CHAPTER IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM

CHAPTER IV
REACTOR VESSEL, INTERNALS, AND REACTOR
COOLANT SYSTEM
September 2004
IV-i
Proposed Draft NUREG-1801
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Proposed Draft NUREG-1801
IV-ii
September 2004
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)
(refined outline to be added when issued for public comment)
Explanation of September 30, 2004 changes in preliminary interim draft chapter
outline and aging management review (AMR) tables: Within the AMR tables, this update
process increases license renewal review efficiency by:
• Consolidating components (combining similar or equivalent components with
matching materials, environment and AMP into a single line-item),
• Increasing consistency between Material/Environment/Aging effects/aging
management Program (MEAP) combinations between systems (some existing
MEAPs had multiple definitions that, based on the aging effect, could be broadened
to envelope these into a singe MEAP),
• Correcting any inconsistencies in the 2001 edition of the GALL Report,
• Updating references to the appropriate aging management programs, and
• Incorporating line-item changes based on approved staff SER positions or interim
staff guidance.
The principal effect of this change is that the tables present the MEAP combinations at a
higher level, and the prior detail within a structure or component line item is no longer
explicitly presented. Consequently, the identifiers for subcomponents within a line item are
no longer presented in the tables. As a result, the introductory listings of these
subcomponents (originally in text preceding each table) have been deleted.
The following AMR tables contain a revised “Item” column and a new column titled “Link”,
which was not contained in the July 2001 revision. The “Item” number is a unique identifier
September 2004
IV-iii
Proposed Draft NUREG-1801
that is used for traceability and, as mentioned above, no longer presents the detailed
subcomponent identification. The link identifies the original item in the current version of the
GALL Report when applicable (items added to this list refer to bases statements not yet
available).
By January 30, 2005, the NRC staff plans to issue a revised GALL Report (NUREG-1801)
and SRP-LR (NUREG-1800) for public comment. NRC anticipates re-numbering the lineitems to provide an improved unique identifier as part of the public comment document.
Also as part of the public comment process, the NRC will issue a NUREG documenting the
basis for the proposed changes to the GALL Report and the SRP-LR. This NUREG bases
document will be an aid for those reviewing the revised documents to understand what was
changed and the basis for the proposed changes.
Proposed Draft NUREG-1801
IV-iv
September 2004
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September 2004
IV-v
Proposed Draft NUREG-1801
A1. REACTOR VESSEL (BOILING WATER REACTOR)
A1.1
Top Head Enclosure
A1.1.1
A1.1.2
A1.1.3
A1.1.4
A1.1.5
A1.2
Vessel Shell
A1.2.1
A1.2.2
A1.2.3
A1.2.4
A1.2.5
A1.2.6
A1.2.7
A1.3
Main Steam
Feedwater
Control Rod Drive (CRD) Return Line
Low Pressure Coolant Injection (LPCI) or Residual Heat Removal (RHR)
Injection Mode
Nozzles Safe Ends
A1.4.1
A1.4.2
A1.4.3
A1.4.4
A1.4.5
A1.5
Vessel Flange
Upper Shell
Intermediate Nozzle Shell
Intermediate Beltline Shell
Lower Shell
Beltline Welds
Attachment Welds
Nozzles
A1.3.1
A1.3.2
A1.3.3
A1.3.4
A1.4
Top Head
Nozzles (Vent, Top Head Spray or Reactor Core Isolation Cooling [RCIC],
and Spare)
Head Flange
Closure Studs and Nuts
Vessel Flange Leak Detection Line
High Pressure Core Spray (HPCS)
Low Pressure Core Spray (LPCS)
CRD Return Line
Recirculating Water (Inlet and Outlet)
LPCI or RHR Injection Mode
Penetrations
A1.5.1
A1.5.2
A1.5.3
A1.5.4
A1.5.5
A1.5.6
CRD Stub Tubes
Instrumentation
Jet Pump Instrument
Standby Liquid Control
Flux Monitor
Drain Line
A1.6
Bottom Head
A1.7
Support Skirt and Attachment Welds
September 2004
IV.A1-1
Proposed Draft NUREG-1801
A1. REACTOR VESSEL (BOILING WATER REACTOR)
Systems, Structures, and Components
This section comprises the boiling water reactor (BWR) pressure vessel and consists of the
vessel shell and flanges; attachment welds; the 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 CRD stub tubes, instrumentation, standby liquid control, flux monitor,
and drain lines; and control rod drive mechanism housings. The support skirt and attachment
welds for vessel supports are also included in the table. Based on Regulatory Guide 1.26,
“Quality Group Classifications and Standards for Water, Steam, and Radioactive-WasteContaining Components of Nuclear Power Plants,” all structures and components that comprise
the reactor vessel are governed by Group A Quality Standards.
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).
Proposed Draft NUREG-1801
IV.A1-2
September 2004
September 2004
IV
A1
Item
R-68
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel (BWR)
IV.A1-3
Structure
and/or
Material
Component
Stainless
IV.A1.4- Nozzle safe
ends
steel, nickel
a
High pressure alloy
core spray
Low pressure
core spray
Control rod
drive return line
Recirculating
water
Low pressure
coolant injection
or RHR
injection mode
Link
Proposed Draft NUREG-1801
R-66
IV.A1.3- Nozzles
Control rod
c
R-65
IV.A1.3- Nozzles
Feedwater
b
Environment
Reactor coolant
Steel (without Reactor coolant
lining/coating
drive return line or with
degraded
lining/coating)
Steel (without Reactor coolant
lining/coating
or with
degraded
lining/coating)
Aging Effect/
Mechanism
Aging Management Program (AMP)
Further
Evaluation
Cracking/ stress
corrosion cracking
and intergranular
stress corrosion
cracking
Chapter XI.M7, “BWR Stress
Corrosion Cracking,” and
Cracking/ cyclic
loading
Chapter XI.M6, “BWR Control Rod
Drive Return Line Nozzle”
No
Cracking/ cyclic
loading
Chapter XI.M5, “BWR Feedwater
Nozzle”
No
No
Chapter XI.M2, “Water Chemistry,” for
BWR water in BWRVIP-29 (EPRI TR103515)
Proposed Draft NUREG-1801
IV
A1
Item
R-67
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel (BWR)
Structure
and/or
Material
Component
Steel
IV.A1.3- Nozzles
Low pressure
e
coolant injection
or RHR
injection mode
Link
Environment
Reactor coolant
and neutron flux
Aging Effect/
Mechanism
Loss of fracture
toughness/ neutron
irradiation
embrittlement
Aging Management Program (AMP)
Further
Evaluation
IV.A1-4
Neutron irradiation embrittlement is a Yes,
time-limited aging analysis (TLAA) to TLAA
be evaluated for the period of
extended operation for all ferritic
materials that have a neutron fluence
greater than 1017 n/cm2 (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 embrittlement
on: (a) the adjusted reference
temperature, the plant’s pressuretemperature limits, (b) the Charpy
upper shelf energy, and (c) the
equivalent margins analyses
performed in accordance with 10 CFR
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
Embrittlement” for acceptable methods
for meeting the requirements of 10
CFR 54.21(c).
September 2004
September 2004
IV
A1
Item
R-69
IV.A1-5
R-04
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel (BWR)
Structure
and/or
Component
IV.A1.5- Penetrations
Control rod
a
drive stub tubes
Instrumentation
Jet pump
instrument
Standby liquid
control
Flux monitor
Drain line
Link
IV.A1.2-b
IV.A1.2-a
IV.A1.3-a
IV.A1.4-b
IV.A1.1-b
IV.A1.5-b
IV.A1.6-a
Piping, piping
components,
and piping
elements
Material
Environment
Stainless
steel, nickel
alloy
Reactor coolant
Proposed Draft NUREG-1801
Steel,
Reactor coolant
stainless
steel, cast
austenitic
stainless
steel, carbon
steel with
nickel-alloy or
stainless steel
cladding,
nickel-alloy
Aging Effect/
Mechanism
Aging Management Program (AMP)
Further
Evaluation
Cracking/ stress
corrosion cracking,
intergranular stress
corrosion cracking,
cyclic loading
Chapter XI.M8, “BWR Penetrations,”
and
Cumulative fatigue
damage/ fatigue
Fatigue is a time-limited aging analysis Yes,
(TLAA) to be performed for the period TLAA
of extended operation, and, for Class 1
components, environmental effects on
fatigue are to be addressed. See the
Standard Review Plan, Section 4.3
“Metal Fatigue,” for acceptable
methods for meeting the requirements
of 10 CFR 54.21(c)(1)(i) and (ii), and
for addressing environmental effects
on fatigue.
No
Chapter XI.M2, “Water Chemistry,” for
BWR water in BWRVIP-29 (EPRI TR103515)
See Chapter X.M1 of this report for
meeting the requirements of
10 CFR 54.21(c)(1)(iii).
Proposed Draft NUREG-1801
IV
A1
Item
R-70
R-60
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel (BWR)
Structure
and/or
Material
Component
Steel
IV.A1.7- Support skirt
and attachment
a
welds
Link
IV.A1.1- Top head
enclosure
c
Closure studs
and nuts
IV.A1-6
R-61
Aging Effect/
Mechanism
Air – indoor
uncontrolled
Cumulative fatigue
damage/ fatigue
Fatigue is a time-limited aging analysis Yes,
(TLAA) to be evaluated for the period TLAA
of extended operation. See the
Standard Review Plan, Section 4.3
“Metal Fatigue,” for acceptable
methods for meeting the requirements
of 10 CFR 54.21(c)(1).
Cracking/ stress
corrosion cracking
and intergranular
stress corrosion
cracking
Chapter XI.M3, “Reactor Head Closure No
Studs”
High strength Air with reactor
low alloy steel coolant leakage
Maximum
tensile
strength <
1172 MPa
(<170 Ksi)
Aging Management Program (AMP)
IV.A1.1- Top head
enclosure
d
Stainless
steel, nickel
alloy
Air with reactor
coolant leakage
Cracking/ stress
corrosion cracking
and intergranular
stress corrosion
cracking
A plant-specific aging management
program is to be evaluated because
existing programs may not be able to
mitigate or detect crack initiation and
growth due to SCC of vessel flange
leak detection line.
IV.A1.1- Top head
enclosure
a
Steel
Reactor coolant
Loss of material/
general, pitting and
crevice corrosion
Chapter XI.M1, “ASME Section XI
No
Inservice Inspection, Subsections IWB,
IWC, and IWD,” for Class 1
components and Chapter XI.M2,
“Water Chemistry,” for BWR water in
BWRVIP-29 (EPRI TR-103515)
Vessel flange
leak detection
line
R-59
Further
Evaluation
Environment
September 2004
(without
cladding)
Top head
Nozzles (vent,
top head spray
or RCIC, and
spare)
Yes,
plant specific
September 2004
IV
A1
Item
R-64
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel (BWR)
Structure
and/or
Component
IV.A1.2- Vessel shell
Attachment
e
welds
Link
Material
Environment
Stainless
steel, nickel
alloy
Reactor coolant
Aging Effect/
Mechanism
Cracking/ stress
corrosion cracking
and intergranular
stress corrosion
cracking
Aging Management Program (AMP)
Chapter XI.M4, “BWR Vessel ID
Attachment Welds,” and
Chapter XI.M2, “Water Chemistry,”
for BWR water in BWRVIP-29 (EPRI
TR-103515)
Further
Evaluation
No
IV.A1-7
Proposed Draft NUREG-1801
Proposed Draft NUREG-1801
IV
A1
Item
R-62
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel (BWR)
Structure
and/or
Component
IV.A1.2- Vessel shell
Intermediate
c
beltline shell
Beltline welds
Link
Material
Environment
Steel (without Reactor coolant
lining/coating and neutron flux
or with
degraded
lining/coating)
Aging Effect/
Mechanism
Loss of fracture
toughness/ neutron
irradiation
embrittlement
Aging Management Program (AMP)
Further
Evaluation
IV.A1-8
September 2004
Neutron irradiation embrittlement is a Yes,
time dependent aging mechanism to TLAA
be evaluated for the period of
extended operation for all ferritic
materials that have a neutron fluence
exceeding 1017 n/cm2 (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 embrittlement
on: (a) the adjusted reference
temperature, the plant’s pressuretemperature 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 50, Appendix G.
Additionally, the applicant is to monitor
axial beltline weld embrittlement. One
acceptable method is to determine that
the mean RTNDT of the axial beltline
welds at the end of the extended
period of operation is less than the
value specified by the staff in its May
7, 2000 letter. See the Standard
Review Plan, Section 4.2 “Reactor
Vessel Neutron Embrittlement” for
acceptable methods for meeting the
requirements of 10 CFR 54.21(c).
September 2004
IV
A1
Item
R-63
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel (BWR)
Structure
and/or
Component
IV.A1.2- Vessel shell
Intermediate
d
beltline shell
Beltline welds
Link
Material
Environment
Steel (without Reactor coolant
lining/coating and neutron flux
or with
degraded
lining/coating)
Aging Effect/
Mechanism
Loss of fracture
toughness/ neutron
irradiation
embrittlement
Aging Management Program (AMP)
Further
Evaluation
Chapter XI.M31, “Reactor Vessel
Surveillance”
Yes, plant
specific
IV.A1-9
Proposed Draft NUREG-1801
A2. REACTOR VESSEL (PRESSURIZED WATER REACTOR)
A2.1
Closure Head
A2.1.1
A2.1.2
A2.1.3
A2.1.4
A2.2
Control Rod Drive (CRD) Head Penetration
A2.2.1
A2.2.2
A2.2.3
A2.3
Inlet
Outlet
Safety Injection (on some)
Nozzle Safe Ends
A2.4.1
A2.4.2
A2.4.3
A2.5
Nozzle
Pressure Housing
Flange Bolting
Nozzles
A2.3.1
A2.3.2
A2.3.3
A2.4
Dome
Head Flange
Stud Assembly
Vessel Flange Leak Detection Line
Inlet
Outlet
Safety Injection (on some)
Shell
A2.5.1
A2.5.2
A2.5.3
A2.5.4
Upper (Nozzle) Shell
Intermediate and Lower Shell
Vessel Flange
Bottom Head
A2.6
Core Support Pads/Core Guide Lugs
A2.7
Penetrations
A2.7.1
A2.7.2
A2.7.3
A2.8
Instrument Tubes (Bottom Head)
Head Vent Pipe (Top Head)
Instrument Tubes (Top Head)
Pressure Vessel Support
A2.8.1
A2.8.2
A2.8.3
September 2004
Skirt Support
Cantilever/Column Support
Neutron Shield Tank
IV.A2-1
Proposed Draft NUREG-1801
A2. REACTOR VESSEL (PRESSURIZED WATER REACTOR)
Systems, Structures, and Components
This section comprises 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.
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 Babcox
and Wilcox designs), the reactor coolant system and connected lines (IV.C2), and the
emergency core cooling system (V.D1).
Proposed Draft NUREG-1801
IV.A2-2
September 2004
September 2004
IV
A2
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel (PWR)
Structure and/or
Component
Item
Link
R-71
IV.A2.1- Closure head
Stud assembly
c
Material
Environment
Aging Effect/
Mechanism
Aging Management Program Further
(AMP)
Evaluation
High strength Air with reactor Cracking/ stress Chapter XI.M3, “Reactor Head No
low alloy steel coolant leakage corrosion cracking Closure Studs”
Maximum
tensile
strength <
1172 MPa
(<170 Ksi)
IV.A2.1- Closure head
Stud assembly
e
R-72
IV.A2.1- Closure head
Stud assembly
d
IV.A2-3
R-73
High strength Air with reactor Cumulative fatigue Fatigue is a time-limited aging Yes
low alloy steel coolant leakage damage/ fatigue analysis (TLAA) to be
TLAA
evaluated for the period of
Maximum
extended operation. See the
tensile
Standard Review Plan, Section
strength <
4.3 “Metal Fatigue,” for
1172 MPa
acceptable methods for
(<170 Ksi)
meeting the requirements of
10 CFR 54.21(c)(1).
High strength Air with reactor Loss of material/ Chapter XI.M3, “Reactor Head No
low alloy steel coolant leakage wear
Closure Studs”
Proposed Draft NUREG-1801
Maximum
tensile
strength <
1172 MPa
(<170 Ksi)
R-74
IV.A2.1- Closure head
Vessel flange leak
f
detection line
Stainless
steel
Air with reactor Cracking/ stress A plant-specific aging
Yes, plant
coolant leakage corrosion cracking management program is to be specific
evaluated because existing
programs may not be capable
of mitigating or detecting crack
initiation and growth due to
SCC in the vessel flange leak
Proposed Draft NUREG-1801
IV
A2
Item
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel (PWR)
Link
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Aging Management Program Further
(AMP)
Evaluation
detection line.
R-78
IV.A2.2- Control rod drive head Stainless
penetration#*#Flange steel
e
Air with reactor Cracking/ stress Chapter XI.M18, “Bolting
coolant leakage corrosion cracking Integrity”
No
IV.A2.2- Control rod drive head Stainless
penetration#*#Flange steel
f
Air with reactor Loss of material/
coolant leakage wear
Chapter XI.M18, “Bolting
Integrity”
No
IV.A2.2- Control rod drive head Stainless
penetration#*#Flange steel
g
Air with reactor Loss of preload/
coolant leakage stress relaxation
Chapter XI.M18, “Bolting
Integrity”
No
IV.A2.2- Control rod drive head Nickel alloy
penetration#*#Nozzle
a
Reactor coolant Cracking/ primary Chapter XI.M1, “ASME Section Yes, plant
water stress
XI Inservice Inspection,
specific
corrosion cracking Subsections IWB, IWC, and
IWD,” for Class 1 components
and Chapter XI.M2, “Water
Chemistry,” for PWR primary
water in EPRI TR-105714 and
plant specific AMP consistent
with applicant commitments to
NRC Order EA-03-009 or any
subsequent regulatory
requirements.
bolting
R-79
IV.A2-4
bolting
R-80
bolting
R-75
September 2004
September 2004
IV
A2
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel (PWR)
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Aging Management Program Further
(AMP)
Evaluation
Item
Link
R-77
IV.A2.2- Control rod drive head Cast
penetration#*#Pressur austenitic
d
Reactor coolant Loss of fracture
>250°C
toughness/
(>482°F)
thermal aging
embrittlement
IV.A2.2- Control rod drive head Stainless
penetration#*#Pressur steel, cast
b
Reactor coolant Cracking/ stress Chapter XI.M1, “ASME Section No
corrosion cracking XI Inservice Inspection,
Subsections IWB, IWC, and
IWD,” for Class 1 components
and
e housing
R-76
e housing
stainless
steel
austenitic
stainless
steel, nickel
alloy
IV.A2-5
Proposed Draft NUREG-1801
R-88
IV.A2.6- Core support pads/core Nickel alloy
guide lugs
a
R-17
IV.A2.8- External surfaces
b
IV.A2.1a
IV.A2.5e
Steel
Chapter XI.M12 “Thermal
Aging Embrittlement of Cast
Austenitic Stainless Steel
(CASS)”
No
Chapter XI.M2, “Water
Chemistry,” for PWR primary
water in EPRI TR-105714
Reactor coolant Cracking/ primary A plant-specific aging
Yes, plant
water stress
management program is to be specific
corrosion cracking evaluated. The applicant is to
provide a plant-specific AMP
or participate in industry
programs to determine
appropriate AMP.
