ML11349A076

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New York State (NYS) Pre-Filed Evidentiary Hearing Exhibit NYS00146B, NUREG-1801, Rev. 1, Volumes 1 and 2, Generic Aging Lessons Learned (GALL) Report (September 2005) (GALL Rev 1)
ML11349A076
Person / Time
Site: Indian Point  Entergy icon.png
Issue date: 09/30/2005
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, Vol. 1, Rev. 1
Download: ML11349A076 (323)
v • d • e


Text

NYS00146B Submitted: December 15, 2011 z III STRUCTURES AND COMPONENT SUPPORTS C

0 B1.1 Class 1 m

Gl Structure Aging Effect! Aging Management Program Further Item Link and!or Material Environment Mechanism (AMP) Evaluation

0 Component (D

< III.B11-8 III.B1.1. Support Galvanized Air with borated Loss of material! Chapter XI. M 10, "Boric Acid No members; steel, water leakage boric acid Corrosion" (TP-3) welds; bolted aluminum corrosion connections; support anchorage to building structure III.B11-9 III.B1.1. Support Stainless steel Air - indoor None None No members; uncontrolled (TP-5) welds; bolted connections; OJ support anchorage to

'" building structure III.B1.1-10 III.B1.1. Support Stainless steel Air with borated None None No members; water leakage (TP-4) welds; bolted connections; support anchorage to building structure (f)

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B1.1 Class 1 3

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'"oo Item Link and!or Component Material Environment Mechanism (AMP) Evaluation U1 III.B1.1-11 III.B1.1. Support Stainless Treated Water Loss of material! Chapter XI.M2, "Water Chemistry," No members; steel; steel < 60C <<140 F) general, pitting, for BWR water, and (TP-10) welds; bolted and crevice connections; corrosion Chapter XI.S3, "ASME Section XI, support Subsection IWF" anchorage to building structure III.B1.1-12 III.B1.1.1-c Support Steel Air - indoor Cumulative Fatigue is a time-limited aging Yes, members; uncontrolled fatigue damage! analysis (TLAA) to be evaluated for TLAA (T-26) welds; bolted fatigue the period of extended operation.

connections; See the Standard Review Plan, support (Only if CLB Section 4.3 "Metal Fatigue," for anchorage to fatigue analysis acceptable methods for meeting the building exists) requirements of 10 CFR 5421(c)(1) structure III.B1.1-13 III.B1.1.1-a Support Steel Air - indoor Loss of material! Chapter XI.S3, "ASME Section XI, No members; uncontrolled or general and Subsection IWF" (T-24) welds; bolted air - outdoor pitting corrosion connections; support anchorage to building structure z

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z III STRUCTURES AND COMPONENT SUPPORTS C

0 B1.1 Class 1 m

Gl Structure Aging Effect! Aging Management Program Further Item Link and!or Material Environment Mechanism (AMP) Evaluation

0 Component (D

< III.B1.1-14 III.B1.1.1-b Support Steel Air with borated Loss of material! Chapter XI. M 10, "Boric Acid No members; water leakage boric acid Corrosion" (T-25) welds; bolted corrosion connections; support anchorage to building structure III.B1.1-15 III.B1.1.3-a Vibration Non-metallic Air - indoor Reduction or Chapter XI.S3, "ASME Section XI, No isolation (eg, Rubber) uncontrolled or loss of isolation Subsection IWF" (T-33) elements air - outdoor function!

radiation OJ hardening, temperature, in humidity, sustained vibratory loading (f)

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III STRUCTURES AND COMPONENT SUPPORTS 3 B1.2 Class 2 and 3 0-

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'"oo Item Link Structure and!or Material Environment Aging Effect! Aging Management Program Further U1 Mechanism (AMP) Evaluation Component III.B1.2-1 III.B1.2.3-a Building Reinforced Air - indoor Reduction in Chapter XI.S6, "Structures Yes, if not concrete at concrete; uncontrolled or concrete anchor Monitoring Program" within the (T-29) locations of Grout air - outdoor capacity due to scope of the expansion local concrete applicant's and grouted degradation! structures anchors; service-induced monitoring grout pads for cracking or other program support base concrete aging plates mechanisms III.B1.2-2 III.B1.2.2-a Constant and Steel Air - indoor Loss of Chapter XI.S3, "ASME Section XI, No variable load uncontrolled or mechanical Subsection IWF" OJ (T-28) spring air - outdoor function!

cD hangers; corrosion, guides; stops distortion, dirt, overload, fatigue due to vibratory and cyclic thermal loads III.B1.2-3 III.B1.2.2-a Sliding Lubrite Air - indoor Loss of Chapter XI.S3, "ASME Section XI, No surfaces uncontrolled or mechanical Subsection IWF" (T-32) air - outdoor function!

corrosion, distortion, dirt, z overload, fatigue C due to vibratory

0 m and cyclic Gl thermal loads
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z III STRUCTURES AND COMPONENT SUPPORTS C

0 B1.2 Class 2 and 3 m

Gl Structure Aging Effect! Aging Management Program Further Item Link and!or Material Environment Mechanism (AMP) Evaluation

0 Component (D

< III.B1.2-4 III.B1.2. Support Aluminum Air - indoor None None No members; uncontrolled (TP-8) welds; bolted connections; support anchorage to building structure III.B12-5 III.B1.2. Support Galvanized Air - indoor None None No members; steel uncontrolled (TP-11) welds; bolted connections; OJ support anchorage to o building structure III.B1.2-6 III.B1.2. Support Galvanized Air with borated Loss of material! Chapter XI. M 10, "Boric Acid No members; steel, water leakage boric acid Corrosion" (TP-3) welds; bolted aluminum corrosion connections; support anchorage to building structure (f)

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B1.2 Class 2 and 3 3

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'"oo Item Link and!or Component Material Environment Mechanism (AMP) Evaluation U1 III.B12-7 III.B1.2. Support Stainless steel Air - indoor None None No members; uncontrolled (TP-5) welds; bolted connections; support anchorage to building structure III.B12-8 III.B1.2. Support Stainless steel Air with borated None None No members; water leakage (TP-4) welds; bolted connections; OJ support anchorage to building structure III.B12-9 III.B1.2.1-c Support Steel Air - indoor Cumulative Fatigue is a time-limited aging Yes, members; uncontrolled fatigue damage! analysis (TLAA) to be evaluated for TLAA (T-26) welds; bolted fatigue the period of extended operation.

connections; See the Standard Review Plan, support (Only if CLB Section 4.3 "Metal Fatigue," for anchorage to fatigue analysis acceptable methods for meeting the building exists) requirements of 10 CFR 5421(c)(1) structure z

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z III STRUCTURES AND COMPONENT SUPPORTS C

0 B1.2 Class 2 and 3 m

Gl Structure Aging Effect! Aging Management Program Further Item Link and!or Material Environment Mechanism (AMP) Evaluation

0 Component (D

< III.B1.2-10 III.B1.2.1-a Support Steel Air - indoor Loss of material! Chapter XI.S3, "ASME Section XI, No members; uncontrolled or general and Subsection IWF" (T-24) welds; bolted air - outdoor pitting corrosion connections; support anchorage to building structure III.B1.2-11 III.B1.2.1-b Support Steel Air with borated Loss of material! Chapter XI. M 10, "Boric Acid No members; water leakage boric acid Corrosion" (T-25) welds; bolted corrosion connections; OJ support anchorage to building structure III.B1.2-12 III.B1.2.2-a Vibration Non-metallic Air - indoor Reduction or Chapter XI.S3, "ASME Section XI, No isolation (eg, Rubber) uncontrolled or loss of isolation Subsection IWF" (T-33) elements air - outdoor function!

radiation hardening, temperature, humidity, sustained vibratory loading (f)

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III STRUCTURES AND COMPONENT SUPPORTS 3 B1,3 Class MC (BWR Containment Supports) 0-

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hangers; corrosion, guides; stops distortion, dirt, overload, fatigue due to vibratory and cyclic thermal loads III.B1,3-3 III.B1,3,2-a Sliding Lubrite Air - indoor Loss of Chapter XI.S3, "ASME Section XI, No surfaces uncontrolled or mechanical Subsection IWF" (T-32) air - outdoor function!

corrosion, distortion, dirt, z overload, fatigue C due to vibratory

0 m and cyclic Gl thermal loads
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z III STRUCTURES AND COMPONENT SUPPORTS C

0 B1.3 Class MC (BWR Containment Supports) m Gl Structure Aging Effect! Aging Management Program Further Item Link and!or Material Environment Mechanism (AMP) Evaluation
0 Component (D

< III.B1.3-4 III.B1.3. Support Aluminum Air - indoor None None No members; uncontrolled (TP-8) welds; bolted connections; support anchorage to building structure III.B13-5 III.B1.3. Support Galvanized Air - indoor None None No members; steel uncontrolled (TP-11) welds; bolted connections; OJ support anchorage to building structure III.B1.3-6 III.B1.3. Support Galvanized Air with borated Loss of material! Chapter XI. M 10, "Boric Acid No members; steel, water leakage boric acid Corrosion" (TP-3) welds; bolted aluminum corrosion connections; support anchorage to building structure (f)

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B1.3 Class MC (BWR Containment Supports) 3 0-

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'"oo Item Link and!or Component Material Environment Mechanism (AMP) Evaluation U1 III.B13-7 III.B1.3. Support Stainless steel Air - indoor None None No members; uncontrolled (TP-5) welds; bolted connections; support anchorage to building structure III.B13-8 III.B1.3. Support Stainless steel Air with borated None None No members; water leakage (TP-4) welds; bolted connections; OJ support anchorage to U1 building structure III.B13-9 III.B1.3.1-b Support Steel Air - indoor Cumulative Fatigue is a time-limited aging Yes, members; uncontrolled fatigue damage! analysis (TLAA) to be evaluated for TLAA (T-26) welds; bolted fatigue the period of extended operation.

connections; See the Standard Review Plan, support (Only if CLB Section 4.3 "Metal Fatigue," for anchorage to fatigue analysis acceptable methods for meeting the building exists) requirements of 10 CFR 5421(c)(1) structure z

C

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z III STRUCTURES AND COMPONENT SUPPORTS C

0 B1,3 Class MC (BWR Containment Supports) m Gl Structure Aging Effect! Aging Management Program Further Item Link and!or Material Environment Mechanism (AMP) Evaluation
0 Component (D

< III.B1,3-10 III.B1,3,1-a Support Steel Air - indoor Loss of material! Chapter XI.S3, "ASME Section XI, No members; uncontrolled or general and Subsection IWF" (T-24) welds; bolted air - outdoor pitting corrosion connections; support anchorage to building structure III.B1,3-11 III.B1,3,2-a Vibration Non-metallic Air - indoor Reduction or Chapter XI.S3, "ASME Section XI, No isolation (eg, Rubber) uncontrolled or loss of isolation Subsection IWF" (T-33) elements air - outdoor function!

radiation OJ hardening, temperature, humidity, sustained vibratory loading (f)

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B2. SUPPORTS FOR CABLE TRAYS, CONDUIT, HVAC DUCTS, TUBETRACK, INSTRUMENT TUBING, NON-ASME PIPING AND COMPONENTS Systems, Structures, and Components This section addresses supports and anchorage for cable trays, conduit, heating, ventilation, and air-conditioning (HVAC) ducts, TubeTrack, instrument tubing, and non-ASME piping and components. Applicable aging effects are identified and the aging management review is presented for each applicable combination of support component and aging effect.

System Interfaces Physical interfaces exist with the structure, system, or component being supported and with the building structural element to which the support is anchored. A primary function of supports is to provide anchorage of the supported element for internal and external design basis events, so that the supported element can perform its intended function.

September 2005 III B2-1 NUREG-1801, Rev. 1 OAGI0000203_335

z III STRUCTURES AND COMPONENT SUPPORTS C

0 B2 Supports for Cable Trays, Conduit, HVAC Ducts, TubeTrack, Instrument Tubing, Non-ASME Piping and Components m

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

Mechanism Evaluation

0 Component (D

< IILB2-1 IILB2.2-a Building Reinforced Air - indoor Reduction in Chapter XLS6, "Structures Monitoring Yes, if not concrete at concrete; uncontrolled or concrete anchor Program" within the (T-29) locations of Grout air - outdoor capacity due to scope of the expansion local concrete applicant's and grouted degradation! structures anchors; grou service-induced monitoring pads for cracking or other program support base concrete aging plates mechanisms IILB2-2 IILB2. Sliding Lubrite, Air - indoor Loss of Chapter XLS6, "Structures Monitoring No support graphitic tool uncontrolled mechanical Program" (TP-1) bearings and steel function!

sliding corrosion, support distortion, dirt, surfaces overload, fatigue due to vibratory and cyclic thermal loads IILB2-3 IILB2. Sliding Lubrite, Air - outdoor Loss of Chapter XLS6, "Structures Monitoring No support graphitic tool mechanical Program" (TP-2) bearings and steel function!

sliding corrosion, support distortion, dirt, surfaces overload, fatigue due to vibratory and cyclic (f) thermal loads (D

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Mechanism Evaluation Component IILB2-4 IILB2. Support Aluminum Air - indoor None None No members; uncontrolled (TP-8) welds; bolted connections; support anchorage to building structure IILB2-5 IILB2. Support Galvanized Air - indoor None None No OJ members; steel uncontrolled (TP-11) welds; bolted

'"w connections; support anchorage to building structure IILB2-6 IILB2. Support Galvanized Air with borated Loss of material! Chapter XL M 10, "Boric Acid No members; steel, water leakage boric acid Corrosion" (TP-3) welds; bolted aluminum corrosion connections; support z anchorage to C

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z III STRUCTURES AND COMPONENT SUPPORTS C

0 B2 Supports for Cable Trays, Conduit, HVAC Ducts, TubeTrack, Instrument Tubing, Non-ASME Piping and Components m

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

Mechanism Evaluation

0 Component (D

< IILB2-7 IILB2. Support Galvanized Air - outdoor Loss of material! Chapter XLS6, "Structures Monitoring No members; steel, pitting and Program" (TP-6) welds; bolted aluminum, crevice corrosion connections; stainless support steel anchorage to building structure IILB2-8 IILB2. Support Stainless Air - indoor None None No members; steel uncontrolled (TP-5) welds; bolted connections; support anchorage to building structure IILB2-9 IILB2. Support Stainless Air with borated None None No members; steel water leakage (TP-4) welds; bolted connections; support anchorage to building structure (f)

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'"oo U1 Structure Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation Component IILB2-10 IILB2.1-a Support Steel Air - indoor Loss of material! Chapter XLS6, "Structures Monitoring Yes, if not members; uncontrolled or general and Program" within the (T-30) welds; bolted air - outdoor pitting corrosion scope of the connections; applicant's support structures anchorage to monitoring building program structure IILB2-11 IILB2.1-b Support Steel Air with borated Loss of material! Chapter XL M 10, "Boric Acid No members; water leakage boric acid Corrosion" (T-25) welds; bolted corrosion connections; support anchorage to building structure z

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This Page Intentionally Left Blank NUREG-1801, Rev. 1 III B2-6 September 2005 OAGI0000203_340

B3. ANCHORAGE OF RACKS, PANELS, CABINETS, AN D ENCLOSURES FOR ELECTRICAL EQUIPMENT AND INSTRUMENTATION Systems, Structures, and Components This section addresses supports and anchorage for racks, panels, cabinets, and enclosures for electrical equipment and instrumentation. Applicable aging effects are identified and the aging management review is presented for each applicable combination of support component and aging effect.

System Interfaces Physical interfaces exist with the structure, system, or component being supported and with the building structural element to which the support is anchored. A primary function of supports is to provide anchorage of the supported element for internal and external design basis events, so that the supported element can perform its intended function.

September 2005 11183-1 NUREG-1801, Rev. 1 OAGI0000203_341

z III STRUCTURES AND COMPONENT SUPPORTS C

0 B3 Anchorage of Racks, Panels, Cabinets, and Enclosures for Electrical Equipment and Instrumentation m

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

Mechanism Evaluation

0 Component (D

< IILB3-1 IILB3.2-a Building Reinforced Air - indoor Reduction in Chapter XLS6, "Structures Monitoring Yes, if not concrete at concrete; uncontrolled or concrete anchor Program" within the (T-29) locations of Grout air - outdoor capacity due to scope of the expansion local concrete applicant's and grouted degradation! structures anchors; grou service-induced monitoring pads for cracking or other program support base concrete aging plates mechanisms IILB3-2 IILB3. Support Aluminum Air - indoor None None No members; uncontrolled (TP-8) welds; bolted connections; support anchorage to building structure IILB3-3 IILB3. Support Galvanized Air - indoor None None No members; steel uncontrolled (TP-11) welds; bolted connections; support anchorage to building structure (f)

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B3 Anchorage of Racks, Panels, Cabinets, and Enclosures for Electrical Equipment and Instrumentation 3

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~ Structure Aging Effect! Further

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Evaluation U1 IILB3-4 IILB3. Support Galvanized Air with borated Loss of material! Chapter XL M 10, "Boric Acid No members; steel, water leakage boric acid Corrosion" (TP-3) welds; bolted aluminum corrosion connections; support anchorage to building structure IILB3-5 IILB3. Support Stainless Air - indoor None None No members; steel uncontrolled (TP-5) welds; bolted connections; support anchorage to building structure IILB3-6 IILB3. Support Stainless Air with borated None None No members; steel water leakage (TP-4) welds; bolted connections; support anchorage to building structure z

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z III STRUCTURES AND COMPONENT SUPPORTS C

0 B3 Anchorage of Racks, Panels, Cabinets, and Enclosures for Electrical Equipment and Instrumentation m

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

Mechanism Evaluation

0 Component (D

< IILB3-7 IILB3,1-a Support Steel Air - indoor Loss of material! Chapter XLS6, "Structures Monitoring Yes, if not members; uncontrolled or general and Program" within the (T-30) welds; bolted air - outdoor pitting corrosion scope of the connections; applicant's support structures anchorage to monitoring building program structure IILB3-8 IILB3,1-b Support Steel Air with borated Loss of material! Chapter XL M 10, "Boric Acid No members; water leakage boric acid Corrosion" (T-25) welds; bolted corrosion connections; OJ support anchorage to

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84. SUPPORTS FOR EMERGENCY DIESEL GENERA TOR (EDG), HVAC SYSTEM COMPONENTS, AND OTHER MISCELLANEOUS MECHANICAL EQUIPMENT Systems, Structures, and Components This section addresses supports and anchorage for the emergency diesel generator (EDG) and HVAC system components, and other miscellaneous mechanical equipment. Applicable aging effects are identified and the aging management review is presented for each applicable combination of support component and aging effect.

System Interfaces Physical interfaces exist with the structure, system, or component being supported and with the building structural element to which the support is anchored. A primary function of supports is to provide anchorage of the supported element for internal and external design basis events, so that the supported element can perform its intended function.

September 2005 III B4-1 NUREG-1801, Rev. 1 OAGI0000203_345

z III STRUCTURES AND COMPONENT SUPPORTS C

0 B4 Supports for Emergency Diesel Generator (EDG), HVAC System Components, and Other Miscellaneous Mechanical Equipment m

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

Mechanism Evaluation

0 Component (D

< IILB4-1 IILB4.3-a Building Reinforced Air - indoor Reduction in Chapter XLS6, "Structures Monitoring Yes, if not concrete at concrete; uncontrolled or concrete anchor Program" within the (T-29) locations of Grout air - outdoor capacity due to scope of the expansion local concrete applicant's and grouted deg radation! structures anchors; service-induced monitoring grout pads for cracking or other program support base concrete aging plates mechanisms IILB4-2 IILB4. Sliding Lubrite, Air - indoor Loss of Chapter XLS6, "Structures Monitoring No support graphitic tool uncontrolled mechanical Program" (TP-1) bearings and steel function!

OJ sliding corrosion, t support surfaces distortion, dirt, overload, fatig ue due to vibratory and cyclic thermal loads IILB4-3 IILB4. Sliding Lubrite, Air - outdoor Loss of Chapter XLS6, "Structures Monitoring No support graphitic tool mechanical Program" (TP-2) bearings and steel function!

sliding corrosion, support distortion, dirt, surfaces overload, fatig ue due to vibratory and cyclic (f) thermal loads (D

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B4 Supports for Emergency Diesel Generator (EDG), HVAC System Components, and Other Miscellaneous Mechanical Equipment 3

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Evaluation U1 IILB4-4 IILB4. Support Aluminum Air - indoor None None No members; uncontrolled (TP-8) welds; bolted connections; support anchorage to building structure IILB4-5 IILB4. Support Galvanized Air - indoor None None No members; steel uncontrolled (TP-11) welds; bolted connections; OJ support t anchorage to building structure IILB4-6 IILB4. Support Galvanized Air with borated Loss of material! Chapter XL M 10, "Boric Acid No members; steel, water leakage boric acid Corrosion" (TP-3) welds; bolted aluminum corrosion connections; support anchorage to building structure z

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z III STRUCTURES AND COMPONENT SUPPORTS C

0 B4 Supports for Emergency Diesel Generator (EDG), HVAC System Components, and Other Miscellaneous Mechanical Equipment m

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

Mechanism Evaluation

0 Component (D

< III.B4-7 III.B4. Support Galvanized Air - outdoor Loss of material! Chapter XI.S6, "Structures Monitoring No members; steel, pitting and Program" (TP-6) welds; bolted aluminum, crevice corrosion connections; stainless steel support anchorage to building structure III.B4-8 III.B4. Support Stainless steel Air - indoor None None No members; uncontrolled (TP-5) welds; bolted connections; OJ support t anchorage to building structure III.B4-9 III.B4. Support Stainless steel Air with borated None None No members; water leakage (TP-4) welds; bolted connections; support anchorage to building structure (f)

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Evaluation U1 IILB4-10 IILB4.1-a Support Steel Air - indoor Loss of material! Chapter XLS6, "Structures Monitoring Yes, if not members; uncontrolled or general and Program" within the (T-30) welds; bolted air - outdoor pitting corrosion scope of the connections; applicant's support structures anchorage to monitoring building program structure IILB4-11 IILB4.1-b Support Steel Air with borated Loss of material! Chapter XL M 10, "Boric Acid No members; water leakage boric acid Corrosion" (T-25) welds; bolted corrosion connections; support anchorage to building structure IILB4-12 IILB4.2-a Vibration Non-metallic Air - indoor Reduction or Chapter XLS6, "Structures Monitoring Yes, if not isolation (eg, Rubber) uncontrolled or loss of isolation Program" within the (T-31) elements air - outdoor function! scope of the radiation applicant's hardening, structures temperature, monitoring humidity, program sustained z vibratory loading C

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OAGI0000203_349

This Page Intentionally Left Blank NUREG-1801, Rev. 1 III B4-6 Septem ber 2005 OAGI0000203_350

85. SUPPORTS FOR PLATFORMS, PIPE WHIP RESTRAINTS, JET IMPINGEMENT SHIELDS, MASONRY WALLS, AND OTHER MISCELLANEOUS STRUCTURES Systems, Structures, and Components This section addresses supports and anchorage for platforms, pipe whip restraints, jet impingement shields, masonry walls, and other miscellaneous structures. Applicable aging effects are identified and the aging management review is presented for each applicable combination of support component and aging effect.

System Interfaces Physical interfaces exist with the structure, system, or component being supported and with the building structural element to which the support is anchored. A primary function of supports is to provide anchorage of the supported element for internal and external design basis events, so that the supported element can perform its intended function.

September 2005 11185-1 NUREG-1801, Rev. 1 OAGI0000203_351

z III STRUCTURES AND COMPONENT SUPPORTS C

0 B5 Supports for Platforms, Pipe Whip Restraints, Jet Impingement Shields, Masonry Walls, and Other Miscellaneous Structures m

Gl Structure OJ Aging Effect! Further o Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation Component IIIB5-1 IIIB5.2-a Building Reinforced Air - indoor Reduction in Chapter XIS6, "Structures Monitoring Yes, if not concrete at concrete; uncontrolled or concrete anchor Program" within the (T-29) locations of Grout air - outdoor capacity due to scope of the expansion local concrete applicant's and grouted degradation! structures anchors; grou service-induced monitoring pads for cracking or other program support base concrete aging plates mechanisms IIIB5-2 IIIB5. Support Aluminum Air - indoor None None No members; uncontrolled (TP-8) welds; bolted connections; support anchorage to building structure IIIB5-3 IIIB5. Support Galvanized Air - indoor None None No members; steel uncontrolled (TP-11) welds; bolted connections; support anchorage to building structure (f)

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Mechanism Evaluation Component IIIB5-4 IIIB5. Support Galvanized Air with borated Loss of material! Chapter XI M 10, "Boric Acid No members; steel, water leakage boric acid Corrosion" (TP-3) welds; bolted aluminum corrosion connections; support anchorage to building structure IIIB5-5 IIIB5. Support Stainless Air - indoor None None No members; steel uncontrolled (TP-5) welds; bolted connections; support anchorage to building structure IIIB5-6 IIIB5. Support Stainless Air with borated None None No members; steel water leakage (TP-4) welds; bolted connections; support anchorage to z building C structure

0 m

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OAGI0000203_353

z III STRUCTURES AND COMPONENT SUPPORTS C

0 B5 Supports for Platforms, Pipe Whip Restraints, Jet Impingement Shields, Masonry Walls, and Other Miscellaneous Structures m

Gl Structure OJ Aging Effect! Further o Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation Component IILB5-7 IILB5,1-a Support Steel Air - indoor Loss of material! Chapter XLS6, "Structures Monitoring Yes, if not members; uncontrolled or general and Program" within the (T-30) welds; bolted air - outdoor pitting corrosion scope of the connections; applicant's support structures anchorage to monitoring building program structure IILB5-8 IILB5,1-b Support Steel Air with borated Loss of material! Chapter XL M 10, "Boric Acid No members; water leakage boric acid Corrosion" (T-25) welds; bolted corrosion connections; support anchorage to building structure (f)

(D Ll ro 3

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CHAPTER IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM September 2005 IV-i NUREG-1801, Rev. 1 OAGI0000203_355

This Page Intentionally Left Blank NUREG-1801, Rev. 1 IV-ii September 2005 OAGI0000203_356

MAJOR PLANT SECTIONS A1. Reactor Vessel (Boiling Water Reactor)

A2. Reactor Vessel (Pressurized Water Reactor)

B1 . Reactor Vessel Internals (Boiling Water Reactor)

B2. Reactor Vessel Internals (PWR) - Westinghouse B3. Reactor Vessel Internals (PWR) - Combustion Engineering B4. Reactor Vessel Internals (PWR) - Babcock and Wilcox C1. Reactor Coolant Pressure Boundary (Boiling Water Reactor)

C2. Reactor Coolant System and Connected Lines (Pressurized Water Reactor)

D1. Steam Generator (Recirculating)

D2. Steam Generator (Once-Through)

E. Common Miscellaneous Material/Environment Combinations September 2005 IV-iii NUREG-1801. Rev. 1 OAGI0000203_357

This Page Intentionally Left Blank NUREG-1801, Rev. 1 IV-iv September 2005 OAGI0000203_358

A1. REACTOR VESSEL (BOILING WATER REACTOR)

Systems, Structures, and Components This section addresses the boiling water reactor (BWR) pressure vessel and consists of the vessel shell and flanges, attachment welds, top and bottom heads, nozzles (including safe ends) for the reactor coolant recirculating system and connected systems (such as high and low pressure core spray, high and low pressure coolant injection, main steam, and feedwater systems), penetrations for control rod drive (CRD) stub tubes, instrumentation, standby liquid control, flux monitor, drain lines, and control rod drive mechanism housings. The support skirt and attachment welds for vessel supports are also included in the following table for the BWR vessel. 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 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 extended period of operation are included in IV.E.

System Interfaces The systems that interface with the reactor vessel include the reactor vessel internals (IV.B1),

the reactor coolant pressure boundary (IV.C1), the emergency core cooling system (V.D2), and standby liquid control system (VII.E2).

September 2005 IVA1-1 NUREG-1801, Rev. 1 OAGI0000203_359

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

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< IV.A1-1 IV.A1.4-a Chapter XI.M7, "BWR Stress Corrosion No Nozzle safe ends Stainless Reactor coolant Cracking! stress (and associated steel; corrosion cracking Cracking," and (R-68) welds) nickel alloy and intergranular stress corrosion Chapter XI.M2, "Water Chemistry," for High pressure cracking BWRwater core spray Low pressure core spray Control rod drive return line Recirculating water Low pressure coolant injection or RHR injection mode IV.A1-2 IV.A1.3-c Nozzles Steel (with Reactor coolant Cracking! cyclic Chapter XI.M6, "BWR Control Rod Drive No or without loading Return Line Nozzle" (R-66) Control rod drive stainless return line steel cladding)

IV.A1-3 IV.A1.3-b Nozzles Steel (with Reactor coolant Cracking! cyclic Chapter XI. M5, "BWR Feedwater No or without loading Nozzle" (R-65) Feedwater stainless steel cladding)

OAGI0000203_360

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D A1 Reactor Vessel (BWR) 3 0-

~

Structure and/or Aging Effect! Further

'"oo Item Link Component Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 IV.A1-4 IV.A1.3-e Nozzles Steel Reactor coolant Loss of fracture Neutron irradiation embritllement is a Yes, and neutron flux toughnessl time-limited aging analysis (TLAA) to TLAA (R-67) Low pressure neutron irradiation be evaluated for the period of extended coolant injection embritllement operation for all ferritic materials that or RHR injection have a neutron fluence greater than mode 1E17 n/cm 2 (E >1 MeV) at the end of the license renewal term. In accordance with approved BWRVIP-74, the TLAA is to evaluate the impact of neutron embritllement on: (a) the adjusted reference temperature values used for calculation of the plant's pressure-temperature limits, (b) the need for inservice inspection of circumferential welds, and (c) the Charpy upper shelf energy or the equivalent margins analyses performed in accordance with 10 CFR Part 50, Appendix G The applicant may choose to demonstrate that the materials of the nozzles are not controlling for the TLAA evaluations. See the standard Review Plan, Section 4.2 "Reactor Vessel Neutron Embritllement" for acceptable methods for meeting the requirements z of 10 CFR 5421(c)

C

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

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0 Component Mechanism Evaluation (D

< IV.A1-5 IV.A1.5-a Chapter XI. MS, "BWR Penetrations,"

Penetrations Stainless Reactor coolant Cracking! stress No steel; corrosion and (R-69) Control rod drive nickel alloy cracking, stub tubes intergranular Chapter XI.M2, "Water Chemistry," for Instrumentation stress corrosion BWRwater Jet pump cracking, cyclic instrument loading Standby liquid control Flux monitor Drain line IV.A1-6 IV.A17-a Pressure vessel Steel Air - indoor Cumulative fatigue Fatigue is a time-limited aging analysis Yes, support skirt and uncontrolled damage! fatigue (TLAA) to be evaluated for the period of TLAA (R-70) attachment welds extended operation. See the Standard Review Plan, Section 4.3 "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR 5421(c)(1)

OAGI0000203_362

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D A1 Reactor Vessel (BWR) 3 0-

~

Structure and!or Aging Effect! Further

'"oo Item Link Component Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 IV.A1-7 IV.A1.3-a Reactor vessel Steel; Reactor coolant Cumulative fatigue Fatigue is a time-limited aging analysis Yes, IV.A1.3-d components: stainless damage! fatigue (TLAA) to be performed for the period of TLAA (R-04) IV.A1.5-b steel; steel extended operation, and, for Class 1 IV.A12-a Flanges; with nickel- components, environmental effects on IV.A1.2-b Nozzles; alloy or fatigue are to be addressed. See the IV.A1.2-b Penetrations; stainless Standard Review Plan, Section 4.3 IV.A1.3-d Safe ends; steel "Metal Fatigue," for acceptable IV.A1.4-b Thermal sleeves; cladding; methods for meeting the requirements IV.A1.1-b Vessel shells, nickel- of 10 CFR 5421(c)(1)

IV.A1.5-b heads and welds alloy IV.A12-a IV.A1.5-b IV.A1.5-b IV.A1.5-b IV.A1.6-a IV.A1.5-b IV.A1.2-b IV.A1.2-b IV.A1-8 IV.A1. Reactor Vessel: Stainless Reactor Coolant Loss of material! Chapter XI.M2, "Water Chemistry," for Yes, detection Flanges, nozzles; steel; steel pitting and crevice BWRwater of aging (RP-25) penetrations; safe with nickel- corrosion effects is to be ends; vessel alloy or The AMP is to be augmented by evaluated shells, heads and stainless verifying the effectiveness of water welds steel chemistry control. See Chapter XI. M32, z cladding; "One-Time Inspection," for an C

0 m nickel- acceptable verification program.

~ alloy OJ

~

OAGI0000203_363

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

0 A1 Reactor Vessel (BWR) m

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0 Component Mechanism Evaluation (D

< IV.A1-9 IV.A1.1-c High-Top head Air with reactor Cracking! stress Chapter XI. M3, "Reactor Head Closure No enclosure strength coolant leakage corrosion cracking Studs" (R-60) low alloy and intergranular Closure studs steel stress corrosion and nuts cracking Maximum tensile strength <

1172 MPa

<<170 ksi)

IV.A1-10 IV.A1.1-d Top head Stainless Air with reactor Cracking! stress A plant-specific aging management Yes, plant-enclosure steel; coolant leakage corrosion cracking program is to be evaluated because specific (R-61) nickel alloy (Internal) and intergranular existing programs may not be capable Vessel flange stress corrosion of mitigating or detecting crack leak detection or cracking initiation and growth due to SCC in the line vessel flange leak detection line.

Reactor Coolant IV.A1-11 IV.A1.1-a Top head Steel Reactor coolant Loss of material! Chapter XI.M2, "Water Chemistry," for Yes, detection enclosure (without general, pitting, BWRwater of aging (R-59) cladding) and crevice effects is to be corrosion The AMP is to be augmented by evaluated Top head verifying the effectiveness of water Nozzles (vent, chemistry control. See Chapter XI. M32, top head spray "One-Time Inspection," for an or RClC, and acceptable verification program.

spare)

IV.A1-12 IV.A12-e Vessel shell Stainless Reactor coolant Cracking! stress Chapter XI. M4, "BWR VessellD No steel; corrosion cracking Attachment Welds," and (R-64) Attachment nickel alloy and intergranular welds stress corrosion Chapter XI.M2, "Water Chemistry," for cracking BWRwater OAGI0000203_364

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D A1 Reactor Vessel (BWR) 3 0-

~

Structure and/or Aging Effect! Further

'"oo Item Link Component Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 IV.A1-13 IV.A1.2-c Vessel shell Steel (with Reactor coolant Loss of fracture Neutron irradiation embritllement is a Yes, or without and neutron flux toughnessl time dependent aging mechanism to be TLAA (R-62) Intermediate stainless neutron irradiation evaluated for the period of extended beltline shell steel embritllement operation for all ferritic materials that Beltline welds cladding) have a neutron fluence greater than 1E17 n/cm 2 (E >1 MeV) at the end of the license renewal term. Aspects of this evaluation may involve a TLAA In accordance with approved BWRVIP-74, the TLAA is to evaluate the impact of neutron embritllement on: (a) the adjusted reference temperature values used for calculation of the plant's pressure-temperature limits, (b) the need for inservice inspection of circumferential welds, and (c) the 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 embritllemenl. 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 z than the value specified by the staff in C

0 its March 7, 2000 letler (ADAMS m

ML031430372) See the Standard

~ Review Plan, Section 4.2 "Reactor OJ

~ Vessel Neutron Embritllement" for acceptable methods for meeting the requirements of 10 CFR 5421(c)

OAGI0000203_365

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

0 A1 Reactor Vessel (BWR) m

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0 Component Mechanism Evaluation (D

< IV.A1-14 IV.A1.2-d Chapter XI. M31, "Reactor Vessel Vessel shell Steel (with Reactor coolant Loss of fracture No or without and neutron flux toughness! Surveillance" (R-63) Intermediate stainless neutron irradiation beltline shell steel embritllement Beltline welds cladding)

OAGI0000203_366

A2. REACTOR VESSEL (PRESSURIZED WA TER 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 extended period of operation are included in IV.E.

