.M22, "Boraflex Monitoring" No I\)

87) storage racks: neutron-I I\) neutron- absorbing absorbing capacity sheets (BWR) due to boraflex degradation VII.A2.AP-236 VII.A2-3(A- Spent fuel Boral@; boron Treated water Reduction of Chapter XI.M40, "Monitoring of No

89) storage racks: steel, and neutron- Neutron-Absorbing Materials other neutron- other materials absorbing than Boraflex" absorbing (excluding capacity; change sheets (BWR) Boraflex) in dimensions and loss of material due to effects of SFP environment 0 VII.A2.A-86 VII.A2-4(A- Spent fuel Boraflex Treated borated Reduction of Chapter XI.M22, "Boraflex Monitoring" No CD C') 86) storage racks: water neutron-CD 0 neutron- absorbing G) 3 0- absorbing capacity CD 0 I\) sheets (PWR) due to boraflex 0 a degradation 0

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88) storage racks: steel, and water neutron- Neutron-Absorbing Materials other neutron- other materials absorbing than Boraflex" absorbing (excluding capacity; change sheets (PWR) Boraflex) in dimensions and loss of material due to effects of SFP environment I\)

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A3. SPENT FUEL POOL COOLING AND CLEANUP (PRESSURIZED WATER REACTOR)

Systems, Structures, and Components This section discusses the pressurized water reactor (PWR) spent fuel pool cooling and cleanup system and consists of piping, valves, heat exchangers, filters, linings, demineralizers, and pumps. The system contains borated water. The system removes heat from the spent fuel pool and transfers heat to the closed-cycle cooling water system, which in turn transfers heat to the open-cycle cooling water system. Based on Regulatory Guide 1.26, "Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants," all components that comprise the PWR spent fuel pool cooling and cleanup system are governed by Group C Quality Standards.

With respect to filters, these items are to be addressed consistent with the Nuclear Regulatory Commission (NRC) position on consumables, provided in the NRC letter from Christopher I.

Grimes to Douglas J. Walters of the Nuclear Energy Institute (NEI), dated March 10,2000.

Specifically, components that function as system filters are typically replaced based on performance or condition monitoring that identifies whether these components are at the end of their qualified lives and may be excluded, on a plant-specific basis, from an aging management review under 10 CFR 54.21(a)(1)(ii). As part of the methodology description, the application should identify the standards that are relied on for replacement, for example, National Fire Protection Association (NFPA) standards for fire protection equipment.

Pump and valve internals perform their intended functions with moving parts or with a change in configuration. They are also subject to replacement based on qualified life or specified time period. Pursuant to 10 CFR 54.21(a)(1), therefore, they are not subject to an aging management review.

Aging management programs for the degradation of external surfaces of components and miscellaneous bolting are included in VI 1.1. Common miscellaneous material/environment combinations where aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation are included in VII.J.

The system piping includes all pipe sizes, including instrument piping.

System Interfaces The systems that interface with the PWR spent fuel cooling and cleanup system are the PWR emergency core cooling system (V.D1), the closed-cycle cooling water system (VII.C2), and the PWR chemical and volume control system (VII.E1).

December 201 0 VII A3-1 NUREG-1801, Rev. 2 OAG10001390_00337

z VII AUXILIARY SYSTEMS c

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100 15) linings water loss of strength Internal Surfaces in Miscellaneous due to elastomer Piping and Ducting Components" degradation VII.A3.A-79 VII.A3-2(A- External Steel Air with borated Loss of material Chapter XI.M1 0, "Boric Acid No

79) surfaces water leakage due to boric acid Corrosion" corrosion VII.A3.AP- VII.A3-3(A- Heat Steel Closed-cycle Loss of material Chapter XI.M21A, "Closed Treated No 189 63) exchanger cooling water due to general, Water Systems" components pitting, crevice, and galvanic

corrosion VII.A3.AP-1 VII.A3- Piping, piping Aluminum Air with borated Loss of material Chapter XI.M1 0, "Boric Acid No CJ.) 4(AP-1) components, water leakage due to boric acid Corrosion" I\)

and piping corrosion elements VII.A3.AP- VII.A3- Piping, piping Copper alloy Closed-cycle Loss of material Chapter XI.M21A, "Closed Treated No 199 5(AP-12) components, cooling water due to general, Water Systems" and piping pitting, crevice, elements and galvanic corrosion VII.A3.AP-43 VII.A3- Piping, piping Copper alloy Closed-cycle Loss of material Chapter XI.M33, "Selective No 6(AP-43) components, (>15% Zn or >8% cooling water due to selective Leaching" and piping AI) leaching elements VII.A3.AP-31 VII.A3- Piping, piping Gray cast iron Treated water Loss of material Chapter XI.M33, "Selective No 7(AP-31 ) components, due to selective Leaching" 0 and piping leaching CD C')

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VII.A3 .AP-79 VII.A3- Piping, piping Steel (with Treated borated Loss of material Chapter XI.M2, "Water Chemistry" No 8(AP-79) components, stainless steel water due to pitting and and piping cladding); crevice corrosion elements stainless steel VII.A3.A-56 VII.A3- Piping, piping Steel (with Treated borated Cracking Chapter XI.M2, "Water Chemistry" No 10(A-56) components, stainless steel or water >60°C due to stress and piping nickel-alloy (>140°F) corrosion elements cladding) cracking z

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A4. SPENT FUEL POOL COOLING AND CLEANUP (BOILING WATER REACTOR)

Systems, Structures, and Components This section discusses the boiling water reactor (BWR) spent fuel pool cooling and cleanup system and consists of piping, valves, heat exchangers, filters, linings, demineralizers, and pumps. The system contains chemically treated oxygenated water. The system removes heat from the spent fuel pool and transfers the heat to the closed-cycle cooling water system, which in turn transfers the heat to the open-cycle cooling water system. Based on Regulatory Guide 1.26, "Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants," all components that comprise the BWR spent fuel pool cooling and cleanup system are governed by Group C Quality Standards.

With respect to filters, these items are to be addressed consistent with the Nuclear Regulatory Commission (NRC) position on consumables, provided in the NRC letter from Christopher I.

Grimes to Douglas J. Walters of the Nuclear Energy Institute (NEI), dated March 10,2000.