Air with borated Loss of material/ Chapter XI.M10, “Boric Acid
No
water leakage boric acid
Corrosion”
corrosion
Proposed Draft NUREG-1801
IV
A2
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel (PWR)
Structure and/or
Component
Item
Link
R-83
IV.A2.4- Nozzle safe ends
Inlet
b
Outlet
Safety injection
R-81
IV.A2.3- Nozzles
Inlet
a
IV.A2-6
Outlet
Safety injection
Aging Effect/
Mechanism
Material
Environment
Stainless
steel, cast
austenitic
stainless
steel, nickel
alloy and
associated
welds and
buttering
Reactor coolant Cracking/ stress
corrosion
cracking, primary
water stress
corrosion cracking
Steel with
Reactor coolant Loss of fracture
stainless
and neutron flux toughness/
steel cladding
neutron irradiation
embrittlement
Aging Management Program Further
(AMP)
Evaluation
Chapter XI.M1, “ASME Section No
XI Inservice Inspection,
Subsections IWB, IWC, and
IWD,” for Class 1 components
and
Chapter XI.M2, “Water
Chemistry,” for PWR primary
water in EPRI TR-105714
September 2004
Neutron irradiation
Yes,
embrittlement is a time-limited TLAA
aging analysis (TLAA) to be
evaluated for the period of
license renewal for all ferritic
materials that have a neutron
fluence greater than
1017 n/cm2 (E >1 MeV) at the
end of the license renewal
term. The TLAA is to evaluate
the impact of neutron
embrittlement on: (a) the
RTPTS value based on the
requirements in
10 CFR 50.61, (b) the adjusted
reference temperature, the
plant’s pressure-temperature
limits, (c) the Charpy upper
shelf energy, and (d) the
equivalent margins analyses
performed in accordance with
10 CFR 50, Appendix G. The
applicant may choose to
demonstrate that the materials
September 2004
IV
A2
Item
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel (PWR)
Link
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Aging Management Program Further
(AMP)
Evaluation
in the inlet, outlet, and safety
injection nozzles are not
controlling for the TLAA
evaluations.
IV.A2-7
Proposed Draft NUREG-1801
R-82
IV.A2.3- Nozzles
Inlet
b
Outlet
Safety injection
Steel with
Reactor coolant Loss of fracture
Chapter XI.M31, “Reactor
stainless
and neutron flux toughness/
Vessel Surveillance”
steel cladding
neutron irradiation
embrittlement
Yes, plant
specific
Proposed Draft NUREG-1801
IV
A2
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel (PWR)
Structure and/or
Component
Aging Management Program Further
(AMP)
Evaluation
IV.A2-8
Environment
IV.A2.7- Penetrations
Head vent pipe (top
b
Nickel alloy
IV.A2.7- Penetrations
Instrument tubes
a
Nickel alloy
Reactor coolant Cracking/ primary Chapter XI.M1, “ASME Section Yes, plant
XI Inservice Inspection,
water stress
specific
corrosion cracking Subsections IWB, IWC, and
IWD,” for Class 1 components
and Chapter XI.M2, “Water
Chemistry,” for PWR primary
water in EPRI TR-105714 and
plant specific AMP consistent
with applicant commitments to
NRC Order EA-03-009 or any
subsequent regulatory
requirements.
Reactor coolant Cracking/ primary Chapter XI.M1, “ASME Section Yes, plant
water stress
XI Inservice Inspection,
specific
corrosion cracking Subsections IWB, IWC, and
IWD,” for Class 1 components
and Chapter XI.M2, “Water
Chemistry,” for PWR primary
water in EPRI TR-105714 and
plant specific AMP consistent
with applicant commitments to
NRC Bulletin BL-03-02 or any
subsequent regulatory
requirements.
Link
R-90
head)
Instrument tubes (top
head)
R-89
Aging Effect/
Mechanism
Material
Item
(bottom head)
September 2004
September 2004
IV
A2
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel (PWR)
Structure and/or
Component
Item
Link
R-04
IV.A2.2-c Piping, piping
IV.A2.3-c components, and
IV.A2.5- piping elements
d
IV.A2.4a
IV.A2.1b
Material
Environment
Steel,
Reactor coolant Cumulative fatigue
stainless
damage/ fatigue
steel, cast
austenitic
stainless
steel, carbon
steel with
nickel-alloy or
stainless
steel
cladding,
nickel-alloy
IV.A2-9
R-91
IV.A2.8- Pressure vessel
support
a
Skirt support
Aging Effect/
Mechanism
Steel
Air – indoor
uncontrolled
Aging Management Program Further
(AMP)
Evaluation
Fatigue is a time-limited aging Yes,
analysis (TLAA) to be
TLAA
performed for the period of
extended operation, and, for
Class 1 components,
environmental effects on
fatigue are to be addressed.
See the Standard Review
Plan, Section 4.3 “Metal
Fatigue,” for acceptable
methods for meeting the
requirements of 10 CFR
54.21(c)(1)(i) and (ii), and for
addressing environmental
effects on fatigue.
Proposed Draft NUREG-1801
See Chapter X.M1 of this
report for meeting the
requirements of
10 CFR 54.21(c)(1)(iii).
Cumulative fatigue Fatigue is a time-limited aging Yes,
damage/ fatigue analysis (TLAA) to be
TLAA
evaluated for the period of
extended operation. See the
Standard Review Plan, Section
4.3 “Metal Fatigue,” for
acceptable methods for
meeting the requirements of
10 CFR 54.21(c)(1).
Proposed Draft NUREG-1801
IV
A2
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel (PWR)
Structure and/or
Component
Item
Link
R-85
IV.A2.5- Vessel shell
Upper shell
b
Material
Environment
Aging Effect/
Mechanism
SA508-Cl 2 Reactor coolant Crack growth/
forgings clad
cyclic loading
Intermediate and lower with stainless
shell
steel using a
(including beltline
high-heatwelds)
input welding
process
Aging Management Program Further
(AMP)
Evaluation
IV.A2-10
Growth of intergranular
Yes
separations (underclad cracks) TLAA
in low-alloy steel forging heat
affected zone under austenitic
stainless steel cladding is a
time-limited aging analysis
(TLAA) to be evaluated for the
period of extended operation
for all the SA 508-Cl 2 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 TimeLimited Aging Analysis,” for
generic guidance for meeting
the requirements of 10 CFR
54.21(c).
September 2004
September 2004
IV
A2
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel (PWR)
Structure and/or
Component
Item
Link
R-84
IV.A2.5- Vessel shell
Upper shell
a
Material
Environment
Steel with
Reactor coolant
stainless
and neutron flux
Intermediate and lower steel cladding
shell
(including beltline
welds)
IV.A2-11
Proposed Draft NUREG-1801
Aging Effect/
Mechanism
Aging Management Program Further
(AMP)
Evaluation
Loss of fracture
toughness/
neutron irradiation
embrittlement
Neutron irradiation
Yes, plant
embrittlement is a time-limited specific
aging analysis (TLAA) to be
evaluated for the period of
license renewal for all ferritic
materials that have a neutron
fluence of greater than
1017 n/cm2 (E >1 MeV) at the
end of the license renewal
term. The TLAA is to evaluate
the impact of neutron
embrittlement on: (a) the
RTPTS value based on the
requirements in
10 CFR 50.61, (b) the adjusted
reference temperature, the
plant’s pressure temperature
limits, (c) the Charpy upper
shelf energy, and (d) the
equivalent margins analyses
performed in accordance with
10 CFR 50, Appendix G. See
the Standard Review Plan,
Section 4.2 “Reactor Vessel
Neutron Embrittlement” for
acceptable methods for
meeting the requirements of
10 CFR 54.21(c).
Proposed Draft NUREG-1801
IV
A2
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel (PWR)
Structure and/or
Component
Item
Link
R-86
IV.A2.5- Vessel shell
Upper shell
c
R-87
IV.A2.5- Vessel shell
Vessel flange
f
Material
Environment
Steel with
Reactor coolant
stainless
and neutron flux
Intermediate and lower steel cladding
shell
(including beltline
welds)
Steel
Aging Effect/
Mechanism
Aging Management Program Further
(AMP)
Evaluation
Loss of fracture
Chapter XI.M31, “Reactor
toughness/
Vessel Surveillance”
neutron irradiation
embrittlement
Reactor coolant Loss of material/
wear
Yes, plant
specific
Chapter XI.M1, “ASME Section No
XI Inservice Inspection,
Subsections IWB, IWC, and
IWD,” for Class 1 components
IV.A2-12
September 2004
B1.
REACTOR VESSEL INTERNALS (BOILING WATER REACTOR)
B1.1
Core Shroud and Core Plate
B1.1.1
B1.1.2
B1.1.3
B1.1.4
B1.1.5
B1.1.6
Core Shroud (Upper, Central, Lower)
Core Plate
Core Plate Bolts
Access Hole Cover
Shroud Support Structure
LPCI Coupling
B1.2
Top Guide
B1.3
Core Spray Lines and Spargers
B1.3.1
B1.3.2
B1.3.3
B1.3.4
B1.4
Jet Pump Assemblies
B1.4.1
B1.4.2
B1.4.3
B1.4.4
B1.4.5
B1.4.6
B1.4.7
B1.4.8
B1.4.9
B1.5
Thermal Sleeve
Inlet Header
Riser Brace Arm
Holddown Beams
Inlet Elbow
Mixing Assembly
Diffuser
Castings
Jet Pump Sensing Line
Fuel Supports and Control Rod Drive (CRD) Assemblies
B1.5.1
B1.5.2
B1.6
Core Spray Lines (Headers)
Spray Ring
Spray Nozzles
Thermal Sleeve
Orificed Fuel Support
CRD Housing
Instrumentation
B1.6.1
B1.6.2
B1.6.3
B1.6.4
September 2004
Intermediate Range Monitor (IRM) Dry Tubes
Low Power Range Monitor (LPRM) Dry Tubes
Source Range Monitor (SRM) Dry Tubes
Incore Neutron Flux Monitor Guide Tubes
IV B1-1
Proposed Draft NUREG-1801
B1.
REACTOR VESSEL INTERNALS (BOILING WATER REACTOR)
Systems, Structures, and Components
This section comprises the boiling water reactor (BWR) vessel internals and consists of the core
shroud 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.
The steam separator and dryer assemblies are not part of the pressure boundary and are
removed during each outage, and they are covered by the plant maintenance program.
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).
Proposed Draft NUREG-1801
IV B1-2
September 2004
September 2004
IV
B1
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (BWR)
Structure and/or
Material
Component
Item
Link
R-95
IV.B1.1- Core shroud and
core
e
Nickel alloy
plate#*#Access
hole cover
(mechanical
covers)
R-94
IV B1-3
IV.B1.1- Core shroud and
core
d
plate#*#Access
hole cover
(welded covers)
Nickel alloy
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
Reactor coolant Cracking/ stress
corrosion cracking,
intergranular stress
corrosion cracking,
irradiation-assisted
stress corrosion
cracking
Chapter XI.M1, “ASME Section XI No
Inservice Inspection, Subsections
IWB, IWC, and IWD,” for Class 1
components and
Reactor coolant Cracking/ stress
corrosion cracking,
intergranular stress
corrosion cracking,
irradiation-assisted
stress corrosion
cracking
Chapter XI.M1, “ASME Section XI No
Inservice Inspection, Subsections
IWB, IWC, and IWD,” for Class 1
components and Chapter XI.M2,
“Water Chemistry,” for BWR
water in BWRVIP-29 (EPRI TR103515)
Chapter XI.M2, “Water
Chemistry,” for BWR water in
BWRVIP-29 (EPRI
TR-103515)
Proposed Draft NUREG-1801
Because cracking initiated in
crevice 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.
Proposed Draft NUREG-1801
IV
B1
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (BWR)
Structure and/or
Material
Component
Item
Link
R-93
IV.B1.1- Core shroud and Stainless
core plate#*#Core steel
b
plate#*#Core plate
bolts (used in early
BWRs)
R-92
IV.B1.1- Core shroud and Stainless
core plate#*#Core steel
a
IV B1-4
shroud (upper,
central, lower)
R-96
IV.B1.1- Core shroud and
core
f
plate#*#Shroud
support structure
(shroud support
cylinder, shroud
support plate,
shroud support
legs)
Nickel alloy
Aging Management Program
(AMP)
Further
Evaluation
Reactor coolant Cracking/ stress
corrosion cracking,
intergranular stress
corrosion cracking,
irradiation-assisted
stress corrosion
cracking
Chapter XI.M9, “BWR Vessel
Internals,” for core plate and
No
Reactor coolant Cracking/ stress
corrosion cracking,
intergranular stress
corrosion cracking,
irradiation-assisted
stress corrosion
cracking
Chapter XI.M9, “BWR Vessel
Internals,” for core shroud and
Reactor coolant Cracking/ stress
corrosion cracking,
intergranular stress
corrosion cracking,
irradiation-assisted
stress corrosion
cracking
Chapter XI.M9, “BWR Vessel
No
Internals,” for shroud support and
Environment
Aging Effect/
Mechanism
Chapter XI.M2, “Water
Chemistry” for BWR water in
BWRVIP-29 (EPRI
TR-103515)
No
Chapter XI.M2, “Water
Chemistry” for BWR water in
BWRVIP-29 (EPRI
TR-103515)
Chapter XI.M2, “Water
Chemistry,” for BWR water in
BWRVIP-29 (EPRI
TR-103515)
September 2004
September 2004
IV
B1
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (BWR)
Structure and/or
Material
Component
Item
Link
R-97
IV.B1.1- Core shroud and
core
g
Stainless
steel
plate#*#Shroud
support structure
(shroud support
cylinder, shroud
support plate,
shroud support
legs)
IV B1-5
R-99
IV.B1.3- Core spray lines
and
a
Stainless
steel
spargers#*#Core
spray lines
(headers)#*#Spray
rings#*#Spray
nozzles#*#Therma
l sleeves
Proposed Draft NUREG-1801
R-104
IV.B1.5- Fuel supports and Stainless
control rod drive
steel
c
assemblies
Control rod drive
housing
Aging Management Program
(AMP)
Further
Evaluation
Reactor coolant Cracking/ stress
corrosion cracking,
intergranular stress
corrosion cracking,
irradiation-assisted
stress corrosion
cracking
Chapter XI.M9, “BWR Vessel
Internals,” for the LPCI coupling
and
No
Reactor coolant Cracking/ stress
corrosion cracking,
intergranular stress
corrosion cracking,
irradiation-assisted
stress corrosion
cracking
Chapter XI.M9, “BWR Vessel
No
Internals,” for core spray internals
and
Reactor coolant Cracking/ stress
corrosion cracking
and intergranular
stress corrosion
cracking
Chapter XI.M9, “BWR Vessel
Internals,” for lower plenum and
Environment
Aging Effect/
Mechanism
Chapter XI.M2, “Water
Chemistry,” for BWR water in
BWRVIP-29 (EPRI
TR-103515)
Chapter XI.M2, “Water
Chemistry,” for BWR water in
BWRVIP-29 (EPRI
TR-103515)
Chapter XI.M2, “Water
Chemistry,” for BWR water in
BWRVIP-29 (EPRI
TR-103515)
No
Proposed Draft NUREG-1801
IV
B1
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (BWR)
Structure and/or
Material
Component
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
Item
Link
R-103
IV.B1.5- Fuel supports and Cast
control rod drive
austenitic
a
Reactor coolant Loss of fracture
toughness/ thermal
aging and neutron
irradiation
embrittlement
Chapter XI.M13, “Thermal Aging No
and Neutron Irradiation
Embrittlement of Cast Austenitic
Stainless Steel (CASS)”
Stainless
IV.B1.6- Instrumentation
Intermediate range steel
a
Reactor coolant Cracking/ stress
corrosion cracking,
intergranular stress
corrosion cracking,
irradiation-assisted
stress corrosion
cracking
Chapter XI. M9, “BWR Vessel
Internals,” for lower plenum and
assemblies
Orificed fuel
support
R-105
stainless
steel
IV B1-6
monitor (IRM) dry
tubes
Source range
monitor (SRM) dry
tubes
Incore neutron flux
monitor guide
tubes
R-101
Chapter XI.M2, “Water
Chemistry,” for BWR water in
BWRVIP-29 (EPRI
TR-103515)
IV.B1.4- Jet pump
assemblies
c
Cast
austenitic
stainless
steel
Reactor coolant Loss of fracture
toughness/ thermal
aging and neutron
irradiation
embrittlement
Chapter XI.M13, “Thermal Aging No
and Neutron Irradiation
Embrittlement of Cast Austenitic
Stainless Steel (CASS)”
IV.B1.4- Jet pump
assemblies
d
Stainless
steel
Reactor coolant Cracking/ cyclic
loading
A plant-specific aging
management program is to be
evaluated.
Castings
September 2004
R-102
No
Jet pump sensing
line
Yes, plant
specific
September 2004
IV
B1
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (BWR)
Structure and/or
Material
Component
Item
Link
R-100
IV.B1.4- Jet pump
assemblies
a
Thermal sleeve
Inlet header
Riser brace arm
Holddown beams
Inlet elbow
Mixing assembly
Diffuser
Castings
Environment
Nickel alloy, Reactor coolant
cast
austenitic
stainless
steel,
stainless
steel
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
Cracking/ stress
corrosion cracking,
intergranular stress
corrosion cracking,
irradiation-assisted
stress corrosion
cracking
Chapter XI.M9, “BWR Vessel
No
Internals,” for jet pump assembly
and
Chapter XI.M2, “Water
Chemistry,” for BWR water in
BWRVIP-29 (EPRI
TR-103515)
IV B1-7
R-53
IV.B1.3-b Reactor vessel
IV.B1.2-b internals
IV.B1.5-b components
IV.B1.4-b
IV.B1.1-c
IV.B1.6-b
Stainless
Reactor coolant Cumulative fatigue
steel, cast
damage/ fatigue
austenitic
stainless
steel, nickel
alloy
Proposed Draft NUREG-1801
For components for which a
Yes,
fatigue analysis has been
TLAA
performed for the
40-year period, fatigue is a timelimited aging analysis (TLAA) to
be evaluated for the period of
extended operation. See the
Standard Review Plan, Section
4.3 “Metal Fatigue,” for
acceptable methods for meeting
the requirements of 10 CFR
54.21(c)(1).
Proposed Draft NUREG-1801
IV
B1
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (BWR)
Structure and/or
Material
Component
Item
Link
R-98
IV.B1.2- Top guide
a
Stainless
steel
Environment
Aging Effect/
Mechanism
Reactor coolant Cracking/ stress
corrosion cracking,
intergranular stress
corrosion cracking,
irradiation-assisted
stress corrosion
cracking
Aging Management Program
(AMP)
Further
Evaluation
IV B1-8
Chapter XI.M9, “BWR Vessel
No
Internals,” for top guide and
Chapter XI.M2, “Water
Chemistry,” for BWR water in
BWRVIP-29 (EPRI TR-103515)
For top guides with neutron
fluence exceeding the IASCC
threshold (5x1020, E>lMeV)
inspect ten (10) percent of the top
guide locations using enhanced
visual inspection technique, EVT1 within 12 years, one-half (5
percent) to be completed within 6
years. Locations selected for
examination will be areas that
have exceeded the neutron
fluence threshold. 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.
September 2004
B2. REACTOR VESSEL INTERNALS (PWR) - WESTINGHOUSE
B2.1
Upper Internals Assembly
B2.1.1
B2.1.2
B2.1.3
B2.1.4
B2.1.5
B2.1.6
B2.1.7
B2.2
RCCA Guide Tube Assemblies
B2.2.1
B2.2.2
B2.2.3
B2.3
Baffle and Former Plates
Baffle/Former Bolts
Lower Internal Assembly
B2.5.1
B2.5.2
B2.5.3
B2.5.4
B2.5.5
B2.5.6
B2.5.7
B2.6
Core Barrel
Core Barrel Flange
Core Barrel Outlet Nozzles
Thermal Shield
Baffle/Former Assembly
B2.4.1
B2.4.2
B2.5
RCCA Guide Tubes
RCCA Guide Tube Bolts
RCCA Guide Tube Support Pins
Core Barrel
B2.3.1
B2.3.2
B2.3.3
B2.3.4
B2.4
Upper Support Plate
Upper Support Column
Upper Support Column Bolts
Upper Core Plate
Upper Core Plate Alignment Pins
Fuel Alignment Pins
Hold-Down Spring
Lower Core Plate
Fuel Alignment Pins
Lower Support Forging or Casting
Lower Support Plate Columns
Lower Support Plate Column Bolts
Radial Support Keys and Clevis Inserts
Clevis Insert Bolts
Instrumentation Support Structures
B2.6.1
B2.6.2
September 2004
Flux Thimble Guide Tubes
Flux Thimbles
IV B2-1
Proposed Draft NUREG-1801
B2. REACTOR VESSEL INTERNALS (PWR) - WESTINGHOUSE
Systems, Structures, and Components
This section comprises 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.