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

September 2005 IV.A2-1 NUREG-1801, Rev. 1 OAGI0000203_367

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

0 A2 Reactor Vessel (PWR) m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< IV.A2-1 Botlom-IV.A2. Stainless Reactor coolant Cracking! stress A plant-specific aging management Yes, plant-mounted guide steel corrosion cracking program is to be evaluated. specific (RP-13) tube IV.A2-2 IV.A2.1-c Closure head High-strength Air with reactor Cracking! stress Chapter XI. M3, "Reactor Head Closure No low alloy coolant leakage corrosion cracking Studs" (R-71) Stud assembly steel Maximum tensile strength <

1172 MPa

<<170 ksi)

IV.A2-3 IV.A2.1-d Closure head High-strength Air with reactor Loss of material! Chapter XI. M3, "Reactor Head Closure No low alloy coolant leakage wear Studs" (R-72) Stud assembly steel Maximum tensile strength <

1172 MPa

<<170 ksi)

IV.A2-4 IV.A2.1-e Closure head Low alloy Air with reactor Cumulative fatigue Fatigue is a time-limited aging analysis Yes, steel coolant leakage damage! fatigue (TLAA) to be evaluated for the period of TLAA (R-73) Stud assembly extended operation. See the Standard Review Plan, Section 4.3 "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR 5421(c)(1)

OAGI0000203_368

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D A2 Reactor Vessel (PWR) 3 0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.A2-5 IV.A2.1-f Closure head Stainless Air with reactor Cracking! stress A plant-specific aging management Yes, plant-steel coolant leakage corrosion cracking program is to be evaluated because specific (R-74) Vessel flange (Internal) existing programs may not be capable leak detection of mitigating or detecting crack line or initiation and growth due to SCC in the vessel flange leak detection line.

Reactor Coolant IV.A2-6 IV.A2.2-e Control rod drive Stainless Air with reactor Cracking! stress Chapter XI. M 18, "Bolting Integrity" No head steel coolant leakage corrosion cracking (R-78) penetration Flange bolting IV.A2-7 IV.A22-f Control rod drive Stainless Air with reactor Loss of material! Chapter XI. M 18, "Bolting Integrity" No head steel coolant leakage wear (R-79) penetration Flange bolting IV.A2-8 IVA22-g Control rod drive Stainless Air with reactor Loss of preload! Chapter XI. M 18, "Bolting Integrity" No head steel coolant leakage thermal effects, (R-80) penetration gasket creep, and self-loosening Flange bolting z

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OAGI0000203_369

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

0 A2 Reactor Vessel (PWR) m

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~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< IV.A2-9 IV.A2.2-a Control rod drive Nickel alloy Reactor coolant Cracking! primary Chapter XI. M1, "ASM E Section XI No head water stress Inservice Inspection, Subsections IWB, (R-75) penetration corrosion cracking IWC, and IWO," for Class 1 components and Chapter XI.M2, 'Water Nozzle and Chemistry," for PWR primary water and welds Chapter XI.M11-A, "Nickel-Alloy Penetration Nozzles Welded to the Upper Reactor Vessel Closure Heads of Pressurized Water Reactors (PWRs Only)"

IV.A2-10 IV.A2.2-d Control rod drive Cast Reactor coolant Loss of fracture Chapter XI.M12, "Thermal Aging No head austenitic >250"C (>482"F) toughness! Embriltlement of Cast Austenitic (R-77) penetration stainless thermal aging Stainless Steel (CASS)"

steel embritllement Pressure housing IV.A2-11 IV.A2.2-b Control rod drive Stainless Reactor coolant Cracking! stress Chapter XI. M2, "Water Chemistry," and No, but head steel; nickel corrosion licensee (R-76) penetration alloy cracking, primary Chapter XI. M1, "ASM E Section XI commitment water stress Inservice Inspection, Subsections IWB, needs to be Pressure corrosion cracking IWC, and IWO" and, confirmed housing For nickel alloy, comply with applicable NRC Orders and provide a commitment in the FSAR supplement to submit a plant-specific AMP to implement applicable (1) Bulletins and Generic Leiters and (2) staff-accepted industry guidelines.

OAGI0000203_370

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D A2 Reactor Vessel (PWR) 3 0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.A2-12 IV.A2.6-a Core support Nickel alloy Reactor coolant Cracking! primary Chapter XI. M2, "Water Chemistry," and No, but pads! core water stress licensee (R-88) guide lugs corrosion cracking Chapter XI. M1, "ASM E Section XI commitment Inservice Inspection, Subsections IWB, needs to be IWC, and IWD" and, confirmed Comply with applicable NRC Orders and provide a commitment in the FSAR supplement to submit a plant-specific AMP to implement applicable (1)

Bulletins and Generic Letters and (2) staff-accepted industry guidelines.

IV.A2-13 IV.A2.8-b External Steel Air with borated Loss of material! Chapter XI.M10, "Boric Acid Corrosion" No IV.A2.5-e surfaces water leakage boric acid (R-17) IV.A2.1-a corrosion IV.A2-14 IV.A2. Flanges; Stainless Reactor Coolant Loss of material! Chapter XI.M2, "Water Chemistry," for No nozzles; steel; steel pitting and crevice PWR primary water (RP-28) penetrations; with nickel- corrosion pressure alloy or housings; safe stainless ends; vessel steel shells, heads cladding; and welds nickel-alloy IV.A2-15 IV.A2.4-b Nozzle safe Stainless Reactor coolant Cracking! stress Chapter XI. M1, "ASM E Section XI No z ends and welds: steel; nickel corrosion Inservice Inspection, Subsections IWB, C

0 (R-83) alloy welds cracking, primary IWC, and IWD," for Class 1 m

~

OJ Inlet Outlet and!or buttering water stress corrosion cracking components and

~ Safety Chapter XI.M2, "Water Chemistry," for injection PWR primary water OAGI0000203_371

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

0 A2 Reactor Vessel (PWR) m

~

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~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< IV.A2-16 IV.A2.3-a Nozzles Steel with Reactor coolant Loss of fracture Neutron irradiation embritllement is a Yes, stainless and neutron flux toughness! time-limited aging analysis (TLAA) to TLAA (R-81) Inlet steel neutron irradiation be evaluated for the period of extended Outlet cladding embritllement operation for all ferritic materials that Safety have a neutron fluence greater than injection 1E17 n!cm 2 (E >1 MeV) at the end of the license renewal term. The TLAA is to evaluate the impact of neutron embritllement 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, and (c) the Charpy upper shelf energy 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.A2-17 IV.A2.3-b Nozzles Steel with Reactor coolant Loss of fracture Chapter XI. M31, "Reactor Vessel No stainless and neutron flux toughness! Surveillance" (R-82) Inlet steel neutron irradiation Outlet cladding embritllement Safety injection OAGI0000203_372

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D A2 Reactor Vessel (PWR) 3 0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.A2-18 IV.A2.7-b Penetrations Nickel alloy Reactor coolant Cracking! primary Chapter XI. M1, "ASM E Section XI No water stress Inservice Inspection, Subsections IWB, (R-90) Head vent pipe corrosion cracking IWC, and IWO," for Class 1 (top head) components and Chapter XI.M2, 'Water Instrument Chemistry," for PWR primary water and tubes (top Chapter XI.M11-A, "Nickel-Alloy head) Penetration Nozzles Welded to the Upper Reactor Vessel Closure Heads of Pressurized Water Reactors (PWRs Only)"

IV.A2-19 IV.A2.7-a Penetrations Nickel alloy Reactor coolant Cracking! primary Chapter XI.M1, "ASME Section XI No, but water stress Inservice Inspection, Subsections IWB, licensee (R-89) Instrument corrosion cracking IWC, and IWD" for Class 1 commitment tubes (bottom components, and needs to be head) confirmed Chapter XI.M2, "Water Chemistry," for PWR primary water and Comply with applicable NRC Orders and provide a commitment in the FSAR supplement to implement applicable (1)

Bulletins and Generic Letters and (2) staff-accepted industry guidelines.

IV.A2-20 IV.A2.8-a Pressure vessel Steel Air - indoor Cumulative fatigue Fatigue is a time-limited aging analysis Yes, z support skirt uncontrolled damage! fatigue (TLAA) to be evaluated for the period of TLAA C

0 (R-70) and attachment extended operation. See the Standard m

welds Review Plan, Section 4.3 "Metal

~

OJ Fatigue," for acceptable methods for

~ meeting the requirements of 10 CFR 5421(c)(1)

OAGI0000203_373

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

0 A2 Reactor Vessel (PWR) m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< IV.A2-21 IV.A2.3-c Reactor vessel Steel; Reactor coolant Cumulative fatigue Fatigue is a time-limited aging analysis Yes, IV.A2.5-d components: stainless damage! fatigue (TLAA) to be performed for the period of TLAA (R-219) IV.A2.4-a steel; steel extended operation, and, for Class 1 IV.A2.4-a Flanges; with nickel- components, environmental effects on IV.A2.4-a Nozzles; alloy or fatigue are to be addressed. See the IV.A2.3-c Penetrations; stainless Standard Review Plan, Section 4.3 IV.A2.5-d Pressure steel "Metal Fatigue," for acceptable IV.A2.2-c housings; cladding; methods for meeting the requirements IV.A2.1-b Safe ends; nickel-alloy of 10 CFR 5421(c)(1)

IV.A2.5-d Thermal IV.A2.5-d sleeves; IV.A2.3-c Vessel shells, IV.A2.2-c heads and welds OAGI0000203_374

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D A2 Reactor Vessel (PWR) 3 0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.A2-22 IV.A2.5-b Vessel shell SA508-C12 Reactor coolant Crack growth! Growth of intergranular separations Yes, forgings clad cyclic loading (underclad cracks) in low-alloy steel TLAA (R-85) Upper shell with forging heat affected zone under Intermediate stainless austenitic stainless steel cladding is a and lower shell steel using a time-limited aging analysis (TLAA) to (including high-heat- be evaluated for the period of extended beltline welds) input welding operation for all the SA 508-CI 2 process forgings where the cladding was 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 and criterion in the ASME Section XI 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 5421(c) z C

0 m

~

OJ

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OAGI0000203_375

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

0 A2 Reactor Vessel (PWR) m

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

Mechanism Evaluation

0 (D

Component

< IV.A2-23 IV.A2.5-a Vessel shell Steel with Reactor coolant Loss of fracture Neutron irradiation embritllement is a Yes, stainless and neutron flux toughnessl time-limited aging analysis (TLAA) to TLAA (R-84) Upper shell steel neutron irradiation be evaluated for the period of extended Intermediate cladding embritllement operation for all ferritic materials that and lower shell have a neutron fluence greater than (including 1E17 n/cm 2 (E >1 MeV) at the end of beltline welds) the license renewal term. The TLAA is to evaluate the impact of neutron embritllement 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, and (c) the Charpy upper shelf energy 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 Embritllement" for acceptable methods for meeting the requirements of 10 CFR 5421(c)

OAGI0000203_376

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D A2 Reactor Vessel (PWR) 3 0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.A2-24 IV.A2.5-c Vessel shell Steel with Reactor coolant Loss of fracture Chapter XI. M31, "Reactor Vessel No stainless and neutron flux toughness! Surveillance" (R-86) Upper shell steel neutron irradiation Intermediate cladding embritllement and lower shell (including beltline welds)

IV.A2-25 IV.A2.5-f Vessel shell Steel Reactor coolant Loss of material! Chapter XI. M1, "ASM E Section XI No wear Inservice Inspection, Subsections IWB, (R-87) Vessel flange IWC, and IWO," for Class 1 components z

C

0 m

~

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This Page Intentionally Left Blank NUREG-1801, Rev. 1 IV A2-12 September 2005 OAGI0000203_378

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 (CRD), and instrument housings, such as the intermediate range monitor (IRM) 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 extended period of operation are included in IV.E.

System Interfaces The systems that interface with the reactor vessel internals include the reactor pressure vessel (IV.A1) and the reactor coolant pressure boundary (IV.C1).

September 2005 IV B1-1 NUREG-1801, Rev. 1 OAGI0000203_379

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

0 B1 Reactor Vessel Internals (BWR) m

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OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation Component IV.B1-1 IV.B1.1-a Core shroud Stainless Reactor coolant Cracking! stress Chapter XI.M9, "BWR Vessel Internals," No (including steel corrosion cracking, for core shroud and (R-92) repairs) and intergranular core plate stress corrosion Chapter XI.M2, "Water Chemistry" for cracking, BWRwater Core shroud irradiation-assisted (upper, stress corrosion central, lower) cracking IV.B1-2 IV.B1.1-f Core shroud Nickel alloy Reactor coolant Cracking! stress Chapter XI.M9, "BWR Vessel Internals," No (including corrosion cracking, for shroud support and (R-96) repairs) and intergranular core plate stress corrosion Chapter XI.M2, "Water Chemistry," for

< cracking, BWRwater OJ Shroud irradiation-assisted r0 support stress corrosion structure cracking (shroud support cylinder, shroud suppon plate, shroud support legs)

IV.B1-3 IVB11-g Core shroud Stainless Reactor coolant Cracking! stress Chapter XI.M9, "BWR Vessel Internals," No and core plate steel corrosion cracking, for the LPCI coupling and (R-97) intergranular LPCI coupling stress corrosion Chapter XI. M2, "Water Chemistry," for cracking, BWRwater irradiation-assisted stress corrosion cracking OAGI0000203_380

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(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B1 Reactor Vessel Internals (BWR) 3 0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B1-4 IV.B1.1-e Core shroud Nickel alloy Reactor coolant Cracking! stress Chapter XI. M1, "ASM E Section XI No and core plate corrosion cracking, Inservice Inspection, Subsections IWB, (R-95) intergranular IWC, and IWO," for Class 1 Access hole stress corrosion components and cover cracking, (mechanical irradiation-assisted Chapter XI.M2, "Water Chemistry," for covers) stress corrosion BWRwater cracking IV.B1-5 IV.B1.1-d Core shroud Nickel alloy Reactor coolant Cracking! stress Chapter XI. M1, "ASM E Section XI No and core plate corrosion cracking, Inservice Inspection, Subsections IWB, (R-94) intergranular IWC, and IWO," for Class 1 Access hole stress corrosion components and Chapter XI.M2, 'Water

< cover cracking, Chemistry," for BWR water OJ (welded irradiation-assisted W

covers) stress corrosion Because cracking initiated in crevice cracking regions is not amenable to visual inspection, for BWRs with a crevice in the access hole covers, an augmented inspection is to include ultrasonic testing (UT) or other demonstrated acceptable inspection of the access hole cover welds.

IV.B1-6 IV.B1.1-b Core shroud Stainless Reactor coolant Cracking! stress Chapter XI.M9, "BWR Vessel Internals," No and core plate steel corrosion cracking, for core plate and z (R-93) intergranular C Core plate stress corrosion Chapter XI.M2, "Water Chemistry" for

0 rn Core plate cracking, BWRwater

~ bolts (used in irradiation-assisted OJ early BWRs) stress corrosion

~

cracking OAGI0000203_381

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

0 B1 Reactor Vessel Internals (BWR) m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation Component IV.B1-7 IV.B1.3-a Core spray line, Stainless Reactor coolant Cracking! stress Chapter XI.M9, "BWR Vessel Internals," No and spargers steel corrosion cracking, for core spray internals and (R-99) intergranular Core spray stress corrosion Chapter XI.M2, "Water Chemistry," for lines (headers) cracking, BWRwater Spray rings irradiation-assisted Spray nozzles stress corrosion Thermal cracking sleeves IV.B1-8 IV.B1.5-c Fuel supports Stainless Reactor coolant Cracking! stress Chapter XI.M9, "BWR Vessel Internals," No

< and control rod steel corrosion cracking for lower plenum and OJ (R-104) drive and intergranular

.h assemblies stress corrosion Chapter XI.M2, "Water Chemistry," for cracking BWRwater Control rod drive housing IV.B1-9 IV.B1.5-a Fuel supports Cast Reactor coolant Loss of fracture Chapter XI.M13, "Thermal Aging and No and control rod austenitic >250°C (>482°F) toughness! Neutron Irradiation Embritllement of (R-103) drive stainless and neutron flux thermal aging and Cast Austenitic Stainless Steel assemblies steel neutron irradiation (CASS)"

embritllement Orificed fuel support OAGI0000203_382

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(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B1 Reactor Vessel Internals (BWR) 3 0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B1-10 IV.B1.6-a Instrumentation Stainless Reactor coolant Cracking! stress Chapter XI.M9, "BWR Vessel Internals," No steel corrosion cracking, for lower plenum and (R-105) Intermediate intergranular range monitor stress corrosion Chapter XI.M2, "Water Chemistry," for (IRM) dry cracking, BWRwater tubes irradiation-assisted Source range stress corrosion monitor (SRM) cracking dry tubes Incore neutron flux monitor guide tubes OJ U,

IV.B1-11 IV.B1.4-c Jet pump Cast Reactor coolant Loss of fracture Chapter XI.M13, "Thermal Aging and No assemblies austenitic >250°C (>482°F) toughness! Neutron Irradiation Embritllement of (R-101) stainless and neutron flux thermal aging and Cast Austenitic Stainless Steel Castings steel neutron irradiation (CASS)"

embritllement IV.B1-12 IV.B1.4-d Jet pump Stainless Reactor coolant Cracking! cyclic A plant-specific aging management Yes, plant-assemblies steel loading program is to be evaluated. specific (R-102)

Jet pump z sensing line C

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OJ

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OAGI0000203_383

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

0 B1 Reactor Vessel Internals (BWR) m

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OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation Component IV.B1-13 IV.B1.4-a Jet pump Nickel alloy; Reactor coolant Cracking! stress Chapter XI.M9, "BWR Vessel Internals," No assemblies stainless corrosion cracking, for jet pump assembly and (R-100) steel intergranular Thermal sleeve stress corrosion Chapter XI.M2, "Water Chemistry," for Inlet header cracking, BWRwater Riser brace irradiation-assisted arm stress corrosion Holddown cracking beams Inlet elbow Mixing assembly

< Diffuser OJ Castings

'" IV.B1-14 IV.B1.6-b Reactor vessel Stainless Reactor coolant Cumulative fatigue Fatigue is a time-limited aging analysis Yes, IV.B1.3-b internals steel; nickel damage! fatigue (TLAA) to be evaluated for the period of TLAA (R-53) IV.B1.4-b components alloy extended operation. See the Standard IV.B1.2-b Review Plan, Section 4.3 "Metal IV.B1.5-b Fatigue," for acceptable methods for IV.B1.1-c meeting the requirements of 10 CFR 5421(c)(1)

IV.B1-15 IVB1. Reactor vessel Stainless Reactor coolant Loss of material! Chapter XI. M1, "ASM E Section XI No internals steel; nickel pitting and crevice Inservice Inspection, Subsections IWB, (RP-26) components alloy corrosion IWC, and IWO," for Class 1 components and Chapter XI.M2, 'Water Chemistry" for BWR water IV.B1-16 IVB1. Steam Dryers Stainless Reactor coolant Cracking! flow- A plant-specific aging management Yes, plant-steel induced vibration program is to be evaluated. specific RP-18)

OAGI0000203_384

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~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B1-17 IV.B1.2-a Top guide Stainless Reactor coolant Cracking! stress Chapter XI.M9, "BWR Vessel Internals," No steel corrosion cracking, for top guide and Chapter XI. M2, "Water (R-98) intergranular Chemistry," for BWR water.

stress corrosion Additionally, for top guides with neutron cracking, fluence exceeding the IASCC threshold irradiation-assisted (5E20, E>IMeV) prior to the period of stress corrosion extended operation, inspect five percent cracking (5%) of the top guide locations using enhanced visual inspection technique, EVT-1 within six years after entering the period of extended operation. An additional 5% of the top guide locations

< will be inspected within twelve years OJ after entering the period of extended

~

operation.

Alternatively, if the neutron fluence for the limiting top guide location is projected to exceed the threshold for IASCC after entering the period of extended operation, inspect 5% of the top guide locations (EVT-1) within six years after the date projected for exceeding the threshold. An additional 5% of the top guide locations will be z

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0 rn date projected for exceeding the

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OJ threshold.

~ The top guide inspection locations are those that have high neutron fluences exceeding the IASCC threshold.

OAGI0000203_385

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Mechanism Evaluation Component The extent and frequency of examination of the top guide is similar to the examination of the control rod drive housing guide tube in BWRVIP-47.

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B2. 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 rod control cluster assemblies (RCCA) 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 extended period of operation are included in IV.E.

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

September 2005 IV B2-1 NUREG-1801, Rev. 1 OAGI0000203_387

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

0 B2 Reactor Vessel Internals (PWR) - Westinghouse m

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Mechanism Evaluation

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Component

< IV.B2-1 IV.B2.4-b Bafflelformer Stainless Reactor coolant Changes in No further aging management review is No, but assembly steel dimensions! void necessary if the applicant provides a licensee (R-124) swelling commitment in the FSAR supplement commitment Baffle and to (1) participate in the industry needs to be former plates programs for investigating and confirmed managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended

< operation, submit an inspection plan for OJ reactor internals to the NRC for review

'"r0 and approval.

OAGI0000203_388

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~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B2-2 IV.B2.4-a Bafflelformer Stainless Reactor coolant Cracking! stress Chapter XI. M2, Water Chemistry" for No, but assembly steel corrosion cracking, PWR primary water licensee (R-123) irradiation-assisted commitment Baffle and stress corrosion No further aging management review is needs to be former plates cracking necessary if the applicant provides a confirmed commitment in the FSAR supplement to (1) participate in the industry programs for investigating and managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and

< (3) upon completion of these programs, OJ but not less than 24 months before

'"W entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

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Mechanism Evaluation

0 (D

Component

< IV.B2-3 IV.B2.4-e Bafflelformer Stainless Reactor coolant Loss of fracture No further aging management review is No, but assembly steel and neutron flux toughness! neutron necessary if the applicant provides a licensee (R-127) irradiation commitment in the FSAR supplement commitment Baffle and embrittlement, void to (1) participate in the industry needs to be former plates swelling programs for investigating and confirmed managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended

< operation, submit an inspection plan for OJ reactor internals to the NRC for review

.h and approval.

IV.B2-4 IV.B2.4-d Bafflelformer Stainless Reactor coolant Changes in No further aging management review is No, but assembly steel dimensions! void necessary if the applicant provides a licensee (R-126) swelling commitment in the FSAR supplement commitment Bafflelformer to (1) participate in the industry needs to be bolts programs for investigating and confirmed managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

OAGI0000203_390

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~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B2-5 IV.B2.4-h Bafflelformer Stainless Reactor coolant Loss of preload! No further aging management review is No, but assembly steel stress relaxation necessary if the applicant provides a licensee (R-129) commitment in the FSAR supplement commitment Bafflelformer to (1) participate in the industry needs to be bolts programs for investigating and confirmed managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended

< operation, submit an inspection plan for OJ reactor internals to the NRC for review

'"U, and approval.

IV.B2-6 IV.B2.4-f Bafflelformer Stainless Reactor coolant Loss of fracture No further aging management review is No, but assembly steel and neutron flux toughness! neutron necessary if the applicant provides a licensee (R-128) irradiation commitment in the FSAR supplement commitment Bafflelformer embrittlement, void to (1) participate in the industry needs to be bolts and swelling programs for investigating and confirmed screws managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and z

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OAGI0000203_391

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Mechanism Evaluation

0 (D

Component

< IV.B2-7 IV.B2.3-b Core barrel Stainless Reactor coolant Changes in No further aging management review is No, but steel dimensions! void necessary if the applicant provides a licensee (R-121) Core barrel swelling commitment in the FSAR supplement commitment (CB) to (1) participate in the industry needs to be CB flange programs for investigating and confirmed (upper) managing aging effects on reactor CB outlet internals; (2) evaluate and implement nozzles the results of the industry programs as Thermal shield applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended

< operation, submit an inspection plan for OJ reactor internals to the NRC for review

'" and approval.

OAGI0000203_392

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0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B2-8 IV.B2.3-a Core barrel Stainless Reactor coolant Cracking! stress Chapter XI. M2, Water Chemistry" for No, but steel corrosion cracking, PWR primary water licensee (R-120) Core barrel irradiation-assisted commitment (CB) stress corrosion No further aging management review is needs to be CB flange cracking necessary if the applicant provides a confirmed (upper) commitment in the FSAR supplement CB outlet to (1) participate in the industry nozzles programs for investigating and Thermal shield managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and

< (3) upon completion of these programs, OJ but not less than 24 months before

'"~ entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

z C

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Mechanism Evaluation

0 (D

Component

< IV.B2-9 IV.B2.3-c Core barrel Stainless Reactor coolant Loss of fracture No further aging management review is No, but steel and neutron flux toughness! neutron necessary if the applicant provides a licensee (R-122) Core barrel irradiation commitment in the FSAR supplement commitment (CB) embrittlement, void to (1) participate in the industry needs to be CB flange swelling programs for investigating and confirmed (upper) managing aging effects on reactor CB outlet internals; (2) evaluate and implement nozzles the results of the industry programs as Thermal shield applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended

< operation, submit an inspection plan for OJ reactor internals to the NRC for review

'"in and approval.

OAGI0000203_394

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(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B2 Reactor Vessel Internals (PWR) - Westinghouse 3

0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B2-10 IV.B2.4-c Core barrel Stainless Reactor coolant Cracking! stress Chapter XI.M2, "Water Chemistry," for No, but assembly steel corrosion cracking, PWR primary water licensee (R-125) irradiation-assisted commitment Bafflelformer stress corrosion No further aging management review is needs to be assembly cracking necessary if the applicant provides a confirmed Bafflelformer commitment in the FSAR supplement bolts and to (1) participate in the industry screws programs for investigating and managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and

< (3) upon completion of these programs, OJ but not less than 24 months before

'"cD entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

z C

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

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Mechanism Evaluation

0 (D

Component

< IV.B2-11 IV.B2.6-b Instrumentation Stainless Reactor coolant Changes in No further aging management review is No, but support steel dimensions! void necessary if the applicant provides a licensee (R-144) structures swelling commitment in the FSAR supplement commitment to (1) participate in the industry needs to be Flux thimble programs for investigating and confirmed guide tubes managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended OJ operation, submit an inspection plan for

~ reactor internals to the NRC for review o and approval.

OAGI0000203_396

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(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B2 Reactor Vessel Internals (PWR) - Westinghouse 3

0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B2-12 IV.B2.6-a Instrumentation Stainless Reactor coolant Cracking! stress Chapter XI.M2, Water Chemistry" for No, but support steel corrosion cracking, PWR primary water licensee (R-143) structures irradiation-assisted commitment stress corrosion No further aging management review is needs to be Flux thimble cracking necessary if the applicant provides a confirmed guide tubes commitment in the FSAR supplement to (1) participate in the industry programs for investigating and managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and OJ (3) upon completion of these programs,

~ but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

IV.B2-13 IV.B2.6-c Instrumentation Stainless Reactor coolant Loss of material! Chapter XI. M37, "Flux Thimble Tube No support steel with or wear Inspection" (R-145) structures without chrome Flux thimble plating tubes z

C

0 rn

~

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

0 B2 Reactor Vessel Internals (PWR) - Westinghouse m

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~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< IV.B2-14 IV.B2.5-i Lower internal Stainless Reactor coolant Loss of preload! No further aging management review is No, but assembly steel; nickel stress relaxation necessary if the applicant provides a licensee (R-137) alloy commitment in the FSAR supplement commitment Clevis insert to (1) participate in the industry needs to be bolts programs for investigating and confirmed managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

IV.B2-15 IV.B2.5-f Lower internal Stainless Reactor coolant Changes in No further aging management review is No, but assembly steel; nickel dimensions! void necessary if the applicant provides a licensee (R-134) alloy swelling commitment in the FSAR supplement commitment Fuel alignment to (1) participate in the industry needs to be pins programs for investigating and confirmed Lower support managing aging effects on reactor plate column internals; (2) evaluate and implement bolts the results of the industry programs as Clevis insert applicable to the reactor internals; and bolts (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

OAGI0000203_398

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B2 Reactor Vessel Internals (PWR) - Westinghouse 3

0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B2-16 IV.B2.5-e Lower internal Stainless Reactor coolant Cracking! stress Chapter XI. M2, Water Chemistry" for No, but assembly steel; nickel corrosion cracking, PWR primary water licensee (R-133) alloy primary water commitment Fuel alignment stress corrosion No further aging management review is needs to be pins cracking, necessary if the applicant provides a confirmed Lower support irradiation-assisted commitment in the FSAR supplement plate column stress corrosion to (1) participate in the industry bolts cracking programs for investigating and Clevis insert managing aging effects on reactor bolts internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and OJ (3) upon completion of these programs,

~ but not less than 24 months before w entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

z C

0 rn

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

0 B2 Reactor Vessel Internals (PWR) - Westinghouse m

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Mechanism Evaluation

0 (D

Component

< IV.B2-17 IVB25-g Lower internal Stainless Reactor coolant Loss of fracture No further aging management review is No, but assembly steel; nickel and neutron flux toughness! neutron necessary if the applicant provides a licensee (R-135) alloy irradiation commitment in the FSAR supplement commitment Fuel alignment embrittlement, void to (1) participate in the industry needs to be pins swelling programs for investigating and confirmed Lower support managing aging effects on reactor plate column internals; (2) evaluate and implement bolts the results of the industry programs as Clevis insert applicable to the reactor internals; and bolts (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

IV.B2-18 IV.B2.5-c Lower internal Stainless Reactor coolant Loss of fracture No further aging management review is No, but assembly steel and neutron flux toughness! neutron necessary if the applicant provides a licensee (R-132) irradiation commitment in the FSAR supplement commitment Lower core embrittlement, void to (1) participate in the industry needs to be plate swelling programs for investigating and confirmed managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

OAG10000203_400

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B2 Reactor Vessel Internals (PWR) - Westinghouse 3

0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B2-19 IV.B2.5-b Lower internal Stainless Reactor coolant Changes in No further aging management review is No, but assembly steel; nickel dimensions! void necessary if the applicant provides a licensee (R-131) alloy swelling commitment in the FSAR supplement commitment Lower core to (1) participate in the industry needs to be plate programs for investigating and confirmed Radial keys managing aging effects on reactor and clevis internals; (2) evaluate and implement inserts the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended OJ operation, submit an inspection plan for

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z C

0 rn

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OAG10000203_401

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

0 B2 Reactor Vessel Internals (PWR) - Westinghouse m

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OJ Structure

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Mechanism Evaluation

0 (D

Component

< IV.B2-20 IV.B2.5-a Chapter XI. M2, Water Chemistry" for Lower internal Stainless Reactor coolant Cracking! stress No, but assembly steel; nickel corrosion cracking, PWR primary water licensee (R-130) alloy primary water commitment Lower core stress corrosion No further aging management review is needs to be plate cracking, necessary if the applicant provides a confirmed Radial keys irradiation-assisted commitment in the FSAR supplement and clevis stress corrosion to (1) participate in the industry inserts cracking programs for investigating and managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and OJ (3) upon completion of these programs,

~ but not less than 24 months before OJ entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

IV.B2-21 IV.B2.5-m Lower internal Cast Reactor coolant Loss of fracture Chapter XI.M13, "Thermal Aging and No assembly austenitic >250"C (>482"F) toughness! thermal Neutron Irradiation Embritllement of (R-140) stainless and neutron flux aging and neutron Cast Austenitic Stainless Steel Lower support steel irradiation (CASS)"

casting embritllement Lower support plate columns OAG10000203_402

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B2 Reactor Vessel Internals (PWR) - Westinghouse 3

0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B2-22 IV.B2.5-n Lower internal Stainless Reactor coolant Loss of fracture No further aging management review is No, but assembly steel and neutron flux toughness! neutron necessary if the applicant provides a licensee (R-141) irradiation commitment in the FSAR supplement commitment Lower support embrittlement, void to (1) participate in the industry needs to be forging swelling programs for investigating and confirmed Lower support managing aging effects on reactor plate columns internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

IV.B2-23 IV.B2.5-1 Lower internal Stainless Reactor coolant Changes in No further aging management review is No, but assembly steel dimensions! void necessary if the applicant provides a licensee (R-139) swelling commitment in the FSAR supplement commitment Lower support to (1) participate in the industry needs to be forging or programs for investigating and confirmed casting managing aging effects on reactor Lower support internals; (2) evaluate and implement plate columns the results of the industry programs as applicable to the reactor internals; and z

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0 m but not less than 24 months before

~ entering the period of extended OJ operation, submit an inspection plan for

~ reactor internals to the NRC for review and approval.

OAG10000203_403

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

0 B2 Reactor Vessel Internals (PWR) - Westinghouse m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< IV.B2-24 IV.B2.5-k Chapter XI. M2, Water Chemistry" for Lower internal Stainless Reactor coolant Cracking! stress No, but assembly steel corrosion cracking, PWR primary water licensee (R-138) irradiation-assisted commitment Lower support stress corrosion No further aging management review is needs to be forging or cracking necessary if the applicant provides a confirmed casting commitment in the FSAR supplement Lower support to (1) participate in the industry plate columns programs for investigating and managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and OJ (3) upon completion of these programs,

~ but not less than 24 months before OJ entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

OAG10000203_404

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B2 Reactor Vessel Internals (PWR) - Westinghouse 3

0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B2-25 IV.B2.5-h Lower internal Stainless Reactor coolant Loss of preload! No further aging management review is No, but assembly steel; nickel stress relaxation necessary if the applicant provides a licensee (R-136) alloy commitment in the FSAR supplement commitment Lower support to (1) participate in the industry needs to be plate column programs for investigating and confirmed bolts managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended OJ operation, submit an inspection plan for

~ reactor internals to the NRC for review CD and approval.

IV.B2-26 IV.B2.5-0 Lower internal Stainless Reactor coolant Loss of material! Chapter XI. M1, "ASM E Section XI No assembly steel wear Inservice Inspection, Subsections IWB, (R-142) IWC, and IWO," for Class 1 Radial keys components and clevis Inserts z

C

0 rn

~

OJ

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OAG10000203_405

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

0 B2 Reactor Vessel Internals (PWR) - Westinghouse m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< IV.B2-27 IV.B2.2-e RCCA guide Stainless Reactor coolant Changes in No further aging management review is No, but tube assemblies steel; nickel dimensions! void necessary if the applicant provides a licensee (R-119) alloy swelling commitment in the FSAR supplement commitment RCCA guide to (1) participate in the industry needs to be tube bolts programs for investigating and confirmed RCCA guide managing aging effects on reactor tube support internals; (2) evaluate and implement pins the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended OJ operation, submit an inspection plan for

'"or0 reactor internals to the NRC for review and approval.