Specifically, components that function as system filters are typically replaced based on performance or condition monitoring that identifies whether these components are at the end of their qualified lives and may be excluded, on a plant-specific basis, from an aging management review under 10 CFR 54.21(a)(1)(ii). As part of the methodology description, the application should identify the standards that are relied on for replacement, for example, National Fire Protection Association (NFPA) standards for fire protection equipment.

Pump and valve internals perform their intended functions with moving parts or with a change in configuration. They are also subject to replacement based on qualified life or a specified time period. Pursuant to 10 CFR 54.21(a)(1), therefore, they are not subject to an aging management review.

Aging management programs for the degradation of external surfaces of components and miscellaneous bolting are included in VI 1.1. Common miscellaneous material/environment combinations where aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation are included in VII.J.

The system piping includes all pipe sizes, including instrument piping.

System Interfaces The systems that interface with the BWR spent fuel cooling and cleanup system are the closed-cycle cooling water system (VII.C2) and the condensate system (VIII.E).

December 201 0 VII A4-1 NUREG-1801, Rev. 2 OAG10001390_00340

z VII AUXILIARY SYSTEMS c

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AS. SUPPRESSION POOL CLEANUP SYSTEM (BOILING WATER REACTOR)

Systems, Structures, and Components This section discusses the suppression pool cleanup system, which maintains water quality in the suppression pool in boiling water reactors (BWRs). The components of this system include piping, filters, valves, and pumps. These components are fabricated of carbon, low-alloy, or austenitic stainless steel. Based on Regulatory Guide 1.26, "Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants," the components that comprise the suppression pool cleanup system are governed by the same Group C Quality Standards Group as the corresponding components in the spent fuel pool cooling and cleanup system (VII.A4).

Pump and valve internals perform their intended functions with moving parts or with a change in configuration. They are also subject to replacement based on qualified life or specified time period. Pursuant to 10 CFR 54.21(a)(1), therefore, they are not subject to an aging management review.

Aging management programs for the degradation of external surfaces of components and miscellaneous bolting are included in VI 1.1. Common miscellaneous material/environment combinations where aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation are included in VII.J.

The system piping includes all pipe sizes, including instrument piping.

System Interfaces The system that interfaces with the suppression pool cleanup system is the BWR containments (lI.B), or BWR emergency core cooling system (V.D2).

Evaluation Summary There are no tables associated with this section because the suppression pool cleanup system in BWRs is similar to the spent fuel pool cooling and cleanup system (VII.A4), and the components in the two systems are identical or very similar. Therefore, the reader is referred to the section for the spent fuel storage pool system for a listing of aging effects, aging mechanisms, and aging management programs that are to be applied to the suppression pool cleanup system components. (The only component in VII.A4 that may not be applicable to the suppression pool cleanup system is the heat exchanger [AMR line-items VII.A4.AP-111, VII.A4.4AP-139, VII.A4.AP-189].)

December 201 0 VII AS-1 NUREG-1801, Rev. 2 OAG10001390_00343

B. OVERHEAD HEAVY LOAD AND LIGHT LOAD (RELATED TO REFUELING)

HANDLING SYSTEMS Systems, Structures, and Components Most commercial nuclear facilities have between fifty and one hundred cranes. Many of these cranes are industrial grade cranes that must meet the requirements of 29 CFR Volume XVII, Part 1910, and Section 1910.179. They do not fall within the scope of 10 CFR Part 54.4 and therefore are not required to be part of the integrated plant assessment (lPA). Normally fewer than ten cranes fall within the scope of 10 CFR Part 54.4. These cranes must comply with the requirements provided in 10 CFR Part 50.65 and Reg. Guide 1.160 for monitoring the effectiveness of maintenance at nuclear power plants.

The Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems (the Program) must demonstrate that the testing and the monitoring of the maintenance programs have been completed to ensure that the structures, systems, and components of these cranes are capable of sustaining their rated loads during the period of extended operation. The inspection is also to evaluate whether the usage of the cranes or hoists has been sufficient to warrant additional fatigue analysis. It should be noted that many of the systems and components of these cranes can be classified as moving parts or as components which change configuration, or they may be subject to replacement based on a qualified life. In any of these cases, they will not fall within the scope of this Aging Management Review (AMR).

The primary components that this program is concerned with are the structural girders and beams that make up the bridge and the trolley.

Based on Regulatory Guide 1.26, "Quality Group Classifications and Standards for Water-,

Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants," all components that comprise the overhead heavy load and light load handling systems are governed by Group C Quality Standards.

Aging management programs for the degradation of external surfaces of components and miscellaneous bolting are included in VII.I. Common miscellaneous material/environment combinations where aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation are included in VII.J.

System Interfaces No other systems discussed in this report interface with the overhead heavy load and light load (related to refueling) handling systems. Physical interfaces exist with the supporting structure.

The direct interface is at the connection to the structure.

December 201 0 VII 8-1 NUREG-1801, Rev. 2 OAG10001390_00344

VII AUXILIARY SYSTEMS z B Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems c

U m Structure G) Aging Effect/ Further I

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07) and structural uncontrolled due to general Overhead Heavy Load and Light Load girders (External) corrosion (Related to Refueling) Handling Systems" VII.B.A-06 VI I. B-2(A- Cranes: Steel Air - indoor, Cumulative Fatigue is a time-limited aging analysis Yes, TLAA

06) structural uncontrolled fatigue damage (TLAA) to be evaluated for the period of girders (External) due to fatigue extended operation for structural girders of cranes that fall within the scope of 10

CFR 54. See SRP-LR Sec. 4.7, "Other III Plant-Specific Time-Limited Aging I

I\) Analyses," for generic guidance for meeting the requirements of 10 CFR 54.21 (c)(1)).

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C1. OPEN-CYCLE COOLING WATER SYSTEM (SERVICE WATER SYSTEM)

Systems, Structures, and Components This section discusses the open-cycle cooling water (OCCW) (or service water) system, which consists of piping, heat exchangers, pumps, flow orifices, basket strainers, and valves, including containment isolation valves. Because the characteristics of an OCCW system may be unique to each facility, the OCCW system is defined as a system or systems that transfer heat from safety-related systems, structures, and components (SSCs) to the ultimate heat sink (UHS),

such as a lake, ocean, river, spray pond, or cooling tower. The AMPs described in this section apply to any such system, provided the service conditions and materials of construction are identical to those identified in the section. The system removes heat from the closed-cycle cooling water system, and, in some plants, other auxiliary systems and components, such as steam turbine bearing oil coolers or miscellaneous coolers in the condensate system. The only heat exchangers addressed in this section are those removing heat from the closed-cycle cooling system. Heat exchangers for removing heat from other auxiliary systems and components are addressed in their respective systems, such as those for the steam turbine bearing oil coolers (VIII.A) and for the condensate system coolers (VIII.E).