System Interfaces
The systems that interface with the reactor vessel internals include the reactor pressure vessel
(IV.A2).
Proposed Draft NUREG-1801
IV B2-2
September 2004
September 2004
IV
B2
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Westinghouse
Structure and/or
Component
Environment
IV.B2.4- Baffle/former
assembly#*#Baffle and
b
Stainless
steel
Reactor coolant Changes in
dimensions/Void
swelling
IV.B2.4- Baffle/former
assembly#*#Baffle and
a
Stainless
steel
Link
R-124
former plates
IV B2-3
R-123
Aging Effect/
Mechanism
Material
Item
former plates
Aging Management Program
(AMP)
Further
Evaluation
Proposed Draft NUREG-1801
Applicant must provide a
No, but
commitment which includes the
licensee
following elements: (1) to
commitment to
participate in industry programs for be confirmed.
investigating and managing aging
effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
Reactor coolant Cracking/ stress
Chapter XI.M2, “Water Chemistry,” No, but
corrosion cracking, for PWR primary water in EPRI
licensee
irradiation-assisted TR-105714 and the applicant must commitment to
stress corrosion
provide a commitment which
be confirmed.
cracking
includes the following elements:
(1) to participate in industry
programs for investigating and
managing aging effects applicable
to Reactor Internals, (2) to
evaluate and implement the results
of the industry programs as
applicable to the Reactor Internals
design and, (3) to submit, for NRC
review and approval an inspection
plan for Reactor Internals, as
based on industry
recommendation, at least 24
months prior to the extended
Proposed Draft NUREG-1801
IV
B2
Item
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Westinghouse
Link
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
period.
IV B2-4
R-127
IV.B2.4- Baffle/former
assembly#*#Baffle and
e
former plates
Stainless
steel
Reactor coolant Loss of fracture
and neutron flux toughness/ neutron
irradiation
embrittlement, void
swelling
September 2004
Applicant must provide a
commitment which includes the
following elements: (1) to
participate in industry programs for
investigating and managing aging
effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
No, but
licensee
commitment to
be confirmed.
September 2004
IV
B2
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Westinghouse
Structure and/or
Component
Material
Item
Link
R-126
Stainless
IV.B2.4- Baffle/former
assembly#*#Baffle/forme steel
d
r bolts
IV B2-5
R-125
Stainless
IV.B2.4- Baffle/former
assembly#*#Baffle/forme steel
c,
r bolts
Environment
Aging Effect/
Mechanism
Reactor coolant Changes in
dimensions/Void
swelling
Aging Management Program
(AMP)
Proposed Draft NUREG-1801
Applicant must provide a
commitment which includes the
following elements: (1) to
participate in industry programs for
investigating and managing aging
effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
Reactor coolant Cracking/ stress
Applicant must provide a
and high fluence corrosion cracking, commitment which includes the
(>1 x 10E21
irradiation-assisted following elements: (1) to
n/cm2 E
stress corrosion
participate in industry programs for
>0.1 MeV)
cracking
investigating and managing aging
effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
Further
Evaluation
No, but
licensee
commitment to
be confirmed.
No, but
licensee
commitment to
be confirmed.
Proposed Draft NUREG-1801
IV
B2
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Westinghouse
Structure and/or
Component
Material
Item
Link
R-128
Stainless
IV.B2.4- Baffle/former
assembly#*#Baffle/forme steel
f
r bolts
R-129
Aging Effect/
Mechanism
alloy
IV B2-6
Stainless
IV.B2.3- Core barrel#*#Core
barrel (CB)#*#CB flange steel
b
(upper)#*#CB outlet
nozzles#*#Thermal
shield
Aging Management Program
(AMP)
Reactor coolant Loss of fracture
A plant-specific aging
and neutron flux toughness/ neutron management program is to be
irradiation
evaluated.
embrittlement
Stainless
Reactor coolant Loss of preload/
IV.B2.4- Baffle/former
assembly#*#Baffle/forme steel, nickel
stress relaxation
h
r bolts
R-121
Environment
Reactor coolant Changes in
dimensions/Void
swelling
Further
Evaluation
Yes, plant
specific
September 2004
A plant-specific aging
Yes, plant
management program is to be
specific
evaluated. Visual inspection (VT-3)
is to be augmented to detect
relevant conditions of stress
relaxation because only the heads
of the baffle/former bolts are
visible, and a plant-specific aging
management program is thus
required.
Applicant must provide a
No, but
commitment which includes the
licensee
following elements: (1) to
commitment to
participate in industry programs for be confirmed.
investigating and managing aging
effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
September 2004
IV
B2
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Westinghouse
Structure and/or
Component
Material
Item
Link
R-120
Stainless
IV.B2.3- Core barrel#*#Core
barrel (CB)#*#CB flange steel
a
(upper)#*#CB outlet
nozzles#*#Thermal
shield
IV B2-7
R-122
Stainless
IV.B2.3- Core barrel#*#Core
barrel
(CB)#*#CB
flange
steel
c
Proposed Draft NUREG-1801
(upper)#*#CB outlet
nozzles#*#Thermal
shield
Environment
Aging Effect/
Mechanism
Reactor coolant Cracking/ stress
corrosion cracking,
irradiation-assisted
stress corrosion
cracking
Aging Management Program
(AMP)
Further
Evaluation
Chapter XI.M2, “Water Chemistry,” No, but
for PWR primary water in EPRI
licensee
TR-105714 and the applicant must commitment to
provide a commitment which
be confirmed.
includes the following elements:
(1) to participate in industry
programs for investigating and
managing aging effects applicable
to Reactor Internals, (2) to
evaluate and implement the results
of the industry programs as
applicable to the Reactor Internals
design and, (3) to submit, for NRC
review and approval an inspection
plan for Reactor Internals, as
based on industry
recommendation, at least 24
months prior to the extended
period.
Reactor coolant Loss of fracture
Applicant must provide a
No, but
>250°C (>482°F) toughness/ neutron commitment which includes the
licensee
and neutron flux irradiation
following elements: (1) to
commitment to
embrittlement, void participate in industry programs for be confirmed.
swelling
investigating and managing aging
effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
Proposed Draft NUREG-1801
IV
B2
Item
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Westinghouse
Link
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
period.
Further
Evaluation
IV B2-8
September 2004
September 2004
IV
B2
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Westinghouse
Structure and/or
Component
Item
Link
R-145
IV.B2.6- Instrumentation support
structures
c
Flux thimble
Aging Effect/
Mechanism
Material
Environment
Stainless
steel
Reactor coolant Loss of material/
wear
Aging Management Program
(AMP)
Further
Evaluation
Chapter XI.M1, “ASME Section XI No
Inservice Inspection, Subsections
IWB, IWC, and IWD,” for Class 1
components and
recommendations of NRC I&E
Bulletin 88-09 “Thimble Tube
Thinning in Westinghouse
Reactors,” described bellow:
IV B2-9
Proposed Draft NUREG-1801
In response to I&E Bulletin 88-09,
an inspection program, with
technical justification, is to be
established and is to include (a) an
appropriate thimble tube wear
acceptance criterion, e.g., percent
through-wall loss, and includes
allowances for inspection
methodology and wear scar
geometry uncertainty, (b) an
appropriate inspection frequency,
e.g., every refueling outage, and
(c) inspection methodology such
as eddy current technique that is
capable of adequately detecting
wear of the thimble tubes. In
addition, corrective actions include
isolation or replacement if a
thimble tube fails to meet the
above acceptance criteria.
Inspection schedule is in
accordance with the guidelines of
I&E Bulletin 88-09.
Proposed Draft NUREG-1801
IV
B2
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Westinghouse
Structure and/or
Component
Environment
IV.B2.6- Instrumentation support
structures
b
Stainless
steel
Reactor coolant Changes in
dimensions/Void
swelling
IV.B2.6- Instrumentation support
structures
a
Stainless
steel
Link
R-144
Flux thimble guide tubes
IV B2-10
R-143
Aging Effect/
Mechanism
Material
Item
Flux thimble guide tubes
Aging Management Program
(AMP)
Further
Evaluation
September 2004
Applicant must provide a
No, but
commitment which includes the
licensee
following elements: (1) to
commitment to
participate in industry programs for be confirmed.
investigating and managing aging
effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
Reactor coolant Cracking/ stress
Chapter XI.M2, “Water Chemistry,” No, but
corrosion cracking, for PWR primary water in EPRI
licensee
irradiation-assisted TR-105714 and the applicant must commitment to
stress corrosion
provide a commitment which
be confirmed.
cracking
includes the following elements:
(1) to participate in industry
programs for investigating and
managing aging effects applicable
to Reactor Internals, (2) to
evaluate and implement the results
of the industry programs as
applicable to the Reactor Internals
design and, (3) to submit, for NRC
review and approval an inspection
plan for Reactor Internals, as
based on industry
recommendation, at least 24
months prior to the extended
September 2004
IV
B2
Item
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Westinghouse
Link
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
period.
IV B2-11
R-137
IV.B2.5-i Lower internal assembly Stainless
Clevis insert bolts
steel, nickel
alloy
Reactor coolant Loss of preload/
stress relaxation
Proposed Draft NUREG-1801
Applicant must provide a
commitment which includes the
following elements: (1) to
participate in industry programs for
investigating and managing aging
effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
No, but
licensee
commitment to
be confirmed.
Proposed Draft NUREG-1801
IV
B2
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Westinghouse
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Item
Link
R-134
Reactor coolant Changes in
IV.B2.5- Lower internal assembly Stainless
Fuel alignment pins
steel, nickel
dimensions/Void
f
IV B2-12
R-133
Aging Management Program
(AMP)
Further
Evaluation
September 2004
Applicant must provide a
No, but
commitment which includes the
licensee
Lower support plate
alloy
swelling
following elements: (1) to
commitment to
column bolts
participate in industry programs for be confirmed.
Clevis insert bolts
investigating and managing aging
effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
Reactor coolant Cracking/ stress
Chapter XI.M2, “Water Chemistry,” No, but
IV.B2.5- Lower internal assembly Stainless
Fuel alignment pins
steel, nickel
corrosion cracking, for PWR primary water in EPRI
licensee
e
Lower support plate
alloy
primary water stress TR-105714 and the applicant must commitment to
column bolts
corrosion cracking, provide a commitment which
be confirmed.
Clevis insert bolts
irradiation-assisted includes the following elements:
stress corrosion
(1) to participate in industry
cracking
programs for investigating and
managing aging effects applicable
to Reactor Internals, (2) to
evaluate and implement the results
of the industry programs as
applicable to the Reactor Internals
design and, (3) to submit, for NRC
review and approval an inspection
plan for Reactor Internals, as
based on industry
recommendation, at least 24
months prior to the extended
September 2004
IV
B2
Item
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Westinghouse
Link
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
period.
IV B2-13
R-135
Reactor coolant Loss of fracture
Applicant must provide a
IV.B2.5- Lower internal assembly Stainless
Fuel alignment pins
steel, nickel and neutron flux toughness/ neutron commitment which includes the
g
Lower support plate
column bolts
Clevis insert bolts
alloy
Proposed Draft NUREG-1801
No, but
licensee
irradiation
following elements: (1) to
commitment to
embrittlement, void participate in industry programs for be confirmed.
swelling
investigating and managing aging
effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
Proposed Draft NUREG-1801
IV
B2
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Westinghouse
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Link
R-132
IV.B2.5- Lower internal assembly Stainless
Lower core plate
steel
c
Reactor coolant Loss of fracture
and neutron flux toughness/ neutron
irradiation
embrittlement, void
swelling
R-131
IV.B2.5- Lower internal assembly Stainless
Lower core plate
steel
b
Reactor coolant Changes in
dimensions/Void
swelling
IV B2-14
Item
Radial keys and clevis
inserts
September 2004
Aging Management Program
(AMP)
Further
Evaluation
Applicant must provide a
commitment which includes the
following elements: (1) to
participate in industry programs for
investigating and managing aging
effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
Applicant must provide a
commitment which includes the
following elements: (1) to
participate in industry programs for
investigating and managing aging
effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
No, but
licensee
commitment to
be confirmed.
No, but
licensee
commitment to
be confirmed.
September 2004
IV
B2
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Westinghouse
Structure and/or
Component
Material
Item
Link
R-130
IV.B2.5- Lower internal assembly Stainless
Lower core plate
steel
a
Radial keys and clevis
inserts
IV B2-15
R-140
IV.B2.5- Lower internal assembly Cast
Lower support casting
austenitic
m
Lower support plate
columns
stainless
steel
Environment
Aging Effect/
Mechanism
Reactor coolant Cracking/ stress
corrosion cracking,
irradiation-assisted
stress corrosion
cracking
Aging Management Program
(AMP)
Further
Evaluation
Proposed Draft NUREG-1801
Chapter XI.M2, “Water Chemistry,” No, but
for PWR primary water in EPRI
licensee
TR-105714 and the applicant must commitment to
provide a commitment which
be confirmed.
includes the following elements:
(1) to participate in industry
programs for investigating and
managing aging effects applicable
to Reactor Internals, (2) to
evaluate and implement the results
of the industry programs as
applicable to the Reactor Internals
design and, (3) to submit, for NRC
review and approval an inspection
plan for Reactor Internals, as
based on industry
recommendation, at least 24
months prior to the extended
period.
Reactor coolant Loss of fracture
Chapter XI.M13, “Thermal Aging No
>250°C (>482°F) toughness/ thermal and Neutron Irradiation
and neutron flux aging and neutron Embrittlement of Cast Austenitic
irradiation
Stainless Steel (CASS)”
embrittlement, void
swelling
Proposed Draft NUREG-1801
IV
B2
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Westinghouse
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Item
Link
R-141
IV.B2.5- Lower internal assembly Stainless
Lower support forging
steel
n
Reactor coolant Loss of fracture
and neutron flux toughness/ neutron
irradiation
embrittlement, void
swelling
IV.B2.5-l Lower internal assembly Stainless
Reactor coolant Changes in
dimensions/Void
swelling
Lower support plate
columns
IV B2-16
R-139
Lower support forging or
casting
Lower support plate
columns
steel, cast
austenitic
stainless
steel
September 2004
Aging Management Program
(AMP)
Further
Evaluation
Applicant must provide a
commitment which includes the
following elements: (1) to
participate in industry programs for
investigating and managing aging
effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
Applicant must provide a
commitment which includes the
following elements: (1) to
participate in industry programs for
investigating and managing aging
effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
No, but
licensee
commitment to
be confirmed.
No, but
licensee
commitment to
be confirmed.
September 2004
IV
B2
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Westinghouse
Structure and/or
Component
Material
Item
Link
R-138
IV.B2.5- Lower internal assembly Stainless
Lower support forging or steel, cast
k
casting
Lower support plate
columns
austenitic
stainless
steel
Environment
Aging Effect/
Mechanism
Reactor coolant Cracking/ stress
corrosion cracking,
irradiation-assisted
stress corrosion
cracking
IV B2-17
R-136
Reactor coolant Loss of preload/
IV.B2.5- Lower internal assembly Stainless
Lower
support
plate
steel,
nickel
stress relaxation
h
column bolts
alloy
Aging Management Program
(AMP)
Further
Evaluation
Proposed Draft NUREG-1801
Chapter XI.M2, “Water Chemistry,” No, but
for PWR primary water in EPRI
licensee
TR-105714 and the applicant must commitment to
provide a commitment which
be confirmed.
includes the following elements:
(1) to participate in industry
programs for investigating and
managing aging effects applicable
to Reactor Internals, (2) to
evaluate and implement the results
of the industry programs as
applicable to the Reactor Internals
design and, (3) to submit, for NRC
review and approval an inspection
plan for Reactor Internals, as
based on industry
recommendation, at least 24
months prior to the extended
period.
Applicant must provide a
No, but
commitment which includes the
licensee
following elements: (1) to
commitment to
participate in industry programs for be confirmed.
investigating and managing aging
effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
Proposed Draft NUREG-1801
IV
B2
Item
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Westinghouse
Link
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
period.
IV B2-18
R-142
IV.B2.5- Lower internal assembly Stainless
Radial keys and clevis
steel
o
Inserts
Reactor coolant Loss of material/
wear
Chapter XI.M1, “ASME Section XI No
Inservice Inspection, Subsections
IWB, IWC, and IWD,” for Class 1
components
September 2004
September 2004
IV
B2
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Westinghouse
Structure and/or
Component
Item
Link
R-118
IV.B2.2- RCCA guide tube
assemblies
d
IV B2-19
R-119
Material
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
Proposed Draft NUREG-1801
Stainless
Reactor coolant Cracking/ stress
Chapter XI.M2, “Water Chemistry,” No, but
steel, nickel
corrosion cracking, for PWR primary water in EPRI
licensee
RCCA guide tube bolts alloy
primary water stress TR-105714 and the applicant must commitment to
RCCA guide tube support
corrosion cracking, provide a commitment which
be confirmed.
pins
irradiation-assisted includes the following elements:
stress corrosion
(1) to participate in industry
cracking
programs for investigating and
managing aging effects applicable
to Reactor Internals, (2) to
evaluate and implement the results
of the industry programs as
applicable to the Reactor Internals
design and, (3) to submit, for NRC
review and approval an inspection
plan for Reactor Internals, as
based on industry
recommendation, at least 24
months prior to the extended
period.
RCCA
guide
tube
Stainless
Reactor
coolant
Changes
in
Applicant must provide a
No, but
IV.B2.2assemblies
steel,
nickel
dimensions/Void
commitment
which
includes
the
licensee
e
RCCA guide tube bolts, alloy
swelling
following elements: (1) to
commitment to
RCCA guide tube support
participate in industry programs for be confirmed.
pins
investigating and managing aging
effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
Proposed Draft NUREG-1801
IV
B2
Item
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Westinghouse
Link
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
period.
IV B2-20
R-117
IV.B2.2- RCCA guide tube
assemblies
b
RCCA guide tubes
Stainless
steel
Reactor coolant Changes in
dimensions/Void
swelling
September 2004
Applicant must provide a
commitment which includes the
following elements: (1) to
participate in industry programs for
investigating and managing aging
effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
No, but
licensee
commitment to
be confirmed.
September 2004
IV
B2
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Westinghouse
Aging Effect/
Mechanism
Item
Link
Structure and/or
Component
Material
Environment
R-116
IV.B2.2- RCCA guide tube
assemblies
a
Stainless
steel
Reactor coolant Cracking/ stress
corrosion cracking,
irradiation-assisted
stress corrosion
cracking
IV B2-21
R-53
Aging Management Program
(AMP)
Further
Evaluation
Proposed Draft NUREG-1801
Chapter XI.M2, “Water Chemistry,” No, but
for PWR primary water in EPRI
licensee
RCCA guide tubes
TR-105714 and the applicant must commitment to
provide a commitment which
be confirmed.
includes the following elements:
(1) to participate in industry
programs for investigating and
managing aging effects applicable
to Reactor Internals, (2) to
evaluate and implement the results
of the industry programs as
applicable to the Reactor Internals
design and, (3) to submit, for NRC
review and approval an inspection
plan for Reactor Internals, as
based on industry
recommendation, at least 24
months prior to the extended
period.
IV.B2.1-c Reactor vessel internals Stainless
Reactor coolant Cumulative fatigue For components for which a
Yes,
IV.B2.1- components
steel, cast
damage/ fatigue
fatigue analysis has been
TLAA
m
austenitic
performed for the
IV.B2.5-p
stainless
40-year period, fatigue is a timeIV.B2.3-d
steel, nickel
limited aging analysis (TLAA) to be
IV.B2.2-c
alloy
evaluated for the period of
IV.B2.5-d
extended operation. See the
IV.B2.2-f
Standard Review Plan, Section 4.3
IV.B2.1-h
“Metal Fatigue,” for acceptable
IV.B2.5-j
methods for meeting the
IV.B2.4-g
requirements of 10 CFR
54.21(c)(1).