OAG10000203_406

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B2 Reactor Vessel Internals (PWR) - Westinghouse 3

0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B2-28 IV.B2.2-d RCCA guide Stainless Reactor coolant Cracking! stress Chapter XI. M2, Water Chemistry" for No, but tube assemblies steel; nickel corrosion cracking, PWR primary water licensee (R-118) alloy primary water commitment RCCA guide stress corrosion No further aging management review is needs to be tube bolts cracking, necessary if the applicant provides a confirmed RCCA guide irradiation-assisted commitment in the FSAR supplement tube support stress corrosion to (1) participate in the industry pins cracking programs for investigating and managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

z C

0 rn

~

OJ

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OAG10000203_407

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

0 B2 Reactor Vessel Internals (PWR) - Westinghouse m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< IV.B2-29 IV.B2.2-b RCCA guide Stainless Reactor coolant Changes in No further aging management review is No, but tube assemblies steel dimensions! void necessary if the applicant provides a licensee (R-117) swelling commitment in the FSAR supplement commitment RCCA guide to (1) participate in the industry needs to be tubes programs for investigating and confirmed managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

OAG10000203_408

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B2 Reactor Vessel Internals (PWR) - Westinghouse 3

0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B2-30 IV.B2.2-a RCCA guide Stainless Reactor coolant Cracking! stress Chapter XI. M2, Water Chemistry" for No, but tube assemblies steel corrosion cracking, PWR primary water licensee (R-116) irradiation-assisted commitment RCCA guide stress corrosion No further aging management review is needs to be tubes cracking necessary if the applicant provides a confirmed commitment in the FSAR supplement to (1) participate in the industry programs for investigating and managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

IV.B2-31 IV.B2.1-m Reactor vessel Stainless Reactor coolant Cumulative fatigue Fatigue is a time-limited aging analysis Yes, IV.B2.2-f internals steel; nickel damage! fatigue (TLAA) to be evaluated for the period of TLAA (R-53) IV.B2.1-c components alloy extended operation. See the Standard IV.B2.2-c Review Plan, Section 4.3 "Metal IV.B2.3-d Fatigue," for acceptable methods for IVB24-g meeting the requirements of 10 CFR z IVB25-p 5421(c)(1)

C IVB25-j

0 m IV.B2.5-d

~

OJ IV.B2.1-h

~

OAG10000203_409

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

0 B2 Reactor Vessel Internals (PWR) - Westinghouse m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< IV.B2-32 IV.B2. Reactor vessel Stainless Reactor coolant Loss of material! Chapter XI.M2, "Water Chemistry," for No internals steel; nickel pitting and crevice PWR primary water (RP-24) components alloy corrosion IV.B2-33 IV.B2.1-d Upper internals Stainless Reactor coolant Loss of preload! No further aging management review is No, but assembly steel stress relaxation necessary if the applicant provides a licensee (R-108) commitment in the FSAR supplement commitment Hold-down to (1) participate in the industry needs to be spring programs for investigating and confirmed managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

IV.B2-34 IV.B2.1-1 Upper internals Stainless Reactor coolant Loss of material! Chapter XI. M1, "ASM E Section XI No assembly steel; nickel wear Inservice Inspection, Subsections IWB, (R-115) alloy IWC, and IWO," for Class 1 Upper core components plate alignment pins OAG10000203_410

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B2 Reactor Vessel Internals (PWR) - Westinghouse 3

0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B2-35 IV.B2.1-f Upper internals Stainless Reactor coolant Changes in No further aging management review is No, but assembly steel dimensions! void necessary if the applicant provides a licensee (R-110) swelling commitment in the FSAR supplement commitment Upper support to (1) participate in the industry needs to be column programs for investigating and confirmed managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended OJ operation, submit an inspection plan for

'"r0 reactor internals to the NRC for review U1 and approval.

z C

0 rn

~

OJ

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OAGI0000203_411

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

0 B2 Reactor Vessel Internals (PWR) - Westinghouse m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< IV.B2-36 IV.B2.1-e Chapter XI. M2, Water Chemistry" for Upper internals Stainless Reactor coolant Cracking! stress No, but assembly steel corrosion cracking, PWR primary water licensee (R-109) irradiation-assisted commitment Upper support stress corrosion No further aging management review is needs to be column cracking necessary if the applicant provides a confirmed commitment in the FSAR supplement to (1) participate in the industry programs for investigating and managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

IV.B2-37 IVB21-g Upper internals Cast Reactor coolant Loss of fracture Chapter XI.M13, "Thermal Aging and No assembly austenitic >250"C (>482"F) toughness! thermal Neutron Irradiation Embritllement of (R-111) stainless and neutron flux aging and neutron Cast Austenitic Stainless Steel Upper support steel irradiation (CASS)"

column embritllement (only cast austenitic stainless steel portions)

OAGI0000203_412

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B2 Reactor Vessel Internals (PWR) - Westinghouse 3

0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B2-38 IV.B2.1-k Upper internals Stainless Reactor coolant Loss of preload! No further aging management review is No, but assembly steel; nickel stress relaxation necessary if the applicant provides a licensee (R-114) alloy commitment in the FSAR supplement commitment Upper support to (1) participate in the industry needs to be column bolts programs for investigating and confirmed managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

IV.B2-39 IVB21-j Upper internals Stainless Reactor coolant Changes in No further aging management review is No, but assembly steel; nickel dimensions! void necessary if the applicant provides a licensee (R-113) alloy swelling commitment in the FSAR supplement commitment Upper support to (1) participate in the industry needs to be column bolts programs for investigating and confirmed Upper core managing aging effects on reactor plate internals; (2) evaluate and implement alignment pins the results of the industry programs as Fuel alignment applicable to the reactor internals; and z

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0 m but not less than 24 months before

~ entering the period of extended OJ operation, submit an inspection plan for

~ reactor internals to the NRC for review and approval.

OAGI0000203_413

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

0 B2 Reactor Vessel Internals (PWR) - Westinghouse m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< IV.B2.1-i Chapter XI. M2, Water Chemistry" for IV.B2-40 Upper internals Stainless Reactor coolant Cracking! stress No, but assembly steel; nickel corrosion cracking, PWR primary water licensee (R-112) alloy primary water commitment Upper support stress corrosion No further aging management review is needs to be column bolts cracking, necessary if the applicant provides a confirmed Upper core irradiation-assisted commitment in the FSAR supplement plate stress corrosion to (1) participate in the industry alignment pins cracking programs for investigating and Fuel alignment managing aging effects on reactor pins internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and OJ (3) upon completion of these programs,

'"r0 but not less than 24 months before OJ entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

OAGI0000203_414

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B2 Reactor Vessel Internals (PWR) - Westinghouse 3

0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B2-41 IV.B2.1-b Upper internals Stainless Reactor coolant Changes in No further aging management review is No, but assembly steel dimensions! void necessary if the applicant provides a licensee (R-107) swelling commitment in the FSAR supplement commitment Upper support to (1) participate in the industry needs to be plate programs for investigating and confirmed Upper core managing aging effects on reactor plate internals; (2) evaluate and implement Hold-down the results of the industry programs as spring applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended OJ operation, submit an inspection plan for

'"r0 reactor internals to the NRC for review CD and approval.

z C

0 rn

~

OJ

~

OAGI0000203_415

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

0 B2 Reactor Vessel Internals (PWR) - Westinghouse m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< IV.B2.1-a Chapter XI. M2, Water Chemistry" for IV.B2-42 Upper internals Stainless Reactor coolant Cracking! stress No, but assembly steel corrosion cracking, PWR primary water licensee (R-106) irradiation-assisted commitment Upper support stress corrosion No further aging management review is needs to be plate cracking necessary if the applicant provides a confirmed Upper core commitment in the FSAR supplement plate to (1) participate in the industry Hold-down programs for investigating and spring managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and OJ (3) upon completion of these programs,

'"oW but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

OAGI0000203_416

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),

shroud assemblies, 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 extended period of operation are included in IV.E.

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

September 2005 IV 83-1 NUREG-1801, Rev. 1 OAGI0000203_417

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

0 B3 Reactor Vessel Internals (PWR) - Combustion Engineering m

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OJ Structure

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Mechanism Evaluation

0 (D

Component

< IV.B3-1 IV.B3.2-e Chapter XIM13, "Thermal Aging and CEA shroud Cast Reactor coolant Loss of fracture No assemblies austenitic >250°C (>482°F) toughness! thermal Neutron Irradiation Embritllement of (R-153) stainless and neutron flux aging and neutron Cast Austenitic Stainless Steel steel irradiation (CASS)"

embritllement IV.B3-2 IV.B3.2-a CEA shroud Stainless Reactor coolant Cracking! stress Chapter XI M2, "Water Chemistry" for No, but assemblies steel corrosion cracking, PWR primary water licensee (R-149) irradiation-assisted commitment stress corrosion No further aging management review is needs to be cracking necessary if the applicant provides a confirmed commitment in the FSAR supplement

< to (1) participate in the industry OJ programs for investigating and managing

~ aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval IV.B3-3 IV.B3.2-d CEA shroud Stainless Reactor coolant Loss of material! Chapter XI M 1, "ASME Section XI No assemblies steel wear Inservice Inspection, Subsections IWB, (R-152) IWC, and IWO," for Class 1 component, CEA shroud extension shaft guides OAGI0000203_418

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B3 Reactor Vessel Internals (PWR) - Combustion Engineering 3

0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B3-4 IV.B3.2-c CEA shroud Stainless Reactor coolant Changes in No further aging management review is No, but assemblies steel; nickel dimensions! void necessary if the applicant provides a licensee (R-151) alloy swelling commitment in the FSAR supplement commitment CEA shrouds to (1) participate in the industry needs to be bolts programs for investigating and managing confirmed aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an

< inspection plan for reactor internals to OJ the NRC for review and approval.

~

z C

0 rn

~

OJ

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OAGI0000203_419

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

0 B3 Reactor Vessel Internals (PWR) - Combustion Engineering m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< IV.B3-5 IV.B3.2-b Chapter XI. M2, "Water Chemistry" for CEA shroud Stainless Reactor coolant Cracking! stress No, but assemblies steel; nickel corrosion cracking, PWR primary water. No further aging licensee (R-150) alloy primary water management review is necessary if the commitment CEA shrouds stress corrosion applicant provides a commitment in the needs to be bolts cracking, FSAR supplement to (1) participate in confirmed irradiation-assisted the industry programs for investigating stress corrosion and managing aging effects on reactor cracking internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before

< entering the period of extended OJ operation, submit an inspection plan for

't reactor internals to the NRC for review and approval.

OAG10000203_420

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B3 Reactor Vessel Internals (PWR) - Combustion Engineering 3

0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B3-6 IVB32-g CEA shroud Stainless Reactor coolant Loss of preload! No further aging management review is No, but assemblies steel; nickel stress relaxation necessary if the applicant provides a licensee (R-154) alloy commitment in the FSAR supplement commitment CEA shrouds to (1) participate in the industry needs to be bolts programs for investigating and managing confirmed aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an

< inspection plan for reactor internals to OJ the NRC for review and approval.

~

IV.B3-7 IV.B3.4-h Core shroud Stainless Reactor coolant Loss of preload! No further aging management review is No, but assembly steel; nickel stress relaxation necessary if the applicant provides a licensee (R-165) alloy commitment in the FSAR supplement commitment Core shroud to (1) participate in the industry needs to be assembly programs for investigating and managing confirmed bolts aging effects on reactor internals; (2)

Core shroud evaluate and implement the results of tie rods the industry programs as applicable to the reactor internals; and (3) upon z

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0 rn less than 24 months before entering the

~ period of extended operation, submit an OJ inspection plan for reactor internals to

~ the NRC for review and approval.

OAGI0000203_421

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

0 B3 Reactor Vessel Internals (PWR) - Combustion Engineering m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< IV.B3-8 IV.B3.4-f Core shroud Stainless Reactor coolant Changes in No further aging management review is No, but assembly steel; nickel dimensions! void necessary if the applicant provides a licensee (R-163) alloy swelling commitment in the FSAR supplement commitment Core shroud to (1) participate in the industry needs to be assembly programs for investigating and managing confirmed bolts (later aging effects on reactor internals; (2) plants are evaluate and implement the results of welded) the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an

< inspection plan for reactor internals to OJ the NRC for review and approval.

~

OAGI0000203_422

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B3 Reactor Vessel Internals (PWR) - Combustion Engineering 3

0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B3-9 IV.B3.4-e Core shroud Stainless Reactor coolant Cracking! stress Chapter XI. M2, "Water Chemistry" for No, but assembly steel; nickel corrosion cracking, PWR primary water licensee (R-162) alloy primary water commitment Core shroud stress corrosion No further aging management review is needs to be assembly cracking, necessary if the applicant provides a confirmed bolts (later irradiation-assisted commitment in the FSAR supplement plants are stress corrosion to (1) participate in the industry welded) cracking programs for investigating and managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon

< completion of these programs, but not OJ less than 24 months before entering the

~ period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

z C

0 rn

~

OJ

~

OAGI0000203_423

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

0 B3 Reactor Vessel Internals (PWR) - Combustion Engineering m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< IV.B3-10 IVB34-g Core shroud Stainless Reactor coolant Loss of fracture No further aging management review is No, but assembly steel; nickel and neutron flux toughness! neutron necessary if the applicant provides a licensee (R-164) alloy irradiation commitment in the FSAR supplement commitment Core shroud em brittlement, void to (1) participate in the industry needs to be assembly swelling programs for investigating and managing confirmed bolts (later aging effects on reactor internals; (2) plants are evaluate and implement the results of welded) the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an

< inspection plan for reactor internals to OJ the NRC for review and approval.

~

OAGI0000203_424

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B3 Reactor Vessel Internals (PWR) - Combustion Engineering 3

0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B3-11 IV.B3.4-a Core shroud Stainless Reactor coolant Cracking! stress Chapter XI. M2, "Water Chemistry" for No, but assembly steel corrosion cracking, PWR primary water licensee (R-159) irradiation-assisted commitment Core shroud stress corrosion No further aging management review is needs to be tie rods (core cracking necessary if the applicant provides a confirmed support plate commitment in the FSAR supplement attached by to (1) participate in the industry welds in later programs for investigating and managing plants) aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon

< completion of these programs, but not OJ o less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

z C

0 rn

~

OJ

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OAGI0000203_425

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

0 B3 Reactor Vessel Internals (PWR) - Combustion Engineering m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< IV.B3-12 IV.B3.4-c Core shroud Stainless Reactor coolant Loss of fracture No further aging management review is No, but assembly steel and neutron flux toughness! neutron necessary if the applicant provides a licensee (R-161) irradiation commitment in the FSAR supplement commitment Core shroud em brittlement, void to (1) participate in the industry needs to be tie rods (core swelling programs for investigating and managing confirmed support plate aging effects on reactor internals; (2) attached by evaluate and implement the results of welds in later the industry programs as applicable to plants) the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an OJ inspection plan for reactor internals to

~ the NRC for review and approval.

o IV.B3-13 IV.B3.4-b Core shroud Stainless Reactor coolant Changes in No further aging management review is No, but assembly steel; nickel dimensions! void necessary if the applicant provides a licensee (R-160) alloy swelling commitment in the FSAR supplement commitment Core shroud to (1) participate in the industry needs to be tie rods (core programs for investigating and managing confirmed support plate aging effects on reactor internals; (2) attached by evaluate and implement the results of welds in later the industry programs as applicable to plants) the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

OAGI0000203_426

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(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B3 Reactor Vessel Internals (PWR) - Combustion Engineering 3

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~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B3-14 IV.B3.3-b Core support Stainless Reactor coolant Changes in No further aging management review is No, but barrel steel dimensions! void necessary if the applicant provides a licensee (R-158) swelling commitment in the FSAR supplement commitment Core support to (1) participate in the industry needs to be barrel upper programs for investigating and managing confirmed flange aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an OJ inspection plan for reactor internals to

~ the NRC for review and approval.

z C

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

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Mechanism Evaluation

0 (D

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< IV.B3-15 IV.B3.3-a Chapter XI. M2, "Water Chemistry" for Core support Stainless Reactor coolant Cracking! stress No, but barrel steel corrosion cracking, PWR primary water licensee (R-155) irradiation-assisted commitment Core support stress corrosion No further aging management review is needs to be barrel upper cracking necessary if the applicant provides a confirmed flange commitment in the FSAR supplement to (1) participate in the industry programs for investigating and managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

OAGI0000203_428

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~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B3-16 IV.B3.3-a Core support Stainless Reactor coolant Loss of fracture No further aging management review is No, but barrel steel and neutron flux toughness! neutron necessary if the applicant provides a licensee (R-157) irradiation commitment in the FSAR supplement commitment Core support em britllement, void to (1) participate in the industry needs to be barrel upper swelling programs for investigating and managing confirmed flange aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an OJ inspection plan for reactor internals to

~ the NRC for review and approval.

w IV.B3-17 IV.B3.3-b Core support Stainless Reactor coolant Loss of material! Chapter XI. M 1, "ASME Section XI No barrel steel wear Inservice Inspection, Subsections IWB, (R-156) IWC, and IWO," for Class 1 component, Core support barrel upper flange Core support barrel alignment z

C keys

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~ assembly austenitic >250°C (>482°F) toughness! thermal Neutron Irradiation Embritllement of OJ (R-171) stainless and neutron flux aging and neutron Cast Austenitic Stainless Steel

~ Core support steel irradiation (CASS)"

column embritllement OAGI0000203_429

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

0 B3 Reactor Vessel Internals (PWR) - Combustion Engineering m

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~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

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< IV.B3-19 IV.B3.5-c Lower internal Stainless Reactor coolant Changes in No further aging management review is No, but assembly steel; nickel dimensions! void necessary if the applicant provides a licensee (R-168) alloy swelling commitment in the FSAR supplement commitment Core support to (1) participate in the industry needs to be plate programs for investigating and managing confirmed Fuel aging effects on reactor internals; (2) alignment evaluate and implement the results of pins the industry programs as applicable to Lower the reactor internals; and (3) upon support completion of these programs, but not structure less than 24 months before entering the beam period of extended operation, submit an assemblies inspection plan for reactor internals to Core support the NRC for review and approval.

column bolts Core support barrel snubber assemblies OAG10000203_430

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(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B3 Reactor Vessel Internals (PWR) - Combustion Engineering 3

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~ Structure Aging Effect! Further

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Evaluation U1 Component IV.B3-20 IV.B3.5-d Lower internal Stainless Reactor coolant Loss of fracture No further aging management review is No, but assembly steel; nickel and neutron flux toughness! neutron necessary if the applicant provides a licensee (R-169) alloy irradiation commitment in the FSAR supplement commitment Core support em brittlement, void to (1) participate in the industry needs to be plate swelling programs for investigating and managing confirmed Fuel aging effects on reactor internals; (2) alignment evaluate and implement the results of pins the industry programs as applicable to Lower the reactor internals; and (3) upon support completion of these programs, but not structure less than 24 months before entering the beam period of extended operation, submit an OJ assemblies inspection plan for reactor internals to

~ Core support the NRC for review and approval.

U1 column bolts Core support barrel snubber assemblies z

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

0 B3 Reactor Vessel Internals (PWR) - Combustion Engineering m

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Mechanism Evaluation

0 (D

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< IV.B3-21 IV.B3.5-a Chapter XI. M2, "Water Chemistry" for Lower internal Stainless Reactor coolant Cracking! stress No, but assembly steel corrosion cracking, PWR primary water licensee (R-166) irradiation-assisted commitment Core support stress corrosion No further aging management review is needs to be plate cracking necessary if the applicant provides a confirmed Lower commitment in the FSAR supplement support to (1) participate in the industry structure programs for investigating and managing beam aging effects on reactor internals; (2) assemblies evaluate and implement the results of Core support the industry programs as applicable to column the reactor internals; and (3) upon OJ Core support completion of these programs, but not

~ barrel less than 24 months before entering the OJ snubber period of extended operation, submit an assemblies inspection plan for reactor internals to the NRC for review and approval.

IV.B3-22 IV.B3.5-e Lower internal Stainless Reactor coolant Loss of material! Chapter XI. M 1, "ASME Section XI No assembly steel; nickel wear Inservice Inspection, Subsections IWB, (R-170) alloy IWC, and IWO," for Class 1 component, Fuel alignment pins Core support barrel snubber assemblies OAGI0000203_432

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(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B3 Reactor Vessel Internals (PWR) - Combustion Engineering 3

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~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B3-23 IV.B3.5-b Lower internal Stainless Reactor coolant Cracking! stress Chapter XI. M2, "Water Chemistry" for No, but Assembly steel; nickel corrosion cracking, PWR primary water licensee (R-167) alloy primary water commitment Fuel stress corrosion No further aging management review is needs to be alignment cracking, necessary if the applicant provides a confirmed pins irradiation-assisted commitment in the FSAR supplement Core support stress corrosion to (1) participate in the industry column bolts cracking programs for investigating and managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

IV.B3-24 IV.B3.4-d Reactor vessel Stainless Reactor coolant Cumulative fatigue Fatigue is a time-limited aging analysis Yes, IVB35-g internals steel; nickel damage! fatigue (TLAA) to be evaluated for the period of TLAA (R-53) IV.B3.2-f components alloy extended operation. See the Standard Review Plan, Section 4.3 "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR z 5421(c)(1)

C

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internals steel; nickel pitting and crevice PWR primary water

~

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~

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

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~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

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< IV.B3-26 IV.B3.1-c Chapter XI. M 1, "ASME Section XI Upper Stainless Reactor coolant Loss of material! No internals steel wear Inservice Inspection, Subsections IWB, (R-148) assembly IWC, and IWO," for Class 1 component, Fuel alignment plate Fuel alignment plate guide lugs and their lugs OJ Hold-down

~ ring OJ IV.B3-27 IV.B3.1-b Upper Stainless Reactor coolant Changes in No further aging management review is No, but internals steel dimensions! void necessary if the applicant provides a licensee (R-147) assembly swelling commitment in the FSAR supplement commitment to (1) participate in the industry needs to be Upper guide programs for investigating and managing confirmed structure aging effects on reactor internals; (2) support plate evaluate and implement the results of Fuel the industry programs as applicable to alignment the reactor internals; and (3) upon plate completion of these programs, but not Fuel less than 24 months before entering the alignment period of extended operation, submit an plate guide inspection plan for reactor internals to lugs and the NRC for review and approval.

guide lug inserts OAGI0000203_434

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B3 Reactor Vessel Internals (PWR) - Combustion Engineering 3

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~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B3-28 IV.B3.1-a Upper Stainless Reactor coolant Cracking! stress Chapter XI. M2, "Water Chemistry" for No, but internals steel corrosion cracking, PWR primary water licensee (R-146) assembly irradiation-assisted commitment stress corrosion No further aging management review is needs to be Upper guide cracking necessary if the applicant provides a confirmed structure commitment in the FSAR supplement support plate to (1) participate in the industry Fuel programs for investigating and managing alignment aging effects on reactor internals; (2) plate evaluate and implement the results of Fuel the industry programs as applicable to alignment the reactor internals; and (3) upon OJ plate guide completion of these programs, but not

~ lugs and less than 24 months before entering the CD guide lug period of extended operation, submit an inserts inspection plan for reactor internals to the NRC for review and approval.

z C

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OAGI0000203_435

This Page Intentionally Left Blank NUREG-1801, Rev. 1 IV.B3-20 September 2005 OAGI0000203_436

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 extended period of operation are included in IV.E.

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

September 2005 IV B4-1 NUREG-1801, Rev. 1 OAGI0000203_437

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM z B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox C

0

~

OJ Item Link Structure and!or Material Environment Ag ing Effect!

Aging Management Program (AMP)

Further

~ Component Mechanism Evaluation IV.B4-1 IV.B4.5-i Bafflelformer Stainless Reactor coolant Loss of fracture No further aging management review is No, but assembly steel and neutron flux toughness! neutron necessary if the applicant provides a licensee (R-128) irradiation commitment in the FSAR supplement commitment Bafflelformer embrittlement, void to (1) participate in the industry needs to be bolts and swelling programs for investigating and confirmed screws managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

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~ Structure Ag ing Effect! Further

'"oo Item Link and/or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B4-2 IV.B4.3-a Control rod Stainless Reactor coolant Cracki ng/ stress Chapter XI. M2, Water Chemistry" for No, but guide tube steel corrosion cracking, PWR primary water licensee (R-180) (CRGT) irradiation-assisted commitment assembly stress corrosion No further aging management review is needs to be cracking necessary if the applicant provides a confirmed CRGT pipe commitment in the FSAR supplement and flange to (1) participate in the industry CRGTspacer programs for investigating and casting managing aging effects on reactor CRGT rod internals; (2) evaluate and implement guide tubes the results of the industry programs as CRGT rod applicable to the reactor internals; and

< guide sectors (3) upon completion of these programs, OJ t but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

z C

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IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM z B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox C

0

~

OJ Item Link Structure and!or Material Environment Ag ing Effect!

Aging Management Program (AMP)

Further

~ Component Mechanism Evaluation IV.B4-3 IV.B4.3-c Control rod Stainless Reactor coolant Changes in No further aging management review is No, but guide tube steel dimensions! void necessary if the applicant provides a licensee (R-182) (CRGT) swelling commitment in the FSAR supplement commitment assembly to (1) participate in the industry needs to be programs for investigating and confirmed CRGT pipe managing aging effects on reactor and flange internals; (2) evaluate and implement CRGTspacer the results of the industry programs as casting applicable to the reactor internals; and CRGTspacer (3) upon completion of these programs, screws but not less than 24 months before Flange-to- entering the period of extended upper grid operation, submit an inspection plan for screws reactor internals to the NRC for review CRGT rod and approva I.

guide tubes CRGT rod guide sectors IV.B4-4 IV.B4.3-d Control rod Cast Reactor coolant Loss of fracture Chapter XI.M13, "Thermal Aging and No guide tube austenitic >250°C (>482°F) toughness! thermal Neutron Irradiation Embritllement of (R-183) (CRGT) stainless and neutron flux aging and neutron Cast Austenitic Stainless Steel assembly steel irradiation (CASS)"

embritllement CRGTspacer casting (f)

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(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox 3

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~ Structure Ag ing Effect! Further

'"oo Item Link and/or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B4-5 IV.B4.3-b Control rod Stainless Reactor coolant Cracki ng/ stress Chapter XI. M2, Water Chemistry" for No, but guide tube steel corrosion cracking, PWR primary water licensee (R-181) (CRGT) irradiation-assisted commitment assembly stress corrosion No further aging management review is needs to be cracking necessary if the applicant provides a confirmed CRGTspacer commitment in the FSAR supplement screws to (1) participate in the industry Flange-to- programs for investigating and upper grid managing aging effects on reactor screws internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and

< (3) upon completion of these programs, OJ 6; but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

z C

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IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM z B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox C

0

~

OJ Item Link Structure and!or Material Environment Ag ing Effect!

Aging Management Program (AMP)

Further

~ Component Mechanism Evaluation IV.B4-6 IV.B4.3-e Control rod Stainless Reactor coolant Loss of preload! No further aging management review is No, but guide tube steel stress relaxation necessary if the applicant provides a licensee (R-184) (CRGT) commitment in the FSAR supplement commitment assembly to (1) participate in the industry needs to be programs for investigating and confirmed Flange-to- managing aging effects on reactor upper grid internals; (2) evaluate and implement screws the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for

<: reactor internals to the NRC for review OJ

-!' and approval.

OJ (f)

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(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox 3

0-

~ Structure Ag ing Effect! Further

'"oo Item Link and/or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B4-7 IVB45-g Core barrel Stainless Reactor coolant Cracki ng/ stress Chapter XI.M2, Water Chemistry," for No, but assembly steel corrosion cracking, PWR primary water licensee (R-125) irradiation-assisted commitment Bafflelformer stress corrosion No further aging management review is needs to be assembly cracking necessary if the applicant provides a confirmed Bafflelformer commitment in the FSAR supplement bolts and to (1) participate in the industry screws programs for investigating and managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and

< (3) upon completion of these programs, OJ but not less than 24 months before t entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

z C

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IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM z B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox C

0

~

OJ Item Link Structure and!or Material Environment Ag ing Effect!

Aging Management Program (AMP)

Further

~ Component Mechanism Evaluation IV.B4-8 IV.B4.5-h Core barrel Stainless Reactor coolant Changes in No further aging management review is No, but assembly steel dimensions! void necessary if the applicant provides a licensee (R-199) swelling commitment in the FSAR supplement commitment Bafflelformer to (1) participate in the industry needs to be bolts and programs for investigating and confirmed screws managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for

<: reactor internals to the NRC for review OJ

-!' and approval.

OJ IV.B4-9 IVB45-j Core barrel Stainless Reactor coolant Loss of preload! No further aging management review is No, but assembly steel stress relaxation necessary if the applicant provides a licensee (R-201) commitment in the FSAR supplement commitment Bafflelformer to (1) participate in the industry needs to be bolts and programs for investigating and confirmed screws managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before (f) entering the period of extended (D

operation, submit an inspection plan for

~ reactor internals to the NRC for review 3

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(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox 3

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~ Structure Ag ing Effect! Further

'"oo Item Link and/or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B4-10 IV.B4.5-a Core barrel Stainless Reactor coolant Cracki ng/ stress Chapter XI.M2, "Water Chemistry" for No, but assembly steel corrosion cracking, PWR primary water licensee (R-193) irradiation-assisted commitment Core barrel stress corrosion No further aging management review is needs to be cylinder (top cracking necessary if the applicant provides a confirmed and bottom commitment in the FSAR supplement flange) to (1) participate in the industry Baffle plates programs for investigating and and formers managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and

< (3) upon completion of these programs, OJ o but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

z C

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IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM z B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox C

0

~

OJ Item Link Structure and!or Material Environment Ag ing Effect!

Aging Management Program (AMP)

Further

~ Component Mechanism Evaluation IV.B4-11 IV.B4.5-c Core barrel Stainless Reactor coolant Changes in No further aging management review is No, but assembly steel; nickel dimensions! void necessary if the applicant provides a licensee (R-195) alloy swelling commitment in the FSAR supplement commitment Core barrel to (1) participate in the industry needs to be cylinder (top programs for investigating and confirmed and bottom managing aging effects on reactor flange) internals; (2) evaluate and implement Lower the results of the industry programs as internals applicable to the reactor internals; and assem bly-to- (3) upon completion of these programs, core barrel but not less than 24 months before bolts entering the period of extended Core barrel-to- operation, submit an inspection plan for

< thermal shield reactor internals to the NRC for review OJ

+/-:

o bolts and approval.

Baffle plates and formers (f)

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(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox 3

0-

~ Structure Ag ing Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B4-12 IV.B4.5-d Core barrel Stainless Reactor coolant Loss of fracture No further aging management review is No, but assembly steel; nickel and neutron flux toughness! neutron necessary if the applicant provides a licensee (R-196) alloy irradiation commitment in the FSAR supplement commitment Core barrel embrittlement, void to (1) participate in the industry needs to be cylinder (top swelling programs for investigating and confirmed and bottom managing aging effects on reactor flange) internals; (2) evaluate and implement Lower the results of the industry programs as internals applicable to the reactor internals; and assem bly-to- (3) upon completion of these programs, core barrel but not less than 24 months before bolts entering the period of extended OJ Core barrel-to- operation, submit an inspection plan for

+/-: thermal shield reactor internals to the NRC for review bolts and approval.

Baffle plates and formers z

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IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM z B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox C

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~

OJ Item Link Structure and/or Material Environment Ag ing Effect!

Aging Management Program (AMP)

Further

~ Component Mechanism Evaluation IV.B4-13 IV.B4.5-b Core barrel Stainless Reactor coolant Cracki ng/ stress Chapter XI. M2, Water Chemistry" for No, but assembly steel; nickel corrosion cracking, PWR primary water licensee (R-194) alloy primary water commitment Lower stress corrosion No further aging management review is needs to be internals cracking, necessary if the applicant provides a confirmed assem bly-to- irradiation-assisted commitment in the FSAR supplement core barrel stress corrosion to (1) participate in the industry bolts cracking programs for investigating and Core barrel-to- managing aging effects on reactor thermal shield internals; (2) evaluate and implement bolts the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

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~ Structure Ag ing Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B4-14 IV.B4.5-e Core barrel Stainless Reactor coolant Loss of preload! No further aging management review is No, but assembly steel; nickel stress relaxation necessary if the applicant provides a licensee (R-197) alloy commitment in the FSAR supplement commitment Lower to (1) participate in the industry needs to be internals programs for investigating and confirmed assem bly-to- managing aging effects on reactor core barrel internals; (2) evaluate and implement bolts the results of the industry programs as Core barrel-to- applicable to the reactor internals; and thermal shield (3) upon completion of these programs, bolts but not less than 24 months before entering the period of extended OJ operation, submit an inspection plan for

+/-:

w reactor internals to the NRC for review and approval.

IV.B4-15 IV.B4.4-f Core support Stainless Reactor coolant Loss of material! Chapter XI.M1, "ASME Section XI No shield steel wear Inservice Inspection, Subsections IWB, (R-190) assembly IWC, and IWO," for Class 1 components Core support shield cylinder (top flange) vent valve assembly z

C locking device

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IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM z B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox C

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~

OJ Item Link Structure and!or Material Environment Ag ing Effect!

Aging Management Program (AMP)

Further

~ Component Mechanism Evaluation IV.B4-16 IV.B4.4-d Core support Stainless Reactor coolant Loss of fracture No further aging management review is No, but shield steel; nickel and neutron flux toughness! neutron necessary if the applicant provides a licensee (R-188) assembly alloy irradiation commitment in the FSAR supplement commitment embrittlement, void to (1) participate in the industry needs to be Core support swelling programs for investigating and confirmed shield cylinder managing aging effects on reactor (top and internals; (2) evaluate and implement bottom flange) the results of the industry programs as Core support applicable to the reactor internals; and shield-to-core (3) upon completion of these programs, barrel bolts but not less than 24 months before Outlet and entering the period of extended vent \,6~e operation, submit an inspection plan for nozzles reactor internals to the NRC for review vent valve and approval.

assembly locking device (f)

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(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox 3

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~ Structure Ag ing Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B4-17 IV.B4.4-c Core support Stainless Reactor coolant Changes in No further aging management review is No, but shield steel; nickel dimensions! void necessary if the applicant provides a licensee (R-187) assembly alloy swelling commitment in the FSAR supplement commitment to (1) participate in the industry needs to be Core support programs for investigating and confirmed shield cylinder managing aging effects on reactor (top and internals; (2) evaluate and implement bottom flange) the results of the industry programs as Core support applicable to the reactor internals; and shield-to-core (3) upon completion of these programs, barrel bolts but not less than 24 months before vent valve entering the period of extended OJ retaining ring operation, submit an inspection plan for

+/-: vent valve reactor internals to the NRC for review U1 assembly and approval.

locking device z

C

0 rn

~

OJ

~

OAGI0000203_451

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM z B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox C

0

~

OJ Item Link Structure and/or Material Environment Ag ing Effect!

Aging Management Program (AMP)

Further

~ Component Mechanism Evaluation IV.B4-18 IV.B4.4-a Core support Stainless Reactor coolant Cracki ng/ stress Chapter XI. M2, Water Chemistry" for No, but shield steel corrosion cracking, PWR primary water licensee (R-185) assembly irradiation-assisted commitment stress corrosion No further aging management review is needs to be Core support cracking necessary if the applicant provides a confirmed shield cylinder commitment in the FSAR supplement (top and to (1) participate in the industry bottom flange) programs for investigating and Outlet and managing aging effects on reactor vent valve internals; (2) evaluate and implement nozzles the results of the industry programs as vent va~e applicable to the reactor internals; and body and (3) upon completion of these programs,

< retaining ring but not less than 24 months before OJ

+/-:

OJ entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

(f)

(D

~

3 0-

~

'"oo U1 OAGI0000203_452

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox 3

0-

~ Structure Ag ing Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B4-19 IV.B4.4-h Core support Stainless Reactor coolant Loss of preload! No further aging management review is No, but shield steel; nickel stress relaxation necessary if the applicant provides a licensee (R-192) assembly alloy commitment in the FSAR supplement commitment to (1) participate in the industry needs to be Core support programs for investigating and confirmed shield-to-core managing aging effects on reactor barrel bolts internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

z C

0 rn

~

OJ

~

OAGI0000203_453

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM z B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox C

0

~

OJ Item Link Structure and!or Material Environment Ag ing Effect!

Aging Management Program (AMP)

Further

~ Component Mechanism Evaluation IV.B4-20 IV.B4.4-b Core support Stainless Reactor coolant Cracki ng! stress Chapter XI. M2, Water Chemistry" for No, but shield steel; nickel corrosion cracking, PWR primary water licensee (R-186) assembly alloy primary water commitment stress corrosion No further aging management review is needs to be Core support cracking, necessary if the applicant provides a confirmed shield-to-core irradiation-assisted commitment in the FSAR supplement barrel bolts stress corrosion to (1) participate in the industry vent valve cracking programs for investigating and assembly managing aging effects on reactor locking device internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs,

< but not less than 24 months before OJ

+/-:

OJ entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

IV.B4-21 IVB4.4-g Core support Cast Reactor coolant Loss of fracture Chapter XI.M13, "Thermal Aging and No shield austenitic >250"C (>482"F) toughness! thermal Neutron Irradiation Embritllement of (R-191) assembly stainless and neutron flux aging and neutron Cast Austenitic Stainless Steel steel irradiation (CASS)"

Outlet and embritllement vent va~e nozzles vent valve body and (f) retaining ring (D

~

3 0-

~

'"oo U1 OAGI0000203_454

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox 3

0-

~ Structure Ag ing Effect! Further

'"oo Item Link and/or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B4-22 IV.B4.7-a Flow distributor Stainless Reactor coolant Cracki ng/ stress Chapter XI. M2, Water Chemistry" for No, but assembly steel corrosion cracking, PWR primary water licensee (R-209) irradiation-assisted commitment Flow stress corrosion No further aging management review is needs to be distributor cracking necessary if the applicant provides a confirmed head and commitment in the FSAR supplement flange to (1) participate in the industry Incore guide programs for investigating and support plate managing aging effects on reactor Clamping ring internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and OJ (3) upon completion of these programs,

+/-: but not less than 24 months before CD entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

z C

0 rn

~

OJ

~

OAGI0000203_455

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM z B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox C

0

~

OJ Item Link Structure and!or Material Environment Ag ing Effect!

Aging Management Program (AMP)

Further

~ Component Mechanism Evaluation IV.B4-23 IV.B4.7-c Flow distributor Stainless Reactor coolant Changes in No further aging management review is No, but assembly steel; nickel dimensions! void necessary if the applicant provides a licensee (R-211) alloy swelling commitment in the FSAR supplement commitment Flow to (1) participate in the industry needs to be distributor programs for investigating and confirmed head and managing aging effects on reactor flange internals; (2) evaluate and implement Shell forging- the results of the industry programs as to-flow applicable to the reactor internals; and distributor (3) upon completion of these programs, bolts but not less than 24 months before Incore guide entering the period of extended support plate operation, submit an inspection plan for

< Clamping ring reactor internals to the NRC for review OJ to and approval.

IV.B4-24 IV.B4.7-d Flow distributor Stainless Reactor coolant Loss of fracture No further aging management review is No, but assembly steel; nickel and neutron flux toughness! neutron necessary if the applicant provides a licensee (R-212) alloy irradiation commitment in the FSAR supplement commitment Flow embrittlement, void to (1) participate in the industry needs to be distributor swelling programs for investigating and confirmed head and managing aging effects on reactor flange internals; (2) evaluate and implement Shell forging- the results of the industry programs as to-flow applicable to the reactor internals; and distributor (3) upon completion of these programs, bolts but not less than 24 months before (f) Incore guide entering the period of extended (D

support plate operation, submit an inspection plan for

~ Clamping ring reactor internals to the NRC for review 3

0- and approval.