Based on Regulatory Guide 1.26, "Quality Group Classifications and Standards for Water-,

Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants," all components that comprise the open-cycle cooling water system are governed by Group C Quality Standards, with the exception of those forming part of the containment penetration boundary which are governed by Group B Quality Standards.

Pump and valve internals perform their intended functions with moving parts or with a change in configuration. They are also subject to replacement based on qualified life or a specified time period. Pursuant to 10 CFR 54.21(a)(1), therefore, they are not subject to an aging management review.

Aging management programs for the degradation of external surfaces of components and miscellaneous bolting are included in VII.I. Common miscellaneous material/environment combinations where aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation are included in VII.J.

The system piping includes all pipe sizes, including instrument piping.

System Interfaces The systems that may interface with the open-cycle cooling water system include the closed-cycle cooling water system (VII.C2), the ultimate heat sink (VII.C3), the emergency diesel generator system (VII.H2), the containment spray system (V.A), the PWR steam generator blowdown system (VI II. F), the condensate system (VIII.E), the auxiliary feedwater system (PWR) (VIII.G), the emergency core cooling system (PWR) (V.D1), and the emergency core cooling system (BWR) (V.D2).

December 201 0 VII C1-1 NUREG-1801, Rev. 2 OAG10001390_00346

VII AUXILIARY SYSTEMS z C1 Open-Cycle Cooling Water System (Service Water System) c

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u CD 75 1 (AP-75) seals and loss of strength Cooling Water System"

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degradation VII.C1.AP- VII.C1- Elastomer: Elastomers Raw water Loss of material Chapter XI.M20, "Open-Cycle No 76 2(AP-76) seals and due to erosion Cooling Water System" components VII.C1.AP- VII.C1- Heat Copper alloy Raw water Loss of material Chapter XI.M20, "Open-Cycle No 179 3(A-65) exchanger due to general, Cooling Water System" components pitting, crevice, galvanic, and microbiolog ically-influenced I

corrosion; fouling I\) that leads to corrosion VII.C1.A-66 VII.C1- Heat Copper alloy (>15% Zn Raw water Loss of material Chapter XI.M33, "Selective No 4 (A-66) exchanger or >8% AI) due to selective Leaching" components leaching VII.C1.AP- VII.C1- Heat Steel Raw water Loss of material Chapter XI.M20, "Open-Cycle No 183 5 (A-64) exchanger due to general, Cooling Water System" components pitting, crevice, galvanic, and microbiolog ically-influenced corrosion; fouling that leads to 0

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other than aluminum or more than tubes 0.20% oxygen or any amount of tin)

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153 exchanger transfer Cooling Water System" tubes due to fouling VII.C1.AP- Piping, piping Aluminum Soil or concrete Loss of material Chapter XI.M41, "Buried and No 173 components, due to pitting and Underground Piping and Tanks" and piping crevice corrosion elements VII.C1.AP- Piping, piping Asbestos cement pipe Soil or concrete Cracking, spalling, Chapter XI.M41, "Buried and No 237 components, corrosion of rebar Underground Piping and Tanks" and piping due to exposure of elements rebar VII.C1.AP- Piping, piping Concrete Soil or concrete Cracking, spalling, Chapter XI.M41, "Buried and No z 178 components, corrosion of rebar Underground Piping and Tanks" c

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u CD 253 components, material material properties Surfaces Monitoring of

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elements chemical attack VII.C1.AP- Piping, piping Concrete; cementitious Air - outdoor Cracking Chapter XI.M36, "External No 251 components, material due to settling Surfaces Monitoring of and piping Mechanical Components" elements VII.C1.AP- Piping, piping Concrete; cementitious Air - outdoor Loss of material Chapter XI.M36, "External No 252 components, material due to abrasion, Surfaces Monitoring of and piping cavitation, Mechanical Components" elements aggressive chemical attack, and leaching I

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elements VII.C1.AP- Piping, piping HDPE Raw water Cracking, Chapter XI.M20, "Open-Cycle No 239 components, (internal) blistering, change Cooling Water System" and piping in color elements due to water absorption VII.C1.AP- Piping, piping HDPE Soil or concrete Cracking, Chapter XI.M41, "Buried and No 175 components, blistering, change Underground Piping and Tanks" and piping in color elements due to water absorption I

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-J I due to aggressive chemical attack VII.C1.AP- VII.C1- Piping, piping Stainless steel Lubricating oil Loss of material Chapter XI.M39, "Lubricating Oil No 138 14(AP-59) components, due to pitting, Analysis," and and piping crevice, and Chapter XI.M32, "One-Time elements microbiolog ically- Inspection" influenced corrosion VII.C1.A-54 VII.C1- Piping, piping Stainless steel Raw water Loss of material Chapter XI.M20, "Open-Cycle No 15(A-54) components, due to pitting and Cooling Water System" and piping crevice corrosion; z elements fouling that leads c to corrosion

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elements corrosion Inspection" VII.C1.AP- VII.C1- Piping, piping Steel (with coating or Raw water Loss of material Chapter XI.M20, "Open-Cycle No 194 19(A-38) components, lining) due to general, Cooling Water System" and piping pitting, crevice, elements and microbiolog ically-influenced corrosion; fouling that leads to corrosion; lining/coating 00 I degradation VII.C1.AP- VII.C1- Piping, piping Steel (with coating or Soil or concrete Loss of material Chapter XI.M41, "Buried and No 198 18(A-01) components, wrapping) due to general, Underground Piping and Tanks" and piping pitting, crevice, elements and microbiolog ically-influenced corrosion VII.C1.AP- Piping, piping Super austenitic Soil or concrete Loss of material Chapter XI.M41, "Buried and No 172 components, due to pitting and Underground Piping and Tanks" and piping crevice corrosion elements VII.C1.AP- Piping, piping Titanium Soil or concrete Loss of material Chapter XI.M41, "Buried and No 0 171 components, due to pitting and Underground Piping and Tanks" CD C') and piping crevice corrosion CD 0 elements G) 3 0-CD 0 I\)

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elements aluminum or more than 0.20% oxygen or any amount of tin)

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C2. CLOSED-CYCLE COOLING WATER SYSTEM Systems, Structures, and Components This section discusses the closed-cycle cooling water (CCCW) system, which consists of piping, radiation elements, temperature elements, heat exchangers, pumps, tanks, flow orifices, and valves, including containment isolation valves. The system contains chemically treated demineralized water. The closed-cycle cooling water system is designed to remove heat from various auxiliary systems and components such as the chemical and volume control system and the spent fuel cooling system to the open-cycle cooling water system (VII.C1). A CCCW system is defined as part of the service water system that does not reject heat directly to a heat sink, has water chemistry control, and is not subject to significant sources of contamination.