Proposed Draft NUREG-1801
IV
B2
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Westinghouse
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Link
R-108
IV.B2.1- Upper internals assembly Stainless
Hold-down spring
steel
d
Reactor coolant Loss of preload/
stress relaxation
R-115
IV.B2.1-l Upper internals assembly Stainless
Reactor coolant Loss of material/
wear
IV B2-22
Item
Upper core plate
alignment pins
steel, nickel
alloy
Aging Management Program
(AMP)
Further
Evaluation
Applicant must provide a
commitment which includes the
following elements: (1) to
participate in industry programs for
investigating and managing aging
effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
Chapter XI.M1, “ASME Section XI
Inservice Inspection, Subsections
IWB, IWC, and IWD,” for Class 1
components
No, but
licensee
commitment to
be confirmed.
No
September 2004
September 2004
IV
B2
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Westinghouse
Structure and/or
Component
Material
Item
Link
R-110
IV.B2.1- Upper internals assembly Stainless
Upper support column
steel, cast
f
austenitic
stainless
steel
IV B2-23
R-109
IV.B2.1- Upper internals assembly Stainless
Upper support column
steel, cast
e
austenitic
stainless
steel
Environment
Aging Effect/
Mechanism
Reactor coolant Changes in
dimensions/Void
swelling
Aging Management Program
(AMP)
Further
Evaluation
Proposed Draft NUREG-1801
Applicant must provide a
No, but
commitment which includes the
licensee
following elements: (1) to
commitment to
participate in industry programs for be confirmed.
investigating and managing aging
effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
Reactor coolant Cracking/ stress
Chapter XI.M2, “Water Chemistry,” No, but
corrosion cracking, for PWR primary water in EPRI
licensee
irradiation-assisted TR-105714 and the applicant must commitment to
stress corrosion
provide a commitment which
be confirmed.
cracking
includes the following elements:
(1) to participate in industry
programs for investigating and
managing aging effects applicable
to Reactor Internals, (2) to
evaluate and implement the results
of the industry programs as
applicable to the Reactor Internals
design and, (3) to submit, for NRC
review and approval an inspection
plan for Reactor Internals, as
based on industry
recommendation, at least 24
months prior to the extended
Proposed Draft NUREG-1801
IV
B2
Item
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Westinghouse
Link
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
period.
IV B2-24
R-111
IV.B2.1- Upper internals assembly Cast
Upper support column
austenitic
g
(only cast austenitic
stainless steel portions)
stainless
steel
Reactor coolant Loss of fracture
>250°C (>482°F) toughness/ thermal
and neutron flux aging and neutron
irradiation
embrittlement, void
swelling
Chapter XI.M13, “Thermal Aging
and Neutron Irradiation
Embrittlement of Cast Austenitic
Stainless Steel (CASS)”
No
September 2004
September 2004
IV
B2
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Westinghouse
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Item
Link
R-114
Reactor coolant Loss of preload/
IV.B2.1- Upper internals assembly Stainless
Upper support column
steel, nickel
stress relaxation
k
bolts
alloy
IV B2-25
R-113
IV.B2.1-j Upper internals assembly Stainless
Upper support column
bolts Upper core plate
alignment pins
Fuel alignment pins
steel, nickel
alloy
Reactor coolant Changes in
dimensions/Void
swelling
Proposed Draft NUREG-1801
Aging Management Program
(AMP)
Further
Evaluation
Applicant must provide a
commitment which includes the
following elements: (1) to
participate in industry programs for
investigating and managing aging
effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
Applicant must provide a
commitment which includes the
following elements: (1) to
participate in industry programs for
investigating and managing aging
effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
No, but
licensee
commitment to
be confirmed.
No, but
licensee
commitment to
be confirmed.
Proposed Draft NUREG-1801
IV
B2
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Westinghouse
Structure and/or
Component
Material
Item
Link
R-112
IV.B2.1-i Upper internals assembly Stainless
Upper support column
bolts Upper core plate
alignment pins
Fuel alignment pins
steel, nickel
alloy
IV B2-26
R-107
IV.B2.1- Upper internals assembly Stainless
Upper support plate
steel
b
Upper core plate
Hold-down spring
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
September 2004
Reactor coolant Cracking/ stress
Chapter XI.M2, “Water Chemistry,” No, but
corrosion cracking, for PWR primary water in EPRI
licensee
primary water stress TR-105714 and the applicant must commitment to
corrosion cracking, provide a commitment which
be confirmed.
irradiation-assisted includes the following elements:
stress corrosion
(1) to participate in industry
cracking
programs for investigating and
managing aging effects applicable
to Reactor Internals, (2) to
evaluate and implement the results
of the industry programs as
applicable to the Reactor Internals
design and, (3) to submit, for NRC
review and approval an inspection
plan for Reactor Internals, as
based on industry
recommendation, at least 24
months prior to the extended
period.
Reactor coolant Changes in
Applicant must provide a
No, but
dimensions/Void
commitment which includes the
licensee
swelling
following elements: (1) to
commitment to
participate in industry programs for be confirmed.
investigating and managing aging
effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
September 2004
IV
B2
Item
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Westinghouse
Link
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
period.
IV B2-27
R-106
IV.B2.1- Upper internals assembly Stainless
Upper support plate
steel
a
Upper core plate
Hold-down spring
Reactor coolant Cracking/ stress
corrosion cracking,
irradiation-assisted
stress corrosion
cracking
Proposed Draft NUREG-1801
Chapter XI.M2, “Water Chemistry,” No, but
for PWR primary water in EPRI
licensee
TR-105714 and the applicant must commitment to
provide a commitment which
be confirmed.
includes the following elements:
(1) to participate in industry
programs for investigating and
managing aging effects applicable
to Reactor Internals, (2) to
evaluate and implement the results
of the industry programs as
applicable to the Reactor Internals
design and, (3) to submit, for NRC
review and approval an inspection
plan for Reactor Internals, as
based on industry
recommendation, at least 24
months prior to the extended
period.
B3. REACTOR VESSEL INTERNALS (PWR) - COMBUSTION ENGINEERING
B3.1
Upper Internals Assembly
B3.1.1
B3.1.2
B3.1.3
B3.1.4
B3.2
Control Element Assembly (CEA) Shroud Assemblies
B3.2.1
B3.2.2
B3.2.3
B3.3
Core Support Barrel
Core Support Barrel Upper Flange
Core Support Barrel Alignment Keys
Core Shroud Assembly
B3.4.1
B3.4.2
B3.4.3
B3.5
CEA Shrouds
CEA Shrouds Bolts
CEA Shrouds Extension Shaft Guides
Core Support Barrel
B3.3.1
B3.3.2
B3.3.3
B3.4
Upper Guide Structure Support Plate
Fuel Alignment Plate
Fuel Alignment Plate Guide Lugs and Guide Lug Inserts
Hold-Down Ring
Core Shroud Assembly
Core Shroud Assembly Bolts
Core Shroud Tie Rods
Lower Internal Assembly
B3.5.1
B3.5.2
B3.5.3
B3.5.4
B3.5.5
B3.5.6
September 2004
Core Support Plate
Fuel Alignment Pins
Lower Support Structure Beam Assemblies
Core Support Column
Core Support Column Bolts
Core Support Barrel Snubber Assemblies
IV B3-1
Proposed Draft NUREG-1801
B3. REACTOR VESSEL INTERNALS (PWR) - COMBUSTION ENGINEERING
Systems, Structures, and Components
This section comprises the Combustion Engineering pressurized water reactor (PWR) vessel
internals and consists of the upper internals assembly, the 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.
System Interfaces
The systems that interface with the reactor vessel internals include the reactor pressure vessel
(IV.A2).
Proposed Draft NUREG-1801
IV B3-2
September 2004
September 2004
IV
B3
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Combustion Engineering
Structure and/or
Component
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Loss of fracture
toughness/ thermal
aging and neutron
irradiation
embrittlement, void
swelling
Chapter XI.M13, “Thermal Aging No
and Neutron Irradiation
Embrittlement of Cast Austenitic
Stainless Steel (CASS)”
Material
Environment
IV.B3.2- CEA Shroud
Assemblies
e
Cast
austenitic
stainless
steel
Reactor coolant
>250°C
(>482°F) and
neutron flux
R-149
IV.B3.2- CEA Shroud
Assemblies
a
Stainless
steel, cast
austenitic
stainless
steel
Reactor coolant Cracking/ stress
corrosion cracking,
irradiation-assisted
stress corrosion
cracking
R-152
Stainless
IV.B3.2- CEA shroud
assemblies#*#CEA
steel
d
Item
Link
R-153
IV B3-3
Proposed Draft NUREG-1801
shroud extension
shaft guides
Reactor coolant Loss of material/
wear
Chapter XI.M2, “Water
Chemistry,” for PWR primary
water in EPRI TR-105714 and
the applicant must provide a
commitment which includes the
following elements: (1) to
participate in industry programs
for investigating and managing
aging effects applicable to
Reactor Internals, (2) to evaluate
and implement the results of the
industry programs as applicable
to the Reactor Internals design
and, (3) to submit, for NRC
review and approval an
inspection plan for Reactor
Internals, as based on industry
recommendation, at least 24
months prior to the extended
period.
Chapter XI.M1, “ASME Section
XI Inservice Inspection,
Subsections IWB, IWC, and
IWD,” for Class 1 components
Further
Evaluation
No, but
licensee
commitment to
be confirmed.
No
Proposed Draft NUREG-1801
IV
B3
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Combustion Engineering
Structure and/or
Component
Material
Item
Link
R-151
Stainless
IV.B3.2- CEA Shroud
Assemblies#*#CE steel, cast
c
IV B3-4
R-150
Environment
Aging Effect/
Mechanism
Reactor coolant Changes in
dimensions/Void
swelling
Aging Management Program
(AMP)
September 2004
Applicant must provide a
commitment which includes the
A shrouds bolts
austenitic
following elements: (1) to
stainless
participate in industry programs
steel, nickel
for investigating and managing
alloy
aging effects applicable to
Reactor Internals, (2) to evaluate
and implement the results of the
industry programs as applicable
to the Reactor Internals design
and, (3) to submit, for NRC
review and approval an
inspection plan for Reactor
Internals, as based on industry
recommendation, at least 24
months prior to the extended
period.
Stainless
Reactor coolant Cracking/ stress
Chapter XI.M2, “Water
IV.B3.2- CEA Shroud
Assemblies#*#CE steel, nickel
corrosion cracking, Chemistry,” for PWR primary
b
A shrouds bolts
alloy
primary water
water in EPRI TR-105714 and
stress corrosion
the applicant must provide a
cracking,
commitment which includes the
irradiation-assisted following elements: (1) to
stress corrosion
participate in industry programs
cracking
for investigating and managing
aging effects applicable to
Reactor Internals, (2) to evaluate
and implement the results of the
industry programs as applicable
to the Reactor Internals design
and, (3) to submit, for NRC
review and approval an
inspection plan for Reactor
Internals, as based on industry
Further
Evaluation
No, but
licensee
commitment to
be confirmed.
No, but
licensee
commitment to
be confirmed.
September 2004
IV
B3
Item
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Combustion Engineering
Link
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
recommendation, at least 24
months prior to the extended
period.
IV B3-5
R-154
Stainless
Reactor coolant Loss of preload/
IV.B3.2- CEA Shroud
Assemblies#*#CE steel, nickel
stress relaxation
g
A shrouds bolts
alloy
Proposed Draft NUREG-1801
Applicant must provide a
commitment which includes the
following elements: (1) to
participate in industry programs
for investigating and managing
aging effects applicable to
Reactor Internals, (2) to evaluate
and implement the results of the
industry programs as applicable
to the Reactor Internals design
and, (3) to submit, for NRC
review and approval an
inspection plan for Reactor
Internals, as based on industry
recommendation, at least 24
months prior to the extended
period.
No, but
licensee
commitment to
be confirmed.
Proposed Draft NUREG-1801
IV
B3
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Combustion Engineering
Structure and/or
Component
Material
Item
Link
R-161
Stainless
IV.B3.4- Core barrel
assembly#*#Core steel
c
barrel cylinder (top
and bottom
flange)#*#Lower
internals assemblyto- core barrel
bolts#*#Core
barrel-to-thermal
shield
bolts#*#Baffle
plates and formers
Environment
Aging Effect/
Mechanism
Reactor coolant Loss of fracture
and neutron flux toughness/ neutron
irradiation
embrittlement, void
swelling
IV B3-6
R-163
Stainless
Reactor coolant Changes in
IV.B3.4- Core shroud
assembly#*#Core steel, nickel
dimensions/Void
f
shroud assembly
bolts (later plants
are welded)
alloy
swelling
September 2004
Aging Management Program
(AMP)
Further
Evaluation
Applicant must provide a
commitment which includes the
following elements: (1) to
participate in industry programs
for investigating and managing
aging effects applicable to
Reactor Internals, (2) to evaluate
and implement the results of the
industry programs as applicable
to the Reactor Internals design
and, (3) to submit, for NRC
review and approval an
inspection plan for Reactor
Internals, as based on industry
recommendation, at least 24
months prior to the extended
period.
Applicant must provide a
commitment which includes the
following elements: (1) to
participate in industry programs
for investigating and managing
aging effects applicable to
Reactor Internals, (2) to evaluate
and implement the results of the
industry programs as applicable
to the Reactor Internals design
and, (3) to submit, for NRC
review and approval an
inspection plan for Reactor
Internals, as based on industry
recommendation, at least 24
months prior to the extended
period.
No, but
licensee
commitment to
be confirmed.
No, but
licensee
commitment to
be confirmed.
September 2004
IV
B3
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Combustion Engineering
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Item
Link
R-162
Stainless
Reactor coolant Cracking/ stress
Chapter XI.M2, “Water
IV.B3.4- Core shroud
assembly#*#Core steel, nickel
corrosion cracking, Chemistry,” for PWR primary
e
shroud assembly
bolts (later plants
are welded)
IV B3-7
R-164
alloy
primary water
stress corrosion
cracking,
irradiation-assisted
stress corrosion
cracking
Proposed Draft NUREG-1801
water in EPRI TR-105714 and
the applicant must provide a
commitment which includes the
following elements: (1) to
participate in industry programs
for investigating and managing
aging effects applicable to
Reactor Internals, (2) to evaluate
and implement the results of the
industry programs as applicable
to the Reactor Internals design
and, (3) to submit, for NRC
review and approval an
inspection plan for Reactor
Internals, as based on industry
recommendation, at least 24
months prior to the extended
period.
Stainless
Reactor coolant Loss of fracture
Applicant must provide a
IV.B3.4- Core shroud
assembly#*#Core steel, nickel and neutron flux toughness/ neutron commitment which includes the
g
shroud assembly alloy
irradiation
following elements: (1) to
bolts (later plants
embrittlement, void participate in industry programs
are welded)
swelling
for investigating and managing
aging effects applicable to
Reactor Internals, (2) to evaluate
and implement the results of the
industry programs as applicable
to the Reactor Internals design
and, (3) to submit, for NRC
review and approval an
inspection plan for Reactor
Internals, as based on industry
Further
Evaluation
No, but
licensee
commitment to
be confirmed.
No, but
licensee
commitment to
be confirmed.
Proposed Draft NUREG-1801
IV
B3
Item
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Combustion Engineering
Link
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
recommendation, at least 24
months prior to the extended
period.
IV B3-8
R-165
Stainless
Reactor coolant Loss of preload/
IV.B3.4- Core shroud
assembly#*#Core
steel,
nickel
stress relaxation
h
shroud assembly
bolts#*#Core
shroud tie rods
alloy
September 2004
Applicant must provide a
commitment which includes the
following elements: (1) to
participate in industry programs
for investigating and managing
aging effects applicable to
Reactor Internals, (2) to evaluate
and implement the results of the
industry programs as applicable
to the Reactor Internals design
and, (3) to submit, for NRC
review and approval an
inspection plan for Reactor
Internals, as based on industry
recommendation, at least 24
months prior to the extended
period.
No, but
licensee
commitment to
be confirmed.
September 2004
IV
B3
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Combustion Engineering
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Item
Link
R-159
Stainless
IV.B3.4- Core shroud
assembly#*#Core steel, cast
a
Reactor coolant Cracking/ stress
corrosion cracking,
irradiation-assisted
stress corrosion
cracking
Stainless
IV.B3.4- Core shroud
assembly#*#Core steel, cast
b
Reactor coolant Changes in
dimensions/Void
swelling
shroud tie rods
austenitic
(core support plate stainless
attached by welds steel
in later plants)
IV B3-9
R-160
Proposed Draft NUREG-1801
shroud tie rods
(core support plate
attached by welds
in later plants)
austenitic
stainless
steel, nickel
alloy
Aging Management Program
(AMP)
Further
Evaluation
Chapter XI.M2, “Water
Chemistry,” for PWR primary
water in EPRI TR-105714 and
the applicant must provide a
commitment which includes the
following elements: (1) to
participate in industry programs
for investigating and managing
aging effects applicable to
Reactor Internals, (2) to evaluate
and implement the results of the
industry programs as applicable
to the Reactor Internals design
and, (3) to submit, for NRC
review and approval an
inspection plan for Reactor
Internals, as based on industry
recommendation, at least 24
months prior to the extended
period.
Applicant must provide a
commitment which includes the
following elements: (1) to
participate in industry programs
for investigating and managing
aging effects applicable to
Reactor Internals, (2) to evaluate
and implement the results of the
industry programs as applicable
to the Reactor Internals design
and, (3) to submit, for NRC
review and approval an
inspection plan for Reactor
Internals, as based on industry
No, but
licensee
commitment to
be confirmed.
No, but
licensee
commitment to
be confirmed.
Proposed Draft NUREG-1801
IV
B3
Item
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Combustion Engineering
Link
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
recommendation, at least 24
months prior to the extended
period.
IV B3-10
R-158
IV.B3.3- Core support barrel Stainless
Core support barrel steel
b
upper flange
Reactor coolant Changes in
dimensions/Void
swelling
September 2004
Applicant must provide a
commitment which includes the
following elements: (1) to
participate in industry programs
for investigating and managing
aging effects applicable to
Reactor Internals, (2) to evaluate
and implement the results of the
industry programs as applicable
to the Reactor Internals design
and, (3) to submit, for NRC
review and approval an
inspection plan for Reactor
Internals, as based on industry
recommendation, at least 24
months prior to the extended
period.
No, but
licensee
commitment to
be confirmed.
September 2004
IV
B3
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Combustion Engineering
Structure and/or
Component
Material
Item
Link
R-155
IV.B3.3- Core support barrel Stainless
Core support barrel steel
a
upper flange
IV B3-11
R-157
IV.B3.3- Core support barrel Stainless
Core support barrel steel
a
upper flange
Environment
Aging Effect/
Mechanism
Reactor coolant Cracking/ stress
corrosion cracking,
irradiation-assisted
stress corrosion
cracking
Aging Management Program
(AMP)
Proposed Draft NUREG-1801
Chapter XI.M2, “Water
Chemistry,” for PWR primary
water in EPRI TR-105714 and
the applicant must provide a
commitment which includes the
following elements: (1) to
participate in industry programs
for investigating and managing
aging effects applicable to
Reactor Internals, (2) to evaluate
and implement the results of the
industry programs as applicable
to the Reactor Internals design
and, (3) to submit, for NRC
review and approval an
inspection plan for Reactor
Internals, as based on industry
recommendation, at least 24
months prior to the extended
period.
Reactor coolant Loss of fracture
Applicant must provide a
and neutron flux toughness/ neutron commitment which includes the
irradiation
following elements: (1) to
embrittlement, void participate in industry programs
swelling
for investigating and managing
aging effects applicable to
Reactor Internals, (2) to evaluate
and implement the results of the
industry programs as applicable
to the Reactor Internals design
and, (3) to submit, for NRC
review and approval an
inspection plan for Reactor
Internals, as based on industry
Further
Evaluation
No, but
licensee
commitment to
be confirmed.
No, but
licensee
commitment to
be confirmed.