~

'"oo U1 OAGI0000203_456

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox 3

0-

~ Structure Ag ing Effect! Further

'"oo Item Link and/or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B4-25 IV.B4.7-b Flow distributor Stainless Reactor coolant Cracki ng/ stress Chapter XI. M2, Water Chemistry" for No, but assembly steel; nickel corrosion cracking, PWR primary water licensee (R-210) alloy primary water commitment Shell forging- stress corrosion No further aging management review is needs to be to-flow cracking, necessary if the applicant provides a confirmed distributor irradiation-assisted commitment in the FSAR supplement bolts stress corrosion to (1) participate in the industry cracking programs for investigating and managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and OJ (3) upon completion of these programs, t but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

z C

0 rn

~

OJ

~

OAGI0000203_457

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM z B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox C

0

~

OJ Item Link Structure and!or Material Environment Ag ing Effect!

Aging Management Program (AMP)

Further

~ Component Mechanism Evaluation IV.B4-26 IV.B4.7-e Flow distributor Stainless Reactor coolant Loss of preload! No further aging management review is No, but assembly steel; nickel stress relaxation necessary if the applicant provides a licensee (R-213) alloy commitment in the FSAR supplement commitment Shell forging- to (1) participate in the industry needs to be to-flow programs for investigating and confirmed distributor managing aging effects on reactor bolts internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for

< reactor internals to the NRC for review OJ t and approval.

'" IV.B4-27 IV.B4.6-h Lower grid Stainless Reactor coolant Loss of material! Chapter XI. M1, "ASME Section XI No assembly steel wear Inservice Inspection, Subsections IWB, (R-208) IWC, and IWO," for Class 1 Fuel assembly components support pads Guide blocks IV.B4-28 IV.B4.6-e Lower grid Cast Reactor coolant Loss of fracture Chapter XI. M13, "Thermal Aging and No assembly austenitic >250°C (>482°F) toughness! thermal Neutron Irradiation Embritllement of (R-206) stainless and neutron flux aging and neutron Cast Austenitic Stainless Steel Incore guide steel irradiation (CASS)"

tube spider embritllement (f)

(D castings

~

3 0-

~

'"oo U1 OAGI0000203_458

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox 3

0-

~ Structure Ag ing Effect! Further

'"oo Item Link and/or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B4-29 IV.B4.6-a Lower grid Stainless Reactor coolant Cracki ng/ stress Chapter XI. M2, Water Chemistry" for No, but assembly steel corrosion cracking, PWR primary water licensee (R-202) irradiation-assisted commitment Lower grid rib stress corrosion No further aging management review is needs to be section cracking necessary if the applicant provides a confirmed Fuel assembly commitment in the FSAR supplement support pads to (1) participate in the industry Lower grid flow programs for investigating and dis!. plate managing aging effects on reactor Orifice plugs internals; (2) evaluate and implement Lower grid and the results of the industry programs as shell forgings applicable to the reactor internals; and OJ Guide blocks (3) upon completion of these programs, tw Shock pads Support post but not less than 24 months before entering the period of extended pipes operation, submit an inspection plan for Incore guide reactor internals to the NRC for review tube spider and approval.

castings z

C

0 rn

~

OJ

~

OAGI0000203_459

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM z B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox C

0

~

OJ Item Link Structure and!or Material Environment Ag ing Effect!

Aging Management Program (AMP)

Further

~ Component Mechanism Evaluation IV.B4-30 IV.B4.6-c Lower grid Stainless Reactor coolant Changes in No further aging management review is No, but assembly steel; nickel dimensions! void necessary if the applicant provides a licensee (R-204) alloy swelling commitment in the FSAR supplement commitment Lower grid rib to (1) participate in the industry needs to be section programs for investigating and confirmed Fuel assembly managing aging effects on reactor support pads internals; (2) evaluate and implement Lower grid rib- the results of the industry programs as to-shell forging applicable to the reactor internals; and screws (3) upon completion of these programs, Lower grid flow but not less than 24 months before dis!. plate entering the period of extended Orifice plugs operation, submit an inspection plan for

< Lower grid and reactor internals to the NRC for review OJ t-" shell forgings Lower and approval.

internals assem bly-to-thermal shield bolts Guide blocks and bolts Shock pads and bolts Support post pipes Incore guide (f) tube spider (D

castings

~

3 0-

~

'"oo U1 OAG10000203_460

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox 3

0-

~ Structure Ag ing Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B4-31 IV.B4.6-d Lower grid Stainless Reactor coolant Loss of fracture No further aging management review is No, but assembly steel; nickel and neutron flux toughness! neutron necessary if the applicant provides a licensee (R-205) alloy irradiation commitment in the FSAR supplement commitment Lower grid rib embrittlement, \,Oid to (1) participate in the industry needs to be section swelling programs for investigating and confirmed Fuel assembly managing aging effects on reactor support pads internals; (2) evaluate and implement Lower grid rib- the results of the industry programs as to-shell forging applicable to the reactor internals; and screws (3) upon completion of these programs, Lower grid flow but not less than 24 months before dis!. plate entering the period of extended OJ Orifice plugs operation, submit an inspection plan for t

U1 Lower grid and shell forgings reactor internals to the NRC for review and approval.

Lower internals assem bly-to-thermal shield bolts Guide blocks and bolts Shock pads and bolts Support post z

C pipes

0 rn

~

OJ

~

OAGI0000203_461

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM z B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox C

0

~

OJ Item Link Structure and/or Material Environment Ag ing Effect!

Aging Management Program (AMP)

Further

~ Component Mechanism Evaluation IV.B4-32 IV.B4.6-b Lower grid Stainless Reactor coolant Cracki ng/ stress Chapter XI. M2, Water Chemistry" for No, but assembly steel; nickel corrosion cracking, PWR primary water licensee (R-203) alloy primary water commitment Lower grid rib- stress corrosion No further aging management review is needs to be to-shell forging cracking, necessary if the applicant provides a confirmed screws irradiation-assisted commitment in the FSAR supplement Lower stress corrosion to (1) participate in the industry internals cracking programs for investigating and assem bly-to- managing aging effects on reactor thermal shield internals; (2) evaluate and implement bolts the results of the industry programs as Guide blocks applicable to the reactor internals; and bolts (3) upon completion of these programs,

< Shock pads but not less than 24 months before OJ t

OJ bolts entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

(f)

(D

~

3 0-

~

'"oo U1 OAGI0000203_462

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox 3

0-

~ Structure Ag ing Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B4-33 IVB46-g Lower grid Stainless Reactor coolant Loss of preload! No further aging management review is No, but assembly steel; nickel stress relaxation necessary if the applicant provides a licensee (R-207) alloy commitment in the FSAR supplement commitment Lower grid rib- to (1) participate in the industry needs to be to-shell forging programs for investigating and confirmed screws managing aging effects on reactor Lower internals; (2) evaluate and implement internals the results of the industry programs as assem bly-to- applicable to the reactor internals; and thermal shield (3) upon completion of these programs, bolts but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

z C

0 rn

~

OJ

~

OAGI0000203_463

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM z B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox C

0

~

OJ Item Link Structure and/or Material Environment Ag ing Effect!

Aging Management Program (AMP)

Further

~ Component Mechanism Evaluation IV.B4-34 IV.B4.1-a Plenum cover Stainless Reactor coolant Cracki ng/ stress Chapter XI. M2, Water Chemistry" for No, but and plenum steel corrosion cracking, PWR primary water licensee (R-172) cylinder irradiation-assisted commitment stress corrosion No further aging management review is needs to be Plenum cover cracking necessary if the applicant provides a confirmed assembly commitment in the FSAR supplement Plenum to (1) participate in the industry cylinder programs for investigating and Reinforcing managing aging effects on reactor plates internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs,

< but not less than 24 months before OJ t

OJ entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

(f)

(D

~

3 0-

~

'"oo U1 OAGI0000203_464

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox 3

0-

~ Structure Ag ing Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B4-35 IV.B4.1-c Plenum cover Stainless Reactor coolant Changes in No further aging management review is No, but and plenum steel dimensions! void necessary if the applicant provides a licensee (R-174) cylinder swelling commitment in the FSAR supplement commitment to (1) participate in the industry needs to be Plenum cover programs for investigating and confirmed assembly managing aging effects on reactor Plenum internals; (2) evaluate and implement cylinder the results of the industry programs as Reinforcing applicable to the reactor internals; and plates (3) upon completion of these programs, Top flange-to- but not less than 24 months before cover bolts entering the period of extended OJ Bottom flange- operation, submit an inspection plan for t

CD to-upper grid screws reactor internals to the NRC for review and approval.

z C

0 rn

~

OJ

~

OAGI0000203_465

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM z B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox C

0

~

OJ Item Link Structure and!or Material Environment Ag ing Effect!

Aging Management Program (AMP)

Further

~ Component Mechanism Evaluation IV.B4-36 IV.B4.1-b Plenum cover Stainless Reactor coolant Cracki ng! stress Chapter XI.M2, "Water Chemistry" for No, but and plenum steel corrosion cracking, PWR primary water licensee (R-173) cylinder irradiation-assisted commitment stress corrosion No further aging management review is needs to be Top flange-to- cracking necessary if the applicant provides a confirmed cover bolts commitment in the FSAR supplement Bottom flange- to (1) participate in the industry to-upper grid programs for investigating and screws managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs,

< but not less than 24 months before OJ to entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

IV.B4-37 IV.B4.3-f Reactor vessel Stainless Reactor coolant Cumulative fatigue Fatigue is a time-limited aging analysis Yes, IV.B4.5-f internals steel; nickel damage! fatigue (TLAA) to be evaluated for the period of TLAA (R-53) IV.B4.6-f components alloy extended operation. See the Standard IV.B4.2-d Review Plan, Section 4.3 "Metal IV.B4.1-d Fatigue," for acceptable methods for IV.B4.4-e meeting the requirements of 10 CFR 5421(c)(1)

IV.B4-38 IVB4. Reactor vessel Stainless Reactor coolant Loss of material! Chapter XI.M2, "Water Chemistry," for No internals steel; nickel pitting and crevice PWR primary water (f)

(D (RP-24) components alloy corrosion

~

3 0-

~

'"oo U1 OAGI0000203_466

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox 3

0-

~ Structure Ag ing Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B4-39 IV.B4.8-b Thermal shield Stainless Reactor coolant Changes in No further aging management review is No, but steel dimensions! void necessary if the applicant provides a licensee (R-215) swelling commitment in the FSAR supplement commitment to (1) participate in the industry needs to be programs for investigating and confirmed managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended OJ operation, submit an inspection plan for

~ reactor internals to the NRC for review and approval.

z C

0 rn

~

OJ

~

OAGI0000203_467

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM z B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox C

0

~

OJ Item Link Structure and/or Material Environment Ag ing Effect!

Aging Management Program (AMP)

Further

~ Component Mechanism Evaluation IV.B4-40 IV.B4.8-a Thermal shield Stainless Reactor coolant Cracki ng/ stress Chapter XI. M2, Water Chemistry" for No, but steel corrosion cracking, PWR primary water licensee (R-214) irradiation-assisted commitment stress corrosion No further aging management review is needs to be cracking necessary if the applicant provides a confirmed commitment in the FSAR supplement to (1) participate in the industry programs for investigating and managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs,

< but not less than 24 months before OJ t entering the period of extended

'" operation, submit an inspection plan for reactor internals to the NRC for review and approval.

(f)

(D

~

3 0-

~

'"oo U1 OAGI0000203_468

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox 3

0-

~ Structure Ag ing Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B4-41 IV.B4.8-c Thermal shield Stainless Reactor coolant Loss of fracture No further aging management review is No, but steel and neutron flux toughness! neutron necessary if the applicant provides a licensee (R-216) irradiation commitment in the FSAR supplement commitment embrittlement, void to (1) participate in the industry needs to be swelling programs for investigating and confirmed managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended OJ operation, submit an inspection plan for tw reactor internals to the NRC for review and approval.

IV.B4-42 IV.B4.2-f Upper grid Stainless Reactor coolant Loss of material! Chapter XI. M1, "ASME Section XI No assembly steel wear Inservice Inspection, Subsections IWB, (R-179) IWC, and IWO," for Class 1 Fuel assembly components support pads Plenum rib pads z

C

0 rn

~

OJ

~

OAGI0000203_469

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM z B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox C

0

~

OJ Item Link Structure and/or Material Environment Ag ing Effect!

Aging Management Program (AMP)

Further

~ Component Mechanism Evaluation IV.B4-43 IV.B4.2-b Upper grid Stainless Reactor coolant Cracki ng/ stress Chapter XI. M2, Water Chemistry" for No, but assembly steel corrosion cracking, PWR primary water licensee (R-176) irradiation-assisted commitment Rib- to-ring stress corrosion No further aging management review is needs to be screws cracking necessary if the applicant provides a confirmed commitment in the FSAR supplement to (1) participate in the industry programs for investigating and managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs,

< but not less than 24 months before OJ t-" entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

(f)

(D

~

3 0-

~

'"oo U1 OAG10000203_470

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox 3

0-

~ Structure Ag ing Effect! Further

'"oo Item Link and/or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.B4-44 IV.B4.2-a Upper grid Stainless Reactor coolant Cracki ng/ stress Chapter XI.M2, Water Chemistry" for No, but assembly steel corrosion cracking, PWR primary water licensee (R-175) irradiation-assisted commitment Upper grid rib stress corrosion No further aging management review is needs to be section cracking necessary if the applicant provides a confirmed Upper grid ring commitment in the FSAR supplement forging to (1) participate in the industry Fuel assembly programs for investigating and support pads managing aging effects on reactor Plenum rib internals; (2) evaluate and implement pads the results of the industry programs as applicable to the reactor internals; and OJ (3) upon completion of these programs, t

U1 but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

z C

0 rn

~

OJ

~

OAGI0000203_471

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM z B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox C

0

~

OJ Item Link Structure and!or Material Environment Ag ing Effect!

Aging Management Program (AMP)

Further

~ Component Mechanism Evaluation IV.B4-45 IV.B4.2-c Upper grid Stainless Reactor coolant Changes in No further aging management review is No, but assembly steel dimensions! void necessary if the applicant provides a licensee (R-177) swelling commitment in the FSAR supplement commitment Upper grid rib to (1) participate in the industry needs to be section programs for investigating and confirmed Upper grid ring managing aging effects on reactor forging internals; (2) evaluate and implement Fuel assembly the results of the industry programs as support pads applicable to the reactor internals; and Plenum rib (3) upon completion of these programs, pads but not less than 24 months before Rib-to-ring entering the period of extended screws operation, submit an inspection plan for

< reactor internals to the NRC for review OJ t

OJ and approval.

IV.B4-46 IV.B4.2-e Upper grid Stainless Reactor coolant Loss of fracture No further aging management review is No, but assembly steel and neutron flux toughness! neutron necessary if the applicant provides a licensee (R-178) irradiation commitment in the FSAR supplement commitment Upper grid rib embrittlement, void to (1) participate in the industry needs to be section swelling programs for investigating and confirmed Upper grid ring managing aging effects on reactor forging internals; (2) evaluate and implement Fuel assembly the results of the industry programs as support pads applicable to the reactor internals; and Plenum rib (3) upon completion of these programs, pads but not less than 24 months before (f) Rib-to-ring entering the period of extended (D

screws operation, submit an inspection plan for

~ reactor internals to the NRC for review 3

0- and approval.

~

'"oo U1 OAGI0000203_472

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 (IC), 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 extended period of operation are included in IV.E.

System Interfaces The systems that interface with the reactor coolant pressure boundary include the reactor pressure vessel (IV.A1), the emergency core cooling system (V.D2), 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 (VIII.B2), and the feedwater system (VIII.D2).

September 2005 IVC1-1 NUREG-1801, Rev. 1 OAGI0000203_473

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

0 C1 Reactor Coolant Pressure Boundary (BWR) m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< IV.C1-1 IV.C1.1-i Chapter XLM1, "ASME Section XI Class 1 piping, Stainless Reactor coolant Cracking! stress No fillings and steel; steel corrosion cracking, Inservice Inspection, Subsections IWB, (R-03) branch intergranular stres, IWC, and IWO," for Class 1 connections corrosion cracking components, and

< nominal pipe (for stainless steel sizes (NPS) 4 only), and thermal Chapter XLM2, 'Water Chemistry," for and mechanical BWR water and loading XLM35, "One-Time Inspection of ASME Code Class 1 Small-bore Piping" IV.C1-2 IVC11-g Class 1 piping, Cast Reactor coolant Loss of fracture Chapter XLM12, "Thermal Aging No piping austenitic >250"C (>482"F) toughness! thermal Embritllement of Cast Austenitic (R-52) com ponents, stainless aging Stainless Steel (CASS)"

and piping steel embrilliement elements OAGI0000203_474

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D C1 Reactor Coolant Pressure Boundary (BWR) 3 0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.C1-3 IV.C1.2-c Class 1 pump Cast Reactor coolant Loss of fracture Chapter XI M 1, "ASME Section XI No IV.C1.3-b casings, and austenitic >250°C (>482°F) toughness! thermal Inservice Inspection, Subsections IWB, (R-08) valve bodies stainless aging IWC, and IWO," for Class 1 components and bonnets steel embritllement 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 thermal aging embritllement of CASS pump casings and valve bodies.

Alternatively, the requirements of ASME Code Case N-481 for pump casings are sufficient for managing the effects of loss of fracture toughness due to thermal aging embritllement of CASS pump casings.

z C

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OAGI0000203_475

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

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Mechanism Evaluation

0 (D

Component

< Chapter XI.M1, "ASME Section XI IV.C1-4 IV.C1.4-a Isolation Stainless Reactor coolant Cracking! stress Yes, detection condenser steel corrosion cracking, Inservice Inspection, Subsections IWB, of aging effect (R-15) components intergranular stres, IWC, and IWO," for Class 1 components isto be corrosion cracking evaluated and Chapter XI. M2, 'Water Chemistry,"

for BWR water.

The AMP in Chapter XI.M1 is to be 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 acceptable verification program is to include temperature and radioactivity monitoring of the shell side water, and eddy current testing of tubes.

OAGI0000203_476

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D C1 Reactor Coolant Pressure Boundary (BWR) 3 0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.C1-5 IV.C1.4-a Isolation Stainless Reactor coolant Cracking! cyclic Chapter XL M1, "ASME Section XI Yes, detection condenser steel; steel loading Inservice Inspection, Subsections IWB, of aging effect (R-225) components IWC, and IWO," for Class 1 components isto be evaluated The AMP in Chapter XLM1 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 is to include temperature and radioactivity monitoring of the shell side water, and eddy current testing of tubes.

IV.C1-6 IV.C1.4-b Isolation Stainless Reactor coolant Loss of material! Chapter XLM2, 'Water Chemistry," for Yes, detection condenser steel; steel general (steel BWRwater of ag ing effects (R-16) components only), pitting and isto be crevice corrosion The AMP is to be augmented by evaluated verifying the effectiveness of water chemistry controL See Chapter XLM32, "One-Time Inspection," for an acceptable verification program.

z C IV.C1-7 IV.C1.1-c Piping, piping Steel Reactor coolant Wall thinning! flow- Chapter XL M17, "Flow-Accelerated No

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~ (R-23) IV.C1.3-a and piping corrosion OJ elements

~

OAGI0000203_477

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

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Mechanism Evaluation

0 (D

Component

< IV.C1-8 IV.C1.1-f Chapter XI. M7, "BWR Stress Corrosion Piping, piping Nickel alloy Reactor coolant Cracking! stress No com ponents, corrosion cracking Cracking," and (R-21) and piping and intergranular elements stress corrosion Chapter XI. M2, "Wffier Chemistry," for greater than or cracking BWRwater equal to 4 NPS IV.C1-9 IV.C1.2-b Piping, piping Stainless Reactor coolant Cracking! stress Chapter XI. M7, "BWR Stress Corrosion No IV.C1.1-f com ponents, steel corrosion cracking Cracking," and (R-20) IV.C1.3-c and piping and intergranular elements stress corrosion Chapter XI.M2, 'Water Chemistry," for greater than or cracking BWRwater equal to 4 NPS IV.C1-10 IV.C1.2-e Pump and Low-alloy System Loss of preload! Chapter XI.M18, "Bolting Integrity" No IV.C1.3-f valve closure steel SA 193 temperature up thermal effects, (R-27) bolting Gr. B7 to 28SOC (550°F) gasket creep, and self-loosening IV.C1-11 IV.C1.2-f Pump and Steel System Cumulative fatigue Fatigue is a time-limited aging analysis Yes, IVC13-g valve closure temperature up damage! fatigue (TLAA) to be performed for the period of TLAA (R-28) bolting to 28SOC (550°F) extended operation; check ASME Code limits for allowable cycles (less than 7000 cycles) of thermal stress range.

See the Standard Review Plan, Section 4.3, "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR 5421(c)(1)

IV.C1-12 IV.C1.3-e Pump and Steel System Loss of material! Chapter XI.M18, "Bolting Integrity" No IV.C1.2-d valve closure temperature up wear (R-26) bolting to 28SOC (550°F)

OAGI0000203_478

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D C1 Reactor Coolant Pressure Boundary (BWR) 3 0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IV.C1-13 IV.C1.2-d Pump and Stainless System Loss of material! Chapter XI.M18, "Bolting Integrity" No IV.C1.3-e valve seal steel; steel temperature up wear (R-29) flange closure to 28SOC (550°F) bolting IV.C1-14 IV.C1. Reactor Steel with Reactor coolant Loss of material! Chapter XI.M2, 'Water Chemistry," for Yes, detection coolant stainless pitting and crevice BWRwater of aging effect (RP-27) pressure steel or corrosion isto be boundary nickel alloy The AMP is to be augmented by evaluated components cladding; verifying the effectiveness of water stainless chemistry control. See Chapter XI.M32, steel; nickel "One-Time Inspection," for an alloy acceptable verification program.

IV.C1-15 IV.C1.1-e Reactor Steel; Reactor coolant Cumulative fatigue Fatigue is a time-limited aging analysis Yes, IV.C1.1-b coolant stainless damage! fatigue (TLAA) to be performed for the period of TLAA (R-220) IV.C1.1-h pressure steel; steel extended operation, and for Class 1 IV.C1.3-d boundary with nickel- components, environmental effects on IV.C1.1-e components: alloy or fatigue are to be addressed. See the IV.C1.1-h Piping, piping stainless Standard Review Plan, Section 4.3 IV.C1.1-d com ponents, steel "Metal Fatigue," for acceptable methods IV.C1.1-h and piping cladding; for meeting the requirements of 10 CFR IV.C1.2-a elements nickel-alloy 5421(c)(1)

IV.C1.2-a IV.C1.1-h IV.C1.2-a z IV.C1.1-h C

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~ IV.C1.3-d OAGI0000203_479

This Page Intentionally Left Blank NUREG-1S01, Rev. 1 IVC1-S September 2005 OAG10000203_480

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 extended period of operation are included in IV.E.

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

September 2005 IVC2-1 NUREG-1801, Rev. 1 OAGI0000203_481

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

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OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< IV.C2-1 IVC21-g Chapter XI. M1, "ASM E Section XI Class 1 piping, Stainless Reactor coolant Cracking! stress No IV.C2.2-h fittings and steel; steel corrosion Inservice Inspection, Subsections IWB, (R-02) branch with cracking, thermal IWC, and IWO," for Class 1 connections < stainless and mechanical components, and NPS 4 steel loading cladding Chapter XI.M2, "Water Chemistry," for PWR water and XI.M35, "One-Time Inspection of ASME Code Class 1 Small-bore Piping" IV.C2-2 IV.C2.5-h Class 1 piping, Stainless Reactor coolant Cracking! stress Chapter XI. M1, "ASM E Section XI No IV.C2.5-m fittings, primary steel; steel corrosion cracking Inservice Inspection, Subsections IWB, (R-07) IV.C2.2-f nozzles, safe with IWC, and IWO," for Class 1 ends, manways, stainless components and and flanges steel cladding Chapter XI.M2, "Water Chemistry," for PWR primary water OAGI0000203_482

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C2 Reactor Coolant System and Connected Lines (PWR)

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

Mechanism Evaluation Component IV,C2-3 IVC22-g Class 1 piping, Cast Reactor coolant Cracking! stress Monitoring and control of primary water Yes, plant-IV,C2,5-i piping austenitic corrosion cracking chemistry in accordance with the specific (R-05) IV,C2,1-e components, stainless guidelines in EPRI TR-105714 (Rev 3 and piping steel or later) minimize the potential of SCC, elements and material selection according to NUREG-0313, Rev, 2 guidelines d

=0,035% C and =7,5% ferrite reduces susceptibility to SCC For CASS components that do not meet either one of the above guidelines, a plant-specific aging

<: management program is to be o

evaluated, The program is to include

'"W (a) adequate inspection methods to ensure detection of cracks, and (b) flaw evaluation methodology for CASS components that are susceptible to thermal aging embritllemenl.

IV,C2-4 IV,C2,5-1 Class 1 piping, Cast Reactor coolant Loss of fracture Chapter XLM12, "Thermal Aging No IV,C2,1-f piping austenitic >250°C (>482°F) toughness! Embritllement of Cast Austenitic (R-52) IV,C2,2-e components, stainless thermal aging Stainless Steel (CASS)"

and piping steel embritllement elements z IV,C2-5 IV,C2,3-b Class 1 pump Stainless Reactor coolant Cracking! stress Chapter XL M1, "ASM E Section XI No C

0 IV,C2A-b casings and steel; steel corrosion cracking Inservice Inspection, Subsections IWB, m (R-09) valve bod ies with IWC, and IWO," for Class 1

~

OJ stainless components and

~ steel cladding Chapter XL M2, "Water Chemistry," for PWR primary water OAGI0000203_483

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

0 C2 Reactor Coolant System and Connected Lines (PWR) m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< IV.C2-6 IV.C2.3-c Chapter XI. M1, "ASM E Section XI Class 1 pump Cast Reactor coolant Loss of fracture No IV.C2A-c casi ngs, and austenitic >250°C (>482°F) toughness! Inservice Inspection, Subsections IWB, (R-08) valve bodies and stainless thermal aging IWC, and IWO," for Class 1 bonnets steel embritllement components 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 thermal aging embritllement of CASS pump casings and valve bodies.

Alternatively, the requirements of ASME Code Case N-481 for pump casings are sufficient for managing the effects of loss of fracture toughness due to thermal aging embritllement of CASS pump casings.

IV.C2-7 IV.C2.5-n Closure bolting High- Air with reactor Cracking! stress Chapter XI. M 18, "Bolting Integrity" No IV.C2.3-e strength low- coolant leakage corrosion cracking (R-11) IV.C2A-e alloy steel, stainless steel IV.C2-8 IVC2A-g Closure bolting Low-alloy Air with reactor Loss of preload! Chapter XI. M 18, "Bolting Integrity" No IVC25-p steel, coolant leakage thermal effects, (R-12) IVC23-g stainless gasket creep, and steel self-loosening OAGI0000203_484

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C2 Reactor Coolant System and Connected Lines (PWR)

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

Mechanism Evaluation Component IV.C2-9 IV.C2.3-f External Steel Air with borated Loss of material! Chapter XI.M10, "Boric Acid Corrosion" No IV.C22-d surfaces water leakage boric acid (R-17) IV.C2.1-d corrosion IV.C2.5-b IV.C2.5-u IV.C2.6-b IV.C2A-f IV.C2.5-0 IV.C2-10 IV.C2.5-w Piping and Stainless System Cumulative fatigue Fatigue is a time-limited aging analysis Yes, IV.C2.3-d components steel; steel temperature up damage! fatigue (TLAA) to be evaluated for the period of TLAA (R-18) IV.C2A-d external to 340°C (644°F) extended operation. See the Standard IV.C2.5-t surfaces and Review Plan, Section 4.3 "Metal bolting Fatigue," for acceptable methods for meeting the requirements of 10 CFR 5421(c)(1)

IV.C2-11 IV.C2. Piping, piping Copper alloy Closed cycle Loss of material! Chapter XI.M21, "Closed-Cycle Cooling No components, cooling water pitting, crevice, Water System" (RP-11) and piping and galvanic elements corrosion IV.C2-12 IV.C2. Piping, piping Copper alloy Closed cycle Loss of material! Chapter XI. M33, "Selective Leaching of No components, >15% Zn cooling water selective leaching Materials" (RP-12) and piping elements z

C

0 m

~

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~

OAGI0000203_485

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

0 C2 Reactor Coolant System and Connected Lines (PWR) m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< IV.C2-13 IV.C2. Piping, piping Nickel alloy Reactor coolant! Cracking! primary Chapter XI. M1, "ASM E Section XI No, but components, steam water stress Inservice Inspection, Subsections IWB, licensee (RP-31) and piping corrosion cracking IWC, and IWD" for Class 1 commitment elements components, and needs to be confirmed Chapter XI.M2, "Water Chemistry" for PWR primary water and Comply with applicable NRC Orders and provide a commitment in the FSAR supplement to implement applicable (1)

Bulletins and Generic Letters and (2) staff-accepted industry guidelines.

IV.C2-14 IV.C2. Piping, piping Steel Closed cycle Loss of material! Chapter XI.M21, "Closed-Cycle Cooling No components, cooling water general, pitting, Water System" (RP-10) and piping and crevice elements corrosion IV.C2-15 IV.C2. Piping, piping Steel with Reactor coolant Loss of material! Chapter XI.M2, "Water Chemistry," for No components, stainless pitting and crevice PWR primary water (RP-23) and piping steel or corrosion elements; nickel alloy flanges; heater cladding; sheaths and stainless sleeves; steel; nickel penetrations; alloy thermal sleeves; vessel shell heads and welds OAGI0000203_486

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C2 Reactor Coolant System and Connected Lines (PWR)

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

Mechanism Evaluation Component IV.C2-16 IV.C2.5-v Pressurizer Stainless Air with metal Cracking! cyclic Chapter XI. M1, "ASM E Section XI No steel; steel temperature up loading Inservice Inspection, Subsections IWB, (R-19) Integral to 288°C (550°F) IWC, and IWD," for Class 1 support components IV.C2-17 IVC25-j Pressurizer Nickel alloy; Reactor coolant Cracking! stress Chapter XI. M2, "Water Chemistry," and No, unless stainless corrosion Chapter XI. M32 "One-Time Inspection" licensee (R-24) Spray head steel cracking, primary and commitment water stress needs to be corrosion cracking For nickel alloy welded spray heads, confirmed comply with applicable NRC Orders and provide a commitment in the FSAR supplement to implement applicable (1)

Bulletins and Generic Letters and (2) staff-accepted industry guidelines.

z C

0 m

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OAGI0000203_487

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

0 C2 Reactor Coolant System and Connected Lines (PWR) m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< IV,C2-18 IV,C2,5-c Chapter XI. M1, "ASM E Section XI Pressurizer Steel with Reactor coolant Cracking! cyclic No IVC25-g components stainless loading In service Inspection, Subsections IWB, (R-58) steel or IWC, and IWO," for Class 1 nickel alloy components and cladding; or stainless Chapter XI.M2, "Water Chemistry," for steel PWR primary water Cracks in the pressurizer cladding 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, IV,C2-19 IVC25-g Pressurizer Steel with Reactor coolant Cracking! stress Chapter XI.M1, "ASME Section XI No IV,C2,5-c components stainless corrosion Inservice Inspection, Subsections IWB, (R-25) steel or cracking, primary IWC, and IWD" for Class 1 nickel alloy water stress components, and cladding; or corrosion cracking stainless Chapter XI.M2, "Water Chemistry," for steel PWR primary water OAGI0000203_488

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C2 Reactor Coolant System and Connected Lines (PWR)

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

Mechanism Evaluation Component IV.C2-20 IV.C2.5-r Pressurizer Stainless Reactor coolant Cracking! stress Chapter XI. M1, "ASM E Section XI No heater sheaths steel corrosion cracking Inservice Inspection, Subsections IWB, (R-217) and sleeves, lwe, and IWO," for Class 1 and heater components and bundle diaphragm plate Chapter XI.M2, "Water Chemistry," for PWR primary water IV.C2-21 IV.C2.5-m Pressurizer Nickel alloy Reactor coolant Cracking! primary Chapter XI. M1, "ASM E Section XI No, but IV.C2.5-k instrumentation or nickel water stress Inservice Inspection, Subsections IWB, licensee (R-06) IV.C2.5-s penetrations, alloy corrosion cracking lwe, and IWD" for Class 1 commitment heater sheaths cladding components, and needs to be

<: and sleeves, confirmed o heater bundle Chapter XI.M2, "Water Chemistry," for

'"cD diaphragm plate, PWR primary water and and manways and flanges For nickel alloy, comply with applicable NRC Orders and provide a commitment in the FSAR supplement to implement applicable (1) Bulletins and Generic Leiters and (2) staff-accepted industry guidelines.

IV.C2-22 IV.C2.6-c Pressurizer Stainless Treated borated Cracking! stress Chapter XI. M1, "ASM E Section XI No relief tank steel; steel water >60°C corrosion cracking Inservice Inspection, Subsections IWB, (R-14) with (>140°F) lwe, and IWO," for Class 1 z Tank shell and stainless components and C

0 heads steel m Flanges and cladding Chapter XI.M2, "Water Chemistry," for

~

OJ nozzles PWR primary water

~

OAGI0000203_489

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

0 C2 Reactor Coolant System and Connected Lines (PWR) m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< IV.C2-23 IV.C2.6-a Pressurizer Steel with Treated borated Cumulative fatigue Fatigue is a time-limited aging analysis Yes, relief tank stainless water damage! fatigue (TLAA) to be evaluated for the period of TLAA (R-13) steel extended operation. See the Standard Tank shell and cladding Review Plan, Section 4.3 "Metal heads Fatigue," for acceptable methods for Flanges and meeting the requirements of 10 CFR nozzles 5421(c)(1)

IV.C2-24 IV.C2. Pressurizer Nickel alloy Reactor coolant! Cracking! primary Chapter XI. M1, "ASM E Section XI No, but surge and steam water stress Inservice Inspection, Subsections IWB, licensee (RP-22) steam space corrosion cracking IWC, and IWD" for Class 1 commitment nozzles, and com ponents, and needs to be

<: welds confirmed o Chapter XI.M2, "Water Chemistry," for t;: PWR primary water and o

Comply with applicable NRC Orders and provide a commitment in the FSAR supplement to implement applicable (1)

Bulletins and Generic Letters and (2) staff-accepted industry guidelines.

OAG10000203_490

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C2 Reactor Coolant System and Connected Lines (PWR)

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

Mechanism Evaluation Component IV,C2-25 IV,C2,5-f Reactor coolant Steel; Reactor coolant Cumulative fatigue Fatigue is a time-limited aging analysis Yes, IV,C2A-a pressure stainless damage! fatigue (TLAA) to be performed for the period of TLAA (R-223) IV,C2,2-a boundary steel; steel extended operation, and, for Class 1 IV,C2,5-f components: with nickel- components, environmental effects on IV,C2,3-a alloy or fatigue are to be addressed, See the IV,C2,2-b Piping, piping stainless Standard Review Plan, Section 4,3 IV,C2,1-b components, steel "Metal Fatigue," for acceptable IV,C2,1-b and piping cladding; methods for meeting the requirements IV,C2,5-a elements; nickel-alloy of 10 CFR 5421(c)(1)

IV,C2,1-a Flanges; IV,C2,1-a Nozzles and IV,C2,5-d safe ends; o IV,C2,2-a Pressurizer t;: IV,C2,2-a vessel shell IV,C2A-a heads and IV,C2,2-b welds; IV,C2,3-a Heater sheaths IVC25-q and sleeves; IV,C2,2-c Penetrations; IV,C2,5-d and IV,C2,5-e Thermal IV,C2,5-f sleeves IV,C2,2-a z

C

0 m

~

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OAGI0000203_491

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

0 C2 Reactor Coolant System and Connected Lines (PWR) m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< IV.C2-26 IV.C2.1-c Chapter XI. M1, "ASM E Section XI Reactor coolant Stainless Reactor coolant Cracking! cyclic No system piping steel; steel loading Inservice Inspection, Subsections IWB, (R-56) and fittings with IWC, and IWO," for Class 1 stainless components Cold leg steel Hot leg cladding Surge line Spray line IV.C2-27 IV.C2.1-c Reactor coolant Stainless Reactor coolant Cracking! stress Chapter XI. M1, "ASM E Section XI No system piping steel; steel corrosion cracking Inservice Inspection, Subsections IWB, (R-30) and fittings with IWC, and IWO," for Class 1 stainless components and Cold leg steel Hot leg cladding Chapter XI.M2, "Water Chemistry," for Surge line PWR primary water Spray line OAGI0000203_492

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 extended period of operation are included in IV.E.