Based on RG 1.26, "Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants," all components in the closed-cycle cooling water system are classified as Group C Quality Standards, with the exception of those forming part of the containment penetration boundary, which are Group B.

The aging management programs (AMPs) for the heat exchanger between the closed-cycle and the open-cycle cooling water systems are addressed in the open-cycle cooling water system (VII.C1). The AMPs for the heat exchangers between the closed-cycle cooling water system and the interfacing auxiliary systems are included in the evaluations of their respective systems, such as those for the pressurized water reactor (PWR) and boiling water reactor (BWR) spent fuel pool cooling and cleanup systems (VII.A3 and VII.A4, respectively) and the PWR chemical and volume control system (VII.E1).

Pump and valve internals perform their intended functions with moving parts or with a change in configuration. They are also subject to replacement based on qualified life or a specified time period. Pursuant to 10 CFR 54.21(a)(1), therefore, they are not subject to an aging management review.

Aging management programs for the degradation of external surfaces of components and miscellaneous bolting are included in VII.I. Common miscellaneous material/environment combinations where aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation are included in VII.J.

The system piping includes all pipe sizes, including instrument piping.

System Interfaces The systems that interface with the closed-cycle cooling water system include the open-cycle cooling water system (VII.C1), the PWR spent fuel pool cooling and cleanup system (VII.A3),

the BWR spent fuel pool cooling and cleanup system (VII.A4), the PWR chemical and volume control system (VII.E1), the BWR reactor water cleanup system (VII.E3), the shutdown cooling system (older BWR, VII.E4), the primary containment heating and ventilation system (VII.F3),

fire protection (VII.G), the emergency diesel generator system (VII.H2), the PWR containment December 201 0 VII C2-1 NUREG-1801, Rev. 2 OAG10001390_00355

spray system (V.A), the PWR and BWR emergency core cooling systems (V.D1 and V.D2), the PWR steam generator blowdown system (VIII. F), the condensate system (VIII. E), and the PWR auxiliary feedwater system (VIII.G).

NUREG-1801, Rev. 2 VII C2-2 December 201 0 OAGI0001390_00356

VII AUXILIARY SYSTEMS oCD C2 Closed-Cycle Cooling Water System C')

CD 3 Structure 0- Aging Effect! Further CD Item Link and/or Material Environment Aging Management Program (AMP)

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C3. ULTIMATE HEAT SINK Systems, Structures, and Components The ultimate heat sink (UHS) consists of a lake, ocean, river, spray pond, or cooling tower. The UHS provides sufficient cooling water for safe reactor shutdown and reactor cooldown via the residual heat removal system or other similar system. Due to the varying configurations of connections to lakes, oceans, and rivers, a plant-specific aging management program (AMP) is required. Appropriate AMPs shall be provided to trend and project (1) deterioration of earthen dams and impoundments; (2) rate of silt deposition; (3) meteorological, climatological, and oceanic data since obtaining the Final Safety Analysis Report (FSAR) data; (4) water level extremes for plants located on rivers; and (5) aging degradation of all upstream and downstream dams affecting the UHS.

The systems, structures, and components included in this section consist of piping, valves, and pumps. The cooling tower is addressed in this report on water-control structures (lII.A6). The ultimate heat sink absorbs heat from the residual heat removal system or other similar system.

Based on Regulatory Guide 1.26, "Quality Group Classifications and Standards for Water-,

Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants," the piping and valves used for the ultimate heat sink are governed by Group C Quality Standards.

Pump and valve internals perform their intended functions with moving parts or with a change in configuration. They are also subject to replacement based on qualified life or a specified time period. Pursuant to 10 CFR 54.21(a)(1), therefore, they are not subject to an aging management review.

Aging management programs for the degradation of external surfaces of components and miscellaneous bolting are included in VII.I. Common miscellaneous material/environment combinations where aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation are included in VII.J.

The system piping includes all pipe sizes, including instrument piping.

System Interfaces The systems that interface with the ultimate heat sink include the open-cycle cooling water system (VII.C1) and the PWR and BWR emergency core cooling systems (V.D1 and V.D2).

December 201 0 VII C3-1 NUREG-1801, Rev. 2 OAG10001390_00360

VII AUXILIARY SYSTEMS z C3 Ultimate Heat Sink c

U m Structure G) Aging Effect/ Aging Management Program Further I

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47) components, (>15% Zn or due to selective Leaching" and piping >8% AI) leaching elements

VII.C3.A-51 VII.C3-4(A- Piping, piping Gray cast iron Raw water Loss of material Chapter XI.M33, "Selective No

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I and piping leaching I\)

elements VII.C3.A-02 VII.C3-5(A- Piping, piping Gray cast iron Soil Loss of material Chapter XI.M33, "Selective No

02) components, due to selective Leaching" and piping leaching elements VII.C3.AP- VII.C3- Piping, piping Nickel alloy Raw water Loss of material Chapter XI.M20, "Open-Cycle No 206 6(AP-53) components, due to general, Cooling Water System" and piping pitting, and crevice elements corrosion VII.C3.A-53 VII.C3-7(A- Piping, piping Stainless steel Raw water Loss of material Chapter XI.M20, "Open-Cycle No

53) components, due to pitting and Cooling Water System" and piping crevice corrosion elements 0

CD VII.C3.AP- VII.C3- Piping, piping Stainless steel Soil Loss of material Chapter XI.M41, "Buried and No C')

0 CD 137 8(AP-56) components, due to pitting and Underground Piping and Tanks" G) 3 0- and piping crevice corrosion CD elements 0 I\)