Proposed Draft NUREG-1801
IV
B3
Item
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Combustion Engineering
Link
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
recommendation, at least 24
months prior to the extended
period.
IV B3-12
R-156
IV.B3.3- Core support
barrel#*#Core
b
Stainless
steel
Reactor coolant Loss of material/
wear
Chapter XI.M1, “ASME Section
XI Inservice Inspection,
Subsections IWB, IWC, and
IWD,” for Class 1 components
IV.B3.5- Lower internal
assembly
f
Cast
austenitic
stainless
steel
Reactor coolant
>250°C
(>482°F) and
neutron flux
Chapter XI.M13, “Thermal Aging No
and Neutron Irradiation
Embrittlement of Cast Austenitic
Stainless Steel (CASS)”
support barrel
upper
flange#*#Core
support barrel
alignment keys
R-171
Core support
column
September 2004
Loss of fracture
toughness/ thermal
aging and neutron
irradiation
embrittlement, void
swelling
No
September 2004
IV
B3
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Combustion Engineering
Structure and/or
Component
Item
Link
R-168
IV.B3.5- Lower internal
assembly
c
Material
Environment
Aging Effect/
Mechanism
IV B3-13
Stainless
Reactor coolant Changes in
steel, cast
dimensions/Void
Core support plate austenitic
swelling
Fuel alignment
stainless
pins
steel, nickel
Lower support
alloy
structure beam
assemblies
Core support
column
Core support
column bolts Core
support barrel
snubber
assemblies
Aging Management Program
(AMP)
Further
Evaluation
Applicant must provide a
commitment which includes the
following elements: (1) to
participate in industry programs
for investigating and managing
aging effects applicable to
Reactor Internals, (2) to evaluate
and implement the results of the
industry programs as applicable
to the Reactor Internals design
and, (3) to submit, for NRC
review and approval an
inspection plan for Reactor
Internals, as based on industry
recommendation, at least 24
months prior to the extended
period.
No, but
licensee
commitment to
be confirmed.
Proposed Draft NUREG-1801
Proposed Draft NUREG-1801
IV
B3
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Combustion Engineering
Structure and/or
Component
Item
Link
R-169
IV.B3.5- Lower internal
assembly
d
Material
Environment
IV B3-14
Stainless
Reactor coolant
steel, nickel and neutron flux
Core support plate alloy
Fuel alignment
pins
Lower support
structure beam
assemblies
Core support
column bolts Core
support barrel
snubber
assemblies
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
Loss of fracture
toughness/ neutron
irradiation
embrittlement, void
swelling
Applicant must provide a
commitment which includes the
following elements: (1) to
participate in industry programs
for investigating and managing
aging effects applicable to
Reactor Internals, (2) to evaluate
and implement the results of the
industry programs as applicable
to the Reactor Internals design
and, (3) to submit, for NRC
review and approval an
inspection plan for Reactor
Internals, as based on industry
recommendation, at least 24
months prior to the extended
period.
No, but
licensee
commitment to
be confirmed.
September 2004
September 2004
IV
B3
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Combustion Engineering
Structure and/or
Component
Aging Effect/
Mechanism
Material
Environment
IV.B3.5- Lower internal
assembly
a
Stainless
steel
Reactor coolant Cracking/ stress
corrosion cracking,
irradiation-assisted
stress corrosion
cracking
IV.B3.5- Lower internal
assembly
e
Stainless
Reactor coolant Loss of material/
steel, nickel
wear
alloy
Item
Link
R-166
Core support plate
Lower support
structure beam
assemblies
Core support
column
Core support barrel
snubber
assemblies
IV B3-15
R-170
Proposed Draft NUREG-1801
Fuel alignment
pins
Core support barrel
snubber
assemblies
Aging Management Program
(AMP)
Further
Evaluation
Chapter XI.M2, “Water
Chemistry,” for PWR primary
water in EPRI TR-105714 and
the applicant must provide a
commitment which includes the
following elements: (1) to
participate in industry programs
for investigating and managing
aging effects applicable to
Reactor Internals, (2) to evaluate
and implement the results of the
industry programs as applicable
to the Reactor Internals design
and, (3) to submit, for NRC
review and approval an
inspection plan for Reactor
Internals, as based on industry
recommendation, at least 24
months prior to the extended
period.
Chapter XI.M1, “ASME Section
XI Inservice Inspection,
Subsections IWB, IWC, and
IWD,” for Class 1 components
No, but
licensee
commitment to
be confirmed.
No
Proposed Draft NUREG-1801
IV
B3
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Combustion Engineering
Structure and/or
Component
Item
Link
R-167
IV.B3.5- Lower internal
Assembly
b
Fuel alignment
pins
Core support
column bolts
IV B3-16
R-54
IV.B3.2-f Reactor vessel
IV.B3.5-g internals
IV.B3.4-d components
Material
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
September 2004
Stainless
Reactor coolant Cracking/ stress
Chapter XI.M2, “Water
steel, nickel
corrosion cracking, Chemistry,” for PWR primary
alloy
primary water
water in EPRI TR-105714 and
stress corrosion
the applicant must provide a
cracking,
commitment which includes the
irradiation-assisted following elements: (1) to
stress corrosion
participate in industry programs
cracking
for investigating and managing
aging effects applicable to
Reactor Internals, (2) to evaluate
and implement the results of the
industry programs as applicable
to the Reactor Internals design
and, (3) to submit, for NRC
review and approval an
inspection plan for Reactor
Internals, as based on industry
recommendation, at least 24
months prior to the extended
period.
Stainless
Reactor coolant Cumulative fatigue For components for which a
steel, cast
damage/ fatigue
fatigue analysis has been
austenitic
performed for the 40-year period,
stainless
fatigue is a time-limited aging
steel, nickel
analysis (TLAA) to be performed
alloy
for the period of extended
operation.
See the Standard Review Plan,
Section 4.3 “Metal Fatigue,” for
acceptable methods for meeting
the requirements of 10 CFR
54.21(c).
Further
Evaluation
No, but
licensee
commitment to
be confirmed.
Yes,
TLAA
September 2004
IV
B3
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Combustion Engineering
Structure and/or
Component
Aging Management Program
(AMP)
Further
Evaluation
Reactor coolant Loss of material/
wear
Chapter XI.M1, “ASME Section
XI Inservice Inspection,
Subsections IWB, IWC, and
IWD,” for Class 1 components
No
Reactor coolant Changes in
dimensions/Void
swelling
Applicant must provide a
commitment which includes the
following elements: (1) to
participate in industry programs
for investigating and managing
aging effects applicable to
Reactor Internals, (2) to evaluate
and implement the results of the
industry programs as applicable
to the Reactor Internals design
and, (3) to submit, for NRC
review and approval an
inspection plan for Reactor
Internals, as based on industry
recommendation, at least 24
months prior to the extended
period.
No, but
licensee
commitment to
be confirmed.
Material
Environment
IV.B3.1- Upper Internals
Assembly
c
Stainless
steel
IV.B3.1- Upper Internals
Assembly
b
Stainless
steel
Item
Link
R-148
Aging Effect/
Mechanism
Fuel alignment
plate
Fuel alignment
plate guide lugs
and their lugs
Hold-down ring
IV B3-17
R-147
Upper guide
structure support
plate
Fuel alignment
plate
Fuel alignment
plate guide lugs
and guide lug
inserts
Proposed Draft NUREG-1801
Proposed Draft NUREG-1801
IV
B3
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) - Combustion Engineering
Structure and/or
Component
Item
Link
R-146
IV.B3.1- Upper Internals
Assembly
a
Upper guide
structure support
plate
Fuel alignment
plate
Fuel alignment
plate guide lugs
and guide lug
inserts
Aging Effect/
Mechanism
Material
Environment
Stainless
steel
Reactor coolant Cracking/ stress
corrosion cracking,
irradiation-assisted
stress corrosion
cracking
IV B3-18
Aging Management Program
(AMP)
Further
Evaluation
Chapter XI.M2, “Water
Chemistry,” for PWR primary
water in EPRI TR-105714 and
the applicant must provide a
commitment which includes the
following elements: (1) to
participate in industry programs
for investigating and managing
aging effects applicable to
Reactor Internals, (2) to evaluate
and implement the results of the
industry programs as applicable
to the Reactor Internals design
and, (3) to submit, for NRC
review and approval an
inspection plan for Reactor
Internals, as based on industry
recommendation, at least 24
months prior to the extended
period.
No, but
licensee
commitment to
be confirmed.
September 2004
B4
REACTOR VESSEL INTERNALS (PWR) - BABCOCK AND WILCOX
B4.1
Plenum Cover and Plenum Cylinder
B4.1.1
B4.1.2
B4.1.3
B4.1.4
B4.1.5
B4.2
Upper Grid Assembly
B4.2.1
B4.2.2
B4.2.3
B4.2.4
B4.2.5
B4.3
Core Support Shield Cylinder (Top and Bottom Flange)
Core Support Shield-to-Core Barrel Bolts
Outlet and Vent Valve Nozzles
Vent Valve Body and Retaining Ring
Vent Valve Assembly Locking Device
Core Barrel Assembly
B4.5.1
B4.5.2
B4.5.3
B4.5.4
B4.5.5
B4.6
CRGT Pipe and Flange
CRGT Spacer Casting
CRGT Spacer Screws
Flange-to-Upper Grid Screws
CRGT Rod Guide Tubes
CRGT Rod Guide Sectors
Core Support Shield Assembly
B4.4.1
B4.4.2
B4.4.3
B4.4.4
B4.4.5
B4.5
Upper Grid Rib Section
Upper Grid Ring Forging
Fuel Assembly Support Pads
Plenum Rib Pads
Rib-to-Ring Screws
Control Rod Guide Tube (CRGT) Assembly
B4.3.1
B4.3.2
B4.3.3
B4.3.4
B4.3.5
B4.3.6
B4.4
Plenum Cover Assembly
Plenum Cylinder
Reinforcing Plates
Top Flange-to-Cover Bolts
Bottom Flange-to-Upper Grid Screws
Core Barrel Cylinder (Top and Bottom Flange)
Lower Internals Assembly-to-Core Barrel Bolts
Core Barrel-to-Thermal Shield Bolts
Baffle Plates and Formers
Baffle/Former Bolts and Screws
Lower Grid (LG) Assembly
B4.6.1
B4.6.2
B4.6.3
B4.6.4
B4.6.5
B4.6.6
September 2004
Lower Grid Rib Section
Fuel Assembly Support Pads
Lower Grid Rib-to-Shell Forging Screws
Lower Grid Flow Distributor Plate
Orifice Plugs
Lower Grid and Shell Forgings
IV B4-1
Proposed Draft NUREG-1801
B4.6.7
B4.6.8
B4.6.9
B4.6.10
B4.6.11
B4.7
Flow Distributor Assembly
B4.7.1
B4.7.2
B4.7.3
B4.7.4
B4.8
Lower Internals Assembly-to-Thermal Shield Bolts
Guide Blocks and Bolts
Shock Pads and Bolts
Support Post Pipes
Incore Guide Tube Spider Castings
Flow Distributor Head and Flange
Shell Forging-to-Flow Distributor Bolts
Incore Guide Support Plate
Clamping Ring
Thermal Shield
Proposed Draft NUREG-1801
IV B4-2
September 2004
B4. REACTOR VESSEL INTERNALS (PWR) - BABCOCK AND WILCOX
Systems, Structures, and Components
This section comprises 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 RadioactiveWaste-Containing Components of Nuclear Power Plants,” all structures and components that
comprise the reactor vessel are governed by Group A or B Quality Standards.
System Interfaces
The systems that interface with the reactor vessel internals include the reactor pressure vessel
(IV.A2).
September 2004
IV B4-3
Proposed Draft NUREG-1801
IV
B4
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) – Babcock & Wilcox
Structure and/or
Component
Material
Item
Link
R-125
Stainless
IV.B4.5- Baffle/former
assembly#*#Baffle/forme steel
g
r bolts
R-180
IV.B4.3- Control rod guide tube
(CRGT)
a
Stainless
steel, cast
assembly#*#CRGT pipe austenitic
and flange#*#CRGT
stainless
spacer casting#*#CRGT steel
rod guide tubes#*#CRGT
rod guide sectors
Proposed Draft NUREG-1801
IV B4-4
Environment
Aging Effect/
Mechanism
Reactor coolant
and high fluence
(>1 x 10E21
n/cm2 E
>0.1 MeV)
Cracking/ stress
corrosion cracking,
irradiation-assisted
stress corrosion
cracking
Aging Management Program
(AMP)
Further
Evaluation
Applicant must provide a
No, but
commitment which includes the
licensee
following elements: (1) to
commitment to
participate in industry programs
be confirmed.
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
Reactor coolant Cracking/ stress
Chapter XI.M2, “Water
No, but
corrosion cracking, Chemistry,” for PWR primary
licensee
irradiation-assisted water in EPRI TR-105714 and the commitment to
stress corrosion
applicant must provide a
be confirmed.
cracking
commitment which includes the
following elements: (1) to
participate in industry programs
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
September 2004
September 2004
IV
B4
Item
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) – Babcock & Wilcox
Link
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
period.
Further
Evaluation
IV B4-5
Proposed Draft NUREG-1801
IV
B4
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) – Babcock & Wilcox
Structure and/or
Component
Item
Link
R-182
IV.B4.3- Control rod guide tube
(CRGT)
c
R-183
IV.B4.3- Control rod guide tube
(CRGT)
d
Material
Stainless
steel, cast
assembly#*#CRGT pipe austenitic
and flange#*#CRGT
stainless
spacer casting#*#CRGT steel
spacer screws#*#Flangeto-upper grid
screws#*#CRGT rod
guide tubes#*#CRGT rod
guide sectors
assembly#*#CRGT
spacer casting
Proposed Draft NUREG-1801
Cast
austenitic
stainless
steel
IV B4-6
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
Reactor coolant Changes in
dimensions/Void
swelling
Applicant must provide a
No, but
commitment which includes the
licensee
following elements: (1) to
commitment to
participate in industry programs
be confirmed.
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
Reactor coolant Loss of fracture
>250°C (>482°F) toughness/ thermal
and neutron flux aging and neutron
irradiation
embrittlement, void
swelling
Chapter XI.M13, “Thermal Aging
and Neutron Irradiation
Embrittlement of Cast Austenitic
Stainless Steel (CASS)”
September 2004
No
September 2004
IV
B4
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) – Babcock & Wilcox
Structure and/or
Component
Aging Effect/
Mechanism
Material
Environment
IV.B4.3- Control rod guide tube
(CRGT)
b
Stainless
steel
Reactor coolant Cracking/ stress
corrosion cracking,
irradiation-assisted
stress corrosion
cracking
IV.B4.3- Control rod guide tube
(CRGT)
e
Stainless
steel
Reactor coolant Loss of preload/
stress relaxation
Item
Link
R-181
assembly#*#CRGT
spacer screws#*#Flangeto-upper grid screws
IV B4-7
R-184
assembly#*#Flange-toupper grid screws
Aging Management Program
(AMP)
Further
Evaluation
Proposed Draft NUREG-1801
Chapter XI.M2, “Water
No, but
Chemistry,” for PWR primary
licensee
water in EPRI TR-105714 and the commitment to
applicant must provide a
be confirmed.
commitment which includes the
following elements: (1) to
participate in industry programs
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
Applicant must provide a
No, but
commitment which includes the
licensee
following elements: (1) to
commitment to
participate in industry programs
be confirmed.
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
IV
B4
Item
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) – Babcock & Wilcox
Link
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
period.
R-199
Stainless
IV.B4.5- Core barrel
assembly#*#Baffle/forme
steel
h
r bolts and screws
Proposed Draft NUREG-1801
IV B4-8
Reactor coolant Changes in
dimensions/Void
swelling
September 2004
Applicant must provide a
No, but
commitment which includes the
licensee
following elements: (1) to
commitment to
participate in industry programs
be confirmed.
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
September 2004
IV
B4
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) – Babcock & Wilcox
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Item
Link
R-198
Stainless
IV.B4.5- Core barrel
assembly#*#Baffle/forme steel
g
Reactor coolant Cracking/ stress
corrosion cracking,
irradiation-assisted
stress corrosion
cracking
IV.B4.5-j Core barrel
Reactor coolant Loss of preload/
stress relaxation
r bolts and screws
IV B4-9
R-201
Proposed Draft NUREG-1801
Stainless
assembly#*#Baffle/forme steel
r bolts and screws
Aging Management Program
(AMP)
Further
Evaluation
A plant-specific aging
management program is to be
evaluated. Historically the VT-3
visual examinations have not
identified baffle/former bolt
cracking because cracking occurs
at the juncture of the bolt head
and shank, which is not
accessible for visual inspection.
However, recent UT examinations
of the baffle/former bolts have
identified cracking in several
plants. The industry is currently
addressing the issue of baffle bolt
cracking in the PWR Materials
Reliability Project, Issues Task
Group (ITG) activities to
determine, develop, and
implement the necessary steps
and plans to manage the
applicable aging effects on a
plant-specific basis.
A plant-specific aging
management program is to be
evaluated.
Yes, plant
specific
Visual inspection (VT-3) is to be
augmented to detect relevant
conditions of stress relaxation
because only the heads of the
baffle/former bolts are visible, and
a plant-specific aging
management program is thus
required.
Yes, plant
specific
IV
B4
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) – Babcock & Wilcox
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Item
Link
R-200
IV.B4.5-i Core barrel
Reactor coolant Loss of fracture
A plant-specific aging
and neutron flux toughness/ neutron management program is to be
irradiation
evaluated.
embrittlement, void
swelling
R-193
Stainless
IV.B4.5- Core barrel
assembly#*#Core barrel steel
a
Reactor coolant Cracking/ stress
corrosion cracking,
irradiation-assisted
stress corrosion
cracking
Stainless
assembly#*#Baffle/forme steel
r bolts and screws
cylinder (top and bottom
flange)#*#Baffle plates
and formers
Proposed Draft NUREG-1801
IV B4-10
September 2004
Further
Evaluation
Yes, plant
specific
Chapter XI.M2, “Water
No, but
Chemistry,” for PWR primary
licensee
water in EPRI TR-105714 and the commitment to
applicant must provide a
be confirmed.
commitment which includes the
following elements: (1) to
participate in industry programs
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
September 2004
IV
B4
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) – Babcock & Wilcox
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Item
Link
R-195
Stainless
Reactor coolant Changes in
IV.B4.5- Core barrel
assembly#*#Core barrel steel, nickel
dimensions/Void
c
cylinder (top and bottom alloy
flange)#*#Lower internals
assembly-to- core barrel
bolts#*#Core barrel-tothermal shield
bolts#*#Baffle plates and
formers
IV B4-11
R-196
swelling
Aging Management Program
(AMP)
Applicant must provide a
No, but
commitment which includes the
licensee
following elements: (1) to
commitment to
participate in industry programs
be confirmed.
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
Stainless
Reactor coolant Loss of fracture
Applicant must provide a
IV.B4.5- Core barrel
assembly#*#Core barrel steel, nickel and neutron flux toughness/ neutron commitment which includes the
d
Proposed Draft NUREG-1801
cylinder (top and bottom alloy
flange)#*#Lower internals
assembly-to- core barrel
bolts#*#Core barrel-tothermal shield
bolts#*#Baffle plates and
formers
Further
Evaluation
irradiation
following elements: (1) to
embrittlement, void participate in industry programs
swelling
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
No, but
licensee
commitment to
be confirmed.
IV
B4
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) – Babcock & Wilcox
Structure and/or
Component
Item
Link
R-194
IV.B4.5- Core barrel
assembly#*#Lower
b
internals assembly-tocore barrel bolts
Core barrel-to-thermal
shield bolts
R-197
IV.B4.5- Core barrel
assembly#*#Lower
e
internals assembly-tocore barrel bolts
Core barrel-to-thermal
shield bolts
Proposed Draft NUREG-1801
Material
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
Stainless
Reactor coolant Cracking/ stress
Chapter XI.M2, “Water
No, but
steel, nickel
corrosion cracking, Chemistry,” for PWR primary
licensee
alloy
irradiation-assisted water in EPRI TR-105714 and the commitment to
stress corrosion
applicant must provide a
be confirmed.
cracking
commitment which includes the
following elements: (1) to
participate in industry programs
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
Stainless
Reactor coolant Loss of preload/
Applicant must provide a
No, but
steel, nickel
stress relaxation
commitment which includes the
licensee
alloy
following elements: (1) to
commitment to
participate in industry programs
be confirmed.