System Interfaces The systems that interface with the steam generators include the reactor coolant system and connected lines (IV.C2), the containment isolation components (V.C), the main steam system (VIII.B1), the feedwater system (V1I1.01), the steam generator blowdown system (VII I. F), and the auxiliary feedwater system (VIII.G).

September 2005 IV 01-1 NUREG-1801, Rev. 1 OAGI0000203_493

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

0 01 Steam Generator (Recirculating)

~

OJ Structure Aging Effect! Further

~ Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation Component IVD1-1 IV.D1.1-i Class 1 piping, Stainless Reactor coolant Cracking! stress Chapter XI. M1, "ASM E Section XI No fittings, primary steel; steel corrosion cracking Inservice Inspection, Subsections IWB, (R-07) nozzles, safe with IWC, and IWO," for Class 1 ends, manways, stainless components and and flanges steel cladding Chapter XI. M2, "Water Chemistry," for PWR primary water IVD1-2 IV.D1.1-1 Closure bolting Steel Air with reactor Cracking! stress Chapter XI. M 18, "Bolting Integrity" No coolant leakage corrosion cracking (R-10)

IVD1-3 IV.D1.1-k External Steel Air with borated Loss of material! Chapter XI.M10, "Boric Acid Corrosion" No IVD11-g surfaces water leakage boric acid R-17) corrosion IVD1-4 IV.D1.1-i Instrument Nickel alloy; Reactor coolant Cracking! primary Chapter XI.M1, "ASME Section XI No, but IVD11-j penetrations steel with water stress Inservice Inspection, Subsections IWB, licensee (R-01) and primary nickel-alloy corrosion cracking IWC, and IWO," for Class 1 commitment side nozzles, cladding components, and needs to be safe ends, and confirmed welds Chapter XI. M2, "Water Chemistry," for PWR primary water and For nickel alloy, comply with applicable NRC Orders and provide a commitment in the FSAR supplement to submit a plant-specific AMP to implement applicable (1) Bulletins and Generic (f)

(D Letters and (2) staff-accepted industry

~ IQuidelines.

3 0-

~

'"oo U1 OAGI0000203_494

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D 01 Steam Generator (Recirculating) 3 0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IVD1-5 IV.D1.1-d Pressure Steel Secondary Wall thinning! Chapter XI.M17, "Flow-Accelerated No boundary and feedwater! steam flow-accelerated Corrosion" (R-37) structural corrosion Steam nozzle and safe end FW nozzle and safe end IVD1-6 IVD1. Primary side Nickel alloy; Reactor coolant Cracking! primary Chapter XI. M2, "Water Chemistry," for No steel with water stress PWR primary water (RP-21) Divider Plate nickel-alloy corrosion cracking cladding IVD1-7 IVD1. Primary side Stainless Reactor coolant Cracking! stress Chapter XI. M2, "Water Chemistry," for No steel corrosion cracking PWR primary water (RP-17) Divider Plate IVD1-8 IV.D1.1-h Recirculating Steel; Reactor coolant Cumulative fatigue Fatigue is a time-limited aging analysis Yes, steam generator stainless damage! fatigue (TLAA) to be performed for the period of TLAA (R-221) components: steel; steel extended operation, and, for Class 1 with nickel- components, environmental effects on Flanges; alloy or fatigue are to be addressed. See the Penetrations; stainless Standard Review Plan, Section 4.3 Nozzles; steel "Metal Fatigue," for acceptable Safe ends, cladding; methods for meeting the requirements z lower heads nickel-alloy of 10 CFR 5421(c)(1)

C and welds

0 m

~

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OAGI0000203_495

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

0 01 Steam Generator (Recirculating)

~

OJ Structure Aging Effect! Further

~ Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation Component IVD1-9 IVD1. Steam Steel Secondary Loss of material! Chapter XI.M19, "Steam Generator No generator feedwater! steam erosion, general, Tubing Integrity" and (RP-16) pitting, and Tube bundle crevice corrosion Chapter XI. M2, "Water Chemistry," for wrapper PWR secondary water IVD1-10 IV.01.1-f Steam Steel System Loss of preload! Chapter XI. M 18, "Bolting Integrity" No generator temperature up to thermal effects, (R-32) closure bolting 340°C (644°F) gasket creep, and self-loosening IVD1-11 IV.01.1-b Steam Steel Secondary Cumulative fatigue Fatigue is a time-limited aging analysis Yes, IV.01.1-a generator feedwater! steam damage! fatigue (TLAA) to be evaluated for the period of TLAA (R-33) components extended operation. See the Standard Review Plan, Section 4.3 "Metal Top head; Fatigue," for acceptable methods for Steam nozzle meeting the requirements of 10 CFR and safe end; 5421(c)(1)

Upper and lower shell; Feedwater and auxiliary feedwater nozzle and safe end; feedwater impingement plate and (f) support (D

~

3 0-

~

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(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D 01 Steam Generator (Recirculating) 3 0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IVD1-12 IV.01.1-c Steam Steel Secondary Loss of material! Chapter XI. M1, "ASM E Section XI Yes, detection generator feedwater! steam general, pitting, Inservice Inspection, Subsections IWB, of aging effects (R-34) components and crevice IWC, and IWO," for Class 2 isto be corrosion components and evaluated Upper and lower shell, Chapter XI. M2, "Water Chemistry," for and transition PWR secondary water. As noted in cone NRC IN 90-04, if general and pitting corrosion of the shell is known to exist, the AMP guidelines in Chapter XI.M1 may 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 to Westinghouse Model 44 and 51 Steam Generators where a high stress region exists at the shell to transition cone weld.

IV.01-13 IV.01.1-e Steam Steel Secondary Loss of material! A plant-specific aging management Yes, plant-generator feedwater erosion program is to be evaluated. specific (R-39) feedwater impingement plate and z support C

0 IVD1-14 IVD1. Steam Chrome Secondary Cracking! stress Chapter XI.M19, "Steam Generator No m

generator plated steel; feedwater! steam corrosion cracking Tubing Integrity" and

~

OJ (RP-14) structural stainless

~ steel; Nickel Chapter XI. M2, "Water Chemistry," for Anti-vibration alloy PWR secondary water bars OAGI0000203_497

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

0 01 Steam Generator (Recirculating)

~

OJ Structure Aging Effect! Further

~ Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation Component IVD1-15 IVD1. Steam Chrome Secondary Loss of material! Chapter XI.M19, "Steam Generator No generator plated steel; feedwater! steam crevice corrosion Tubing Integrity" and (RP-15) structural stainless and frelting steel; Nickel Chapter XI. M2, "Water Chemistry," for Anti-vibration alloy PWR secondary water bars IVD1-16 IV.01.2-h Steam Steel Secondary Wall thinning! Chapter XI.M19, "Steam Generator No generator feedwater! steam flow-accelerated Tubing Integrity" and (R-41) structural corrosion Chapter XI. M2, "Water Chemistry," for Tube support PWR secondary water laltice bars IVD1-17 IV.01.2-k Steam Steel Secondary Ligament Chapter XI.M19, "Steam Generator No generator feedwater! steam cracking! Tubing Integrity" and (R-42) structural corrosion Chapter XI. M2, "Water Chemistry," for Tube support PWR secondary water plates IVD1-18 IV012-j Tube plugs Nickel alloy Reactor coolant Cracking! primary Chapter XI.M19, "Steam Generator No IV.01.2-i water stress Tubing Integrity" and (R-4Q) corrosion cracking Chapter XI. M2, "Water Chemistry," for PWR primary water (f)

(D

~

3 0-

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(f)

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~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IVD1-19 IVD12-g Tubes Nickel alloy Secondary Denting! corrosion Chapter XLM19, "Steam Generator No feedwater! steam of carbon steel Tubing Integrity" and (R-43) tube support plate Chapter XL M2, "Water Chemistry," for PWR secondary water.

For plants that could experience denting at the upper support plates, the applicant should evaluate potential for rapidly propagating cracks and then develop and take corrective actions consistent with Bulletin 88-02, "Rapidly Propagating Cracks in SG Tubes."

IVD1-20 IV.D1.2-a Tubes and Nickel alloy Reactor coolant Cracking! primary Chapter XLM19, "Steam Generator No sleeves water stress Tubing Integrity" and (R-44) corrosion cracking Chapter XL M2, "Water Chemistry," for PWR primary water IVD1-21 IV.D1.2-d Tubes and Nickel alloy Reactor coolant Cumulative fatigue Fatigue is a time-limited aging analysis Yes, sleeves and secondary damage! fatigue (TLAA) to be evaluated for the period of TLAA (R-46) feedwater!steam extended operation. See the Standard Review Plan, Section 4.3 "Metal Fatigue," for acceptable methods for z meeting the requirements of 10 CFR C 5421(c)(1)

0 m

IVD1-22 IV.D1.2-c Tubes and Nickel alloy Secondary Cracking! Chapter XLM19, "Steam Generator No

~

OJ sleeves feedwater! steam intergranular Tubing Integrity" and

~ (R-48) attack Chapter XL M2, "Water Chemistry," for PWR secondary water OAGI0000203_499

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

0 01 Steam Generator (Recirculating)

~

OJ Structure Aging Effect! Further

~ Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation Component IVD1-23 IV.D1.2-b Tubes and Nickel alloy Secondary Cracking! outer Chapter XI.M19, "Steam Generator No sleeves feedwater! steam diameter stress Tubing Integrity" and (R-47) corrosion cracking Chapter XI. M2, "Water Chemistry," for PWR secondary water IVD1-24 IV.D1.2-e Tubes and Nickel alloy Secondary Loss of material! Chapter XI.M19, "Steam Generator No sleeves feedwater! steam fretting and wear Tubing Integrity" and (R-49)

Chapter XI. M2, "Water Chemistry," for PWR secondary water IVD1-25 IV.D1.2-f Tubes and Nickel alloy Secondary Loss of material! Chapter XI.M19, "Steam Generator No sleeves feedwater! steam wastage and Tubing Integrity" and (R-50) (exposed to pitting corrosion phosphate Chapter XI. M2, "Water Chemistry," for chemistry) PWR secondary water IVD1-26 IV.D1.3-a Upper assembly Steel Secondary Wall thinning! A plant-specific aging management Yes, plant-and separators feedwater! steam flow-accelerated program is to be evaluated. Reference specific (R-51) corrosion NRC IN 91-19, "Steam Generator Feedwater Feedwater Distribution Piping inlet ring and Damage."

support (f)

(D

~

3 0-

~

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02. 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 extended period of operation are included in IV.E.

System Interfaces The systems that interface with the steam generators include the reactor coolant system and connected lines (IV.C2), the main steam system (VIII.B1), the feedwater system (V1I1.01), the steam generator blowdown system (VII I. F), and the auxiliary feedwater system (VIII.G).

September 2005 IV 02-1 NUREG-1801, Rev. 1 OAGI0000203_501

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

0 D2 Steam Generator (Once-Through) m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< IVD2-1 IVD21-j Chapter XI.M10, "Boric Acid Corrosion" No External Steel Air with borated Loss of material!

IV.D2.1-b surfaces water leakage boric acid (R-17) corrosion IVD2-2 IV.D2.1-h Instrument Nickel alloy; Reactor coolant Cracking! primary Chapter XI.M1, "ASME Section XI No, but penetrations anc steel with water stress Inservice Inspection, Subsections IWB, licensee (R-01) primary side nickel-alloy corrosion cracking IWC, and IWO," for Class 1 commitment nozzles, safe cladding com ponents, and needs to be ends, and welds confirmed Chapter XI. M2, "Water Chemistry," for PWR primary water and For nickel alloy, comply with applicable NRC Orders and provide a commitment in the FSAR supplement to submit a plant-specific AMP to implement applicable (1) Bulletins and Generic Leiters and (2) staff-accepted industry IQuidelines.

IVD2-3 IV.D2.1-c Once-through Steel; Reactor coolant Cumulative fatigue Fatigue is a time-limited aging analysis Yes, steam generator stainless damage! fatigue (TLAA) to be performed for the period of TLAA (R-222) components: steel; steel extended operation, and, for Class 1 with nickel- components, environmental effects on Primary side alloy or fatigue are to be addressed. See the nozzles stainless Standard Review Plan, Section 4.3 Safe ends and steel "Metal Fatigue," for acceptable welds cladding; methods for meeting the requirements nickel-alloy of 10 CFR 5421(c)(1)

OAGI0000203_502

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D D2 Steam Generator (Once-Through) 3 0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IVD2-4 IV.D2.1-a Primary side Steel with Reactor coolant Cracking! stress Chapter XI.M1, "ASME Section XI No, but components stainless corrosion Inservice Inspection, Subsections IWB, licensee (R-35) steel or cracking, primary IWC, and IWO," for Class 1 commitment Upper and nickel alloy water stress com ponents, and needs to be lower heads cladding corrosion cracking confirmed Tube sheets Chapter XI. M2, "Water Chemistry," for and tube-to- PWR primary water and tube sheet welds For nickel alloy, comply with applicable NRC Orders and provide a commitment in the FSAR supplement to implement applicable (1) Bulletins and Generic

<: Leiters and (2) staff-accepted industry o

guidelines.

'"W IVD2-5 IV.D2.1-1 Secondary Steel Air with leaking Loss of material! Chapter XI.M1, "ASME Section XI No manways and secondary-side erosion Inservice Inspection, Subsections IWB, (R-31) hand holes water and!or IWC, and IWO," for Class 2 (cover only) steam components IVD2-6 IV.D2.1-k Steam generato Steel System Loss of preload! Chapter XI. M 18, "Bolting Integrity" No closure bolting temperature up to thermal effects, (R-32) 340°C (644°F) gasket creep, and self-loosening z

C

0 m

~

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OAGI0000203_503

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

0 D2 Steam Generator (Once-Through) m

~

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~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< IVD2-7 IV.D2.1-f Chapter XI.M17, "Flow-Accelerated Steam generato Steel Secondary Wall thinning! No components feedwater! steam flow-accelerated Corrosion" (R-38) corrosion feedwater and auxiliary feedwater nozzles and safe ends Steam nozzles and safe ends IVD2-8 IV.D2.1-e Steam generato Steel Secondary Loss of material! Chapter XI. M2, "Water Chemistry," for Yes, detection components feedwater! steam general, pitting, PWR secondary water of aging effects (R-224) and crevice isto be Shell corrosion The AMP is to be augmented by evaluated assembly verifying the effectiveness of water chem istry control. See Chapter XI.M32, "One-Time Inspection," for an acceptable verification proQram.

IVD2-9 IV.D2.1-i Steam generato Nickel alloy Secondary Cracking! stress Chapter XI.M2, Water Chemistry" and No components feedwater! steam corrosion cracking (R-36) Chapter XI. M32 "One-Time Inspection" Such as or Chapter XI.M1, "ASME Section XI secondary sidE Inservice Inspection, Subsections IWB, nozzles (vent, IWC, and IWD" drain, and instrumentatio n)

OAGI0000203_504

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D D2 Steam Generator (Once-Through) 3 0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IVD2-10 IVD21-g Steam generato Steel Secondary Cumulative fatigue Fatigue is a time-limited aging analysis Yes, IV.D2.1-d components feedwater! steam damage! fatigue (TLAA) to be evaluated for the period of TLAA (R-33) extended operation. See the standard Top head; Review Plan, Section 4.3 "Metal Steam nozzle Fatigue," for acceptable methods for and safe end; meeting the requirements of 10 CFR Upper and 5421(c)(1) lower shell; FW and AFW nozzle and safe end; FW impingement plate and support IVD2-11 IVD2. Steam generato Steel Secondary Ligament Chapter XI.M19, "Steam Generator No structural feedwater! steam cracking! Tubing Integrity" and (R-42) corrosion Tube support Chapter XI. M2, "Water Chemistry," for plates PWR secondary water IVD2-12 IVD22-g Tube plugs Nickel alloy Reactor coolant Cracking! primary Chapter XI.M19, "Steam Generator No z IV.D2.2-f water stress Tubing Integrity" and C

0 (R-40) corrosion cracking m Chapter XI. M2, "Water Chemistry," for

~

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~

OAGI0000203_505

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

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~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< IVD2-13 IVD2. Tubes Nickel alloy Secondary Denting! corrosion Chapter XI.M19, "Steam Generator No feedwater! steam of carbon steel Tubing Integrity" and (R-226) tube support plate Chapter XI. M2, "Water Chemistry," for PWR secondary water.

IVD2-14 IV.D2.2-a Tubes and Nickel alloy Reactor coolant Cracking! primary Chapter XI.M19, "Steam Generator No sleeves water stress Tubing Integrity" and (R-44) corrosion cracking Chapter XI. M2, "Water Chemistry," for PWR primary water IVD2-15 IV.D2.2-e Tubes and Nickel alloy Reactor coolant Cumulative fatigue Fatigue is a time-limited aging analysis Yes, sleeves and secondary damage! fatigue (TLAA) to be evaluated for the period of TLAA (R-46) feedwater!steam extended operation. See the standard Review Plan, Section 4.3 "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR 5421(c)(1)

IVD2-16 IV.D2.2-c Tubes and Nickel alloy Secondary Cracking! Chapter XI.M19, "Steam Generator No sleeves feedwater! steam intergranular Tubing Integrity" and (R-48) attack Chapter XI. M2, "Water Chemistry," for PWR secondary water IVD2-17 IV.D2.2-b Tubes and Nickel alloy Secondary Cracking! outer Chapter XI.M19, "Steam Generator No sleeves feedwater! steam diameter stress Tubing Integrity" and (R-47) corrosion cracking Chapter XI. M2, "Water Chemistry," for PWR secondary water OAGI0000203_506

(f)

(D IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM (D D2 Steam Generator (Once-Through) 3 0-

~ Structure Aging Effect! Further

'"oo Item Link and/or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component IVD2-18 IV.D2.2-d Tubes and Nickel alloy Secondary Loss of materiall Chapter XI.M19, "Steam Generator No sleeves feedwaterl steam fretting and wear Tubing Integrity" and (R-49)

Chapter XI. M2, "Water Chemistry," for PWR secondary water z

C

0 m

~

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OAGI0000203_507

This Page Intentionally Left Blank NUREG-1801, Rev. 1 IVD2-8 September 2005 OAGI0000203_508

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 extended period of 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.)

September 2005 IV E-1 NUREG-1801, Rev. 1 OAGI0000203_509

z C

0 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM m

E Common Miscellaneous Material Environment Combinations

~

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0 Item Link and/or Material Environment Aging Management Program (AMP)

(D Mechanism Evaluation

< Component IV.E-1 IV.E. Piping, piping Nickel alloy Air - indoor None None No components, uncontrolled (RP-03) and piping (External) elements IV.E-2 IV.E. Piping, piping Stainless Air - indoor None None No components, steel uncontrolled (RP-04) and piping (External) elements IV.E-3 IV.E. Piping, piping Stainless Air with borated None None No components, steel water leakage (RP-05) and piping m elements r0 IV.E-4 IV.E. Piping, piping Stainless Concrete None None No components, steel (RP-06) and piping elements IV.E-5 IV.E. Piping, piping Stainless Gas None None No components, steel (RP-07) and piping elements IV.E-6 IV.E. Piping, piping Steel Concrete None None No components, (RP-01) and piping elements OAGI0000203_510

CHAPTER V ENGINEERED SAFETY FEATURES September 2005 V-i NUREG-1801, Rev. 1 OAGI0000203_511

This Page Intentionally Left Blank NUREG-1801, Rev. 1 V-ii September 2005 OAGI0000203_512

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

B. Standby Gas Treatment System (Boiling Water Reactors)

C. Containment Isolation Components D1. Emergency Core Cooling System (Pressurized Water Reactors)

D2. Emergency Core Cooling System (Boiling Water Reactors)

E. External Surfaces of Components and Miscellaneous Bolting F. Common Miscellaneous Material/Environment Combinations September 2005 V-iii NUREG-1801, Rev. 1 OAGI0000203_513

This Page Intentionally Left Blank NUREG-1801, Rev. 1 V-iv September 2005 OAGI0000203_514

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 extended period of 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).

September 2005 VA-1 NUREG-1801. Rev. 1 OAGI0000203_515

z V ENGINEERED SAFETY FEATURES C

0 A Containment Spray System (PWR) m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< V.A-1 VA5-a Air - indoor Ducting, piping Steel Loss of material! Chapter XI. M36, "External Surfaces No VA2-a and uncontrolled general corrosion Monitoring" (E-26) components (External) external surfaces V.A-2 VA Encapsulation Steel Air - indoor Loss of material! Chapter XI. M38, "Inspection of Internal No Components uncontrolled general, pitting, and Surfaces in Miscellaneous Piping and (EP-42) (Internal) crevice corrosion Ducting Components" V.A-3 VA Encapsulation Steel Air with borated Loss of material! Chapter XI. M38, "Inspection of Internal No Components water leakage general, pitting, Surfaces in Miscellaneous Piping and (EP-43) (Internal) crevice and boric Ducting Components"

< acid corrosion

>> V.A-4 VA6-d External Steel Air with borated Loss of material! Chapter XI.M10, "Boric Acid Corrosion" No r0 VA4-b surfaces water leakage boric acid corrosion (E-28) VA3-b VA1-b VA5-b V.A-5 VA Heat Copper alloy Closed cycle Loss of material! Chapter XI. M21, "Closed-Cycle Cooling No exchanger cooling water pitting, crevice, and Water System" (EP-13) components galvanic corrosion V.A-6 VA Heat Copper allo Closed cycle Loss of material! Chapter XI. M33, "Selective Leaching of No exchanger >15% Zn cooling water selective leaching Materials" (EP-37) components V.A-7 VA6-c Heat Stainless Closed cycle Loss of material! Chapter XI.M21, "Closed-Cycle Cooling No exchanger steel cooling water pitting and crevice Water System" (E-19) components corrosion OAGI0000203_516

(f)

(D V ENGINEERED SAFETY FEATURES (D A Containment Spray System (PWR) 3 0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component V.A-S VA6-a Heat Stainless Raw water Loss of material! Chapter XL M20, "Open-Cycle Cooling No exchanger steel pitting, crevice, and Water System" (E-20) components microbiologically influenced corrosion, and fouling V.A-9 VA6-c Heat Steel Closed cycle Loss of material! Chapter XLM21, "Closed-Cycle Cooling No exchanger cooling water general, pitting, Water System" (E-17) components crevice, and galvanic corrosion V.A-10 VA6-a Heat Steel Raw water Loss of material! Chapter XI.M20, "Open-Cycle Cooling No exchanger general, pitting, Water System"

>> (E-1S) components crevice, galvanic, W and microbiologically influenced corrosion, and fouling V.A-11 VA Heat Copper alloy Closed cycle Reduction of heat Chapter XLM21, "Closed-Cycle Cooling No exchanger cooling water transfer! fouling Water System" (EP-39) tubes V.A-12 VA Heat Copper alloy Lubricating oil Reduction of heat Chapter XL M39, "Lubricating Oil Yes, detection exchanger transfer! fouling Analysis" of aging (EP-47) tubes effects is to be z The AMP is to be augmented by evaluated C

0 verifying the effectiveness of the m

lubricating oil analysis program. See

~

OJ Chapter XLM32, "One-Time

~ Inspection," for an acceptable verification program.

OAGI0000203_517

z V ENGINEERED SAFETY FEATURES C

0 A Containment Spray System (PWR) m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< V.A-13 Chapter XI.M21, "Closed-Cycle Cooling No VA Heat Stainless Closed cycle Reduction of heat exchanger steel cooling water transfer! fouling Water System" (EP-35) tubes V.A-14 VA Heat Stainless Lubricating oil Reduction of heat Chapter XI. M39, "Lubricating Oil Yes, detection exchanger steel transfer! fouling Analysis" of aging (EP-50) tubes effects is to be The AMP is to be augmented by evaluated verifying the effectiveness of the lubricating oil analysis program. See Chapter XI.M32, "One-Time Inspection," for an acceptable verification program.

< V.A-15 VA6-b Heat Stainless Raw water Reduction of heat Chapter XI. M20, "Open-Cycle Cooling No

.h exchanger steel transfer! fouling Water System" (E-21) tubes V.A-16 VA Heat Stainless Treated water Reduction of heat Chapter XI.M2, "Water Chemistry" Yes, detection exchanger steel transfer! fouling of aging (EP-34) tubes The AMP is to be augmented by effects is to be verifying the effectiveness of water evaluated chemistry control. See Chapter XI.M32, "One-Time Inspection," for an acceptable verification program.

V.A-17 VA Heat Steel Lubricating oil Reduction of heat Chapter XI. M39, "Lubricating Oil Yes, detection exchanger transfer! fouling Analysis" of aging (EP-40) tubes effects is to be The AMP is to be augmented by evaluated verifying the effectiveness of the lubricating oil analysis program. See Chapter XI.M32, "One-Time Inspection," for an acceptable verification proQram.

OAGI0000203_518

(f)

(D V ENGINEERED SAFETY FEATURES (D A Containment Spray System (PWR) 3 0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component V.A-18 VA Motor Cooler Gray cast Treated water Loss of material! Chapter XL M33, "Selective Leaching of No iron selective leaching Materials" (E-43)

V.A-19 VA2-a Piping and Steel Air - indoor Loss of material! Chapter XL M38, "Inspection of Internal No VA5-a components uncontrolled general corrosion Surfaces in Miscellaneous Piping and (E-29) internal (Internal) Ducting Components" surfaces V.A-20 VA Piping, piping Copper alloy Closed cycle Loss of material! Chapter XLM21, "Closed-Cycle Cooling No components, cooling water pitting, crevice, and Water System" (EP-36) and piping galvanic corrosion elements V.A-21 VA Piping, piping Copper alloy Lubricating oil Loss of material! Chapter XL M39, "Lubricating Oil Yes, detection components, pitting and crevice Analysis" of aging (EP-45) and piping corrosion effects is to be elements The AMP is to be augmented by evaluated verifying the effectiveness of the lubricating oil analysis program. See Chapter XLM32, "One-Time Inspection," for an acceptable verification program.

V.A-22 VA Piping, piping Copper allo Closed cycle Loss of material! Chapter XL M33, "Selective Leaching of No components, >15% Zn cooling water selective leaching Materials" (EP-27) and piping elements z

C V.A-23 VA Piping, piping Stainless Closed cycle Loss of material! Chapter XLM21, "Closed-Cycle Cooling No

0 rn components, steel cooling water pitting and crevice Water System"

~ (EP-33) and piping corrosion OJ elements

~

OAGI0000203_519

z V ENGINEERED SAFETY FEATURES C

0 A Containment Spray System (PWR) m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< V.A-24 Chapter XI.M21, "Closed-Cycle Cooling No VA Piping, piping Stainless Closed cycle Cracking! stress com ponents, steel cooling water corrosion cracking Water System" (EP-44) and piping >60°C (>140°F) elements V.A-25 VA Piping, piping Steel Lubricating oil Loss of material! Chapter XI. M39, "Lubricating Oil Yes, detection components, general, pitting, and Analysis" of aging (EP-46) and piping crevice corrosion effects is to be elements The AMP is to be augmented by evaluated verifying the effectiveness of the lubricating oil analysis program. See Chapter XI.M32, "One-Time Inspection," for an acceptable

< verification program.

'" V.A-26 (EP-53)

VA Piping, piping components, piping Stainless steel Condensation (Internal)

Loss of material! A plant-specific aging management pitti ng and crevice program is to be evaluated.

corrosion Yes, plant-specific elements internal surfaces, and tanks V.A-27 VA Piping, piping Stainless Treated borated Loss of material! Chapter XI.M2, "Water Chemistry," for No components, steel water pitting and crevice PWR primary water (EP-41) piping corrosion elements, and tanks V.A-28 VA1-a Piping, piping Stainless Treated borated Cracking! stress Chapter XI.M2, "Water Chemistry," for No VA3-a components, steel water >60°C corrosion cracking PWR primary water (E-12) VA4-a piping (>140°F)

VA1-c elements, and tanks OAGI0000203_520

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 (N FPA) 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 extended period of operation, are included in V.F.

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

September 2005 V B-1 NUREG-1801, Rev. 1 OAGI0000203_521

z V ENGINEERED SAFETY FEATURES C

0 B Standby Gas Treatment System (BWR) m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

0 Mechanism Evaluation (D

Component

< V.B-1 V.B.2-a Air - indoor Ducting and Steel Loss of material! Chapter XL M38, "Inspection of Internal No components uncontrolled general corrosion Surfaces in Miscellaneous Piping and (E-25) internal (Internal) Ducting Components" surfaces V.B-2 V.B.1-a Ducting Steel Air - indoor Loss of material! Chapter XL M36, "External Surfaces No closure bolting uncontrolled general corrosion Monitoring" (E-40) (External)

V.B-3 V.B.2-a Ducting, Steel Air - indoor Loss of material! Chapter XL M36, "External Surfaces No V.B.1-a piping and uncontrolled general corrosion Monitoring" (E-26) components (External) external

< surfaces OJ r0 V.B-4 V.B.2-b Elastomer Elastomers Air - indoor Hardening and loss A plant-specific aging management Yes, plant-V.B.1-b seals and uncontrolled of strength! program is to be evaluated. specific (E-06) components elastomer degradation V.B-5 V.B. Heat Copper alloy Closed cycle Loss of material! Chapter XL M33, "Selective Leaching of No exchanger >15% Zn cooling water selective leaching Materials" (EP-37) components V.B-6 V.B. Piping, piping Copper alloy Closed cycle Loss of material! Chapter XLM21, "Closed-Cycle Cooling No components, cooling water pitting, crevice, and Water System" (EP-36) and piping galvanic corrosion elements V.B-7 V.B. Piping, piping Copper alloy Closed cycle Loss of material! Chapter XL M33, "Selective Leaching of No components, >15% Zn cooling water selective leaching Materials" (EP-27) and piping elements OAGI0000203_522

(f)

(D V ENGINEERED SAFETY FEATURES (D B Standby Gas Treatment System (BWR) 3 0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component V.B-8 V.B. Piping, piping Gray cast Soil Loss of material! Chapter XI. M33, "Selective Leaching of No components, iron selective leaching Materials" (EP-54) and piping elements V.B-9 V.B. Piping, piping Steel (with or Soil Loss of material! Chapter XI. M28, "Buried Piping and No components, without general, pitting, Tanks Surveillance," or (E-42) and piping coating or crevice, and elements wrapping) microbiologically Chapter XI. M34, "Buried Piping and Yes, detection influenced Tanks Inspection" of aging corrosion effects and operating experience are

< to be further OJ W evaluated z

C

0 m

~

OJ

~

OAGI0000203_523

This Page Intentionally Left Blank NUREG-1801, Rev. 1 V 8-4 September 2005 OAGI0000203_524

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 I LA 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 extended period of operation, are included in V.F.

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

September 2005 V C-1 NUREG-1801, Rev. 1 OAGI0000203_525

z V ENGINEERED SAFETY FEATURES C

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OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< V.C-1 V.G.1-a Air - indoor Containment Steel Loss of material! Chapter XL M36, "External Surfaces No isolation piping uncontrolled general corrosion Monitoring" (E-35) and (External) components external surfaces V.C-2 V.G.1-a Containment Steel Condensation Loss of material! Chapter XL M36, "External Surfaces No isolation piping (External) general corrosion Monitoring" (E-30) and components external

< surfaces 2 V.C-3 V.G.1-b Containment Stainless isolation piping steel Raw water Loss of material! Chapter XL M20, "Open-Cycle Cooling pitting, crevice, and Water System" No (E-34) and microbiologically components influenced internal corrosion, and surfaces fouling V.C-4 V.G.1-b Containment Stainless Treated water Loss of material! Chapter XLM2, "Water Chemistry" Yes, detection isolation piping steel pitting and crevice of aging (E-33) and corrosion The AMP is to be augmented by effects is to be components verifying the effectiveness of water evaluated internal chem istry controL See Chapter surfaces XLM32, "One-Time Inspection," for an acceptable verification program.

OAGI0000203_526

(f)

(D V ENGINEERED SAFETY FEATURES (D C Containment Isolation Components 3

0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component V.C-5 V.G.1-a Containment Steel Raw water Loss of material! Chapter XL M20, "Open-Cycle Cooling No isolation piping general, pitting, Water System" (E-22) and crevice, and components microbiologically internal influenced surfaces corrosion, and fouling V.C-6 V.G.1-a Containment Steel Treated water Loss of material! Chapter XLM2, "Water Chemistry" Yes, detection isolation piping general, pitting, and of aging (E-31) and crevice corrosion The AMP is to be augmented by effects is to be components verifying the effectiveness of water evaluated internal chem istry controL See Chapter

< surfaces XLM32, "One-Time Inspection," for an 8 acceptable verification program.

V.C-7 VC Piping, piping Stainless Closed cycle Loss of material! Chapter XL M21, "Closed-Cycle Cooling No components, steel cooling water pitting and crevice Water System" (EP-33) and piping corrosion elements V.C-8 VC Piping, piping Stainless Closed cycle Cracki ng! stress Chapter XL M21, "Closed-Cycle Cooling No components, steel cooling water corrosion cracking Water System" (EP-44) and piping >60°C (>140°F) elements V.C-9 VC Piping, piping Steel Closed cycle Loss of material! Chapter XL M21, "Closed-Cycle Cooling No components, cooling water general, pitting, and Water System" z (EP-48) and piping crevice corrosion C

0 elements rn

~

OJ

~

OAGI0000203_527

This Page Intentionally Left Blank NUREG-1801, Rev. 1 V C-4 September 2005 OAGI0000203_528

01. 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. They 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 extended period of 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 (IV.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).

September 2005 V 01-1 NUREG-1801, Rev. 1 OAGI0000203_529

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OJ Structure

~ Aging Effect/ Aging Management Program Further Item Link and!or Material Environment Mechanism (AMP) Evaluation

0 (D

Component

< V.D1-1 V.D1.6-d Chapter XLM10, "Boric Acid Corrosion" No External Steel Air with borated Loss of material!

V.D1.5-b surfaces water leakage boric acid corrosion (E-28) V.D1.1-d V.D1.2-b V.D1.8-b V.D1.4-c V.D17-a V.D1.3-a V.D1-2 V.D1. Heat Copper alloy Closed cycle Loss of material! Chapter XLM21, "Closed-Cycle No exchanger cooling water pitting, crevice, and Cooling Water System" (EP-13) components galvanic corrosion V.D1-3 V.D1. Heat Copper alloy Closed cycle Loss of material! Chapter XL M33, "Selective Leaching of No exchanger >15% Zn cooling water selective leaching Materials" (EP-37) components V.D1-4 V.D1.5-a Heat Stainless Closed cycle Loss of material! Chapter XLM21, "Closed-Cycle No V.D1.6-a exchanger steel cooling water pitting and crevice Cooling Water System" (E-19) components corrosion V.D1-5 V.D1.6-b Heat Stainless Raw water Loss of material! Chapter XL M20, "Open-Cycle Cooling No exchanger steel pitting, crevice, and Water System" (E-20) components microbiologically influenced corrosion, and fouling V.D1-6 V.D1.6-a Heat Steel Closed cycle Loss of material! Chapter XLM21, "Closed-Cycle No V.D1.5-a exchanger cooling water general, pitting, Cooling Water System" (E-17) components crevice, and galvanic corrosion OAGI0000203_530

(f)

(D V ENGINEERED SAFETY FEATURES (D 01 Emergency Core Cooling System (PWR) 3 0-

~ Structure Aging Effect/ Aging Management Program Further

'"oo Item Link and!or Material Environment Mechanism (AMP) Evaluation U1 Component V.D1-7 V.D1.6-b Heat Steel Raw water Loss of material! Chapter XL M20, "Open-Cycle Cooling No exchanger general, pitting, Water System" (E-1S) components crevice, galvanic, and microbiologically influenced corrosion, and fouling V.D1-S V.D1. Heat Copper alloy Lubricating oil Reduction of heat Chapter XL M39, "Lubricating Oil Yes, detection exchanger transfer! fouling Analysis" of aging (EP-47) tubes effects is to be The AMP is to be augmented by evaluated verifying the effectiveness of the lubricating oil analysis program. See Chapter XLM32, "One-Time Inspection," for an acceptable verification program.

V.D1-19 V.D1. Heat Stainless Closed cycle Reduction of heat Chapter XLM21, "Closed-Cycle No exchanger steel cooling water transfer! fouling Cooling Water System" (EP-35) tubes V.D1-10 V.D1. Heat Stainless Lubricating oil Reduction of heat Chapter XL M39, "Lubricating Oil Yes, detection exchanger steel transfer! fouling Analysis" of aging (EP-50) tubes effects is to be The AMP is to be augmented by evaluated z verifying the effectiveness of the C

0 lubricating oil analysis program. See m Chapter XLM32, "One-Time

~

OJ Inspection," for an acceptable

~

verification proqram.