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0-CD Structure I\) Aging Effect/ Aging Management Program Further a Item Link and/or Material Environment Mechanism (AMP) Evaluation a Component VII.C3.AP- VII.C3- Piping, piping Steel (with Raw water Loss of material Chapter XI.M20, "Open-Cycle No 194 10(A-38) components, coating or due to general, Cooling Water System" and piping lining) pitting, crevice, and elements microbiologically-influenced corrosion; fouling that leads to corrosion; lin ing/coating degradation VII.C3.AP- VII.C3-9(A- Piping, piping Steel (with Soil Loss of material Chapter XI.M41, "Buried and No 198 01) components, coating or due to general, Underground Piping and Tanks" and piping wrapping) pitting, crevice, and elements microbiologically-influenced corrosion VII.C3.AP- Piping, piping Stainless steel Air - outdoor Cracking Chapter XI.M36, "External Surfaces Yes, 209 components, due to stress Monitoring of Mechanical environmental and piping corrosion cracking Components" conditions elements; need to be tanks evaluated VII.C3.AP- Piping, piping Stainless steel Air - outdoor Loss of material Chapter XI.M36, "External Surfaces Yes, 221 components, due to pitting and Monitoring of Mechanical environmental and piping crevice corrosion Components" conditions elements; need to be z tanks evaluated c

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D. COMPRESSED AIR SYSTEM Systems, Structures, and Components This section discusses the compressed air system, which consists of piping, valves (including containment isolation valves), air receivers, pressure regulators, filters, and dryers. The system components and piping are located in various buildings at most nuclear power plants. Based on Regulatory Guide 1.26, "Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants," all components of the compressed air system are classified as Group D Quality Standards, with the exception of those forming part of the containment penetration boundary, which are Group B. However, the cleanliness of these components and high air quality is to be maintained because the air provides the motive power for instruments and active components (some of them safety-related) that may not function properly if nonsafety Group D equipment is contaminated.

With respect to filters, these items are to be addressed consistent with the Nuclear Regulatory Commission (NRC) position on consumables, provided in the NRC letter from Christopher I.

Grimes to Douglas J. Walters of the Nuclear Energy Institute (NEI), dated March 10,2000.

Specifically, components that function as system filters are typically replaced based on performance or condition monitoring that identifies whether these components are at the end of their qualified lives and may be excluded, on a plant-specific basis, from an aging management review under 10 CFR 54.21(a)(1)(ii). As part of the methodology description, the application should identify the standards that are relied on for replacement, for example, National Fire Protection Association (NFPA) standards for fire protection equipment.

Pump and valve internals perform their intended functions with moving parts or with a change in configuration. They are also subject to replacement based on qualified life or a specified time period. Pursuant to 10 CFR 54.21(a)(1), therefore, they are not subject to an aging management review.

Aging management programs for the degradation of external surfaces of components and miscellaneous bolting are included in VI 1.1. Common miscellaneous material/environment combinations where aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation are included in VII.J.

The system piping includes all pipe sizes, including instrument piping.

System Interfaces Various other systems discussed in this report may interface with the compressed air system.

December 201 0 VII D-1 NUREG-1801, Rev. 2 OAG10001390_00363

VII AUXILIARY SYSTEMS z D Compressed Air System c

U m Structure and! Aging Effect! Further G)

Item Link Material Environment Aging Management Program (AMP)

I or Component Mechanism Evaluation 00 a VIID.AP-121 VII.D-1 (A- Closure bolting Steel; Condensation Loss of material Chapter XI.M18, "Bolting Integrity" No

u 103) stainless due to general CD

< steel (steel only),

I\) pitting, and crevice corrosion VIID.A-80 VII.D-3(A- Piping and Steel Air - indoor, Loss of material Chapter XI.M36, "External Surfaces No

80) components uncontrolled due to general Monitoring of Mechanical Components" (External (External) corrosion surfaces)

VIID.AP-240 Piping, piping Copper Condensation Loss of material Chapter XI.M24, "Compressed Air No components, alloy due to general, Monitoring" and piping pitting, and elements crevice corrosion

VIID.AP-81 VII.D- Piping, piping Stainless Condensation Loss of material Chapter XI.M24, "Compressed Air No o I 4(AP-81 ) components, steel (Internal) due to pitting and Monitoring" I\)

and piping crevice corrosion elements VIID.A-26 VII.D-2(A- Piping, piping Steel Condensation Loss of material Chapter XI.M24, "Compressed Air No

26) components, (Internal) due to general Monitoring" and piping and pitting elements: corrosion compressed air system VIID.AP-209 Piping, piping Stainless Air - outdoor Cracking Chapter XI.M36, "External Surfaces Yes, components, steel due to stress Monitoring of Mechanical Components" environmental and piping corrosion conditions elements; cracking need to be tanks evaluated 0 VIID.AP-221 Piping, piping Stainless Air - outdoor Loss of material Chapter XI.M36, "External Surfaces Yes, CD C')

CD components, steel due to pitting and Monitoring of Mechanical Components" environmental 0

G) 3 0-and piping crevice corrosion conditions CD elements; need to be 0

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E1. CHEMICAL AND VOLUME CONTROL SYSTEM (PRESSURIZED WATER REACTOR)

Systems, Structures, and Components This section discusses a portion of the pressurized water reactor (PWR) chemical and volume control system (CVCS). The portion of the PWR CVCS covered in this section extends from the isolation valves associated with the reactor coolant pressure boundary (and Code change as discussed below) to the volume control tank. This portion of the PWR CVCS consists of high-and low-pressure piping and valves (including the containment isolation valves), regenerative and letdown heat exchangers, pumps, basket strainers, and the volume control tank. The system contains chemically treated borated water; the shell side of the letdown heat exchanger contains closed-cycle cooling water (treated water).

Based on Regulatory Guide 1.26, "Quality Group Classifications and Standards for Water-,

Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants," all components that comprise the CVCS are governed by Group C Quality Standards. Portions of the CVCS extending from the reactor coolant system up to and including the isolation valves associated with reactor coolant pressure boundary are governed by Group A Quality Standards and covered in IV.C2.

Pump and valve internals perform their intended functions with moving parts or with a change in configuration. They are also subject to replacement based on qualified life or a specified time period. Pursuant to 10 CFR 54.21(a)(1), therefore, they are not subject to an aging management review.

Aging management programs for the degradation of external surfaces of components and miscellaneous bolting are included in VII.I. Common miscellaneous material/environment combinations where aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation are included in VII.J.