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
IV B4-12
September 2004
September 2004
IV
B4
Item
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) – Babcock & Wilcox
Link
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
period.
IV B4-13
R-190
IV.B4.4- Core support shield
assembly
f
Core support shield
cylinder
(top flange)
VV assembly locking
device
Stainless
steel
Reactor coolant Loss of material/
wear
Chapter XI.M1, “ASME Section XI No
Inservice Inspection, Subsections
IWB, IWC, and IWD,” for Class 1
components
Proposed Draft NUREG-1801
IV
B4
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) – Babcock & Wilcox
Structure and/or
Component
Item
Link
R-188
IV.B4.4- Core support shield
assembly
d
Material
Environment
Stainless
Reactor coolant
steel, nickel and neutron flux
Core support shield
alloy, PH
cylinder (top and bottom Stainless
flange)
Steel forging
Core support shield-tocore barrel bolts
Outlet and vent valve
(VV) nozzles
VV assembly locking
device
Proposed Draft NUREG-1801
IV B4-14
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Loss of fracture
toughness/ neutron
irradiation
embrittlement, void
swelling
Applicant must provide a
No, but
commitment which includes the
licensee
following elements: (1) to
commitment to
participate in industry programs
be confirmed.
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
September 2004
Further
Evaluation
September 2004
IV
B4
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) – Babcock & Wilcox
Structure and/or
Component
Item
Link
R-187
IV.B4.4- Core support shield
assembly
c
R-185
IV.B4.4- Core support shield
assembly
a
Material
Environment
Aging Effect/
Mechanism
Stainless
Reactor coolant Changes in
steel, nickel
dimensions/Void
Core support shield
alloy, PH
swelling
cylinder (top and bottom Stainless
flange)
Steel forging
Core support shield-tocore barrel bolts
VV retaining ring
VV assembly locking
device
IV B4-15
Proposed Draft NUREG-1801
Stainless
Reactor coolant
steel, PH
Core support shield
stainless
cylinder (top and bottom steel forging,
flange)
CASS
Outlet and vent valve
(VV) nozzles
VV body and retaining
ring
Cracking/ stress
corrosion cracking,
irradiation-assisted
stress corrosion
cracking
Aging Management Program
(AMP)
Further
Evaluation
Applicant must provide a
No, but
commitment which includes the
licensee
following elements: (1) to
commitment to
participate in industry programs
be confirmed.
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
Chapter XI.M2, “Water
No, but
Chemistry,” for PWR primary
licensee
water in EPRI TR-105714 and the commitment to
applicant must provide a
be confirmed.
commitment which includes the
following elements: (1) to
participate in industry programs
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
IV
B4
Item
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) – Babcock & Wilcox
Link
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
24 months prior to the extended
period.
R-192
IV.B4.4- Core support shield
assembly
h
Core support shield-tocore barrel bolts
Proposed Draft NUREG-1801
Stainless
Reactor coolant Loss of preload/
steel, nickel
stress relaxation
alloy
IV B4-16
September 2004
Applicant must provide a
No, but
commitment which includes the
licensee
following elements: (1) to
commitment to
participate in industry programs
be confirmed.
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
September 2004
IV
B4
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) – Babcock & Wilcox
Structure and/or
Component
Item
Link
R-186
IV.B4.4- Core support shield
assembly
b
Core support shield-tocore barrel bolts
VV assembly locking
device
IV B4-17
R-191
IV.B4.4- Core support shield
assembly
g
Proposed Draft NUREG-1801
Outlet and vent valve
nozzles
VV body and retaining
ring
Material
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
Stainless
Reactor coolant Cracking/ stress
Chapter XI.M2, “Water
No, but
steel, nickel
corrosion cracking, Chemistry,” for PWR primary
licensee
alloy
irradiation-assisted water in EPRI TR-105714 and the commitment to
stress corrosion
applicant must provide a
be confirmed.
cracking
commitment which includes the
following elements: (1) to
participate in industry programs
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
Cast
Reactor coolant Loss of fracture
Chapter XI.M13, “Thermal Aging No
austenitic
>250°C (>482°F) toughness/ thermal and Neutron Irradiation
stainless
and neutron flux aging and neutron Embrittlement of Cast Austenitic
steel
irradiation
Stainless Steel (CASS)”
embrittlement, void
swelling
IV
B4
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) – Babcock & Wilcox
Structure and/or
Component
Material
Item
Link
R-209
IV.B4.7- Flow distributor assembly Stainless
Flow distributor head and steel
a
R-212
Environment
Aging Effect/
Mechanism
Reactor coolant Cracking/ stress
corrosion cracking,
irradiation-assisted
stress corrosion
cracking
Aging Management Program
(AMP)
Further
Evaluation
Chapter XI.M2, “Water
No, but
Chemistry,” for PWR primary
licensee
flange
water in EPRI TR-105714 and the commitment to
Incore guide support
applicant must provide a
be confirmed.
plate
commitment which includes the
Clamping ring
following elements: (1) to
participate in industry programs
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
Flow
distributor
assembly
Stainless
Reactor
coolant
Loss
of
fracture
Applicant must provide a
No, but
IV.B4.7Flow
distributor
head
and
steel,
nickel
and
neutron
flux
toughness/
neutron
commitment
which
includes
the
licensee
d
flange
alloy
irradiation
following elements: (1) to
commitment to
Shell forging-to-flow
embrittlement, void participate in industry programs
be confirmed.
distributor bolts
swelling
for investigating and managing
Incore guide support
aging effects applicable to Reactor
plate
Internals, (2) to evaluate and
Clamping ring
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
Proposed Draft NUREG-1801
IV B4-18
September 2004
September 2004
IV
B4
Item
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) – Babcock & Wilcox
Link
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
period.
IV B4-19
R-211
Reactor coolant Changes in
IV.B4.7- Flow distributor assembly Stainless
Flow
distributor
head
and
steel,
nickel
dimensions/Void
c
flange
Shell forging-to-flow
distributor bolts
Incore guide support
plate Clamping ring
alloy
swelling
Proposed Draft NUREG-1801
Applicant must provide a
No, but
commitment which includes the
licensee
following elements: (1) to
commitment to
participate in industry programs
be confirmed.
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
IV
B4
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) – Babcock & Wilcox
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Item
Link
R-210
Reactor coolant Cracking/ stress
Chapter XI.M2, “Water
IV.B4.7- Flow distributor assembly Stainless
Shell forging-to-flow
steel, nickel
corrosion cracking, Chemistry,” for PWR primary
b
R-213
Further
Evaluation
No, but
licensee
distributor bolts
alloy
irradiation-assisted water in EPRI TR-105714 and the commitment to
stress corrosion
applicant must provide a
be confirmed.
cracking
commitment which includes the
following elements: (1) to
participate in industry programs
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
Flow
distributor
assembly
Stainless
Reactor
coolant
Loss
of
preload/
Applicant must provide a
No, but
IV.B4.7Shell
forging-to-flow
steel,
nickel
stress
relaxation
commitment
which
includes
the
licensee
e
distributor bolts
alloy
following elements: (1) to
commitment to
participate in industry programs
be confirmed.
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
Proposed Draft NUREG-1801
IV B4-20
September 2004
September 2004
IV
B4
Item
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) – Babcock & Wilcox
Link
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
period.
IV B4-21
R-208
IV.B4.6- Lower grid assembly
Fuel assembly support
h
Stainless
steel
Reactor coolant Loss of material/
wear
Chapter XI.M1, “ASME Section XI No
Inservice Inspection, Subsections
IWB, IWC, and IWD,” for Class 1
components
Reactor coolant Loss of fracture
>250°C (>482°F) toughness/ thermal
and neutron flux aging and neutron
irradiation
embrittlement, void
swelling
Chapter XI.M13, “Thermal Aging
and Neutron Irradiation
Embrittlement of Cast Austenitic
Stainless Steel (CASS)”
pads Guide blocks
R-206
Proposed Draft NUREG-1801
Cast
IV.B4.6- Lower grid assembly
Incore guide tube spider austenitic
e
castings
stainless
steel
No
IV
B4
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) – Babcock & Wilcox
Structure and/or
Component
Item
Link
R-202
IV.B4.6- Lower grid assembly
Lower grid rib section
a
Material
Stainless
steel, cast
Fuel assembly support
austenitic
pads
stainless
Lower grid flow dist. plate steel
Orifice plugs
Lower grid and shell
forgings
Guide blocks
Shock pads
Support post pipes
Incore guide tube spider
castings
Proposed Draft NUREG-1801
IV B4-22
Environment
Aging Effect/
Mechanism
Reactor coolant Cracking/ stress
corrosion cracking,
irradiation-assisted
stress corrosion
cracking
September 2004
Aging Management Program
(AMP)
Further
Evaluation
Chapter XI.M2, “Water
No, but
Chemistry,” for PWR primary
licensee
water in EPRI TR-105714 and the commitment to
applicant must provide a
be confirmed.
commitment which includes the
following elements: (1) to
participate in industry programs
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
September 2004
IV
B4
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) – Babcock & Wilcox
Structure and/or
Component
Item
Link
R-204
IV.B4.6- Lower grid assembly
Lower grid rib section
c
IV B4-23
Fuel assembly support
pads
Lower grid rib-to-shell
forging screws
Lower grid flow dist. plate
Orifice plugs
Lower grid and shell
forgings
Lower internals
assembly-to- thermal
shield bolts
Guide blocks and bolts
Shock pads and bolts
Support post pipes
Incore guide tube spider
castings
Material
Environment
Aging Effect/
Mechanism
Stainless
Reactor coolant Changes in
steel, cast
dimensions/Void
austenitic
swelling
stainless
steel, nickel
alloy
Aging Management Program
(AMP)
Further
Evaluation
Applicant must provide a
No, but
commitment which includes the
licensee
following elements: (1) to
commitment to
participate in industry programs
be confirmed.
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
Proposed Draft NUREG-1801
IV
B4
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) – Babcock & Wilcox
Structure and/or
Component
Item
Link
R-205
IV.B4.6- Lower grid assembly
Lower grid rib section
d
Material
Fuel assembly support
pads
Lower grid rib-to-shell
forging screws
Lower grid flow dist. plate
Orifice plugs
Lower grid and shell
forgings
Lower internals
assembly-to-thermal
shield bolts
Guide blocks and bolts
Shock pads and bolts
Support post pipes
Proposed Draft NUREG-1801
Environment
Aging Effect/
Mechanism
Stainless
Reactor coolant Loss of fracture
steel, nickel and neutron flux toughness/ neutron
alloy
irradiation
embrittlement, void
swelling
IV B4-24
September 2004
Aging Management Program
(AMP)
Further
Evaluation
Applicant must provide a
No, but
commitment which includes the
licensee
following elements: (1) to
commitment to
participate in industry programs
be confirmed.
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
September 2004
IV
B4
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) – Babcock & Wilcox
Structure and/or
Component
Item
Link
R-203
IV.B4.6- Lower grid assembly
Lower grid rib-to-shell
b
forging screws
Lower internals
assembly-to- thermal
shield bolts
Guide blocks and bolts
Shock pads and bolts
IV B4-25
R-207
IV.B4.6- Lower grid assembly
Lower grid rib-to-shell
g
Proposed Draft NUREG-1801
forging screws
Lower internals
assembly-to-thermal
shield bolts
Material
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
Stainless
Reactor coolant Cracking/ stress
Chapter XI.M2, “Water
No, but
steel, nickel
corrosion cracking, Chemistry,” for PWR primary
licensee
alloy
irradiation-assisted water in EPRI TR-105714 and the commitment to
stress corrosion
applicant must provide a
be confirmed.
cracking
commitment which includes the
following elements: (1) to
participate in industry programs
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
Stainless
Reactor coolant Loss of preload/
Applicant must provide a
No, but
steel, nickel
stress relaxation
commitment which includes the
licensee
alloy
following elements: (1) to
commitment to
participate in industry programs
be confirmed.
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
IV
B4
Item
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) – Babcock & Wilcox
Link
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
period.
R-172
IV.B4.1- Plenum cover and
plenum cylinder
a
Stainless
steel
Plenum cover assembly
Plenum cylinder
Reinforcing plates
Proposed Draft NUREG-1801
IV B4-26
Reactor coolant Cracking/ stress
corrosion cracking,
irradiation-assisted
stress corrosion
cracking
September 2004
Chapter XI.M2, “Water
No, but
Chemistry,” for PWR primary
licensee
water in EPRI TR-105714 and the commitment to
applicant must provide a
be confirmed.
commitment which includes the
following elements: (1) to
participate in industry programs
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
September 2004
IV
B4
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) – Babcock & Wilcox
Aging Effect/
Mechanism
Item
Link
Structure and/or
Component
Material
Environment
R-174
IV.B4.1- Plenum cover and
plenum cylinder
c
Stainless
steel
Reactor coolant Changes in
dimensions/Void
swelling
IV.B4.1- Plenum cover and
plenum cylinder
b
Stainless
steel
Plenum cover assembly
Plenum cylinder
Reinforcing plates
Top flange-to-cover bolts
Bottom flange-to-upper
grid screws
IV B4-27
R-173
Top flange-to-cover bolts
Bottom flange-to-upper
grid screws
Aging Management Program
(AMP)
Further
Evaluation
Proposed Draft NUREG-1801
Applicant must provide a
No, but
commitment which includes the
licensee
following elements: (1) to
commitment to
participate in industry programs
be confirmed.
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
Reactor coolant Cracking/ stress
Chapter XI.M2, “Water
No, but
corrosion cracking, Chemistry,” for PWR primary
licensee
irradiation-assisted water in EPRI TR-105714 and the commitment to
stress corrosion
applicant must provide a
be confirmed.
cracking
commitment which includes the
following elements: (1) to
participate in industry programs
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
IV
B4
Item
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) – Babcock & Wilcox
Link
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
period.
R-54
IV.B4.6-f Reactor vessel internals
IV.B4.5-f components
IV.B4.1-d
IV.B4.2-d
IV.B4.3-f
Proposed Draft NUREG-1801
Stainless
Reactor coolant Cumulative fatigue
steel, cast
damage/ fatigue
austenitic
stainless
steel, nickel
alloy
IV B4-28
September 2004
For components for which a
Yes,
fatigue analysis has been
TLAA
performed for the 40-year period,
fatigue is a time-limited aging
analysis (TLAA) to be performed
for the period of extended
operation.
See the Standard Review Plan,
Section 4.3 “Metal Fatigue,” for
acceptable methods for meeting
the requirements of 10 CFR
54.21(c).
September 2004
IV
B4
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) – Babcock & Wilcox
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Item
Link
R-189
IV.B4.4- Reactor vessel internals Stainless
components
steel, cast
e
Reactor coolant Cumulative fatigue
damage/ fatigue
IV.B4.8- Thermal shield
b
Reactor coolant Changes in
dimensions/Void
swelling
austenitic
stainless
steel, nickel
alloy, PH
Stainless
Steel forging
IV B4-29
R-215
Stainless
steel
Aging Management Program
(AMP)
Further
Evaluation
Proposed Draft NUREG-1801
For components for which a
Yes,
fatigue analysis has been
TLAA
performed for the
40-year period, fatigue is a timelimited aging analysis (TLAA) to
be performed for the period of
extended operation, and, for
Class 1 components,
environmental effects on fatigue
are to be addressed. See the
Standard Review Plan, Section
4.3 “Metal Fatigue,” for acceptable
methods for meeting the
requirements of 10 CFR 54.21(c).
Applicant must provide a
No, but
commitment which includes the
licensee
following elements: (1) to
commitment to
participate in industry programs
be confirmed.
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
IV
B4
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) – Babcock & Wilcox
Structure and/or
Component
Aging Effect/
Mechanism
Material
Environment
IV.B4.8- Thermal shield
c
Stainless
steel
Reactor coolant Loss of fracture
and neutron flux toughness/ neutron
irradiation
embrittlement, void
swelling
IV.B4.8- Thermal shield
a
Stainless
steel
Item
Link
R-216
R-214
Proposed Draft NUREG-1801
IV B4-30
Aging Management Program
(AMP)
Further
Evaluation
Applicant must provide a
No, but
commitment which includes the
licensee
following elements: (1) to
commitment to
participate in industry programs
be confirmed.
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
Reactor coolant Cracking/ stress
Chapter XI.M2, “Water
No, but
corrosion cracking, Chemistry,” for PWR primary
licensee
irradiation-assisted water in EPRI TR-105714 and the commitment to
stress corrosion
applicant must provide a
be confirmed.
cracking
commitment which includes the
following elements: (1) to
participate in industry programs
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
September 2004
September 2004
IV
B4
Item
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) – Babcock & Wilcox
Link
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
period.
IV B4-31
R-179
IV.B4.2- Upper grid assembly
Fuel assembly support
f
pads Plenum rib pads
Stainless
steel
Reactor coolant Loss of material/
wear
Chapter XI.M1, “ASME Section XI No
Inservice Inspection, Subsections
IWB, IWC, and IWD,” for Class 1
components
Proposed Draft NUREG-1801
IV
B4
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) – Babcock & Wilcox
Structure and/or
Component
Item
Link
R-176
IV.B4.2- Upper grid assembly
Rib- to-ring screws
b
Proposed Draft NUREG-1801
Aging Effect/
Mechanism
Material
Environment
Stainless
steel
Reactor coolant Cracking/ stress
corrosion cracking,
irradiation-assisted
stress corrosion
cracking
IV B4-32
September 2004
Aging Management Program
(AMP)
Further
Evaluation
Chapter XI.M2, “Water
No, but
Chemistry,” for PWR primary
licensee
water in EPRI TR-105714 and the commitment to
applicant must provide a
be confirmed.
commitment which includes the
following elements: (1) to
participate in industry programs
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
September 2004
IV
B4
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) – Babcock & Wilcox
Structure and/or
Component
Aging Effect/
Mechanism
Material
Environment
IV.B4.2- Upper grid assembly
Upper grid rib section
a
Stainless
steel
Reactor coolant Cracking/ stress
corrosion cracking,
irradiation-assisted
stress corrosion
cracking
IV.B4.2- Upper grid assembly
Upper grid rib section
c
Stainless
steel
Reactor coolant Changes in
dimensions/Void
swelling
Item
Link
R-175
Upper grid ring forging
Fuel assembly support
pads
Plenum rib pads
IV B4-33
R-177
Proposed Draft NUREG-1801
Upper grid ring forging
Fuel assembly support
pads
Plenum rib pads
Rib-to-ring screws
Aging Management Program
(AMP)
Further
Evaluation
Chapter XI.M2, “Water
No, but
Chemistry,” for PWR primary
licensee
water in EPRI TR-105714 and the commitment to
applicant must provide a
be confirmed.
commitment which includes the
following elements: (1) to
participate in industry programs
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
Applicant must provide a
No, but
commitment which includes the
licensee
following elements: (1) to
commitment to
participate in industry programs
be confirmed.
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
IV
B4
Item
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Vessel Internals (PWR) – Babcock & Wilcox
Link
Structure and/or
Component
Material
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
period.
R-178
IV.B4.2- Upper grid assembly
Upper grid rib section
e
Stainless
steel
Upper grid ring forging
Fuel assembly support
pads
Plenum rib pads
Rib-to-ring screws
Proposed Draft NUREG-1801
IV B4-34
Reactor coolant Loss of fracture
toughness/ neutron
irradiation
embrittlement, void
swelling
September 2004
Applicant must provide a
No, but
commitment which includes the
licensee
following elements: (1) to
commitment to
participate in industry programs
be confirmed.
for investigating and managing
aging effects applicable to Reactor
Internals, (2) to evaluate and
implement the results of the
industry programs as applicable to
the Reactor Internals design and,
(3) to submit, for NRC review and
approval an inspection plan for
Reactor Internals, as based on
industry recommendation, at least
24 months prior to the extended
period.