OAGI0000203_531

z V ENGINEERED SAFETY FEATURES C

0 01 Emergency Core Cooling System (PWR) m

~

OJ Structure

~ Aging Effect/ Aging Management Program Further Item Link and!or Material Environment Mechanism (AMP) Evaluation

0 (D

Component

< V.D1-11 V.D1.6-c Chapter XL M20, "Open-Cycle Cooling Heat Stainless Raw water Reduction of heat No exchanger steel transfer! fouling Water System" (E-21) tubes V.D1-12 V.D1. Heat Steel Lubricating oil Reduction of heat Chapter XL M39, "Lubricating Oil Yes, detection exchanger transfer! fouling Analysis" of aging (EP-40) tubes effects is to be The AMP is to be augmented by evaluated verifying the effectiveness of the lubricating oil analysis program. See Chapter XLM32, "One-Time Inspection," for an acceptable verification program.

V.D1-13 V.D1. Motor Cooler Gray cast Treated water Loss of material! Chapter XL M33, "Selective Leaching of No iron selective leaching Materials" (E-43)

V.D1-14 V.D1.2-c Orifice Stainless Treated borated Loss of material! A plant-specific aging management Yes, plant-(miniflow steel water erosion program is to be evaluated for erosion specific (E-24) recirculation) of the orifice due to extended use of the centrifugal HPSI pump for normal charging. See Licensee Event Report 50-275!94-023 for evidence of erosion.

V.D1-15 V.D1.8-c Partially Stainless Raw water Loss of material! A plant-specific aging management Yes, plant-encased tanks steel pitting and crevice program is to be evaluated for pitting specific (E-01) with breached corrosion and crevice corrosion of tank bottom moisture because moisture and water can barrier egress under the tank due to cracking of the perimeter seal from weathering.

OAGI0000203_532

(f)

(D V ENGINEERED SAFETY FEATURES (D 01 Emergency Core Cooling System (PWR) 3 0-

~ Structure Aging Effect/ Aging Management Program Further

'"oo Item Link and!or Material Environment Mechanism (AMP) Evaluation U1 Component V.D1-16 V.D1.1-b Piping, piping Cast Treated borated Loss of fracture Chapter XIM12, "Thermal Aging No com ponents, austenitic water >250°C toughness! thermal Embrilliement of Cast Austenitic (E-47) and piping stainless (>482°F) aging embrilliement Stainless Steel (CASS)"

elements steel V.D1-17 V.D1. Piping, piping Copper alloy Closed cycle Loss of material! Chapter XIM21, "Closed-Cycle No com ponents, cooling water pilling, crevice, and Cooling Water System" (EP-36) and piping galvanic corrosion elements V.D1-19 V.D1. Piping, piping Copper alloy Lubricating oil Loss of material! Chapter XI M39, "Lubricating Oil Yes, detection com ponents, pilling and crevice Analysis" of aging (EP-45) and piping corrosion effects is to be elements The AMP is to be augmented by evaluated verifying the effectiveness of the lubricating oil analysis program. See Chapter XIM32, "One-Time Inspection," for an acceptable verification program.

V.D1-19 V.D1. Piping, piping Copper alloy Closed cycle Loss of material! Chapter XI M33, "Selectil.e Leaching of No com ponents, >15% Zn cooling water selective leaching Materials" (EP-27) and piping elements V.D1-20 V.D1. Piping, piping Gray cast Closed cycle Loss of material! Chapter XI M33, "Selective Leaching of No com ponents, iron cooling water selective leaching Materials" (EP-52) piping z

C elements

0 m V.D1-21 V.D1. Piping, piping Gray cast Soil Loss of material! Chapter XI M33, "Selective Leaching of No

~ com ponents, iron selective leaching Materials" OJ (EP-54) and piping

~ elements OAGI0000203_533

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0 01 Emergency Core Cooling System (PWR) m

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OJ Structure

~ Aging Effect/ Aging Management Program Further Item Link and!or Material Environment Mechanism (AMP) Evaluation

0 (D

Component

< V.D1-22 Chapter XLM21, "Closed-Cycle V.D1. Piping, piping Stainless Closed cycle Loss of material! No com ponents, steel cooling water pitting and crevice Cooling Water System" (EP-33) and piping corrosion elements V.D1-23 V.D1. Piping, piping Stainless Closed cycle Cracking! stress Chapter XLM21, "Closed-Cycle No com ponents, steel cooling water corrosion cracking Cooling Water System" (EP-44) and piping >60°C (>140°F) elements V.D1-24 V.D1. Piping, piping Stainless Lubricating oil Loss of material! Chapter XL M39, "Lubricating Oil Yes, detection com ponents, steel pitting and crevice Analysis" of aging (EP-51) and piping corrosion effects is to be elements The AMP is to be augmented by evaluated verifying the effectiveness of the lubricating oil analysis program. See Chapter XLM32, "One-Time Inspection," for an acceptable verification program.

V.D1-25 V.D1. Piping, piping Stainless Raw water Loss of material! Chapter XL M20, "Open-Cycle Cooling No com ponents, steel pitting, crevice, and Water System" (EP-55) and piping microbiologically elements influenced corrosion V.D1-26 V.D1. Piping, piping Stainless Soil Loss of material! A plant-specific aging management Yes, plant-com ponents, steel pitting and crevice program is to be evaluated. specific (EP-31) and piping corrosion elements OAGI0000203_534

(f)

(D V ENGINEERED SAFETY FEATURES (D 01 Emergency Core Cooling System (PWR) 3 0-

~ Structure Aging Effect/ Aging Management Program Further

'"oo Item Link and!or Material Environment Mechanism (AMP) Evaluation U1 Component V.D1-27 V.D1.1-c Piping, piping Stainless Treated borated Cumulative fatigue Fatigue is a time-limited aging Yes, V.D1.4-a com ponents, steel water damage! fatigue analysis (TLAA) to be evaluated for the TLAA (E-13) and piping period of extended operation. Seethe elements Standard Review Plan, Section 4.3 "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR 5421(c)(1)

V.D1-28 V.D1. Piping, piping Steel Lubricating oil Loss of material! Chapter XL M39, "Lubricating Oil Yes, detection com ponents, general, pitting, and Analysis" of aging (EP-46) and piping crevice corrosion effects is to be elements The AMP is to be augmented by evaluated verifying the effectiveness of the lubricating oil analysis program. See Chapter XLM32, "One-Time Inspection," for an acceptable verification program.

V.D1-29 V.D1. Piping, piping Stainless Condensation Loss of material! A plant-specific aging management Yes, plant-com ponents, steel (Internal) pitting and crevice program is to be evaluated. specific (EP-53) piping corrosion elements internal surfaces, and tanks V.D1-30 V.D1. Piping, piping Stainless Treated borated Loss of material! Chapter XLM2, "Water Chemistry," for No z

C com ponents, steel water pitting and crevice PWR primary water

0 rn (EP-41) piping corrosion

~ elements, and OJ tanks

~

OAGI0000203_535

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OJ Structure

~ Aging Effect/ Aging Management Program Further Item Link and!or Material Environment Mechanism (AMP) Evaluation

0 (D

Component

< V.D1-31 V.D1.1-a Chapter XI.M2, "Water Chemistry," for No Piping, piping Stainless Treated borated Cracking! stress V.D12-a com ponents, steel water >60°C corrosion cracking PWR primary water (E-12) V.D1.8-a piping (>140°F)

V.D1.7-b elements, and V.D1.4-b tanks V.D1-32 V.D1. Pump Casings Steel with Treated borated Loss of material! A plant-specific aging management Yes, verify stainless water cladding breach program is to be evaluated. that plant-(EP-49) steel specific cladding Reference NRC Information Notice 94- program 63, "Boric Acid Corrosion of Charging addresses Pump Casings Caused by Cladding cladding Cracks." breach V.D1-33 V.D1.7-b Safety Steel with Treated borated Cracking! stress Chapter XI.M2, "Water Chemistry," for No injection tank stainless water >60°C corrosion cracking PWR primary water (E-38) (accumulator) steel (>140°F) cladding OAGI0000203_536

D2. 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.

They 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 extended period of operation, are included in VLF.

System Interfaces The systems that interface with the emergency core cooling system include the reactor vessel (IV.A1), the reactor coolant pressure boundary (IV.C1), the feedwater system (VIILD2), the condensate system (VilLE), the closed-cycle cooling water system (VII.C2), the open-cycle cooling water system (VILC1), and the ultimate heat sink (VILC3).

September 2005 V 02-1 NUREG-1801, Rev. 1 OAGI0000203_537

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Mechanism Evaluation

0 (D

Component

< V.D2-1 V.D2.5-b Air - indoor A plant-specific aging management Drywell and Steel Loss of material! Yes, plant-suppression uncontrolled general corrosion program is to be evaluated. specific (E-04) chamber (Internal) and fouling spray system (internal surfaces):

Flow orifice Spray nozzles V.D2-2 V.D2.5-a Ducting, Steel Air - indoor Loss of material! Chapter XI. M36, "External Surfaces No V.D2.1-e piping and uncontrolled general corrosion Monitoring" (E-26) components (External) external surfaces V.D2-3 V.D2. Heat Copper alloy Closed cycle Loss of material! Chapter XI.M21, "Closed-Cycle Cooling No exchanger cooling water pitting, crevice, and Water System" (EP-13) components galvanic corrosion V.D2-4 V.D2. Heat Copper alloy Closed cycle Loss of material! Chapter XI. M33, "Selective Leaching of No exchanger >15% Zn cooling water selective leaching Materials" (EP-37) components V.D2-5 V.D2.4-c Heat Stainless Closed cycle Loss of material! Chapter XI.M21, "Closed-Cycle Cooling No exchanger steel cooling water pitting and crevice Water System" (E-19) components corrosion V.D2-6 V.D2.4-a Heat Stainless Raw water Loss of material! Chapter XI. M20, "Open-Cycle Cooling No exchanger steel pitting, crevice, and Water System" (E-20) components microbiologically influenced corrosion, and fouling OAGI0000203_538

V ENGINEERED SAFETY FEATURES D2 Emergency Core Cooling System (BWR)

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

Mechanism Evaluation Component V.D2-7 V.D2.4-c Heat Steel Closed cycle Loss of material! Chapter XI.M21, "Closed-Cycle Cooling No exchanger cooling water general, pitting, Water System" (E-17) components crevice, and galvanic corrosion V.D2-S V.D2.4-a Heat Steel Raw water Loss of material! Chapter XI. M20, "Open-Cycle Cooling No exchanger general, pitting, Water System" (E-1S) components crevice, galvanic, and microbiologically influenced corrosion, and fouling V.D2-9 V.D2. Heat Copper alloy Lubricating oil Reduction of heat Chapter XI. M39, "Lubricating Oil Yes, detection exchanger transfer! fouling Analysis" of aging (EP-47) tubes effects is to be The AMP is to be augmented by evaluated verifying the effectiveness of the lubricating oil analysis program. See Chapter XI.M32, "One-Time Inspection," for an acceptable verification program.

V.D2-10 V.D2. Heat Stainless Closed cycle Reduction of heat Chapter XI. M21, "Closed-Cycle Cooling No exchanger steel cooling water transfer! fouling Water System" (EP-35) tubes z

C

0 m

~

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Mechanism Evaluation

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< V.D2-11 V.D2. Heat Stainless Lubricating oil Reduction of heat Chapter XL M39, "Lubricating Oil Yes, detection exchanger steel transfer! fouling Analysis" of aging (EP-50) tubes effects is to be The AMP is to be augmented by evaluated verifying the effectiveness of the lubricating oil analysis program. See Chapter XLM32, "One-Time Inspection," for an acceptable verification program.

V.D2-12 V.D2.4-b Heat Stainless Raw water Reduction of heat Chapter XL M20, "Open-Cycle Cooling No exchanger steel transfer! fouling Water System" (E-21) tubes V.D2-13 V.D2. Heat Stainless Treated water Reduction of heat Chapter XLM2, "Water Chemistry" Yes, detection exchanger steel transfer! fouling of aging (EP-34) tubes The AMP is to be augmented by effects is to be verifying the effectiveness of water evaluated chemistry controL See Chapter XLM32, "One-Time Inspection," for an acceptable verification program.

V.D2-14 V.D2. Heat Steel Lubricating oil Reduction of heat Chapter XL M39, "Lubricating Oil Yes, detection exchanger transfer! fouling Analysis" of aging (EP-40) tubes effects is to be The AMP is to be augmented by evaluated verifying the effectiveness of the lubricating oil analysis program. See Chapter XLM32, "One-Time Inspection," for an acceptable verification program.

V.D2-15 V.D2.4-b Heat Steel Raw water Reduction of heat Chapter XL M20, "Open-Cycle Cooling No exchanger transfer! fouling Water System" (E-23) tubes OAGI0000203_540

V ENGINEERED SAFETY FEATURES D2 Emergency Core Cooling System (BWR)

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

Mechanism Evaluation Component V.D2-16 V.D2.5-a Piping and Steel Air - indoor Loss of material! Chapter XL M38, "Inspection of Internal No components uncontrolled general corrosion Surfaces in Miscellaneous Piping and (E-29) internal (Internal) Ducting Components" surfaces V.D2-17 V.D2.1-e Piping and Steel Condensation Loss of material! Chapter XL M38, "Inspection of Internal No components (Internal) general, pitting, and Surfaces in Miscellaneous Piping and (E-27) internal crevice corrosion Ducting Components" surfaces V.D2-18 V.D2. Piping, piping Aluminum Air with borated Loss of material! Chapter XLM10, "Boric Acid Corrosion" No com ponents, water leakage boric acid corrosion (EP-2) and piping elements V.D2-19 V.D2. Piping, piping Aluminum Treated water Loss of material! Chapter XL M2, "Water Chemistry" for Yes, detection com ponents, pitting and crevice BWRwater of aging (EP-26) and piping corrosion effects is to be elements The AMP is to be augmented by evaluated verifying the effectiveness of water chem istry controL See Chapter XLM32, "One-Time Inspection," for an acceptable verification program.

V.D2-20 V.D2.1-d Piping, piping Cast Treated water Loss of fracture Chapter XLM12, "Thermal Aging No com ponents, austenitic >250"C toughness! thermal Embritllement of Cast Austenitic (E-11) and piping stainless (>482"F) aging embrittlement Stainless Steel (CASS)"

elements steel z

C V.D2-21 Chapter XLM21, "Closed-Cycle Cooling No

0 V.D2. Piping, piping Copper alloy Closed cycle Loss of material!

rn com ponents, cooling water pitting, crevice, and Water System"

~

OJ (EP-36) and piping galvanic corrosion elements

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~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

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< V.D2-22 V.D2. Piping, piping Copper alloy Lubricating oil Loss of material! Chapter XL M39, "Lubricating Oil Yes, detection com ponents, pitting and crevice Analysis" of aging (EP-45) and piping corrosion effects is to be elements The AMP is to be augmented by evaluated verifying the effectiveness of the lubricating oil analysis program. See Chapter XLM32, "One-Time Inspection," for an acceptable verification program.

V.D2-23 V.D2. Piping, piping Copper alloy Closed cycle Loss of material! Chapter XL M33, "Selective Leaching of No com ponents, >15% Zn cooling water selective leaching Materials" (EP-27) and piping elements V.D2-24 V.D2. Piping, piping Gray cast Soil Loss of material! Chapter XL M33, "Selective Leaching of No com ponents, iron selective leaching Materials" (EP-54) and piping elements V.D2-25 V.D2. Piping, piping Stainless Closed cycle Loss of material! Chapter XLM21, "Closed-Cycle Cooling No com ponents, steel cooling water pitting and crevice Water System" (EP-33) and piping corrosion elements V.D2-26 V.D2. Piping, piping Stainless Closed cycle Cracking! stress Chapter XLM21, "Closed-Cycle Cooling No com ponents, steel cooling water corrosion cracking Water System" (EP-44) and piping >60"C (>140"F) elements V.D2-27 V.D2. Piping, piping Stainless Soil Loss of material! A plant-specific aging management Yes, plant-com ponents, steel pitting and crevice program is to be evaluated. specific (EP-31) and piping corrosion elements OAGI0000203_542

V ENGINEERED SAFETY FEATURES D2 Emergency Core Cooling System (BWR)

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

Mechanism Evaluation Component V.D2-28 V.D2. Piping, piping Stainless Treated water Loss of material! Chapter XI. M2, "Water Chemistry" for Yes, detection com ponents, steel pitting and crevice BWRwater of aging (EP-32) and piping corrosion effects is to be elements The AMP is to be augmented by evaluated verifying the effectiveness of water chem istry control. See Chapter XI.M32, "One-Time Inspection," for an acceptable verification program.

V.D2-29 V.D2.1-c Piping, piping Stainless Treated water Cracking! stress Chapter XI.M7, "BWR stress No V.D2.3-c com ponents, steel >60°C (> 140°F) corrosion cracking Corrosion Cracking," and (E-37) and piping and intergranular elements stress corrosion Chapter XI.M2, "Water Chemistry," for cracking BWRwater V.D2-30 V.D2. Piping, piping Steel Lubricating oil Loss of material! Chapter XI. M39, "Lubricating Oil Yes, detection com ponents, general, pitting, and Analysis" of aging (EP-46) and piping crevice corrosion effects is to be elements The AMP is to be augmented by evaluated verifying the effectiveness of the lubricating oil analysis program. See Chapter XI.M32, "One-Time Inspection," for an acceptable verification program.

V.D2-31 V.D2.1-f Piping, piping Steel Steam Wall thinning! flow- Chapter XI.M17, "Flow-Accelerated No com ponents, accelerated Corrosion" z (E-07) and piping corrosion C

0 elements m

~

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Mechanism Evaluation

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< V.D2-32 V.D2.1-b Fatigue is a time-limited aging Piping, piping Steel Treated water Cumulative fatigue Yes, com ponents, damage! fatigue analysis (TLAA) to be evaluated for the TLAA (E-10) and piping period of extended operation. See the elements Standard Review Plan, Section 4.3 "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR 5421(c)(1)

V.D2-33 V.D2.3-b Piping, piping Steel Treated water Loss of material! Chapter XI.M2, "Water Chemistry," for Yes, detection V.D2.1-a com ponents, general, pitting, and BWRwater of aging (E-08) V.D2.2-a and piping crevice corrosion effects is to be elements The AMP is to be augmented by evaluated verifying the effectiveness of water chem istry control. See Chapter XI.M32, "One-Time Inspection," for an acceptable verification program.

V.D2-34 V.D2.3-a Piping, piping Steel Treated water Wall thinning! flow- Chapter XI.M17, "Flow-Accelerated No com ponents, accelerated Corrosion" (E-09) and piping corrosion elements V.D2-35 V.D2.1-e Piping, piping Stainless Condensation Loss of material! A plant-specific aging management Yes, plant-com ponents, steel (Internal) pitting and crevice program is to be evaluated. specific (E-14) and piping corrosion elements internal surfaces OAGI0000203_544

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.)

September 2005 V E-1 NUREG-1801, Rev. 1 OAGI0000203_545

V ENGINEERED SAFETY FEATURES z E External Surfaces of Components and Miscellaneous Bolting C

0

~

OJ Item Link Structure and!or Material Environment Aging Effect!

Aging Management Program (AMP)

Further

~ Mechanism Evaluation Component V.E-1 V.E. Bolting Steel Air - outdoor Loss of material! Chapter XI.M18, "Bolting Integrity" No (External) general, pitting, (EP-1) and crevice corrosion V.E-2 V.E. Bolting Steel Air with borated Loss of material! Chapter XI. M 10, "Boric Acid Corrosion" No water leakage boric acid (E-41) corrosion V.E-3 V.E.2-b Closure bolting High- Air with steam Cracking! cyclic Chapter XI.M18, "Bolting Integrity" No strength or water leakage loading, stress (E-03) steel corrosion cracking V.E-4 V.E. Closure bolting Steel Air - indoor Loss of material! Chapter XI.M18, "Bolting Integrity" No uncontrolled general, pitting,

< (EP-25) (External) and crevice rn r0 corrosion V.E-5 V.E. Closure bolting Steel Air - indoor Loss of preload! Chapter XI.M18, "Bolting Integrity" No uncontrolled thermal effects, (EP-24) (External) gasket creep, and self-loosening V.E-6 V.E.2-a Closure bolting Steel Air with steam Loss of material! Chapter XI.M18, "Bolting Integrity" No or water leakage general corrosion E-02)

V.E-7 V.E. External Steel Air - indoor Loss of material! Chapter XI. M36, "External Surfaces No surfaces uncontrolled general corrosion Monitoring" (E-44) (External)

V.E-8 V.E. External Steel Air - outdoor Loss of material! Chapter XI. M36, "External Surfaces No (f) surfaces (External) general corrosion Monitoring" (D

(E-45)

~ V.E-9 V.E.1-a External Steel Air with borated Loss of material! Chapter XI. M 10, "Boric Acid Corrosion" No 3

0- surfaces water leakage boric acid

~ (E-28) corrosion

'"oo U1 OAGI0000203_546

V ENGINEERED SAFETY FEATURES E External Surfaces of Components and Miscellaneous Bolting Structure Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation Component V.E-10 V.E.1-b External Steel Condensation Loss of material! Chapter XI. M36, "External Surfaces No surfaces (External) general corrosion Monitoring" (E-46)

V.E-11 V.E. Piping, piping Copper allo; Air with borated Loss of material! Chapter XI. M1 0, "Boric Acid Corrosion" No com ponents, >15% Zn water leakage boric acid (EP-38) and piping corrosion elements rn W

z C

0 rn

~

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~

OAGI0000203_547

This Page Intentionally Left Blank NUREG-1801, Rev. 1 V E-4 September 2005 OAGI0000203_548

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 extended period of 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 (BWRs). (For example, see System Interfaces in V.A to V.D2 for details.)

September 2005 V F-1 NUREG-1801. Rev. 1 OAGI0000203_549

V ENGINEERED SAFETY FEATURES z F Common Miscellaneous Material/Environment Combinations C

0

~

OJ Item Link Structure and/or Material Environment Aging Effect!

Aging Management Program (AMP)

Further

~ Mechanism Evaluation Component V.F-1 V.F. Ducting Galvanized Air - indoor None None No steel controlled (EP-14) (External)

V.F-2 V.F. Piping, piping Aluminum Air - indoor None None No components, uncontrolled (EP-3) and piping (Internal/External) elements V.F-3 V.F. Piping, piping Copper alloy Air - indoor None None No components, uncontrolled (EP-10) and piping (External) elements V.F-4 V.F. Piping, piping Copper alloy Gas None None No components, (EP-9) and piping elements V.F-5 V.F. Piping, piping Copper alloy Air with borated None None No components, <15% Zn water leakage (EP-12) and piping elements V.F-6 V.F. Piping Glass Air - indoor None None No elements uncontrolled (EP-15) (External)

V.F-7 V.F. Piping Glass Lubricating oil None None No elements (f)

(EP-16)

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(D V ENGINEERED SAFETY FEATURES (D F Common Miscellaneous Material/Environment Combinations 3

~ Structure

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Further Evaluation U1 Component V.F-8 V.F. Piping Glass Raw water None None No elements (EP-28)

V.F-9 V.F. Piping Glass Treated borated None None No elements water (EP-3O)

V.F-10 V.F. Piping Glass Treated water None None No elements (EP-29)

-n W V.F-11 V.F. Piping, piping Nickel alloy Air - indoor None None No components, uncontrolled (EP-17) and piping (External) elements V.F-12 V.F. Piping, piping Stainless Air - indoor None None No components, steel uncontrolled (EP-18) and piping (External) elements V.F-13 V.F. Piping, piping Stainless Air with borated None None No components, steel water leakage (EP-19) and piping z elements C

0 m V.F-14 V.F. Piping, piping Stainless Concrete None None No

~ components, steel OJ (EP-20) and piping

~ elements OAGI0000203_551

V ENGINEERED SAFETY FEATURES z F Common Miscellaneous Material/Environment Combinations C

0

~

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Aging Management Program (AMP)

Further

~ Mechanism Evaluation Component V.F-15 V.F. Piping, piping Stainless Gas None None No components, steel (EP-22) and piping elements V.F-16 V.F. Piping, piping Steel Air - indoor None None No components, controlled (EP-4) and piping (External) elements V.F-17 V.F. Piping, piping Steel Concrete None None No components, (EP-5) and piping elements

< V.F-18 V.F. Piping, piping Steel Gas None None No

-n

.h components, (EP-7) and piping elements (f)

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CHAPTER VI ELECTRICAL COMPONENTS September 2005 VI-i NUREG-1801, Rev. 1 OAGI0000203_553

This Page Intentionally Left Blank NUREG-1801, Rev. 1 VI-ii September 2005 OAGI0000203_554

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 September 2005 VI-iii NUREG-1801, Rev. 1 OAGI0000203_555

This Page Intentionally Left Blank NUREG-1801, Rev. 1 VI-iv September 2005 OAGI0000203_556

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 <<1000V) and medium-voltage (2 kVto 35 kV). High voltage (>35 kV) power cables and connections have unique, specialized constructions and must be evaluated on an application specific basis.

This section also addresses components that are relied upon to meet the station blackout (SBO) requirements for restoration of offsite power. The plant system portion of the offsite power system that is used to connect the plant to the offsite power source is included in the SBO restoration equipment scope. This path typically includes the switchyard 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 (including bus ducts or cables), and associated control circuits and structures.

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 Structures and Components Supports section.

September 2005 VIA-1 NUREG-1801, Rev. 1 OAGI0000203_557

z VI ELECTRICAL COMPONENTS C

0 A Equipment Not Subject to 10 CFR 50.49 Environmental Qualification Requirements m

Gl Structure and!or Aging Effect! Aging Management Program Further Item Link Material Environment Component Mechanism (AMP) Evaluation VI.A-1 VIA able Connections Various f"ir - indoor and Loosening of bolted Chapter XI.E6, "Electrical No (Metallic Parts) metals used outdoor connections due to Connections Not Subject to 10 (LP-12) for electrical thermal cycling, ohmic CFR 50.49 Environmental contacts heating, electrical Qualification Requirements" transients, vibration, chemical contamination, corrosion, and oxidation VI.A-2 VIA1-a Gonductor insulation Various f"dverse localized Embrittlement, cracking, Chapter XI. E1, "Electrical No or electrical cables organic environment melting, discoloration, Cables and Connections Not (L-01) and connections polymers caused by heat, swelling, or loss of Subject to 10 CFR 50.49 (eg, EPR, radiation, or dielectric strength Environmental Qualification SR, EPDM, moisture in the leading to reduced Requirements" XLPE) presence of insulation resistance oxygen (IR); electrical failure!

degradation of organics (Thermal!

thermoxidative),

rad iolysis and photolysis (UV sensitive materials only) of organics; radiation-induced oxidation, and moisture intrusion (f)

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(D VI ELECTRICAL COMPONENTS (D

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0-

~ Structure and!or Aging Effect! Aging Management Program Further Item Link Material Environment

'"oo Component Mechanism (AMP) Evaluation U1 VI.A-3 VIA1-b onductor insulation Various f"dverse localized Embrittlement, cracking, Chapter XI.E2, "Electrical No or electrical cables organic environment melting, discoloration, Cables and Connections Not (L-02) and connections used polymers caused by heat, swelling, or loss of Subject to 10 CFR 50.49 in instrumentation (eg, EPR, radiation, or dielectric strength Environmental Qualification ircuits that are SR, EPDM, moisture in the leading to reduced Requirements Used in sensitive to reduction XLPE) presence of insulation resistance Instrumentation Circuits" in conductor oxygen (IR); electrical failure!

insulation resistance degradation of organics (IR) (Thermal!

thermoxidative),

rad iolysis and photolysis (UV sensitive materials only) of organics; radiation-induced oxidation, and moisture intrusion VI.A-4 VIA1-c onductor insulation Various f"dverse localized Localized damage and Chapter XI.E3, "Inaccessible No or inaccessible organic environment breakdown of insulation Medium Voltage Cables Not (L-03) medium-voltage (2kV polymers caused by leading to electrical Subject to 10 CFR 50.49 o 35kV) cables (eg, EPR, exposure to failure! moisture Environmental Qualification z (eg, installed in SR, EPDM, moisture and intrusion, water trees Requirements" C conduit or direct XLPE) ~/Oitage

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z VI ELECTRICAL COMPONENTS C

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Gl Structure and!or Aging Effect! Aging Management Program Further Item Link Material Environment Component Mechanism (AMP) Evaluation VI.A-5 VIA2-a onnector contacts Various f"ir with borated Corrosion of connector Chapter XI. M 10, "Boric Acid No or electrical metals used fNater leakage contact surfaces! Corrosion" (L-04) onnectors exposed for electrical intrusion of borated o borated water contacts water leakage VI.A-6 VIA Fuse Holders (Not Insulation f"dverse localized Embrittlement, cracking, Chapter XI. E1, "Electrical No Part of a Larger material - environment melting, discoloration, Cables and Connections Not (LP-03) Assembly); bakelite, caused by heat, swelling, or loss of Subject to 10 CFR 50.49 Insulation phenolic radiation, or dielectric strength Environmental Qualification melamine or moisture in the leading to reduced Requirements" ceramic, presence of insulation resistance molded oxygen or > 60- (IR); electrical failure!

polycarbonate year service degradation (Thermal!

and other limiting therm oxidative) of emperature organics!thermoplastics, radiation-induced oxidation, moisture intrusion and ohmic heating (f)

(D (D

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A Equipment Not Subject to 10 CFR 50.49 Environmental Qualification Requirements 3

0-

~ Structure and/or Aging Effect! Aging Management Program Further Item Link Material Environment

'"oo Component Mechanism (AMP) Evaluation U1 VI.A-7 VIA Fuse Holders (Not Insulation f"ir - indoor None None No Part of a Larger material - uncontrolled (LP-02) Assembly); bakelite, (I nternal/External)

Insulation phenolic melamine or ceramic, molded polycarbonate and other VI.A-8 VIA Fuse Holders (Not Copper alloy f"ir - indoor Fatigue/ohmic heating, Chapter XI.E5, "Fuse Holders" No Part of a Larger thermal cycling, (LP-01) Assembly); electrical transients, Metallic Clamp frequent manipulation, vibration, chemical contamination, corrosion, and oxidation VI.A-9 VIA High voltage Porcelain, f"ir - outdoor Degradation of insulator A plant-specific aging fYes, plant-insulators Malleable quality/presence of any management program is to be ~pecific (LP-07) iron, salt deposits or surface evaluated for plants located aluminum, contamination such that the potential exists for galvanized salt deposits or surface steel, cement contamination (e.g., in the z

C vicinity of salt water bodies or

0 industrial pollution) m Gl
0 (D

OAGI0000203_561

z VI ELECTRICAL COMPONENTS C

0 A Equipment Not Subject to 10 CFR 50.49 Environmental Qualification Requirements m

Gl Structure and!or Aging Effect! Aging Management Program Further Item Link Material Environment Component Mechanism (AMP) Evaluation VI.A-10 VIA High voltage Porcelain, f"ir - outdoor Loss of material! A plant-specific aging fYes, plant-insulators Malleable mechanical wear due to management program is to be ~pecific (LP-11) iron, wind blowing on evaluated.

aluminum, transmission conductors galvanized steel, cement VI.A-11 VIA Metal enclosed bus Aluminum! f"ir - indoor and Loosening of bolted Chapter XI.E4, "Metal Enclosed No Silver Plated outdoor connections! thermal Bus" (LP-04) Bus!connections Aluminum cycling and ohmic Copper! Silver heating Plated Copper; Stainless steel, steel VI.A-12 VIA Metal enclosed bus Elastomers f"ir - indoor and Hardening and loss of Chapter XI.S6, "Structures No outdoor strength! elastomer Monitoring Program" (LP-10) Enclosure degradation assemblies VI.A-13 VIA Metal enclosed bus Steel; f"ir - indoor and Loss of material! Chapter XI.S6, "Structures No galvanized outdoor general corrosion Monitoring Program" (LP-06) Enclosure steel assemblies (f)

(D (D

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(D VI ELECTRICAL COMPONENTS (D

A Equipment Not Subject to 10 CFR 50.49 Environmental Qualification Requirements 3

0-

~ Structure and/or Aging Effect! Aging Management Program Further Item Link Material Environment

'"oo Component Mechanism (AMP) Evaluation U1 VI.A-14 VIA Metal enclosed bus Porcelain, f"ir - indoor and Embrittlement, cracking, Chapter XI.E4, "Metal Enclosed No xenoy, outdoor melting, discoloration, Bus" (LP-05) Insulation/insulators thermo-plastic swelling, or loss of organic dielectric strength polymers leading to reduced insulation resistance (IR); electrical failurel thermal/thermoxidative degradation of organics/thermoplastics, radiation-induced oxidation; moisture/debris intrusion, and ohmic heating VI.A-15 VIA Switch yard bus and Aluminum, f"ir - outdoor Loss of materiall wind A plant-specific aging [Yes, plant-connections copper, induced abrasion and management program is to be ~pecific (LP-09) bronze, fatigue evaluated.

stainless Loss of cond uctor steel, strengthl corrosion galvanized Increased resistance of steel connectionl oxidation or loss of pre-load z

C

0 m

Gl

0 (D

OAGI0000203_563

z VI ELECTRICAL COMPONENTS C

0 A Equipment Not Subject to 10 CFR 50.49 Environmental Qualification Requirements m

Gl Structure and!or Aging Effect! Aging Management Program Further Item Link Material Environment Component Mechanism (AMP) Evaluation VI.A-16 VIA ransmission Aluminum, f"ir - outdoor Loss of material! wind A plant-specific aging fYes, plant-conductors and steel induced abrasion and management program is to be ~pecific (LP-08) onnections fatigue evaluated.

Loss of cond uctor strength! corrosion Increased resistance of connection! oxidation or loss of pre-load (f)

(D (D

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~

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B. EQUIPMENT SUBJECT TO 10 CFR 50.49 ENVIRONMENTAL QUALIFICATION REQUIREMENTS Systems, Structures and Components The Nuclear Regulatory Commission (N RC) 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.

September 2005 VI B-1 NUREG-1801, Rev. 1 OAGI0000203_565

z VI ELECTRICAL COMPONENTS C

0 B Equipment Subject to 10 CFR 50.49 Environmental Qualification Requirements m

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

Mechanism Evaluation

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< VI.B-1 VI.B.1-a Electrical Various Adverse Various EQ is a time-limited aging analysis Yes, equipment polymeric localized degradation! (TLAA) to be evaluated for the period TLAA (L-05) subject to and metallic environment various of extended operation. See the 10 CFR 50.49 materials caused by heat, mechanisms Standard Review Plan, Section 4.4, EQ radiation, "Environmental Qualification (EQ) of requirements oxygen, Electrical Equipment," for acceptable moisture, or methods for meeting the requirements voltage of 10 CFR 5421(c)(1)(i) and (ii)

See Chapter X. E1, "Environmental Qualification (EQ) of Electric Components," of this report for meeting the requirements of 10 CFR 54.21(c)(1)(iii)

(f)

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'"oo U1 OAGI0000203_566

CHAPTER VII AUXILIARY SYSTEMS September 2005 VII-i NUREG-1801, Rev. 1 OAGI0000203_567

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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)

A5. 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)

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 External Surfaces of Components and Miscellaneous Bolting J. Common Miscellaneous Material/Environment Combinations September 2005 VII-iii NUREG-1801. Rev. 1 OAGI0000203_569

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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 extended period of operation are included in VII.J.

System Interfaces No other systems discussed in this report interface with those used for new fuel storage.

September 2005 VIIA1-1 NUREG-1801, Rev. 1 OAGI0000203_571

z C

0 VII AUXILIARY SYSTEMS m

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Mechanism Evaluation Component VII.A1-1 VII.A1.1-a Structural Steel Air - indoor Loss of material! Chapter XI.S6, "Structures Monitoring No Steel uncontrolled general, pitting, Program" (A-94) (External) and crevice corrosion (f)

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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 include 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 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 extended period of operation are included in VII.J.

System Interfaces No other systems discussed in this report interface with those used for spent fuel storage.