The system piping includes all pipe sizes, including instrument piping.

System Interfaces The systems that interface with the chemical and volume control system include the reactor coolant system (lV.C2), the emergency core cooling system (V.D1), the spent fuel pool cooling system (VII.A3), and the closed-cycle cooling water system (VII.C2).

December 201 0 VII E1-1 NUREG-1801, Rev. 2 OAG10001390_00365

VII AUXILIARY SYSTEMS z E1 Chemical and Volume Control System (PWR) c

U m

G) Structure and/or Aging Effect/ Aging Management Further I

Item Link Material Environment 00 Component Mechanism Program (AMP) Evaluation a

VII.E1.A-79 VII.E1- External surfaces Steel Air with borated Loss of material Chapter XI.M1 0, "Boric No

u CD 1(A-79) water leakage due to boric acid Acid Corrosion"

< corrosion I\)

VII.E1.AP- VII.E1- Heat exchanger Copper alloy Closed-cycle Loss of material Chapter XI.M21 A, No 203 2(AP-34) components cooling water due to general, "Closed Treated Water pitting, crevice, and Systems" galvanic corrosion VII.E1.AP- VII.E1- Heat exchanger Copper alloy Treated water Loss of material Chapter XI.M33, No 65 3(AP-65) components (>15% Zn or >8% due to selective "Selective Leaching" AI) leaching VII.E1.AP- VII.E1- Heat exchanger Stainless steel Treated borated Cracking Chapter XI.M2, "Water No 118 5(A-84) components water >60°C due to stress Chemistry," and

(>140°F) corrosion cracking Chapter XI.M32, "One-I I\)

Time Inspection" VII.E1.AP- VII.E1- Heat exchanger Steel Closed-cycle Loss of material Chapter XI.M21 A, No 189 6(A-63) components cooling water due to general, "Closed Treated Water pitting, crevice, and Systems" galvanic corrosion VII.E1.A- VII.E1- Heat exchanger Stainless steel Treated borated Cumulative fatigue Fatigue is a time-limited Yes, TLAA 100 4(A-100) components and water damage aging analysis (TLAA) to tubes due to fatigue be evaluated for the period of extended operation. See the SRP, Section 4.3 "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR 0

CD C')

54.21 (c)(1).

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a VII.E1.A-69 VII.E1- Heat exchanger Stainless steel Treated borated Cracking Chapter XI.M2, "Water Yes, plant-9(A-69) components, water >60°C due to stress Chemistry." specific non-regenerative (>140°F) corrosion cracking; The AMP is to be cyclic loading augmented by verifying the absence of cracking due to stress corrosion cracking and cyclic loading. An acceptable verification program is to include temperature and radioactivity monitoring of the shell side water, and eddy current testing of tubes.

I CJ.)

VII.E1.AP- VII.E1- High-pressure Stainless steel Treated borated Cracking Chapter XI.M1, "ASME No 115 7(A-76) pump, casing water due to cyclic loading Section XI Inservice Inspection, Subsections IWB, IWC, and IWD" VII.E1.AP- VII.E1- High-pressure Stainless steel Treated borated Cracking Chapter XI.M2, "Water No 114 7(A-76) pump, casing water >60°C due to stress Chemistry," and

(>140°F) corrosion cracking Chapter XI.M32, "One-Time Inspection" VII.E1.AP- VII.E1- High-pressure Steel, high- Air with steam or Cracking Chapter XI.M18, "Bolting No 122 8(A-104) pump, closure strength water leakage due to stress Integrity" bolting corrosion cracking; z cyclic loading c

u VII.E1.AP- VII.E1- Piping, piping Aluminum Air with borated Loss of material Chapter XI.M1 0, "Boric No m

G) 1 10(AP-1) components, and water leakage due to boric acid Acid Corrosion" I

piping elements corrosion 00 0 a G)

VII.E1.AP- VII.E1- Piping, piping Copper alloy Closed-cycle Loss of material Chapter XI.M21 A, No

U 199 11 (AP-12) components, and cooling water due to general, "Closed Treated Water CD 0

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VII.E1.AP- VII.E1- Piping, piping Copper alloy Lubricating oil Loss of material Chapter XI.M39, No

u CD 133 12(AP-47) components, and due to pitting and "Lubricating Oil Analysis,"

< piping elements crevice corrosion and I\)

Chapter XI.M32, "One-Time Inspection" VII.E1.AP- VII.E1- Piping, piping Copper alloy Closed-cycle Loss of material Chapter XI.M33, No 43 13(AP-43) components, and (>15% Zn or >8% cooling water due to selective "Selective Leaching" piping elements AI) leaching VII.E1.AP- VII.E1- Piping, piping Gray cast iron Treated water Loss of material Chapter XI.M33, No 31 14(AP-31) components, and due to selective "Selective Leaching" piping elements leaching VII.E1.AP- VII.E1- Piping, piping Stainless steel Lubricating oil Loss of material Chapter XI.M39, No 138 15(AP-59) components, and due to pitting, "Lubricating Oil Analysis,"

I

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piping elements crevice, and and microbiologically- Chapter XI.M32, "One-influenced corrosion Time Inspection" VII.E1.A-57 VII.E1- Piping, piping Stainless steel Treated borated Cumulative fatigue Fatigue is a time-limited Yes, TLAA 16(A-57) components, and water damage aging analysis (TLAA) to piping elements due to fatigue be evaluated for the period of extended operation. See the SRP, Section 4.3 "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR 54.21 (c)(1).

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a VII.E1.A-34 VII.E1- Piping, piping Steel Air - indoor, Cumulative fatigue Fatigue is a time-limited Yes, TLAA 18(A-34) components, and uncontrolled damage aging analysis (TLAA) to piping elements due to fatigue be evaluated for the period of extended operation. See the SRP, Section 4.3 "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR 54.21 (c)(1).