This Page Intentionally Left Blank
September 2004
IV B4-35
Proposed Draft NUREG-1801
C1. REACTOR COOLANT PRESSURE BOUNDARY (BOILING WATER REACTOR)
C1.1
Piping and Fittings
C1.1.1
C1.1.2
C1.1.3
C1.1.4
C1.1.5
C1.1.6
C1.1.7
C1.1.8
C1.1.9
C1.1.10
C1.1.11
Main Steam
Feedwater
High Pressure Coolant Injection (HPCI) System
Reactor Core Isolation Cooling (RCIC) System
Recirculation
Residual Heat Removal (RHR) System
Low Pressure Coolant Injection (LPCI) System
Low Pressure Core Spray (LPCS) System
High Pressure Core Spray (HPCS) System
Lines to Isolation Condenser
Lines to Reactor Water Cleanup (RWC) and Standby Liquid Control
(SLC) Systems
C1.1.12 Steam Line to HPCI and RCIC Pump Turbine
C1.1.13 Small Bore Piping Less than NPS 4
C1.2
Recirculation Pump
C1.2.1
C1.2.2
C1.2.3
C1.2.4
C1.3
Valves
C1.3.1
C1.3.2
C1.3.3
C1.3.4
C1.4
Casing
Cover
Seal Flange
Closure Bolting
Body
Bonnet
Seal Flange
Closure Bolting
Isolation Condenser
C1.4.1
C1.4.2
C1.4.3
C1.4.4
September 2004
Tubing
Tubesheet
Channel Head
Shell
IV C1-1
Proposed Draft NUREG-1801
C1. REACTOR COOLANT PRESSURE BOUNDARY (BOILING WATER REACTOR)
Systems, Structures, and Components
This section comprises 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 system (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.
Therefore, they are not subject to an aging management review, pursuant to 10 CFR
54.21(a)(1).
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).
Proposed Draft NUREG-1801
IV C1-2
September 2004
September 2004
IV
C1
Item
R-03
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Coolant Pressure Boundary (BWR)
Structure
and/or
Material
Component
IV.C1.1-i Class 1 piping, Stainless
fittings and
steel, Steel
branch
connections <
NPS 4
Link
Environment
Aging Effect/
Mechanism
Reactor coolant Cracking/ stress
corrosion cracking
and intergranular
stress corrosion
cracking
Aging Management Program
(AMP)
Chapter XI.M1, “ASME Section XI
Inservice Inspection, Subsections
IWB, IWC, and IWD,” for Class 1
components and
Further
Evaluation
Yes,
parameters
monitored/
inspected and
detection of
Chapter XI.M2, “Water Chemistry,” aging effects
for BWR water in BWRVIP-29
are to be
(EPRI TR-103515)
evaluated
IV C1-3
Inspection in accordance with
ASME Section XI does not require
volumetric examination of pipes
less than NPS 4. A plant-specific
destructive examination or a
nondestructive examination (NDE)
that permits inspection of the inside
surfaces of the piping is to be
conducted to ensure that cracking
has not occurred and the
component intended function will
be maintained during the extended
period of operation.
Proposed Draft NUREG-1801
Proposed Draft NUREG-1801
IV
C1
Item
R-55
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Coolant Pressure Boundary (BWR)
Structure
and/or
Material
Component
IV.C1.1-i Class 1 piping, Stainless
fittings and
steel, Steel
branch
connections <
NPS 4
Link
IV C1-4
R-52
September 2004
IV.C1.1- Class 1 piping, Cast
piping
austenitic
g
components,
and piping
elements
stainless
steel
Environment
Aging Effect/
Mechanism
Reactor coolant Cracking/ thermal
and mechanical
loading
Aging Management Program
(AMP)
Further
Evaluation
Chapter XI.M1, “ASME Section XI
Inservice Inspection, Subsections
IWB, IWC, and IWD,” for Class 1
components
Yes,
parameters
monitored/
inspected and
detection of
Inspection in accordance with
aging effects
ASME Section XI does not require are to be
volumetric examination of pipes
evaluated
less than NPS 4. A plant-specific
destructive examination or a
nondestructive examination (NDE)
that permits inspection of the inside
surfaces of the piping is to be
conducted to ensure that cracking
has not occurred and the
component intended function will
be maintained during the extended
period of operation.
The AMPs are to be augmented by
verifying that service-induced weld
cracking is not occurring in the
small-bore piping less than NPS 4,
including pipe, fittings, and branch
connections. See Chapter XI.M32,
“One-Time Inspection” for an
acceptable verification method.
Reactor coolant Loss of fracture
Chapter XI.M12, “Thermal Aging
No
>250°C (>482°F) toughness/ thermal Embrittlement of Cast Austenitic
aging embrittlement Stainless Steel (CASS)”
September 2004
IV
C1
Item
R-08
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Coolant Pressure Boundary (BWR)
Structure
and/or
Component
IV.C1.3-b Class 1 pump
IV.C1.2-c casings, and
valve bodies
and bonnets
Link
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Material
Environment
Cast
austenitic
stainless
steel
Reactor coolant Loss of fracture
Chapter XI.M1, “ASME Section XI
>250°C (>482°F) toughness/ thermal Inservice Inspection, Subsections
aging embrittlement IWB, IWC, and IWD,” for Class 1
components
IV C1-5
For pump casings and valve
bodies, screening for susceptibility
to thermal aging is not required.
The ASME Section XI inspection
requirements are sufficient for
managing the effects of loss of
fracture toughness due to thermal
aging embrittlement 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
embrittlement of CASS pump
casings.
Further
Evaluation
No
Proposed Draft NUREG-1801
Proposed Draft NUREG-1801
IV
C1
Item
R-15
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Coolant Pressure Boundary (BWR)
Structure
and/or
Component
IV.C1.4- Isolation
condenser
a
tube side
components
Link
Aging Effect/
Mechanism
Material
Environment
Stainless
steel, Steel
Reactor coolant Cracking/ stress
corrosion cracking
and intergranular
stress corrosion
cracking
Aging Management Program
(AMP)
Further
Evaluation
Chapter XI.M1, “ASME Section XI
Inservice Inspection, Subsections
IWB, IWC, and IWD,” for Class 1
components and
Yes, plant
specific
Chapter XI.M2, “Water Chemistry,”
for BWR water in BWRVIP-29
(EPRI
TR-103515)
IV C1-6
The AMP in Chapter XI.M1 is to be
augmented to detect cracking due
to stress corrosion cracking and
cyclic loading or loss of material
due to pitting and crevice corrosion,
and verification of the effectiveness
of the program is required 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.
September 2004
September 2004
IV
C1
Item
R-16
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Coolant Pressure Boundary (BWR)
Structure
and/or
Component
IV.C1.4- Isolation
condenser
b
tube side
components
Link
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Material
Environment
Stainless
steel, Steel
Reactor coolant Loss of material/
Chapter XI.M1, “ASME Section XI
general, pitting and Inservice Inspection, Subsections
crevice corrosion
IWB, IWC, and IWD,” for Class 1
components and
Further
Evaluation
Yes, plant
specific
Chapter XI.M2, “Water Chemistry,”
for BWR water in BWRVIP-29
(EPRI
TR-103515)
IV C1-7
Proposed Draft NUREG-1801
R-23
IV.C1.3-a Piping, piping Steel
IV.C1.1-a components,
IV.C1.1-c and piping
elements
The AMP in Chapter XI.M1 is to be
augmented to detect cracking due
to stress corrosion cracking and
cyclic loading or loss of material
due to pitting and crevice corrosion,
and verification of the effectiveness
of the program is required 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.
Reactor coolant Wall thinning/ flow- Chapter XI.M17, “Flow-Accelerated No
accelerated
Corrosion”
corrosion
Proposed Draft NUREG-1801
IV
C1
Item
R-04
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Coolant Pressure Boundary (BWR)
Structure
and/or
Component
IV.C1.1-e Piping, piping
IV.C1.1-h components,
IV.C1.2-a and piping
IV.C1.1-b elements
IV.C1.3-d
IV.C1.1-d
Link
Material
Environment
Aging Effect/
Mechanism
Steel,
Reactor coolant Cumulative fatigue
stainless
damage/ fatigue
steel, cast
austenitic
stainless
steel, carbon
steel with
nickel-alloy or
stainless
steel
cladding,
nickel-alloy
IV C1-8
R-21
IV.C1.1- Piping, piping Nickel alloy
components,
f
and piping
elements
greater than or
equal to 4 NPS
R-22
September 2004
IV.C1.3-c Piping, piping Stainless
IV.C1.1-f components, steel
and piping
elements
greater than or
equal to 4 NPS
Reactor coolant Cracking/ stress
corrosion cracking
and intergranular
stress corrosion
cracking
Reactor coolant Cracking/ stress
corrosion cracking
and intergranular
stress corrosion
cracking
Aging Management Program
(AMP)
Further
Evaluation
Fatigue is a time-limited aging
Yes,
analysis (TLAA) to be performed for TLAA
the period of extended operation,
and, for Class 1 components,
environmental effects on fatigue are
to be addressed. See the Standard
Review Plan, Section 4.3 “Metal
Fatigue,” for acceptable methods
for meeting the requirements of 10
CFR 54.21(c)(1)(i) and (ii), and for
addressing environmental effects
on fatigue.
See Chapter X.M1 of this report for
meeting the requirements of
10 CFR 54.21(c)(1)(iii).
Chapter XI.M7, “BWR Stress
No
Corrosion Cracking” and
Chapter XI.M2, “Water Chemistry,”
for BWR water in BWRVIP-29
(EPRI
TR-103515)
Chapter XI.M7, “BWR Stress
No
Corrosion Cracking” and
Chapter XI.M2, “Water Chemistry,”
for BWR water in BWRVIP-29
(EPRI
TR-103515)
September 2004
IV
C1
Item
R-20
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Coolant Pressure Boundary (BWR)
Structure
and/or
Material
Component
IV.C1.1-f Piping, piping Stainless
IV.C1.3-c components, steel, cast
IV.C1.2-b and piping
austenitic
elements
stainless
greater than or steel
equal to 4 NPS
Link
Environment
Aging Effect/
Mechanism
Reactor coolant Cracking/ stress
corrosion cracking
and intergranular
stress corrosion
cracking
Aging Management Program
(AMP)
Further
Evaluation
Chapter XI.M7, “BWR Stress
Corrosion Cracking” and
No
Chapter XI.M2, “Water Chemistry,”
for BWR water in BWRVIP-29
(EPRI
TR-103515)
Chapter XI.M18, “Bolting Integrity” No
IV C1-9
Proposed Draft NUREG-1801
R-27
IV.C1.3-f Pump and
IV.C1.2-e valve closure
bolting
HighSystem
Loss of preload/
strength low- temperature up stress relaxation
alloy steel SA to 288°C (550°F)
193 Gr. B7
R-28
IV.C1.2-f Pump and
IV.C1.3-g valve closure
bolting
Steel
System
Cumulative fatigue Fatigue is a time-limited aging
Yes,
temperature up damage/ fatigue
analysis (TLAA) to be performed for TLAA
to 288°C (550°F)
the period of extended operation;
check 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 54.21(c).
R-26
IV.C1.2-d Pump and
IV.C1.3-e valve closure
bolting
Steel
System
Loss of material/
temperature up wear
to 288°C (550°F)
Chapter XI.M18, “Bolting Integrity” No
R-29
IV.C1.3-e Pump and
Stainless
IV.C1.2-d valve seal
steel, Steel
flange closure
bolting
Air with metal
Loss of material/
temperature up wear
to 288°C (550°F)
Chapter XI.M18, “Bolting Integrity” No
Proposed Draft NUREG-1801
IV C1-10
This Page Intentionally Left Blank
September 2004
C2. REACTOR COOLANT SYSTEM AND CONNECTED LINES
(PRESSURIZED WATER REACTOR)
C2.1
Reactor Coolant System Piping and Fittings
C2.1.1
C2.1.2
C2.1.3
C2.1.4
C2.1.5
C2.2
Connected Systems Piping and Fittings
C2.2.1
C2.2.2
C2.2.3
C2.2.4
C2.2.5
C2.2.6
C2.2.7
C2.2.8
C2.3
Casing
Cover
Closure Bolting
Valves (Check, Control, Hand, Motor-Operated, Relief, and Containment Isolation)
C2.4.1
C2.4.2
C2.4.3
C2.5
Residual Heat Removal (RHR) or Low Pressure Injection System
(Decay Heat Removal [DHR]/ Shutdown System)
Core Flood System (CFS)
High Pressure Injection System (Makeup & Letdown Functions)
Chemical and Volume Control System
Sampling System
Drains and Instrument Lines
Nozzles and Safe Ends
Small-Bore Piping, Fittings, and Branch Connections Less than NPS 4 in
Connected Systems
Reactor Coolant Pump
C2.3.1
C2.3.2
C2.3.3
C2.4
Cold Leg
Hot Leg
Surge Line
Spray Line
Small-Bore RCS Piping, Fittings, and Branch Connections
Less than NPS 4
Body
Bonnet
Closure Bolting
Pressurizer
C2.5.1
C2.5.2
C2.5.3
C2.5.4
C2.5.5
C2.5.6
C2.5.7
C2.5.8
C2.5.9
C2.5.10
C2.5.11
C2.5.12
September 2004
Shell/Heads
Spray Line Nozzle
Surge Line Nozzle
Spray Head
Thermal Sleeves
Instrument Penetrations
Safe Ends
Manway and Flanges
Manway and Flange Bolting
Heater Sheaths and Sleeves
Support Keys, Skirt, and Shear Lugs
Integral Support
IV C2-1
Proposed Draft NUREG-1801
C2.6
Pressurizer Relief Tank
C2.6.1
C2.6.2
Tank Shell and Heads
Flanges and Nozzles
Proposed Draft NUREG-1801
IV C2-2
September 2004
C2. REACTOR COOLANT SYSTEM AND CONNECTED LINES (PRESSURIZED WATER
REACTOR)
Systems, Structures, and Components
This section comprises 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 (RHR)
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 spray (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.
Therefore, they are not subject to an aging management review, pursuant to 10 CFR
54.21(a)(1).
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 2004
IV C2-3
Proposed Draft NUREG-1801
Proposed Draft NUREG-1801
IV
C2
Item
R-02
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Coolant System and Connected Lines (PWR)
Structure
and/or
Material
Environment
Component
IV.C2.1-g Class 1 piping, Stainless steel Reactor coolant
IV.C2.2-h fittings and
branch
connections <
NPS 4
Link
Aging Effect/
Mechanism
Cracking/ stress
corrosion cracking
Aging Management Program (AMP)
Chapter XI.M1, “ASME Section XI
Inservice Inspection, Subsections IWB,
IWC, and IWD,” for Class 1
components and
Further
Evaluation
Yes,
parameters
monitored/
inspected and
detection of
Chapter XI.M2, “Water Chemistry,” for aging effects
PWR primary water in EPRI TRare to be
105714
evaluated
IV C2-4
Inspection in accordance with ASME
Section XI does not require volumetric
examination of pipes less than NPS 4.
A plant-specific destructive
examination or a nondestructive
examination (NDE) that permits
inspection of the inside surfaces of the
piping is to be conducted to ensure that
cracking has not occurred and the
component intended function will be
maintained during the extended period
of operation.
September 2004
September 2004
IV
C2
Item
R-57
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Coolant System and Connected Lines (PWR)
Structure
and/or
Component
IV.C2.2-h Class 1 piping,
IV.C2.1-g fittings and
branch
connections <
NPS 4
Link
Material
Environment
Stainless
Reactor coolant
steel/ steel
with stainless
steel cladding
IV C2-5
Proposed Draft NUREG-1801
R-07
IV.C2.2-f
IV.C2.5-h
IV.C2.5m
IV.C2.5-r
Class 1 piping,
fittings and
primary
nozzles, safe
ends,
manways, and
flanges
Stainless
Reactor coolant
steel, steel
with stainless
steel or nickelalloy cladding,
nickel-alloy
Aging Effect/
Mechanism
Cracking/ thermal
and mechanical
loading
Aging Management Program (AMP)
Further
Evaluation
Chapter XI.M1, “ASME Section XI
Inservice Inspection, Subsections IWB,
IWC, and IWD,” for Class 1
components
Yes,
parameters
monitored/
inspected and
detection of
Inspection in accordance with ASME aging effects
Section XI does not require volumetric are to be
examination of pipes less than NPS 4. evaluated
A plant-specific destructive
examination or a nondestructive
examination (NDE) that permits
inspection of the inside surfaces of the
piping is to be conducted to ensure that
cracking has not occurred and the
component intended function will be
maintained during the extended period
of operation.
The AMPs are to be augmented by
verifying that service-induced weld
cracking is not occurring in the smallbore piping less than NPS 4, including
pipe, fittings, and branch connections.
See Chapter XI.M32, “One-Time
Inspection” for an acceptable
verification method.
Cracking/ stress
Chapter XI.M1, “ASME Section XI
No
corrosion cracking, Inservice Inspection, Subsections IWB,
primary water stress IWC, and IWD,” for Class 1
corrosion cracking
components and
Chapter XI.M2, “Water Chemistry,” for
PWR primary water in EPRI TR105714
Proposed Draft NUREG-1801
IV
C2
Item
R-05
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Coolant System and Connected Lines (PWR)
Structure
and/or
Component
IV.C2.5-i Class 1 piping,
IV.C2.1-e piping
IV.C2.2-g components,
and piping
elements
Link
Material
Environment
Cast
Reactor coolant
austenitic
stainless steel
Aging Effect/
Mechanism
Cracking/ stress
corrosion cracking
Aging Management Program (AMP)
Further
Evaluation
Monitoring and control of primary water Yes, plant
chemistry in accordance with the
specific
guidelines in EPRI TR-105714 (Rev. 3
or later revisions or update) minimize
the potential of SCC, and material
selection according to the NUREG0313, Rev. 2 guidelines of ≤0.035% C
and ≥7.5% ferrite has reduced
susceptibility to SCC.
IV C2-6
For CASS components that do not
meet either one of the above
guidelines, a plant-specific aging
management program is to be
evaluated. The program is to include
(a) adequate inspection methods to
ensure detection of cracks, and (b) flaw
evaluation methodology for CASS
components that are susceptible to
thermal aging embrittlement.
R-52
IV.C2.2-e Class 1 piping, Cast
Reactor coolant Loss of fracture
IV.C2.5-l piping
austenitic
>250°C (>482°F) toughness/ thermal
IV.C2.1-f components, stainless steel
aging embrittlement
and piping
elements
Chapter XI.M12, “Thermal Aging
Embrittlement of Cast Austenitic
Stainless Steel (CASS)”
No
September 2004
September 2004
IV
C2
Item
R-09
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Coolant System and Connected Lines (PWR)
Structure
and/or
Component
IV.C2.3-b Class 1 pump
IV.C2.4-b casings and
valve bodies
Link
Material
Environment
CASS, carbon Reactor coolant
steel with
stainless steel
cladding
Aging Effect/
Mechanism
Cracking/ stress
corrosion cracking
Aging Management Program (AMP)
Further
Evaluation
Monitoring and control of primary water No
chemistry in accordance with the
guidelines in EPRI TR-105714 (Rev. 3
or later revisions or update) minimize
the potential of SCC, and material
selection according to the NUREG0313, Rev. 2 guidelines of ≤0.035% C
and ≥7.5% ferrite has reduced
susceptibility to SCC.
IV C2-7
For CASS components that do not
meet either one of the above
guidelines, see Chapter XI.M1, “ASME
Section XI, Subsections IWB, IWC, and
IWD.”