September 2005 VII A2-1 NUREG-1801, Rev. 1 OAGI0000203_573

z VII AUXILIARY SYSTEMS C

0 A2 Spent Fuel Storage m

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< VII.A2-1 VII.A2. Piping, piping Stainless Treated borated Loss of material! Chapter XI.M2, 'Water Chemistry," for No com ponents, Steel; Steel water pitting and crevice PWR primary water (AP-79) and piping with corrosion elements stainless steel cladding VII.A2-2 VII.A2.1-a Spent fuel Boraflex Treated water Reduction of Chapter XI. M22, "Boraflex Monitoring" No storage racks neutron-absorbing (A-87) capacity! boraflex Neutron- degradation absorbing sheets -

BWR VII.A2-3 VII.A2.1-b Spent fuel Boral, Treated water Reduction of A plant-specific aging management Yes, plant-storage racks boron steel neutron-absorbing program is to be evaluated. specific (A-89) capacity and loss Neutron- of material!

absorbing general corrosion sheets -

BWR VII.A2-4 VII.A2.1-a Spent fuel Boraflex Treated borated Reduction of Chapter XI. M22, "Boraflex Monitoring" No storage racks water neutron-absorbing (A-86) capacity! boraflex Neutron- degradation absorbing sheets -

PWR OAGI0000203_574

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(D VII AUXILIARY SYSTEMS (D A2 Spent Fuel Storage 3

0-

~ Structure Aging Effect! Further

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Evaluation U1 Component VII.A2-5 VII.A2.1-b Spent fuel Boral, Treated borated Reduction of A plant-specific aging management Yes, plant-storage racks boron steel water neutron-absorbing program is to be evaluated. specific (A-SS) capacity and loss Neutron- of materiall absorbing general corrosion sheets -

PWR VII.A2-6 VII.A2.1-c Spent fuel Stainless Treated water Cracki ngl stress Chapter XI.M2, 'Water Chemistry," for No storage racks steel >60"C (>140"F) corrosion cracking BWRwater (A-96)

Storage racks - BWR VII.A2-7 VII.A2.1-c Spent fuel Stainless Treated borated Cracki ngl stress Chapter XI.M2, 'Water Chemistry," for No storage racks steel water>60"C corrosion cracking PWR primary water (A-97) (>140"F)

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OAGI0000203_575

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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 (N RC) position on consumables, provided in the NRC letter from Christopher I.

Grimes to Douglas J. Walters of the Nuclear Energy Institute (NEO, 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 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 extended period of 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).

September 2005 VII A3-1 NUREG-1801, Rev. 1 OAGI0000203_577

z VII AUXILIARY SYSTEMS C

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~

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< A plant-specific aging management VII.A3-1 VII.A3.3-a Elastomer Elastomers Treated borated Hardening and Yes, plant-VII.A3.2-d lining water loss of strength! program that determines and assesses specific (A-15) VII.A3.5-c elastomer the qualified life of the linings in the VII.A3.5-a degradation environment is to be evaluated.

VII.A3.3-d VII.A3.2-a VII.A3-2 VII.A3.1-a External Steel Air with borated Loss of material! Chapter XI. M 10, "Boric Acid Corrosion" No VII.A3.5-b surfaces water leakage boric acid (A-79) VII.A3.3-c corrosion VII.A3.2-c VII.A3.2-b VII.A3.4-b VII.A3.6-a VII.A3-3 VII.A3.4-a Heat Steel Closed cycle Loss of material! Chapter XI.M21, "Closed-Cycle Cooling No exchanger cooling water general, pitting, Water System" (A-63) components crevice, and galvanic corrosion OAGI0000203_578

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~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component VII.A3-4 VII.A3. Piping, piping Aluminum Air with borated Loss of material! Chapter XI. M 10, "Boric Acid Corrosion" No com ponents, water leakage boric acid (AP-1) and piping corrosion elements VII.A3-5 VII.A3. Piping, piping Copper alloy Closed cycle Loss of material! Chapter XI.M21, "Closed-Cycle Cooling No com ponents, cooling water pitting, crevice, Water System" (AP-12) and piping and galvanic elements corrosion VII.A3-6 VII.A3. Piping, piping Copper alloy Closed cycle Loss of material! Chapter XI. M33, "Selective Leaching of No com ponents, >15% Zn cooling water selective leaching Materials" (AP-43) and piping elements VII.A3-7 VII.A3. Piping, piping Gray cast Treated water Loss of material! Chapter XI. M33, "Selective Leaching of No com ponents, iron selective leaching Materials" (AP-31) and piping elements VII.A3-8 VII.A3. Piping, piping Stainless Treated borated Loss of material! Chapter XI.M2, "Water Chemistry," for No com ponents, Steel; Steel water pitting and crevice PWR primary water (AP-79) and piping with stainless corrosion elements steel claddin, VII.A3-9 VII.A3.5-a Piping, piping Steel with Treated borated Loss of material! Chapter XI. M2, "Water Chemistry," for Yes, detection VII.A3.3-a com ponents, elastomer water pitting and crevice PWR primary water of aging effect (A-39) VII.A3.2-a and piping lining corrosion (only for isto be z elements steel after lining The AMP is to be augmented by evaluated C

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~

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~ acceptable verification program.

OAGI0000203_579

z VII AUXILIARY SYSTEMS C

0 A3 Spent Fuel Pool Cooling and Cleanup (PWR) m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

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Component

< VII.A3,3-b VII.A3-10 Piping, piping Steel with Treated borated Cracking! stress Chapter XI.M2, "Water Chemistry," for No com ponents, stainless water >60°C corrosion cracki ng PWR primary water (A-56) and piping steel claddin, (>140°F) elements OAGI0000203_580

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 (N RC) position on consumables, provided in the NRC letter from Christopher L Grimes to Douglas J. Walters of the Nuclear Energy Institute (NEO, 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 (N FPA) 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 VILI. 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 extended period of 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 (VILC2) and the condensate system (VilLE).

September 2005 VII A4-1 NUREG-1801, Rev. 1 OAGI0000203_581

z VII AUXILIARY SYSTEMS C

0 A4 Spent Fuel Pool Cooling and Cleanup (BWR) m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

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Component VII.A4-1 VII.A4.5-b Elastomer Elastomers Treated water Hardening and loss A plant-specific aging management Yes, plant-VII.A4.3-b lining of strength! program that determines and assesses specific (A-16) VII.A4.3-a elastomer the qualified life of the linings in the VII.A4.2-b degradation environment is to be evaluated.

VII.A4.5-a VII.A4.2-a VII.A4-2 VII.A4.4-b Heat Stainless Treated water Loss of material! Chapter XI.M2, "Water Chemistry," for Yes, detection exchanger steel; steel Pitting and crevice BWRwater of aging effect (A-70) components with stainless corrosion isto be steel claddin[ The AMP is to be augmented by evaluated verifying the effectiveness of water chem istry control. See Chapter XI. M32, "One-Time Inspection," for an acceptable verification program.

VII.A4-3 VII.A4.4-a Heat Steel Closed cycle Loss of material! Chapter XI.M21, "Closed-Cycle Cooling No exchanger cooling water general, pitting, Water System" (A-63) components crevice, and galvanic corrosion VII.A4-4 VII.A4. Heat Stainless Treated water Reduction of heat Chapter XI. M2, "Water Chemistry" Yes, detection exchanger steel transfer! fouling of ag ing effects (AP-62) tubes The AMP is to be augmented by isto be verifying the effectiveness of water evaluated chem istry control. See Chapter XI. M32, "One-Time Inspection," for an acceptable verification program.

OAGI0000203_582

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(D VII AUXILIARY SYSTEMS (D A4 Spent Fuel Pool Cooling and Cleanup (BWR) 3 0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component VII.A4-5 VII.A4. Piping, piping Aluminum Treated water Loss of material! Chapter XI.M2, "Water Chemistry" Yes, detection com ponents, pitting and crevice of aging effect (AP-38) and piping corrosion The AMP is to be augmented by isto be elements verifying the effectiveness of water evaluated chem istry control. See Chapter XI. M32, "One-Time Inspection," for an acceptable verification program.

VII.A4-6 VII.A4. Piping, piping Copper alloy Closed cycle Loss of material! Chapter XI.M21, "Closed-Cycle Cooling No com ponents, cooling water pitting, crevice, and Water System" (AP-12) and piping galvanic corrosion elements VII.A4-7 VII.A4. Piping, piping Copper alloy Treated water Loss of material! Chapter XI.M2, "Water Chemistry," for Yes, detection com ponents, pitting, crevice, and BWRwater of aging effect (AP-64) and piping galvanic corrosion isto be elements The AMP is to be augmented by evaluated verifying the effectiveness of water chem istry control. See Chapter XI. M32, "One-Time Inspection," for an acceptable verification program.

VII.A4-8 VII.A4. Piping, piping Copper alloy Closed cycle Loss of material! Chapter XI. M33, "Selective Leaching of No com ponents, >15% Zn cooling water selective leaching Materials" (AP-43) and piping elements VII.A4-9 VII.A4. Piping, piping Copper alloy Treated water Loss of material! Chapter XI. M33, "Selective Leaching of No z com ponents, >15% Zn selective leaching Materials" C

0 (AP-32) and piping m

elements

~

OJ VII.A4-10 VII.A4. Piping, piping Gray cast Treated water Loss of material! Chapter XI. M33, "Selective Leaching of No

~ selective leaching Materials" components, iron (AP-31) and piping elements OAGI0000203_583

z VII AUXILIARY SYSTEMS C

0 A4 Spent Fuel Pool Cooling and Cleanup (BWR) m

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Mechanism Evaluation

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< VII.A4.1-a VII.A4-11 Piping, piping Stainless Treated water Loss of material! Chapter XI.M2, "Water Chemistry," for Yes, detection VII.A4.6-a com ponents, steel pitting and crevice BWRwater of aging effect (A-58) and piping corrosion isto be elements The AMP is to be augmented by evaluated verifying the effectiveness of water chem istry control. See Chapter XI. M32, "One-Time Inspection," for an acceptable \erification program.

VII.A4-12 VII.A4.5-a Piping, piping Steel with Treated water Loss of material! Chapter XI.M2, "Water Chemistry," for Yes, detection VII.A4.2-a com ponents, elastomer pitting and crevice BWRwater of aging effect (A-40) VII.A4.3-a and piping lining or corrosion (only for isto be elements stainless steel after The AMP is to be augmented by evaluated steel claddin[ lining!cladding verifying the effectiveness of water degradation) chem istry control. See Chapter XI. M32, "One-Time Inspection," for an acceptable verification program.

OAGI0000203_584

A5. 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 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 extended period of 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 containment (II.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 [VII.A4.4].)

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B. OVERHEAD HEAW 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 (IPA). 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 extended period of 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.

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z VII AUXILIARY SYSTEMS C

0 B Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

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< VII.B-1 VII.B.2-a Cranes - rails Air - indoor Chapter XI. M23, "Inspection of Overhead No Steel Loss of material!

uncontrolled wear Heavy Load and Light Load (Related to (A-05) (External) Refueling) Handling Systems" VII.B-2 VII.B.1-a Cranes - Steel Air - indoor Cumulative fatigue Fatigue is a time-limited aging analysis Yes, Structural uncontrolled damage! fatigue (TLAA) to be evaluated for the period of TLAA (A-06) girders (External) extended operation for structural girders of cranes that fall within the scope of 10 CFR 54. See the Standard Review Plan, Section 4.7, "Other Plant-Specific Time-Limited Aging Analyses," for generic guidance for meeting the requirements of 10 CFR 5421 (c)(1)

VII.B-3 VII.B.1-b Cranes - Steel Air - indoor Loss of material! Chapter XI. M23, "Inspection of Overhead No Structural uncontrolled General corrosion Heavy Load and Light Load (Related to (A-Ol) girders (External) Refueling) Handling Systems" OAGI0000203_588

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 (VilLA) and for the condensate system coolers (VilLE).

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 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 extended period of operation are included in VILJ.

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 (VILC2), the ultimate heat sink (VILC3), the emergency diesel generator system (VILH2), the containment spray system (V.A), the PWR steam generator blowdown system (VIILF), the condensate system (VilLE), the auxiliary feedwater system (PWR)

(VIILG), the emergency core cooling system (PWR) (V.D1), and the emergency core cooling system (BWR) (V.D2).

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z VII AUXILIARY SYSTEMS C

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OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

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0 (D

Component

< VIIC1-1 VllCt Elastomer Elastomers Raw water Hardening and loss Chapter XI M20, "Open-Cycle Cooling No seals and of strength! Water System" (AP-75) components elastomer degradation VIIC1-2 VllCt Elastomer Elastomers Raw water Loss of material! Chapter XI M20, "Open-Cycle Cooling No seals and erosion Water System" (AP-76) components VIIC1-3 VIIC1.3-a Heat Copper alloy Raw water Loss of material! Chapter XI M20, "Open-Cycle Cooling No exchanger pitting, crevice, Water System" (A-65) components galvanic, and microbiologically influenced corrosion, and fouling VIIC1-4 VIIC1.3-a Heat Copper alloy Raw water Loss of material! Chapter XI M33, "Selective Leaching of No exchanger >15% Zn selective leaching Materials" (A-66) components VIIC1-5 VIIC1.3-a Heat Steel Raw water Loss of material! Chapter XI M20, "Open-Cycle Cooling No exchanger general, pitting, Water System" (A-64) components crevice, galvanic, and microbiologically influenced corrosion, and fouling VIIC1-6 VIIC1.3-b Heat Copper alloy Raw water Reduction of heat Chapter XI M20, "Open-Cycle Cooling No exchanger transfer! fouling Water System" (A-72) tubes VIIC1-7 VllCt Heat Stainless Raw water Reduction of heat Chapter XI M20, "Open-Cycle Cooling No exchanger steel transfer! fouling Water System" (AP-61) tubes OAGI0000203_590

(f)

(D VII AUXILIARY SYSTEMS (D C1 Open-Cycle Cooling Water System (Service Water System) 3 0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component VIIC1-8 VllCt Piping, piping Copper alloy Lubricating oil Loss of material! Chapter XI M39, "Lubricating Oil Yes, detection components, pitting and crevice Analysis" of aging (AP-47) and piping corrosion effects is to be elements The AMP is to be augmented by evaluated verifying the effectiveness of the lubricating oil analysis program. See Chapter XIM32, "One-Time Inspection," for an acceptable verification program.

VIIC1-9 VIIC1.2-a Piping, piping Copper alloy Raw water Loss of material! Chapter XI M20, "Open-Cycle Cooling No VIIC1.1-a components, pitting, crevice, and Water System" (A-44) and piping microbiologically elements influenced corrosion, and foulinQ VIIC1-10 VIIC1.1-a Piping, piping Copper alloy Raw water Loss of material! Chapter XI M33, "Selective Leaching of No VIIC1.2-a components, >15% Zn selective leaching Materials" (A-47) and piping elements VIIC1-11 VIIC1.5-a Piping, piping Gray cast iron Raw water Loss of material! Chapter XI M33, "Selective Leaching of No components, selective leaching Materials" (A-51) and piping elements VIIC1-12 VIIC1.1-c Piping, piping Gray cast iron Soil Loss of material! Chapter XI M33, "Selective Leaching of No z components, selective leaching Materials" C

0 (A-02) and piping m

elements

~

OJ VIIC1-13 VllCt Piping, piping Nickel alloy Raw water Loss of material! Chapter XI M20, "Open-Cycle Cooling No

~ components, pitting and crevice Water System" (AP-53) and piping corrosion elements OAGI0000203_591

z VII AUXILIARY SYSTEMS C

0 C1 Open-Cycle Cooling Water System (Service Water System) m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< VIIC1-14 VllCt Piping, piping Stainless Lubricating oil Loss of material! Chapter XI M39, "Lubricating Oil Yes, detection components, steel pitting, crevice, and Analysis" of aging (AP-59) and piping microbiologically effects is to be elements influenced corrosion The AMP is to be augmented by evaluated verifying the effectiveness of the lubricating oil analysis program. See Chapter XIM32, "One-Time Inspection," for an acceptable verification program.

VIIC1-15 VIIC1.2-a Piping, piping Stainless Raw water Loss of material! Chapter XI M20, "Open-Cycle Cooling No VIIC1.6-a components, steel pitting and crevice Water System" (A-54) VIIC1.1-a and piping corrosion, and VIIC1.4-a elements fouling VIIC1-16 VllCt Piping, piping Stainless Soil Loss of material! A plant-specific aging management Yes, plant-components, steel pitting and crevice program is to be evaluated. specific (AP-56) and piping corrosion elements VIIC1-17 VllCt Piping, piping Steel Lubricating oil Loss of material! Chapter XI M39, "Lubricating Oil Yes, detection components, general, pitting, and Analysis" of aging (AP-30) and piping crevice corrosion effects is to be elements The AMP is to be augmented by evaluated verifying the effectiveness of the lubricating oil analysis program. See Chapter XIM32, "One-Time Inspection," for an acceptable verification proqram.

OAGI0000203_592

(f)

(D VII AUXILIARY SYSTEMS (D C1 Open-Cycle Cooling Water System (Service Water System) 3 0-

~ Structure Aging Effect! Further

'"oo Item Link and/or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component VIIC1-18 VIIC1.1-b Piping, piping Steel (with or Soil Loss of material/ Chapter XI M28, "Buried Piping and No components, without general, pitting, Tanks Surveillance," or (A-01) and piping coating or crevice, and elements wrapping) microbiologically Chapter XI M34, "Buried Piping and Yes, detection influenced corrosion Tanks Inspection" of aging effects and operating experience are to be further evaluated VIIC1-19 VIIC1.6-a Piping, piping Steel (with or Raw water Loss of material/ Chapter XI M20, "Open-Cycle Cooling No VIIC1.2-a components, without general, pitting, Water System" (A-38) VIIC1.1-a and piping lining/coating crevice, and VIIC1.5-a elements or with microbiologically degraded influenced Ilining/coatin g) corrosion, fouling, and lining/coating degradation z

C

0 m

~

OJ

~

OAGI0000203_593

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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 (AM Ps) 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 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 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 I. 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 extended period of 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 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 spray system (V.A), the PWR and BWR emergency core cooling systems (V.D1 and V.D2), the PWR September 2005 VII C2-1 NUREG-1801, Rev. 1 OAGI0000203_595

steam generator blowdown system (VIILF), the condensate system (VilLE), and the PWR auxiliary feedwater system (VIILG).

NUREG-1801, Rev. 1 VII C2-2 September 2005 OAGI0000203_596

(f)

(D VII AUXILIARY SYSTEMS (D C2 Closed-Cycle Cooling Water System 3

0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component VIIC2-1 VIIC2. Heat Steel Closed cycle Loss of material! Chapter XIM21, "Closed-Cycle Cooling No exchanger cooling water general, pitting, Water System" (A-63) components crevice, and galvanic corrosion VIIC2-2 VIIC2. Heat Copper Alloy Closed cycle Reduction of heat Chapter XIM21, "Closed-Cycle Cooling No exchanger cooling water transfer! fouling Water System" (AP-SO) tubes VIIC2-3 VIIC2. Heat Stainless Closed cycle Reduction of heat Chapter XIM21, "Closed-Cycle Cooling No exchanger steel cooling water transfer! fouling Water System" (AP-63) tubes VIIC2-4 VIIC2. Piping, piping Copper alloy Closed cycle Loss of material! Chapter XIM21, "Closed-Cycle Cooling No com ponents, cooling water pitting, crevice, and Water System" (AP-12) and piping galvanic corrosion elements VIIC2-5 VIIC2. Piping, piping Copper alloy Lubricating oil Loss of material! Chapter XI M39, "Lubricating Oil Yes, detection com ponents, pitting and crevice Analysis" of aging effect (AP-47) and piping corrosion isto be elements The AMP is to be augmented by evaluated verifying the effectiveness of the lubricating oil analysis program. See Chapter XIM32, "One-Time Inspection,"

for an acceptable verification program.

VIIC2-6 VIIC2. Piping, piping Copper alloy Closed cycle Loss of material! Chapter XI M33, "Selective Leaching of No z com ponents, >15% Zn cooling water selective leaching Materials" C (AP-43) and piping

0 m elements

~

OJ VIIC2-7 VIIC2. Piping, piping Copper alloy Treated water Loss of material! Chapter XI M33, "Selective Leaching of No

~ com ponents, >15% Zn selective leaching Materials" (AP-32) and piping elements OAGI0000203_597

z VII AUXILIARY SYSTEMS C

0 C2 Closed-Cycle Cooling Water System m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< VIIC2-8 Chapter XI M33, "Selective Leaching of No VIIC2.3-a Piping, piping Gray cast Closed cycle Loss of material!

com ponents, iron cooling water selective leaching Materials" (A-50) and piping elements VIIC2-9 VIIC2. Piping, piping Gray cast Treated water Loss of material! Chapter XIM33, "Selective Leaching of No com ponents, iron selective leaching Materials" (AP-31) and piping elements VIIC2-10 VIIC2.2-a Piping, piping Stainless Closed cycle Loss of material! Chapter XIM21, "Closed-Cycle Cooling No com ponents, steel cooling water pitting and crevice Water System" (A-52) and piping corrosion elements VIIC2-11 VIIC2. Piping, piping Stainless Closed cycle Cracking! stress Chapter XIM21, "Closed-Cycle Cooling No com ponents, steel cooling water corrosion cracking Water System" (AP-60) and piping >60°C (> 140°F) elements VIIC2-12 VIIC2. Piping, piping Stainless Lubricating oil Loss of material! Chapter XI M39, "Lubricating Oil Yes, detection com ponents, steel pitting, crevice, and Analysis" of aging effect (AP-59) and piping microbiologically isto be elements influenced corrosion The AMP is to be augmented by evaluated verifying the effectiveness of the lubricating oil analysis program. See Chapter XIM32, "One-Time Inspection,"

for an acceptable verification program.

OAGI0000203_598

(f)

(D VII AUXILIARY SYSTEMS (D C2 Closed-Cycle Cooling Water System 3

0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component VIIC2-13 VIIC2. Piping, piping Steel Lubricating oil Loss of material! Chapter XI M39, "Lubricating Oil Yes, detection com ponents, general, pitting, and Analysis" of aging effect (AP-30) and piping crevice corrosion isto be elements The AMP is to be augmented by evaluated verifying the effectiveness of the lubricating oil analysis program. See Chapter XIM32, "One-Time Inspection,"

for an acceptable verification program.

VIIC2-14 VIIC2.3-a Piping, piping Steel Closed cycle Loss of material! Chapter XIM21, "Closed-Cycle Cooling No VIIC2.2-a com ponents, cooling water general, pitting, and Water System" (A-25) VIIC2.5-a piping crevice corrosion VIIC2.4-a elements, and VIIC2.1-a tanks z

C

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OJ

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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 (1II.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 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 extended period of 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).

September 2005 VII C3-1 NUREG-1801, Rev. 1 OAGI0000203_601

z VII AUXILIARY SYSTEMS C

0 C3 Ultimate Heat Sink m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< VIIC3-1 Chapter XI M20, "Open-Cycle Cooling VIIC3. Heat Stainless Raw water Reduction of heat No exchanger steel transfer! fouling Water System" (AP-61) tubes VIIC3-2 VIIC3.1-a Piping, piping Copper alloy Raw water Loss of material! Chapter XI M20, "Open-Cycle Cooling No VIIC3.2-a components, pitting and crevice Water System" (A-43) and piping corrosion elements VIIC3-3 VIIC3.2-a Piping, piping Copper alloy Raw water Loss of material! Chapter XI M33, "Selective Leaching of No VIIC3.1-a components, >15% Zn selective leaching Materials" (A-47) and piping elements VIIC3-4 VIIC3. Piping, piping Gray cast iron Raw water Loss of material! Chapter XI M33, "Selective Leaching of No components, selective leaching Materials" (A-51) and piping elements VIIC3-5 VIIC3. Piping, piping Gray cast iron Soil Loss of material! Chapter XI M33, "Selective Leaching of No components, selective leaching Materials" (A-02) and piping elements VIIC3-6 VIIC3. Piping, piping Nickel alloy Raw water Loss of material! Chapter XI M20, "Open-Cycle Cooling No components, pitting and crevice Water System" (AP-53) and piping corrosion elements VIIC3-7 VIIC3.2-a Piping, piping Stainless Raw water Loss of material! Chapter XI M20, "Open-Cycle Cooling No components, steel pitting and crevice Water System" (A-53) and piping corrosion elements OAGI0000203_602

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(D VII AUXILIARY SYSTEMS (D C3 Ultimate Heat Sink 3

0-

~ Structure Aging Effect! Further

'"oo Item Link and/or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component VIIC3-8 VIIC3. Piping, piping Stainless Soil Loss of material/ A plant-specific aging management Yes, plant-components, steel pitting and crevice program is to be evaluated. specific (AP-56) and piping corrosion elements VIIC3-9 VIIC3. Piping, piping Steel (with or Soil Loss of material/ Chapter XI M28, "Buried Piping and No components, without general, pitting, Tanks Surveillance," or (A-01) and piping coating or crevice, and elements wrapping) microbiologically Chapter XI M34, "Buried Piping and Yes, detection influenced corrosion Tanks Inspection" of aging effects and operating experience are to be further evaluated VIIC3-10 VIIC3.1-a Piping, piping Steel (with or Raw water Loss of material/ Chapter XI M20, "Open-Cycle Cooling No VIIC3.2-a components, without general, pitting, Water System" (A-38) VIIC3.3-a and piping lining/coating crevice, and elements or with microbiologically degraded influenced Ilining/coatin g) corrosion, fouling, and lining/coating degradation z

C

0 m

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OJ

~

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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 receiver, 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 (N RC) position on consumables, provided in the NRC letter from Christopher I.

Grimes to Douglas J. Walters of Nuclear Energy Institute (NEO, 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 (N FPA) 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 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 extended period of 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.

September 2005 VII [)'1 NUREG-1801, Rev. 1 OAGI0000203_605

z VII AUXILIARY SYSTEMS C

0 0 Compressed Air System m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation ii)l Component VIID-1 VILD.2-a Closure bolting Steel Condensation Loss of material! Chapter XLM18, "Bolting Integrity" No general, pitting, (A-103) and crevice corrosion VIID-2 VILD.5-a Compressed Steel Condensation Loss of material! Chapter XL M24, "Compressed Air No VILD.6-a air system (Internal) general and pitting Monitoring" (A-26) VILD.3-a corrosion VILD.2-a Piping, piping VILD.4-a com ponents, VILD.1-a and piping elements VIID-3 VILD.3-a Piping and Steel Air - indoor Loss of material! Chapter XL M36, "External Surfaces No VILD.5-a components uncontrolled general corrosion Monitoring" (A-80) VILD.6-a external (External)

VILD.4-a surfaces VILD.2-a VILD.1-a VIID-4 VIID Piping, piping Stainless Condensation Loss of material! Chapter XLM24, "Compressed Air No components, steel (Internal) pitting and crevice Monitoring" (AP-81) and piping corrosion elements OAGI0000203_606

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 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 extended period of 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 (IV.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).

September 2005 VII E1-1 NUREG-1801, Rev. 1 OAGI0000203_607

z VII AUXILIARY SYSTEMS C

0 E1 Chemical and Volume Control System (PWR) m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< VII.E1-1 Chapter XI.M10, "Boric Acid Corrosion" No VII.E1.9-a External Steel Air with borated Loss of material!

VII.E1.3-b surfaces water leakage boric acid corrosion (A-79) VII.E1.10-a VII.E1.5-b VI I. E1.7-b VII.E1.6-a VII.E1.1-b VII.E1.4-a VII.E1.2-a VII.E1.8-d VII.E1-2 VII.E1 Heat Copper alloy Closed cycle Loss of material! Chapter XI. M21, "Closed-Cycle Cooling No exchanger cooling water pitting, crevice, and Water System" (AP-34) components galvanic corrosion VII.E1-3 VII.E1. Heat Copper alloy Treated water Loss of material! Chapter XI. M33, "Selective Leaching of No exchanger >15% Zn selective leaching Materials" (AP-65) components OAGI0000203_608

(f)

(D VII AUXILIARY SYSTEMS (D E1 Chemical and Volume Control System (PWR) 3 0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component VII.E1-4 VII.E1.8-a Heat Stainless Treated borated Cumulative fatigue Fatigue is a time-limited aging analysis Yes, exchanger steel water damage! fatigue (TLAA) to be evaluated for the period of TLAA (A-100) components extended operation. See the Standard Review Plan, Section 4.3 "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR 5421(c)(1)

VII.E1-5 VI I. E1.7-c Heat Stainless Treated borated Cracking! stress Chapter XI. M2, "Water Chemistry," for Yes, plant-exchanger steel water >60°C corrosion cracking, PWR primary water specific (A-84) components (>140°F) cyclic loading The AMP is to be augmented by verifying the absence of cracking due to stress corrosion cracking and cyclic loading. A plant specific aging management program is to be evaluated.

VII.E1-6 VII.E1.8-c Heat Steel Closed cycle Loss of material! Chapter XI. M21, "Closed-Cycle Cooling No exchanger cooling water general, pitting, Water System" (A-63) components crevice, and galvanic corrosion VII.E1-7 VII.E1.5-a High-pressure Stainless Treated borated Cracking! stress Chapter XI.M2, "Water Chemistry," for Yes, plant-pump steel water corrosion cracking, PWR primary water specific (A-76) cyclic loading Casing The AMP is to be augmented by z verifying the absence of cracking due C

0 to stress corrosion cracking and cyclic m

~

OJ loading. A plant specific aging management program is to be

~ evaluated.

OAGI0000203_609

z VII AUXILIARY SYSTEMS C

0 E1 Chemical and Volume Control System (PWR) m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< VII.E1-8 High-pressure High- Chapter XI.M18, "Bolting Integrity" VII.E1.5-a Air with steam 0 Cracking! cyclic Yes, if the pump strength water leakage loading, stress bolts are not (A-104) steel corrosion cracking The AMP is to be augmented by replaced during Closure appropriate inspection to detect maintenance bolting cracking if the bolts are not otherwise replaced during maintenance.

VII.E1-9 VII.E1.8-b Non- Stainless Treated borated Cracking! stress Chapter XI. M2, "Water Chemistry," for Yes, plant-regenerative steel water >60°C corrosion cracking, PWR primary water. specific (A-69) heat (>140°F) cyclic loading exchanger The AMP is to be augmented by components verifying the absence of cracking due to stress corrosion cracking and cyclic loading, or loss of material due to pitting and crevice corrosion. An acceptable verification program is to include temperature and radioactivity monitoring of the shell side water, and eddy current testinQ of tubes.

VII.E1-10 VII.E1. Piping, piping Aluminum Air with borated Loss of material! Chapter XI.M10, "Boric Acid Corrosion" No com ponents, water leakage boric acid corrosion (AP-1) and piping elements VII.E1-11 VII.E1. Piping, piping Copper alloy Closed cycle Loss of material! Chapter XI. M21, "Closed-Cycle Cooling No com ponents, cooling water pitting, crevice, and Water System" (AP-12) and piping galvanic corrosion elements OAGI0000203_610

(f)

(D VII AUXILIARY SYSTEMS (D E1 Chemical and Volume Control System (PWR) 3 0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component VII.E1-12 VII.E1. Piping, piping Copper alloy Lubricating oil Loss of material! Chapter XI. M39, "Lubricating Oil Yes, detection com ponents, pitting and crevice Analysis" of aging effects (AP-47) and piping corrosion isto be elements The AMP is to be augmented by evaluated verifying the effectiveness of the lubricating oil analysis program. See Chapter XI. M32, "One-Time Inspection,"

for an acceptable verification program.

VII.E1-13 VII.C2. Piping, piping Copper alloy Closed cycle Loss of material! Chapter XI. M33, "Selective Leaching of No com ponents, >15% Zn cooling water selective leaching Materials" (AP-43) and piping elements VII.E1-14 VII.E1. Piping, piping Gray cast Treated water Loss of material! Chapter XI. M33, "Selective Leaching of No com ponents, iron selective leaching Materials" (AP-31) and piping elements VII.E1-15 VII.E1. Piping, piping Stainless Lubricating oil Loss of material! Chapter XI. M39, "Lubricating Oil Yes, detection com ponents, steel pitting, crevice, and Analysis" of aging effects (AP-59) and piping microbiologically isto be elements influenced corrosion The AMP is to be augmented by evaluated verifying the effectiveness of the z lubricating oil analysis program. See C

0 Chapter XL M32, "One-Time Inspection,"

m

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OJ for an acceptable verification program.

~

OAGI0000203_611

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~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

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Component

< VII.E1-16 VII.E1.1-a Fatigue is a time-limited aging analysis Yes, Piping, piping Stainless Treated lxlrated Cumulative fatigue VII.E1.8-a com ponents, steel water damage! fatigue (TLAA) to be evaluated for the period of TLAA (A-57) VI I. E1.7-a and piping extended operation. See the Standard VII.E1.3-a elements Review Plan, Section 4.3 "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR 5421(c)(1)

VII.E1-17 VII.E1. Piping, piping Stainless Treated borated Loss of material! Chapter XI. M2, "Water Chemistry," for No com ponents, Steel; Steel water pitting and crevice PWR primary water (AP-79) and piping with corrosion elements stainless steel cladding VII.E1-18 VII.E1.1-a Piping, piping Steel Air - indoor Cumulative fatigue Fatigue is a time-limited aging analysis Yes, VII.E1.8-a com ponents, uncontrolled damage! fatigue (TLAA) to be evaluated for the period of TLAA (A-34) VI I. E1.7-a and piping extended operation. See the Standard VII.E1.3-a elements Review Plan, Section 4.3 "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR 5421(c)(1)

VII.E1-19 VII.E1. Piping, piping Steel Lubricating oil Loss of material! Chapter XI. M39, "Lubricating Oil Yes, detection components, general, pitting, and Analysis" of aging effects (AP-30) and piping crevice corrosion isto be elements The AMP is to be augmented by evaluated verifying the effectiveness of the lubricating oil analysis program. See Chapter XI.M32, "One-Time Inspection,"

for an acceptable verification program.

OAGI0000203_612

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(D VII AUXILIARY SYSTEMS (D E1 Chemical and Volume Control System (PWR) 3 0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component VII.E1-20 VII.E1. Piping, piping Stainless Treated borated Cracking! stress Chapter XI. M2, "Water Chemistry," for No com ponents, steel water >60°C corrosion cracking PWR primary water (AP-82) piping (>140°F) elements, and tanks VII.E1-21 VII.E1. Pump Casings Steel with Treated borated Loss of material! A plant-specific aging management Yes, verify stainless water cladding breach program is to be evaluated. plant -specific (AP-85) steel program cladding Reference NRC Information Notice 94- addresses 63, "Boric Acid Corrosion of Charging cladding Pump Casings Caused by Cladding breach 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 (Na2B10016*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 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 extended period of 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 (IV.A1). If used, the SLC system would inject sodium pentaborate solution into the pressure vessel near the bottom of the reactor core.

September 2005 VII E2-1 NUREG-1801, Rev. 1 OAGI0000203_615

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< VII.E2-1 VII.E2. Piping, piping Stainless Sodium Loss of material! Chapter XI.M2, "Water Chemistry," for Yes, detection com ponents, steel penta borate pitting and crevice BWRwater of aging effect (AP-73) and piping solution corrosion isto be elements The AMP is to be augmented by evaluated verifying the effectiveness of water chemistry control. See Chapter XI.M32, "One-Time Inspection," for an acceptable verification program.

VII.E2-2 VII.E2.4-a Piping, piping Stainless Sodium Cracki ng! stress Chapter XI.M2, 'Water Chemistry," for Yes, detection VII.E2.1-a com ponents, steel penta borate corrosion cracking BWRwater of aging effect (A-59) VII.E2.2-a and piping solution >60°C isto be VII.E2.3-a elements (>140°F) The AMP is to be augmented by evaluated verifying the effectiveness of water chemistry control. See Chapter XI.M32, "One-Time Inspection," for an acceptable verification program.

OAGI0000203_616

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 for 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 VILI. 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 extended period of 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 (IV.C1), the closed-cycle cooling water system (VILC2), and the condensate system (VilLE).

September 2005 VII E3-1 NUREG-1801, Rev. 1 OAGI0000203_617

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Mechanism Evaluation

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Component

< Chapter XI. M21, "Closed-Cycle Cooling No VII.E3-1 VII.E3.4-b Heat Stainless Closed cycle Loss of material!

exchanger steel; steel cooling water microbiologically Water System" (A-67) components with influenced corrosion stainless steel cladding VII.E3-2 VII.E3.4-a Heat Stainless Closed cycle Cracking! stress Chapter XI. M21, "Closed-Cycle Cooling No exchanger steel; steel cooling water corrosion cracking Water System" (A-68) components with >60°C (>140°F) stainless steel claddinQ VII.E3-3 VII.E3.4-a Heat Stainless Treated water Cracking! stress A plant-specific aging management Yes, plant-exchanger steel; steel >60°C (>140°F) corrosion cracking program is to be evaluated. specific (A-71) components with stainless steel cladding VII.E3-4 VII.E3. Heat Steel Closed cycle Loss of material! Chapter XI. M21, "Closed-Cycle Cooling No exchanger cooling water general, pitting, Water System" (A-63) components crevice, and galvanic corrosion VII.E3-5 VII.E3. Heat Stainless Closed cycle Reduction of heat Chapter XI. M21, "Closed-Cycle Cooling No exchanger steel cooling water transfer! fouling Water System" (AP-63) tubes OAGI0000203_618

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(D VII AUXILIARY SYSTEMS (D E3 Reactor Water Cleanup System 3

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~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component VII.E3-6 VII.E3. Heat Stainless Treated water Reduction of heat Chapter XI.M2, 'Water Chemistry" Yes, detection exchanger steel transfer! fouling of aging effect (AP-62) tubes The AMP is to be augmented by isto be verifying the effectiveness of water evaluated chem istry control. See Chapter XI.M32, "One-Time Inspection," for an acceptable verification program.