VII.E1.AP- VII.E1- Piping, piping Steel Lubricating oil Loss of material Chapter XI.M39, No 127 19(AP-30) components, and due to general, "Lubricating Oil Analysis,"

piping elements pitting, and crevice and I

corrosion Chapter XI.M32, "One-0"1 Time Inspection" VII.E1.AP- VII.E1- Piping, piping Steel (with Treated borated Loss of material Chapter XI.M2, "Water No 79 17(AP-79) components, and stainless steel water due to pitting and Chemistry" piping elements cladding); stainless crevice corrosion steel VII.E1.AP- Piping, piping Stainless steel Air - outdoor Cracking Chapter XI.M36, "External Yes, 209 components, and due to stress Surfaces Monitoring of environmental piping elements; corrosion cracking Mechanical Components" conditions need tanks to be evaluated VII.E1.AP- Piping, piping Stainless steel Air - outdoor Loss of material Chapter XI.M36, "External Yes, 221 components, and due to pitting and Surfaces Monitoring of environmental z piping elements; crevice corrosion Mechanical Components" conditions need c

u tanks to be evaluated m

G) VII.E1.AP- VII.E1- Piping, piping Stainless steel Treated borated Cracking Chapter XI.M2, "Water No I

00 82 20(AP-82) components, and water >60°C due to stress Chemistry" 0 a piping elements; (>140°F) corrosion cracking G) tanks

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VII.E1.AP- VII.E1- Pump Casings Steel (with Treated borated Loss of material A plant-specific aging Yes, verify that

u CD 85 21 (AP-85) stainless steel or water due to cladding management program is plant-specific

< nickel-alloy breach to be evaluated. program I\)

cladding) Reference NRC add resses clad Information Notice 94-63, cracking "Boric Acid Corrosion of Charging Pump Casings Caused by Cladding Cracks."

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E2. STANDBY LIQUID CONTROL SYSTEM (BOILING WATER REACTOR)

Systems, Structures, and Components This section discusses the portion of the standby liquid control (SLC) system extending from the containment isolation valve to the solution storage tank. The system serves as a backup reactivity control system in all boiling water reactors (BWRs). The major components of this system are the piping, the solution storage tank, the solution storage tank heaters, valves, and pumps. All of the components from the storage tank to the explosive actuated discharge valve operate in contact with a sodium pentaborate (Na2B1Q016"10H20) solution.

Based on Regulatory Guide 1.26, "Quality Group Classifications and Standards for Water-,

Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants," all components that comprise the standby liquid control system are governed by Group B Quality Standards. The portions of the standby liquid control system extending from the reactor coolant pressure boundary up to and including the containment isolation valves are governed by Group A Quality Standards and are covered in IV.C1.

Pump and valve internals perform their intended functions with moving parts or with a change in configuration. They are also subject to replacement based on qualified life or a specified time period. Pursuant to 10 CFR 54.21(a)(1), therefore, they are not subject to an aging management review.

Aging management programs for the degradation of external surfaces of components and miscellaneous bolting are included in VII.I. Common miscellaneous material/environment combinations where aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation are included in VII.J.

The system piping includes all pipe sizes, including instrument piping.

System Interfaces The system that interfaces with the SLC system is the BWR reactor pressure vessel (lV.A 1). If used, the SLC system would inject sodium pentaborate solution into the pressure vessel near the bottom of the reactor core.

December 201 0 VII E2-1 NUREG-1801, Rev. 2 OAG10001390_00371

VII AUXILIARY SYSTEMS z E2 Standby Liquid Control System (BWR) c

U m Structure G) Aging Effect! Further I

Item Link and/or Material Environment Aging Management Program (AMP) 00 Mechanism Evaluation a Component VII.E2.AP-141 VII.E2- Piping, piping Stainless Sodium Loss of material Chapter XI.M2, "Water Chemistry," and No

u CD 1(AP-73) components, steel penta borate due to pitting and Chapter XI.M32, "One-Time Inspection"

< and piping solution crevice corrosion I\)

elements VII.E2.AP-181 VI I. E2-2 (A- Piping, piping Stainless Sodium Cracking Chapter XI.M2, "Water Chemistry," and No

59) components, steel penta borate due to stress Chapter XI.M32, "One-Time Inspection" and piping solution >60°C corrosion cracking elements (>140°F) m I\)

I I\)

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E3. REACTOR WATER CLEANUP SYSTEM (BOILING WATER REACTOR)

Systems, Structures, and Components This section discusses the reactor water cleanup (RWCU) system, which provides for cleanup and particulate removal from the recirculating reactor coolant in all boiling water reactors (BWRs). Some plants may not include the RWCU system in the scope of license renewal, while other plants may include the RWCU system because it is associated with safety-related functions.

Based on Regulatory Guide 1.26, "Quality Group Classifications and Standards for Water-,

Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants," the portion of the RWCU system extending from the reactor coolant recirculation system up to and including the containment isolation valves are covered in IV.C1. The remainder of the system outboard of the isolation valves is governed by Group C Quality Standards. In this table, only aging management programs for RWCU-related piping and components outboard of the isolation valves are evaluated. The aging management program for containment isolation valves in the RWCU system is evaluated in IV.C1, which concerns the reactor coolant pressure boundary in BWRs.

Pump and valve internals perform their intended functions with moving parts or with a change in configuration. They are also subject to replacement based on qualified life or specified time period. Pursuant to 10 CFR 54.21(a)(1), therefore, they are not subject to an aging management review.

Aging management programs for the degradation of external surfaces of components and miscellaneous bolting are included in VII.I. Common miscellaneous material/environment combinations where aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation are included in VII.J.

The system piping includes all pipe sizes, including instrument piping.

System Interfaces The systems that interface with the BWR reactor water cleanup system include the reactor coolant pressure boundary (lV.C1), the closed-cycle cooling water system (VII.C2), and the condensate system (VIII.E).

December 201 0 VII E3-1 NUREG-1801, Rev. 2 OAG10001390_00373

VII AUXILIARY SYSTEMS z E3 Reactor Water Cleanup System c

U m Structure G) Aging Effect/ Further I

Item Link and/or Material Environment Aging Management Program (AMP) 00 Mechanism Evaluation a Component VII.E3.AP- VII.E3-1 (A- Heat Stainless Closed-cycle Loss of material Chapter XI.M21A, "Closed Treated No

u CD 191 67) exchanger steel; steel cooling water due to Water Systems"

< components with microbiologically-I\)

stainless influenced steel corrosion cladding VII.E3.AP- VII.E3-2(A- Heat Stainless Closed-cycle Cracking Chapter XI.M21A, "Closed Treated No 192 68) exchanger steel; steel cooling water due to stress Water Systems" components with >60°C (>140°F) corrosion cracking stainless steel cladding

VII.E3.AP- VII.E3-3(A- Heat Stainless Treated water Cracking Chapter XI.M2, "Water Chemistry," No m

CJ.)