VII.E3-7 VII.E3. Piping, piping Aluminum Treated water Loss of material! Chapter XI.M2, 'Water Chemistry" Yes, detection com ponents, pitting and crevice of aging effect (AP-38) and piping corrosion The AMP is to be augmented by isto be elements verifying the effectiveness of water evaluated chem istry control. See Chapter XI.M32, "One-Time Inspection," for an acceptable verification program.

VII.E3-8 VII.E3. Piping, piping Copper alloy Closed cycle Loss of material! Chapter XI. M21, "Closed-Cycle Cooling No com ponents, cooling water pitting, crevice, and Water System" (AP-12) and piping galvanic corrosion elements VII.E3-9 VII.E3. Piping, piping Copper alloy Treated water Loss of material! Chapter XI. M2, "Water Chemistry," for Yes, detection com ponents, pitting, crevice, and BWRwater of aging effect (AP-64) and piping galvanic corrosion is to be elements The AMP is to be augmented by evaluated verifying the effectiveness of water chem istry control. See Chapter z XI.M32, "One-Time Inspection," for an C

0 acceptable verification program.

m VII.E3-10 VII.E3. Piping, piping Copper alloy Closed cycle Loss of material! Chapter XI. M33, "Selective Leaching of No

~

OJ com ponents, >15% Zn cooling water selective leaching Materials"

~ (AP-43) and piping elements OAGI0000203_619

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OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

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Component

< Chapter XI. M33, "Selective Leaching of No VII.E3-11 VII.E3. Piping, piping Copper alloy Treated water Loss of material!

com ponents, >15% Zn selective leaching Materials" (AP-32) and piping elements VII.E3-12 VII.E3. Piping, piping Gray cast Treated water Loss of material! Chapter XI. M33, "Selective Leaching of No com ponents, iron selective leaching Materials" (AP-31) and piping elements VII.E3-13 VII.E3. Piping, piping Stainless Closed cycle Cracking! stress Chapter XI. M21, "Closed-Cycle Cooling No com ponents, steel cooling water corrosion cracking Water System" (AP-60) and piping >60°C (>140°F) elements VII.E3-14 VII.E3.1-b Piping, piping Stainless Treated water Cumulative fatigue Fatigue is a time-limited aging analysis Yes, VII.E3.2-b com ponents, steel damage! fatigue (TLAA) to be evaluated for the period of TLAA (A-62) and piping extended operation. See the standard elements Review Plan, Section 4.3 "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR 5421(c)(1)

VII.E3-15 VII.E3. Piping, piping Stainless Treated water Loss of material! Chapter XI. M2, "Water Chemistry," for Yes, detection com ponents, steel pitting and crevice BWRwater of aging effect (A-58) and piping corrosion isto be elements The AMP is to be augmented by evaluated verifying the effectiveness of water chemistry control. See Chapter XI.M32, "One-Time Inspection," for an acceptable verification program.

OAGI0000203_620

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(D VII AUXILIARY SYSTEMS (D E3 Reactor Water Cleanup System 3

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~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component VII.E3-16 VII.E3.1-a Piping, piping Stainless Treated water Cracking! stress Chapter XI. M25, "BWR Reactor Water No VII.E3.2-a com ponents, steel >60°C (>140°F) corrosion cracking, Cleanup System" (A-60) and piping intergranular stress elements corrosion cracking VII.E3-17 VII.E3.2-c Piping, piping Steel Air - indoor Cumulative fatigue Fatigue is a time-limited aging analysis Yes, com ponents, uncontrolled damage! fatigue (TLAA) to be evaluated for the period of TLAA (A-34) and piping extended operation. See the standard elements Review Plan, Section 4.3 "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR 5421(c)(1)

VII.E3-18 VII.E3. Piping, piping Steel Treated water Loss of material! Chapter XI. M2, "Water Chemistry," for Yes, detection com ponents, general, pitting, and BWRwater of aging effect (A-35) and piping crevice corrosion isto be elements The AMP is to be augmented by evaluated verifying the effectiveness of water chem istry control. See Chapter XI.M32, "One-Time Inspection," for an acceptable verification program.

VII.E3-19 VII.E3.3-d Regenerative Stainless Treated water Cracking! stress A plant-specific aging management Yes, plant-heat steel >60°C (>140°F) corrosion cracking program is to be evaluated. specific (A-85) exchanger components z

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E4. SHUTDOWN COOLING SYSTEM (OLOER 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 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 extended period of 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 (IV.C1) and the closed-cycle cooling water system (VII.C2).

September 2005 VII E4-1 NUREG-1801, Rev. 1 OAGI0000203_623

z VII AUXILIARY SYSTEMS C

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Mechanism Evaluation

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Component

< Chapter XI. M21, "Closed-Cycle Cooling No VII.E4-1 VII.E4.4-a Heat Stainless Closed cycle Loss of material!

exchanger steel; steel cooling water microbiologically Water System" (A-67) components with influenced corrosion stainless steel cladding VII.E4-2 VII.E4.4-a Heat Steel Closed cycle Loss of material! Chapter XI. M21, "Closed-Cycle Cooling No exchanger cooling water general, pitting, Water System" (A-63) components crevice, and galvanic corrosion VII.E4-3 VII.E4. Heat Stainless Closed cycle Reduction of heat Chapter XI. M21, "Closed-Cycle Cooling No exchanger steel cooling water transfer! fouling Water System" (AP-63) tubes VII.E4-4 VII.E4. Piping, piping Aluminum Treated water Loss of material! Chapter XI.M2, "Water Chemistry" Yes, detection com ponents, pitting and crevice of aging effect (AP-38) and piping corrosion The AMP is to be augmented by isto be elements verifying the effectiveness of water evaluated chem istry control. See Chapter XI.M32, "One-Time Inspection," for an acceptable verification program.

VII.E4-5 VII.E4. Piping, piping Copper alloy Closed cycle Loss of material! Chapter XI. M21, "Closed-Cycle Cooling No com ponents, cooling water pitting, crevice, and Water System" (AP-12) and piping galvanic corrosion elements OAGI0000203_624

(f)

(D VII AUXILIARY SYSTEMS (D E4 Shutdown Cooling System (Older BWR) 3 0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component VII.E4-6 VII.E4. Piping, piping Copper alloy Lubricating oil Loss of material! Chapter XI. M39, "Lubricating Oil Yes, detection com ponents, pitting and crevice Analysis" of aging effect (AP-47) and piping corrosion isto be elements The AMP is to be augmented by evaluated verifying the effectiveness of the lubricating oil analysis program. See Chapter XI. M32, "One-Time Inspection,"

for an acceptable verification program.

VII.E4-7 VII.E4. Piping, piping Copper alloy Treated water Loss of material! Chapter XI. M2, "Water Chemistry," for Yes, detection com ponents, pitting, crevice, and BWRwater of aging effect (AP-64) and piping galvanic corrosion isto be elements The AMP is to be augmented by evaluated verifying the effectiveness of water chem istry control. See Chapter XI.M32, "One-Time Inspection," for an acceptable verification program.

VII.E4-8 VII.E4. Piping, piping Copper alloy Closed cycle Loss of material! Chapter XI. M33, "Selective Leaching of No com ponents, >15% Zn cooling water selective leaching Materials" (AP-43) and piping elements VII.E4-9 VII.E4. Piping, piping Copper alloy Treated water Loss of material! Chapter XI. M33, "Selective Leaching of No com ponents, >15% Zn selective leaching Materials" (AP-32) and piping elements z VII.E4-10 VII.E4. Piping, piping Gray cast Treated water Loss of material! Chapter XI. M33, "Selective Leaching of No C

0 com ponents, iron selective leaching Materials" m

(AP-31) and piping

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Mechanism Evaluation

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< Chapter XI. M21, "Closed-Cycle Cooling No VII.E4-11 VII.E4. Piping, piping Stainless Closed cycle Cracking! stress com ponents, steel cooling water corrosion cracking Water System" (AP-60) and piping >60°C (>140°F) elements VII.E4-12 VII.E4. Piping, piping Stainless Lubricating oil Loss of material! Chapter XI. M39, "Lubricating Oil Yes, detection com ponents, steel pitting, crevice, and Analysis" of aging effect (AP-59) and piping microbiologically isto be elements influenced corrosion The AMP is to be augmented by evaluated verifying the effectiveness of the lubricating oil analysis program. See Chapter XI. M32, "One-Time Inspection,"

for an acceptable verification program.

m VII.E4-13 VII.E4.1-b Piping, piping Stainless Treated water Cumulative fatigue Fatigue is a time-limited aging analysis Yes, t (A-62) com ponents, steel and piping damage! fatigue (TLAA) to be evaluated for the period of extended operation. See the Standard TLAA elements Review Plan, Section 4.3 "Metal Fatigue," for acceptable methods for meeting the requirements of 10 CFR 5421(c)(1)

VII.E4-14 VII.E4.1-a Piping, piping Stainless Treated water Loss of material! Chapter XI. M2, "Water Chemistry," for Yes, detection com ponents, steel pitting and crevice BWRwater of aging effect (A-58) and piping corrosion isto be elements The AMP is to be augmented by evaluated verifying the effectiveness of water chem istry control. See Chapter XI.M32, "One-Time Inspection," for an acceptable verification program.

OAGI0000203_626

(f)

(D VII AUXILIARY SYSTEMS (D E4 Shutdown Cooling System (Older BWR) 3 0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component VII.E4-15 VII.E4.3-a Piping, piping Stainless Treated water Cracking! stress Chapter XI.M7, "BWR Stress Corrosion No VII.E4.1-c com ponents, steel >60°C (> 140°F) corrosion cracking Cracking," and (A-61) and piping elements Chapter XI. M2, "Water Chemistry," for BWRwater VII.E4-16 VII.E4. Piping, piping Steel Lubricating oil Loss of material! Chapter XI. M39, "Lubricating Oil Yes, detection com ponents, general, pitting, and Analysis" of aging effect (AP-30) and piping crevice corrosion isto be elements The AMP is to be augmented by evaluated verifying the effectiveness of the lubricating oil analysis program. See Chapter XI. M32, "One-Time Inspection,"

for an acceptable verification program.

VII.E4-17 VII.E4.1-a Piping, piping Steel Treated water Loss of material! Chapter XI.M2, "Water Chemistry," for Yes, detection VII.E4.2-a com ponents, general, pitting, and BWRwater of aging effect (A-35) and piping crevice corrosion isto be elements The AMP is to be augmented by evaluated verifying the effectiveness of water chem istry control. See Chapter XI.M32, "One-Time Inspection," for an acceptable verification program.

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F1. CONTROL ROOM AREA VENTILATION SYSTEM Systems, Structures, and Components This section discusses the control room area ventilation system (with warm moist air as the normal environment), which contains ducts, piping and fittings, equipment frames and housings, flexible collars and seals, filters, and heating and cooling air handlers. 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 control room area ventilation system are governed by Group B Quality Standards.

With respect to filters and seals, 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 Nuclear Energy Institute (NEO, dated March 10, 2000. Specifically, components that function as system filters and seals 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.

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 extended period of operation are included in VII.J.

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

System Interfaces The system that interfaces with the control room area ventilation system is the auxiliary and radwaste area ventilation system (VII.F2). The cooling coils receive their cooling water from other systems, such as the hot water heating system or the chilled water cooling system.

September 2005 VII F1-1 NUREG-1801, Rev. 1 OAGI0000203_629

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< VILF1-1 A plant-specific aging management VILF1.4-a Ducting and Stainless Condensation Loss of material! Yes, plant-components steel pitting and crevice program is to be evaluated. specific (A-D9) corrosion VILF1-2 VILF1.4-a Ducting and Steel Air - indoor Loss of material! Chapter XL M36, "External Surfaces No VILF1.1-a components uncontrolled general corrosion Monitoring" (A-1D) external (External) surfaces VILF1-3 VILF1.4-a Ducting and Steel Condensation Loss of material! Chapter XLM38, "Inspection of Internal No VILF1.1-a components (Internal) general, pitting, Surfaces in Miscellaneous Piping and (A-D8) internal crevice, and (for drip Ducting Components" surfaces pans and drain lines) microbiologically influenced corrosion VILF1-4 VILF1.1-a Ducting Steel Air - indoor Loss of material! Chapter XL M36, "External Surfaces No closure bolting uncontrolled general corrosion Monitoring" (A-1D5) (External)

VILF1-5 VILF1.1-c Elastomer Elastomers Air - indoor Loss of material! A plant-specific aging management Yes, plant-seals and uncontrolled wear program is to be evaluated. specific (A-73) components (External)

VILF1-6 VILF1.1-c Elastomer Elastomers Air - indoor Loss of material! A plant-specific aging management Yes, plant-seals and uncontrolled wear program is to be evaluated. specific (A-18) components (Internal)

VILF1-7 VILF1.4-b Elastomer Elastomers Air - indoor Hardening and loss A plant-specific aging management Yes, plant-VILF1.1-b seals and uncontrolled of strength! program is to be evaluated. specific (A-17) components (Internal! elastomer External) degradation VILF1-8 VILF1. Heat Copper alloy Closed cycle Loss of material! Chapter XL M21, "Closed-Cycle Cooling No exchanger cooling water pitting, crevice, and Water System" (AP-34) components galvanic corrosion OAGI0000203_630

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(D VII AUXILIARY SYSTEMS (D F1 Control Room Area Ventilation System 3

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~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component VILF1-9 VILF1. Heat Copper alloy Treated water Loss of material! Chapter XL M33, "Selective Leaching of No exchanger >15% Zn selective leaching Materials" (AP-65) components VILF1-10 VILF1. Heat Steel Air - indoor Loss of material! Chapter XL M36, "External Surfaces No exchanger uncontrolled general, pitting, and Monitoring" (AP-41) components (External) crevice corrosion VILF1-11 VILF1. Heat Steel Closed cycle Loss of material! Chapter XL M21, "Closed-Cycle Cooling No exchanger cooling water general, pitting, Water System" (A-63) components crevice, and galvanic corrosion VILF1-12 VILF1. Heat Copper Alloy Closed cycle Reduction of heat Chapter XL M21, "Closed-Cycle Cooling No exchanger cooling water transfer! fouling Water System" (AP-SO) tubes VILF1-13 VILF1. Heat Steel Closed cycle Reduction of heat Chapter XL M21, "Closed-Cycle Cooling No exchanger cooling water transfer! fouling Water System" (AP-77) tubes VILF1-14 VILF1. Piping, piping Aluminum Condensation Loss of material! A plant-specific aging management Yes, plant-components, pitting and crevice program is to be evaluated. specific (AP-74) and piping corrosion elements VILF1-15 VILF1. Piping, piping Copper alloy Closed cycle Loss of material! Chapter XL M21, "Closed-Cycle Cooling No components, cooling water pitting, crevice, and Water System" (AP-12) and piping galvanic corrosion z elements C

0 VILF1-16 VILF1.2-a A plant-specific aging management m Piping, piping Copper alloy Condensation Loss of material! Yes, plant-

~

OJ (A-46) components, and piping (External) pitting and crevice corrosion program is to be evaluated. specific

~ elements OAGI0000203_631

z VII AUXILIARY SYSTEMS C

0 F1 Control Room Area Ventilation System m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< VILF1-17 Chapter XL M33, "Selective Leaching of No VILF1. Piping, piping Copper alloy Closed cycle Loss of material!

components, >15% Zn cooling water selective leaching Materials" (AP-43) and piping elements VILF1-18 VILF1. Piping, piping Gray cast Treated water Loss of material! Chapter XL M33, "Selective Leaching of No components, iron selective leaching Materials" (AP-31) and piping elements VILF1-19 VILF1. Piping, piping Steel Lubricating oil Loss of material! Chapter XL M39, "Lubricating Oil Yes, detection components, general, pitting, and Analysis" of aging (AP-3D) and piping crevice corrosion effects is to be elements The AMP is to be augmented by evaluated verifying the effectiveness of the lubricating oil analysis program. See Chapter XLM32, "One-Time Inspection,"

for an acceptable verification proQram.

VILF1-2D VILF1.3-a Piping, piping Steel Closed cycle Loss of material! Chapter XL M21, "Closed-Cycle Cooling No components, cooling water general, pitting, and Water System" (A-25) piping crevice corrosion elements, and tanks OAGI0000203_632

F2. Auxiliary and Radwaste Area Ventilation System Systems, Structures, and Components This section discusses the auxiliary and radwaste areas ventilation systems (with warm moist air as the normal environment) and contains ducts, piping and fittings, equipment frames and housings, flexible collars and seals, filters, and heating and cooling air handlers. 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 auxiliary and radwaste area ventilation system are governed by Group B Quality Standards.

With respect to filters and seals, these items 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 Nuclear Energy Institute (NEO, dated March 10,2000. Specifically, components that function as system filters and seals 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 (N FPA) standards for fire protection equipment.

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 extended period of operation are included in VII.J.

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

System Interfaces The systems that interface with the auxiliary and radwaste area ventilation system are the control room area ventilation system (VII.F1) and the diesel generator building ventilation system (VII.F4). The cooling coils receive their cooling water from other systems, such as the hot water heating system or the chilled water cooling system.

September 2005 VII F2-1 NUREG-1801, Rev. 1 OAGI0000203_633

z VII AUXILIARY SYSTEMS C

0 F2 Auxiliary and Radwaste Area Ventilation System m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< VILF2-1 A plant-specific aging management VILF2.4-a Ducting and Stainless Condensation Loss of material! Yes, plant-components steel pitting and crevice program is to be evaluated. specific (A-D9) corrosion VILF2-2 VILF2.1-a Ducting and Steel Air - indoor Loss of material! Chapter XL M36, "External Surfaces No VILF2.4-a components uncontrolled general corrosion Monitoring" (A-1D) external (External) surfaces VILF2-3 VILF2.1-a Ducting and Steel Condensation Loss of material! Chapter XLM38, "Inspection of Internal No VILF2.4-a components (Internal) general, pitting, Surfaces in Miscellaneous Piping and (A-D8) internal crevice, and (for drip Ducting Components" surfaces pans and drain lines) microbiologically influenced corrosion VILF2-4 VILF2.1-a Ducting Steel Air - indoor Loss of material! Chapter XL M36, "External Surfaces No closure bolting uncontrolled general corrosion Monitoring" (A-1D5) (External)

VILF2-5 VILF2.1-c Elastomer Elastomers Air - indoor Loss of material! A plant-specific aging management Yes, plant-seals and uncontrolled wear program is to be evaluated. specific (A-73) components (External)

VILF2-6 VILF2.1-c Elastomer Elastomers Air - indoor Loss of material! A plant-specific aging management Yes, plant-seals and uncontrolled wear program is to be evaluated. specific (A-18) components (Internal)

VILF2-7 VILF2.1-b Elastomer Elastomers Air - indoor Hardening and loss A plant-specific aging management Yes, plant-VILF2.4-b seals and uncontrolled of strength! program is to be evaluated. specific (A-17) components (Internal! elastomer External) degradation VILF2-8 VILF2. Heat Steel Air - indoor Loss of material! Chapter XL M36, "External Surfaces No exchanger uncontrolled general, pitting, and Monitoring" (AP-41) components (External) crevice corrosion OAGI0000203_634

(f)

(D VII AUXILIARY SYSTEMS (D F2 Auxiliary and Radwaste Area Ventilation System 3

0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component VILF2-9 VILF2. Heat Steel Closed cycle Loss of material! Chapter XL M21, "Closed-Cycle Cooling No exchanger cooling water general, pitting, Water System" (A-63) components crevice, and galvanic corrosion VILF2-10 VILF2. Heat Copper Alloy Closed cycle Reduction of heat Chapter XL M21, "Closed-Cycle Cooling No exchanger cooling water transfer! fouling Water System" (AP-SO) tubes VILF2-11 VILF2. Heat Steel Closed cycle Reduction of heat Chapter XL M21, "Closed-Cycle Cooling No exchanger cooling water transfer! fouling Water System" (AP-77) tubes VILF2-12 VILF2. Piping, piping Aluminum Condensation Loss of material! A plant-specific aging management Yes, plant-components, pitting and crevice program is to be evaluated. specific (AP-74) and piping corrosion elements VILF2-13 VILF2. Piping, piping Copper alloy Closed cycle Loss of material! Chapter XL M21, "Closed-Cycle Cooling No components, cooling water pitting, crevice, and Water System" (AP-12) and piping galvanic corrosion elements VILF2-14 VILF2.2-a Piping, piping Copper alloy Condensation Loss of material! A plant-specific aging management Yes, plant-components, (External) pitting and crevice program is to be evaluated. specific (A-46) and piping corrosion elements VILF2-15 VILF2. Piping, piping Copper alloy Closed cycle Loss of material! Chapter XL M33, "Selective Leaching of No z components, >15% Zn cooling water selective leaching Materials" C

0 (AP-43) and piping rn elements

~

OJ VILF2-16 VILF2. Piping, piping Gray cast Treated water Loss of material! Chapter XL M33, "Selective Leaching of No

~ com ponents, iron selective leaching Materials" (AP-31) and piping elements OAGI0000203_635

z VII AUXILIARY SYSTEMS C

0 F2 Auxiliary and Radwaste Area Ventilation System m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< VILF2-17 VILF2. Piping, piping Steel Lubricating oil Loss of material! Chapter XL M39, "Lubricating Oil Yes, detection components, general, pitting, and Analysis" of aging (AP-30) and piping crevice corrosion effects is to be elements The AMP is to be augmented by evaluated verifying the effectiveness of the lubricating oil analysis program. See Chapter XLM32, "One-Time Inspection,"

for an acceptable verification program.

VILF2-18 VILF2.3-a Piping, piping Steel Closed cycle Loss of material! Chapter XLM21, "Closed-Cycle Cooling No components, cooling water general, pitting, and Water System" (A-25) piping crevice corrosion elements, and tanks OAGI0000203_636

F3. PRIMARY CONTAINMENT HEATING AND VENTILATION SYSTEM Systems, Structures, and Components This section discusses the primary containment heating and ventilation system (with warm moist air as the normal environment), which contains ducts, piping and fittings, equipment frames and housings, flexible collars and seals, filters, and heating and cooling air handlers.

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 primary containment heating and ventilation system are governed by Group C Quality Standards.

With respect to filters and seals, 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 (NEO, dated March 10, 2000. Specifically, components that function as system filters and seals 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.

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 extended period of operation are included in VII.J.

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

System Interfaces The systems that interface with the primary containment heating and ventilation system are the closed-cycle cooling water system (VII.C2) and the PWR and BWR containments (II.A and II.B, respectively). The cooling coils receive their cooling water from other systems, such as the hot water heating system or the chilled water cooling system.

September 2005 VII F3-1 NUREG-1801, Rev. 1 OAGI0000203_637

z VII AUXILIARY SYSTEMS C

0 F3 Primary Containment Heating and Ventilation System m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< VILF3-1 A plant-specific aging management VILF3.4-a Ducting and Stainless Condensation Loss of material! Yes, plant-components steel pitting and crevice program is to be evaluated. specific (A-D9) corrosion VILF3-2 VILF3.1-a Ducting and Steel Air - indoor Loss of material! Chapter XL M36, "External Surfaces No VILF3.4-a components uncontrolled general corrosion Monitoring" (A-1D) external (External) surfaces VILF3-3 VILF3.4-a Ducting and Steel Condensation Loss of material! Chapter XLM38, "Inspection of Internal No VILF3.1-a components (Internal) general, pitting, Surfaces in Miscellaneous Piping and (A-D8) internal crevice, and (for drip Ducting Components" surfaces pans and drain lines) microbiologically influenced corrosion VILF3-4 VILF3.1-a Ducting Steel Air - indoor Loss of material! Chapter XL M36, "External Surfaces No closure bolting uncontrolled general corrosion Monitoring" (A-1D5) (External)

VILF3-5 VILF3.1-c Elastomer Elastomers Air - indoor Loss of material! A plant-specific aging management Yes, plant-seals and uncontrolled wear program is to be evaluated. specific (A-73) components (External)

VILF3-6 VILF3.1-c Elastomer Elastomers Air - indoor Loss of material! A plant-specific aging management Yes, plant-seals and uncontrolled wear program is to be evaluated. specific (A-18) components (Internal)

VILF3-7 VILF3.1-b Elastomer Elastomers Air - indoor Hardening and loss A plant-specific aging management Yes, plant-VILF3.4-b seals and uncontrolled of strength! program is to be evaluated. specific (A-17) components (Internal! elastomer External) degradation VILF3-8 VILF3. Heat Copper alloy Closed cycle Loss of material! Chapter XL M21, "Closed-Cycle Cooling No exchanger cooling water pitting, crevice, and Water System" (AP-34) components galvanic corrosion OAGI0000203_638

(f)

(D VII AUXILIARY SYSTEMS (D F3 Primary Containment Heating and Ventilation System 3

0-

~ Structure Aging Effect! Further

'"oo Item Link and!or Material Environment Mechanism Aging Management Program (AMP)

Evaluation U1 Component VILF3-9 VILF3. Heat Copper alloy Treated water Loss of material! Chapter XL M33, "Selective Leaching of No exchanger >15% Zn selective leaching Materials" (AP-65) components VILF3-10 VILF3. Heat Steel Air - indoor Loss of material! Chapter XL M36, "External Surfaces No exchanger uncontrolled general, pitting, and Monitoring" (AP-41) components (External) crevice corrosion VILF3-11 VILF3. Heat Steel Closed cycle Loss of material! Chapter XL M21, "Closed-Cycle Cooling No exchanger cooling water general, pitting, Water System" (A-63) components crevice, and galvanic corrosion VILF3-12 VILF3. Heat Copper Alloy Closed cycle Reduction of heat Chapter XLM21, "Closed-Cycle Cooling No exchanger cooling water transfer! fouling Water System" (AP-SO) tubes VILF3-13 VILF3. Heat Steel Closed cycle Reduction of heat Chapter XL M21, "Closed-Cycle Cooling No exchanger cooling water transfer! fouling Water System" (AP-77) tubes VILF3-14 VILF3. Piping, piping Aluminum Condensation Loss of material! A plant-specific aging management Yes, plant-components, pitting and crevice program is to be evaluated. specific (AP-74) and piping corrosion elements VILF3-15 VILF3. Piping, piping Copper alloy Closed cycle Loss of material! Chapter XL M21, "Closed-Cycle Cooling No components, cooling water pitting, crevice, and Water System" (AP-12) and piping galvanic corrosion z elements C

0 VILF3-16 VILF3.2-a A plant-specific aging management m Piping, piping Copper alloy Condensation Loss of material! Yes, plant-

~

OJ (A-46) components, and piping (External) pitting and crevice corrosion program is to be evaluated. specific

~ elements OAGI0000203_639

z VII AUXILIARY SYSTEMS C

0 F3 Primary Containment Heating and Ventilation System m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< VILF3-17 Chapter XL M33, "Selective Leaching of No VILF3. Piping, piping Copper alloy Closed cycle Loss of material!

components, >15% Zn cooling water selective leaching Materials" (AP-43) and piping elements VILF3-18 VILF3. Piping, piping Gray cast Closed cycle Loss of material! Chapter XL M33, "Selective Leaching of No components, iron cooling water selectil. leaching Materials" (A-50) and piping elements VILF3-19 VILF3. Piping, piping Steel Lubricating oil Loss of material! Chapter XL M39, "Lubricating Oil Yes, detection components, general, pitting, and Analysis" of aging (AP-30) and piping crevice corrosion effects is to be elements The AMP is to be augmented by evaluated verifying the effectiveness of the lubricating oil analysis program. See Chapter XLM32, "One-Time Inspection,"

for an acceptable verification proQram.

VILF3-20 VILF3.3-a Piping, piping Steel Closed cycle Loss of material! Chapter XL M21, "Closed-Cycle Cooling No components, cooling water general, pitting, and Water System" (A-25) piping crevice corrosion elements, and tanks OAGI0000203_640

F4. DIESEL GENERATOR BUILDING VENTILATION SYSTEM Systems, Structures, and Components This section discusses the diesel generator building ventilation system (with warm moist air as the normal environment), which contains ducts, piping and fittings, equipment frames and housings, flexible collars and seals, and heating and cooling air handlers. 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 diesel generator building ventilation system are governed by Group C Quality Standards.

With respect to seals, these items are to be addressed consistent with the Nuclear Regulatory Commission (N RC) position on consumables, provided in the NRC letter from Christopher I.

Grimes to Douglas J. Walters of Nuclear Energy Institute (NEO, dated March 10,2000.

Specifically, components that function as system seals 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 (N FPA) standards for fire protection equipment.

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 extended period of operation are included in VII.J.

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

System Interfaces The system that interfaces with the diesel generator building system is the auxiliary and radwaste area ventilation system (VII.F2). The cooling coils receive their cooling water from other systems, such as the hot water heating system or the chilled water cooling system.

September 2005 VII F4-1 NUREG-1801, Rev. 1 OAGI0000203_641

z VII AUXILIARY SYSTEMS C

0 F4 Diesel Generator Building Ventilation System m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< VILF4-1 VILF4.1-a Air - indoor Ducting and Steel Loss of material! Chapter XL M36, "External Surfaces No components uncontrolled general corrosion Monitoring" (A-10) external (External) surfaces VILF4-2 VILF4.1-a Ducting and Steel Condensation Loss of material! Chapter XLM38, "Inspection of Internal No components (Internal) general, pitting, Surfaces in Miscellaneous Piping and (A-08) internal crevice, and (for drip Ducting Components" surfaces pans and drain lines) microbiologically influenced corrosion VILF4-3 VILF4.1-a Ducting Steel Air - indoor Loss of material! Chapter XL M36, "External Surfaces No closure bolting uncontrolled general corrosion Monitoring" (A-105) (External)

VILF4-4 VILF4.1-c Elastomer Elastomers Air - indoor Loss of material! A plant-specific aging management Yes, plant-seals and uncontrolled wear program is to be evaluated. specific (A-73) components (External)

VILF4-5 VILF4.1-c Elastomer Elastomers Air - indoor Loss of material! A plant-specific aging management Yes, plant-seals and uncontrolled wear program is to be evaluated. specific (A-18) components (Internal)

VILF4-6 VILF4.1-b Elastomer Elastomers Air - indoor Hardening and loss A plant-specific aging management Yes, plant-seals and uncontrolled of strength! program is to be evaluated. specific (A-17) components (Internal! elastomer External) degradation VILF4-7 VILF4. Heat Steel Air - indoor Loss of material! Chapter XL M36, "External Surfaces No exchanger uncontrolled general, pitting, and Monitoring" (AP-41) components (External) crevice corrosion OAGI0000203_642

VII AUXILIARY SYSTEMS (f)

F4 Diesel Generator Building Ventilation System (D

~ Structure 3 Aging Effect! Further 0- Item Link and!or Material Environment Aging Management Program (AMP)

~ Mechanism Evaluation Component

'"oo VILF4-8 VILF4. Heat Steel Closed cycle Loss of material! Chapter XLM21, "Closed-Cycle Cooling No U1 exchanger cooling water general, pitting, Water System" (A-63) components crevice, and galvanic corrosion VILF4-9 VILF4. Heat Steel Closed cycle Reduction of heat Chapter XL M21, "Closed-Cycle Cooling No exchanger cooling water transfer! fouling Water System" (AP-77) tubes VILF4-10 VILF4. Piping, piping Aluminum Condensation Loss of material! A plant-specific aging management Yes, plant-components, pitting and crevice program is to be evaluated. specific (AP-74) and piping corrosion elements VILF4-11 VILF4. Piping, piping Copper alloy Closed cycle Loss of material! Chapter XL M21, "Closed-Cycle Cooling No components, cooling water pitting, crevice, and Water System" (AP-12) and piping galvanic corrosion elements VILF4-12 VILF4.2-a Piping, piping Copper alloy Condensation Loss of material! A plant-specific aging management Yes, plant-components, (External) pitting and crevice program is to be evaluated. specific (A-46) and piping corrosion elements VILF4-13 VILF4. Piping, piping Copper alloy Closed cycle Loss of material! Chapter XL M33, "Selective Leaching of No components, >15% Zn cooling water selective leaching Materials" (AP-43) and piping elements VILF4-14 VILF4. Piping, piping Gray cast Treated water Loss of material! Chapter XL M33, "Selective Leaching of No components, iron selective leaching Materials" z

C (AP-31) and piping

0 elements

~

OJ

~

OAGI0000203_643

z VII AUXILIARY SYSTEMS C

0 F4 Diesel Generator Building Ventilation System m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< VILF4-15 VILF4. Piping, piping Steel Lubricating oil Loss of material! Chapter XL M39, "Lubricating Oil Yes, detection components, general, pitting, and Analysis" of aging (AP-30) and piping crevice corrosion effects is to be elements The AMP is to be augmented by evaluated verifying the effectiveness of the lubricating oil analysis program. See Chapter XLM32, "One-Time Inspection,"

for an acceptable verification program.

VILF4-16 VILF4.3-a Piping, piping Steel Closed cycle Loss of material! Chapter XL M21, "Closed-Cycle Cooling No components, cooling water general, pitting, and Water System" (A-25) piping crevice corrosion elements, and tanks OAGI0000203_644

G. FIRE PROTECTION Systems, Structures, and Components This section discusses the fire protection systems for both boiling water reactors (BWRs) and pressurized water reactors (PWRs), which consist of several Class 1 structures, mechanical systems, and electrical components. The Class 1 structures include the intake structure, the turbine building, the auxiliary building, the diesel generator building, and the primary containment.

Structural components include fire barrier walls, ceilings, floors, fire doors, and penetration seals. Mechanical systems include the high pressure service water system, the reactor coolant pump oil collect system, and the diesel fire system. Mechanical components include piping and fittings, filters, fire hydrants, mulsifiers, pumps, sprinklers, strainers, and valves (including containment isolation 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 fire protection system are governed by Group C Quality Standards.

With respect to filters, seals, portable fire extinguishers, and fire hoses, these items 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 Nuclear Energy Institute (NEO, dated March 10, 2000. Specifically, components that function as system filters, seals, portable fire extinguishers, and fire hoses 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 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 extended period of operation are included in VII.J.

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

System Interfaces The systems and structures that interface with the fire protection system include various Class 1 structures and component supports (liLA and III.B), the electrical components (VI.A and VI.B), the closed-cycle cooling water system (VII.C2), and the diesel fuel oil system (VII.H1).

September 2005 VII G-1 NUREG-1801, Rev. 1 OAGI0000203_645

z VII AUXILIARY SYSTEMS C

0 G Fire Protection m

~

OJ Structure

~ Aging Effect! Further Item Link and!or Material Environment Aging Management Program (AMP)

Mechanism Evaluation

0 (D

Component

< VII.G-1 Air - indoor VII.G.3-a Fire barrier Elastomers Increased hardness, Chapter XI. M26, "Fire Protection" No VII.G.1-a penetration uncontrolled shrinkage and loss (A-19) VII.G.2-a seals of strength!

VII.G.4-a weathering VII.G-2 VII.G.1-a Fire barrier Elastomers Air - outdoor Increased hardness, Chapter XI. M26, "Fire Protection" No VII.G.3-a penetration shrinkage and loss (A-20) VII.G.4-a seals of strength!

VII.G.2-a weathering VII.G-3 VII.G.2-d Fire rated Steel Air - indoor Loss of material! Chapter XI. M26, "Fire Protection" No VII.G.5-c doors uncontrolled wear (A-21) VII.G.3-d VII.G.4-d VII.G.1-d VII.G-4 VII.G.1-d Fire rated Steel Air - outdoor Loss of material! Chapter XI. M26, "Fire Protection" No VII.G.2-d doors wear (A-22) VII.G.4-d VII.G.3-d VII.G-5 VII.G. Heat Steel Air - indoor Loss of material! Chapter XI.M36, "External Surfaces No exchanger uncontrolled general, pitting, and Monitoring" (AP-41) components (External) crevice corrosion VII.G-6 VII.G. Heat Steel Air - outdoor Loss of material! Chapter XI. M36, "External Surfaces No exchanger (External) general, pitting, and Monitoring" (AP-40) components crevice corrosion VII.G-7 VII.G. Heat Stainless Raw water Reduction of heat Chapter XI. M20, "Open-Cycle Cooling No exchanger steel transfer! fouling Water System" (AP-61) tubes VII.G-8 VII.G. Piping, piping Aluminum Raw water Loss of material! Chapter XI. M26, "Fire Protection" No components, pitting and crevice (AP-83) and piping corrosion elements OAGI0000203_646

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