112 71 ) exchanger steel; steel >60°C (>140°F) due to stress and I

I\) components with corrosion cracking Chapter XI.M32, "One-Time stainless Inspection" steel cladding VII.E3.AP- VII.E3-4(A- Heat Steel Closed-cycle Loss of material Chapter XI.M21A, "Closed Treated No 189 63) exchanger cooling water due to general, Water Systems" components pitting, crevice, and galvanic corrosion VII.E3.AP- VII.E3- Heat Stainless Closed-cycle Reduction of heat Chapter XI.M21A, "Closed Treated No 188 5(AP-63) exchanger steel cooling water transfer Water Systems" tubes due to fouling VII.E3.AP- VII.E3- Heat Stainless Treated water Reduction of heat Chapter XI.M2, "Water Chemistry," No 139 6(AP-62) exchanger steel transfer and 0 tubes due to fouling Chapter XI.M32, "One-Time CD C')

CD Inspection" 0 3 G) 0-VII.E3.AP- VII.E3- Piping, piping Aluminum Treated water Loss of material Chapter XI.M2, "Water Chemistry," No CD 130 7(AP-38) components, due to pitting and and 0 I\)

0 a and piping crevice corrosion Chapter XI.M32, "One-Time 0 elements Inspection"

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CD Mechanism Evaluation I\) Component a

VII.E3.AP- VII.E3- Piping, piping Copper alloy Closed-cycle Loss of material Chapter XI.M21A, "Closed Treated No a

199 8(AP-12) components, cooling water due to general, Water Systems" and piping pitting, crevice, and elements galvanic corrosion VII.E3.AP- VII.E3- Piping, piping Copper alloy Treated water Loss of material Chapter XI.M2, "Water Chemistry," No 140 9(AP-64) components, due to general, and and piping pitting, crevice, and Chapter XI.M32, "One-Time elements galvanic corrosion Inspection" VII.E3.AP-43 VII.E3- Piping, piping Copper alloy Closed-cycle Loss of material Chapter XI.M33, "Selective Leaching" No 10(AP-43) components, (>15% Zn or cooling water due to selective and piping >8% AI) leaching elements

VII.E3.AP-32 VII.E3- Piping, piping Copper alloy Treated water Loss of material Chapter XI.M33, "Selective Leaching" No mCJ.)

11 (AP-32) components, (>15% Zn or due to selective I

CJ.) and piping >8% AI) leaching elements VII.E3.AP-31 VII.E3- Piping, piping Gray cast Treated water Loss of material Chapter XI.M33, "Selective Leaching" No 12(AP-31) components, iron due to selective and piping leaching elements VII.E3.AP- VII.E3- Piping, piping Stainless Closed-cycle Cracking Chapter XI.M21A, "Closed Treated No 186 13(AP-60) components, steel cooling water due to stress Water Systems" and piping >60°C (>140°F) corrosion cracking elements VII.E3.A-62 VII.E3- Piping, piping Stainless Treated water Cumulative fatigue Fatigue is a time-limited aging Yes, TLAA z 14(A-62) components, steel damage analysis (TLAA) to be evaluated for c

u and piping due to fatigue the period of extended operation. See m elements the SRP, Section 4.3 "Metal Fatigue,"

G)

I for acceptable methods for meeting 0 a 00 the requirements of 10 CFR G) 54.21 (c)(1).

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VII AUXILIARY SYSTEMS z E3 Reactor Water Cleanup System c

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Item Link and/or Material Environment Aging Management Program (AMP) 00 Mechanism Evaluation a Component VII.E3.AP- VII.E3- Piping, piping Stainless Treated water Loss of material Chapter XI.M2, "Water Chemistry," No

u CD 110 15(A-58) components, steel due to pitting and and

< and piping crevice corrosion Chapter XI.M32, "One-Time I\)

elements Inspection" VII.E3.AP- VII.E3- Piping, piping Stainless Treated water Cracking Chapter XI.M2, "Water Chemistry," No 283 16(A-60) components, steel >60°C (>140°F) due to stress and and piping corrosion cracking, Chapter XI.M25, "BWR Reactor Water elements intergranular stress Cleanup System" corrosion cracking VII.E3.A-34 VII.E3- Piping, piping Steel Air - indoor, Cumulative fatigue Fatigue is a time-limited aging Yes, TLAA 17(A-34) components, uncontrolled damage analysis (TLAA) to be evaluated for and piping due to fatigue the period of extended operation. See elements the SRP, Section 4.3 "Metal Fatigue,"

for acceptable methods for meeting the requirements of 10 CFR 54.21 (c)(1).

VII.E3.AP- VII.E3- Piping, piping Steel Treated water Loss of material Chapter XI.M2, "Water Chemistry," No 106 18(A-35) components, due to general, and and piping pitting, and crevice Chapter XI.M32, "One-Time elements corrosion Inspection" VII.E3.AP- VII.E3- Regenerative Stainless Treated water Cracking Chapter XI.M2, "Water Chemistry," No 120 19(A-85) heat steel >60°C (>140°F) due to stress and exchanger corrosion cracking Chapter XI.M32, "One-Time components Inspection" 0

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E4. SHUTDOWN COOLING SYSTEM (OLDER BWR)

Systems, Structures, and Components This section discusses the shutdown cooling (SOC) system for older vintage boiling water reactors (BWRs) and consists of piping and fittings, the SOC system pump, the heat exchanger, and valves.

Based on Regulatory Guide 1.26, "Quality Group Classifications and Standards for Water-,

Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants," all components that comprise the SOC system are governed by Group B Quality Standards.

Portions of the SOC system extending from the reactor coolant pressure boundary up to and including the containment isolation valves are governed by Group A Quality Standards and are covered in IV.C1.

Pump and valve internals perform their intended functions with moving parts or with a change in configuration. They are also subject to replacement based on qualified life or a specified time period. Pursuant to 10 CFR 54.21(a)(1), therefore, they are not subject to an aging management review.

Aging management programs for the degradation of external surfaces of components and miscellaneous bolting are included in VII.I. Common miscellaneous material/environment combinations where aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation are included in VII.J.

The system piping includes all pipe sizes, including instrument piping.

System Interfaces The systems that interface with the SOC system include the reactor coolant pressure boundary (lV.C1) and the closed-cycle cooling water system (VII.C2).

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