Proposed Draft NUREG-1801
Proposed Draft NUREG-1801
IV
C2
Item
R-08
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Coolant System and Connected Lines (PWR)
Structure
and/or
Component
IV.C2.4-c Class 1 pump
IV.C2.3-c casings, and
valve bodies
and bonnets
Link
Material
Environment
Aging Effect/
Mechanism
Cast
Reactor coolant Loss of fracture
austenitic
>250°C (>482°F) toughness/ thermal
stainless steel
aging embrittlement
Aging Management Program (AMP)
Further
Evaluation
Chapter XI.M1, “ASME Section XI
No
Inservice Inspection, Subsections IWB,
IWC, and IWD,” for Class 1
components
IV C2-8
For pump casings and valve bodies,
screening for susceptibility to thermal
aging is not required. The ASME
Section XI inspection requirements are
sufficient for managing the effects of
loss of fracture toughness due to
thermal aging embrittlement 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 embrittlement of
CASS pump casings.
September 2004
R-11
IV.C2.3-e Closure bolting High-strength Air with reactor
IV.C2.5-n
low-alloy
coolant leakage
IV.C2.4-e
steel,
stainless steel
Cracking/ stress
corrosion cracking
Chapter XI.M18, “Bolting Integrity”
No
R-12
IV.C2.5-p Closure bolting High-strength Air with reactor
IV.C2.3-g
low-alloy
coolant leakage
IV.C2.4-g
steel,
stainless steel
Loss of preload/
stress relaxation
Chapter XI.M18, “Bolting Integrity”
No
September 2004
IV
C2
Item
R-17
R-18
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Coolant System and Connected Lines (PWR)
Structure
and/or
Component
IV.C2.1-d External
IV.C2.5-b surfaces
IV.C2.2-d
IV.C2.6-b
IV.C2.5-u
IV.C2.5-o
IV.C2.3-f
IV.C2.4-f
Link
IV C2-9
IV.C2.3-d Piping and
IV.C2.4-d components
IV.C2.5-w external
IV.C2.5-t surfaces and
bolting
Material
Environment
Aging Effect/
Mechanism
Steel
Air with borated
water leakage
Loss of material/
boric acid corrosion
Stainless
steel, Steel
System
Cumulative fatigue
temperature up to damage/ fatigue
340°C (644°F)
Aging Management Program (AMP)
Further
Evaluation
Chapter XI.M10, “Boric Acid Corrosion” No
Fatigue is a time-limited aging analysis Yes,
(TLAA) to be performed for the period TLAA
of extended operation. See the
Standard Review Plan, Section 4.3
“Metal Fatigue,” for acceptable
methods for meeting the requirements
of 10 CFR 54.21(c)(1)(i) and (ii).
See Chapter X.M1 of this report for
meeting the requirements of
10 CFR 54.21(c)(1)(iii).
Proposed Draft NUREG-1801
Proposed Draft NUREG-1801
IV
C2
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Coolant System and Connected Lines (PWR)
Item
Link
R-04
IV.C2.2-b
IV.C2.5-d
IV.C2.5-a
IV.C2.4-a
IV.C2.3-a
IV.C2.2-c
IV.C2.1-b
IV.C2.1-a
IV.C2.2-a
IV.C2.5-f
IV.C2.5-e
IV.C2.5-q
Structure
and/or
Component
Piping, piping
components,
and piping
elements
Material
Environment
Steel,
Reactor coolant
stainless
steel, cast
austenitic
stainless
steel, carbon
steel with
nickel-alloy or
stainless steel
cladding,
nickel-alloy
Aging Effect/
Mechanism
Cumulative fatigue
damage/ fatigue
Aging Management Program (AMP)
Further
Evaluation
Fatigue is a time-limited aging analysis Yes,
(TLAA) to be performed for the period TLAA
of extended operation, and, for Class 1
components, environmental effects on
fatigue are to be addressed. See the
Standard Review Plan, Section 4.3
“Metal Fatigue,” for acceptable
methods for meeting the requirements
of 10 CFR 54.21(c)(1)(i) and (ii), and
for addressing environmental effects
on fatigue.
IV C2-10
See Chapter X.M1 of this report for
meeting the requirements of
10 CFR 54.21(c)(1)(iii).
R-19
Stainless
IV.C2.5- Pressurizer
Integral support steel, Steel
v
R-24
IV.C2.5-j Pressurizer
Spray head
Air with metal
Cracking/ cyclic
temperature up to loading
288°C (550°F)
Nickel alloy, Reactor coolant
cast austenitic
stainless
steel,
stainless steel
Cracking/ primary
water stress
corrosion cracking
Chapter XI.M1, “ASME Section XI
No
Inservice Inspection, Subsections IWB,
IWC, and IWD,” for Class 1
components
A plant-specific aging management
program is to be evaluated.
Yes, plant
specific
September 2004
September 2004
IV
C2
Item
R-58
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Coolant System and Connected Lines (PWR)
Structure
and/or
Component
IV.C2.5-c Pressurizer
IV.C2.5-g components
Link
Material
Environment
Steel with
Reactor coolant
stainless steel
or nickel alloy
cladding; or
stainless steel
Aging Effect/
Mechanism
Cracking/ cyclic
loading
Aging Management Program (AMP)
Further
Evaluation
Chapter XI.M1, “ASME Section XI
No
Inservice Inspection, Subsections IWB,
IWC, and IWD,” for Class 1
components and
Chapter XI.M2, “Water Chemistry,” for
PWR primary water in EPRI TR105714
IV C2-11
R-25
IV.C2.5-g Pressurizer
IV.C2.5-c components
Proposed Draft NUREG-1801
Steel with
Reactor coolant
stainless steel
or nickel alloy
cladding; or
stainless steel
Cracking/ stress
corrosion cracking
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 ISI program, the existing AMP is
adequate for managing the effect of
pressurizer clad cracking.
Chapter XI.M1, “ASME Section XI
No
Inservice Inspection, Subsections IWB,
IWC, and IWD,” for Class 1
components and
Chapter XI.M2, “Water Chemistry,” for
PWR primary water in EPRI TR105714
Proposed Draft NUREG-1801
IV
C2
Item
R-06
IV C2-12
R-14
R-13
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Coolant System and Connected Lines (PWR)
Structure
and/or
Material
Component
IV.C2.5-k Pressurizer
Nickel alloy
IV.C2.5-s instrumentation
penetrations
and heater
sheaths and
sleeves
Link
Environment
Aging Effect/
Mechanism
Aging Management Program (AMP)
Further
Evaluation
Reactor coolant
Cracking/ primary
water stress
corrosion cracking
Chapter XI.M1, “ASME Section XI
Yes, plant
Inservice Inspection, Subsections IWB, specific
IWC, and IWD,” for Class 1
components and Chapter XI.M2,
“Water Chemistry,” for PWR primary
water in EPRI TR-105714 and plant
specific AMP consistent with applicant
commitments to NRC Bulletin BL-04-01
or any subsequent regulatory
requirements.
Stainless
Treated borated
steel/ steel
water >60°C
Tank shell and with stainless (>140°F)
heads
steel cladding
Flanges and
nozzles
Cracking/ stress
corrosion cracking
Chapter XI.M1, “ASME Section XI
No
Inservice Inspection, Subsections IWB,
IWC, and IWD,” for Class 2
components and
IV.C2.6- Pressurizer
relief tank
c
IV.C2.6- Pressurizer
relief tank
a
Steel with
Treated borated
stainless steel water
Tank shell and cladding
heads
Flanges and
nozzles
Same as above
Cumulative fatigue
damage/ fatigue
Chapter XI.M2, “Water Chemistry,” for
PWR primary water in EPRI TR105714
Fatigue is a time-limited aging analysis Yes,
(TLAA) to be performed for the period TLAA
of extended operation. See the
Standard Review Plan, Section 4.3
“Metal Fatigue,” for acceptable
methods for meeting the requirements
of
10 CFR 54.21(c)(1)(i) and (ii).
September 2004
See Chapter X.M1 of this report, for
meeting the requirements of
10 CFR 54.21(c)(1)(iii).
September 2004
IV
C2
Item
R-56
R-30
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Reactor Coolant System and Connected Lines (PWR)
Structure
and/or
Component
IV.C2.1- Reactor
coolant system
c
piping and
fittings
Cold leg
Hot leg
Surge line
Spray line
Link
Material
Stainless
Reactor coolant
steel/ steel
with stainless
steel cladding
Stainless
IV.C2.1- Reactor
coolant
system
steel/ steel
c
IV C2-13
piping and
fittings
Cold leg
Hot leg
Surge line
Spray line
Environment
with stainless
steel cladding
Reactor coolant
Aging Effect/
Mechanism
Aging Management Program (AMP)
Further
Evaluation
Cracking/ cyclic
loading
Chapter XI.M1, “ASME Section XI
No
Inservice Inspection, Subsections IWB,
IWC, and IWD,” for Class 1
components
Cracking/ stress
corrosion cracking
Chapter XI.M1, “ASME Section XI
No
Inservice Inspection, Subsections IWB,
IWC, and IWD,” for Class 1
components and
Chapter XI.M2, “Water Chemistry,” for
PWR primary water in EPRI TR105714
Proposed Draft NUREG-1801
This Page Intentionally Left Blank
Proposed Draft NUREG-1801
IV C2-14
September 2004
D1. STEAM GENERATOR (RECIRCULATING)
D1.1
Pressure Boundary and Structural
D1.1.1
D1.1.2
D1.1.3
D1.1.4
D1.1.5
D1.1.6
D1.1.7
D1.1.8
D1.1.9
D1.1.10
D1.1.11
D1.2
Tube Bundle
D1.2.1
D1.2.2
D1.2.3
D1.2.4
D1.3
Top Head
Steam Nozzle and Safe End
Upper and Lower Shell
Transition Cone
Feedwater Nozzle and Safe End
Feedwater Impingement Plate and Support
Secondary Manway and Handhole Bolting
Lower Head
Primary Nozzles and Safe Ends
Instrument Nozzles
Primary Manway (Cover and Bolting)
Tubes and Sleeves
Tube Support Lattice Bars (Combustion Engineering)
Tube Plugs
Tube Support Plates
Upper Assembly and Separators
D1.3.1
September 2004
Feedwater Inlet Ring and Support
IV D1-1
Proposed Draft NUREG-1801
D1. STEAM GENERATOR (RECIRCULATING)
Systems, Structures, and Components
This section consists of 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.
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 (VIII.D1), the steam generator blowdown system (VIII.F), and
the auxiliary feedwater system (VIII.G).
Proposed Draft NUREG-1801
IV D1-2
September 2004
September 2004
IV
D1
Item
R-07
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Steam Generator (Recirculating)
Structure
and/or
Component
IV.D1.1-i Class 1 piping,
fittings and
primary
nozzles, safe
ends,
manways, and
flanges
Link
Material
Environment
Stainless
Reactor coolant
steel, steel
with stainless
steel or
nickel-alloy
cladding,
nickel-alloy
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
Cracking/ stress
corrosion cracking,
primary water
stress corrosion
cracking
Chapter XI.M1, “ASME Section XI
Inservice Inspection, Subsections
IWB, IWC, and IWD,” for Class 1
components and
No
Chapter XI.M2, “Water Chemistry,”
for PWR primary water in EPRI TR105714
Chapter XI.M18, “Bolting Integrity” No
IV D1-3
R-10
IV.D1.1-l Closure bolting Steel
Air with reactor
coolant leakage
Cracking/ stress
corrosion cracking
R-17
IV.D1.1-g External
IV.D1.1-k surfaces
Steel
Air with borated
water leakage
Loss of material/
Chapter XI.M10, “Boric Acid
boric acid corrosion Corrosion”
R-01
IV.D1.1-j Instrument
Nickel alloy
Reactor coolant
Cracking/ primary
water stress
corrosion cracking
penetrations
and primary
side nozzles
No
Proposed Draft NUREG-1801
Chapter XI.M1, “ASME Section XI Yes, plant
Inservice Inspection, Subsections specific
IWB, IWC, and IWD,” for Class 1
components and Chapter XI.M2,
“Water Chemistry,” for PWR
primary water in EPRI TR-105714
and plant specific AMP consistent
with applicant commitments to
NRC Orders, Bulletins and Generic
Letters associated with nickel
alloys.
Proposed Draft NUREG-1801
IV
D1
Item
R-04
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Steam Generator (Recirculating)
Structure
and/or
Component
IV.D1.1- Piping, piping
components,
h
and piping
elements
Link
Material
Steel,
Reactor coolant
stainless
steel, cast
austenitic
stainless
steel, carbon
steel with
nickel-alloy or
stainless
steel
cladding,
nickel-alloy
IV D1-4
R-37
Steel
IV.D1.1- Pressure
boundary and
d
structural
Steam nozzle
and safe end
FW nozzle and
safe end
R-32
September 2004
IV.D1.1- Steam
generator
f
closure bolting
Environment
Steel
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
Cumulative fatigue Fatigue is a time-limited aging
Yes,
damage/ fatigue
analysis (TLAA) to be performed
TLAA
for the period of extended
operation, and, for Class 1
components, environmental effects
on fatigue are to be addressed.
See the Standard Review Plan,
Section 4.3 “Metal Fatigue,” for
acceptable methods for meeting
the requirements of 10 CFR
54.21(c)(1)(i) and (ii), and for
addressing environmental effects
on fatigue.
See Chapter X.M1 of this report for
meeting the requirements of
10 CFR 54.21(c)(1)(iii).
Secondary
Wall thinning/ flow- Chapter XI.M17, “Flow-Accelerated No
feedwater/steam accelerated
Corrosion”
corrosion
System
Loss of preload/
temperature up to stress relaxation
340°C (644°F)
Chapter XI.M18, “Bolting Integrity” No
September 2004
IV
D1
Item
R-33
R-39
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Steam Generator (Recirculating)
Structure
and/or
Component
IV.D1.1-a Steam
IV.D1.1-b generator
components
Link
IV.D1.1- Steam
generator
e
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
IV D1-5
Material
Environment
Steel
Secondary
Cumulative fatigue Fatigue is a time-limited aging
feedwater/steam damage/ fatigue
analysis (TLAA) to be evaluated for
the period of extended operation.
See the Standard Review Plan,
Section 4.3, “Metal Fatigue” for
acceptable methods for meeting
the requirements of 10 CFR
54.21(c).
Secondary
Loss of material/
A plant-specific aging management
feedwater
erosion
program is to be evaluated.
Yes,
TLAA
Secondary
Loss of material/
Chapter XI.M1, “ASME Section XI
feedwater/steam general, pitting and Inservice Inspection, Subsections
crevice corrosion
IWB, IWC, and IWD,” for Class 2
components and
Yes,
detection of
aging effects is
to be
evaluated
Steel
Yes, plant
specific
feedwater
impingement
plate and
support
R-34
Steel
IV.D1.1- Steam
generator shell
c
Proposed Draft NUREG-1801
assembly (for
OTSG), upper
and lower
shell, and
transition cone
(for
recirculating
steam
generator)
Chapter XI.M2, “Water Chemistry,”
for PWR secondary water in EPRI
TR-102134
As noted in NRC Information
Notice IN 90-04, general and pitting
corrosion of the shell exists, the
AMP guidelines in Chapter XI.M1
may not be sufficient to detect
general and pitting corrosion, and
additional inspection procedures
are to be developed, if required.
Proposed Draft NUREG-1801
IV
D1
Item
R-40
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Steam Generator (Recirculating)
Structure
and/or
Component
IV.D1.2-i Tube plugs
IV.D1.2-j
Link
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Cracking/ primary
water stress
corrosion cracking
Chapter XI.M19, “Steam Generator No
Tubing Integrity” and Chapter
XI.M2, “Water Chemistry,” for PWR
primary water in EPRI TR-105714
Material
Environment
Nickel alloy
Reactor coolant
Secondary
Loss of material/
feedwater/steam flow-accelerated
corrosion
IV.D1.2- Tube support
lattice bars
h
Steel
R-42
IV.D1.2- Tube support
plates
k
Steel
IV D1-6
R-41
Applicant must provide a
commitment to submit, for NRC
review and approval, an inspection
plan for tube support lattice bars as
based upon staff approved NEI 9706 guidelines, or other alternative
regulatory basis for steam
generator degradation
management, at least 24 months
prior to the extended period.
Secondary
Ligament cracking/ Chapter XI.M19, “Steam Generator
feedwater/steam corrosion
Tubing Integrity” and
Chapter XI.M2, “Water Chemistry,”
for PWR secondary water in EPRI
TR-102134
Further
Evaluation
No, but
licensee
commitment to
be confirmed.
No
September 2004
September 2004
IV
D1
Item
R-43
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Steam Generator (Recirculating)
Structure
and/or
Component
IV.D1.2- Tubes
Link
IV D1-7
IV.D1.2- Tubes and
sleeves
a
Aging Management Program
(AMP)
Further
Evaluation
Environment
Nickel alloy
Secondary
Denting/ corrosion Chapter XI.M19, “Steam Generator No
feedwater/steam of carbon steel tube Tubing Integrity” and
support plate
Chapter XI.M2, “Water Chemistry,”
for PWR secondary water in EPRI
TR-102134.
g
R-44
Aging Effect/
Mechanism
Material
Nickel alloy
Reactor coolant
Cracking/ primary
water stress
corrosion cracking
For plants where analyses were
completed in response to NRC
Bulletin 88-02 “Rapidly Propagating
Cracks in SG Tubes,” the results of
those analyses have to be
reconfirmed for the period of
license renewal.
Chapter XI.M19, “Steam Generator No
Tubing Integrity” and
Chapter XI.M2, “Water Chemistry,”
for PWR primary water in EPRI TR105714
.
Proposed Draft NUREG-1801
Proposed Draft NUREG-1801
IV
D1
Item
R-45
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Steam Generator (Recirculating)
Structure
and/or
Component
IV.D1.2- Tubes and
sleeves
d
Link
IV.D1.2- Tubes and
sleeves
c
Aging Management Program
(AMP)
Further
Evaluation
Environment
Nickel alloy
Reactor coolant Cumulative fatigue Fatigue is a time-limited aging
Yes,
and secondary
damage/ fatigue
analysis (TLAA) to be performed
TLAA
feedwater/steam
for the period of extended
operation, and, for Class 1
components, environmental effects
on fatigue are to be addressed.
See the Standard Review Plan,
Section 4.3 “Metal Fatigue,” for
acceptable methods for meeting
the requirements of
10 CFR 54.21(c)(1)(i) and (ii), and
for addressing environmental
effects on fatigue.
IV D1-8
R-48
Aging Effect/
Mechanism
Material
Nickel alloy
See Chapter X.M1 of this report for
meeting the requirements of
10 CFR 54.21(c)(1)(iii).
Secondary
Cracking/
Chapter XI.M19, “Steam Generator no
feedwater/steam intergranular attack Tubing Integrity” and
Chapter XI.M2, “Water Chemistry,”
for PWR secondary water in EPRI
TR-102134
R-47
IV.D1.2- Tubes and
sleeves
b
Nickel alloy
September 2004
Secondary
Cracking/ outer
feedwater/steam diameter stress
corrosion cracking
Chapter XI.M19, “Steam Generator no
Tubing Integrity” and
Chapter XI.M2, “Water Chemistry,”
for PWR secondary water in EPRI
TR-102134
September 2004
IV
D1
Item
R-49
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Steam Generator (Recirculating)
Structure
and/or
Component
IV.D1.2- Tubes and
sleeves
e
Link
Aging Effect/
Mechanism
Material
Environment
Nickel alloy
Secondary
Loss of material/
feedwater/steam fretting and wear
Aging Management Program
(AMP)
Further
Evaluation
Chapter XI.M19, “Steam Generator No
Tubing Integrity” and
Chapter XI.M2, “Water Chemistry,”
for PWR secondary water in EPRI
TR-102134
R-50
IV.D1.2- Tubes and
sleeves
f
Nickel alloy
IV D1-9
(exposed to
phosphate
chemistry)
R-51
Steel
IV.D1.3- Upper
assembly and
a
separators
Feedwater inlet
ring and
support
Secondary
Loss of material/
Chapter XI.M19, “Steam Generator No
feedwater/steam wastage and pitting Tubing Integrity” and
corrosion
Chapter XI.M2, “Water Chemistry,”
for PWR secondary water in EPRI
TR-102134
Secondary
Loss of material/
feedwater/steam flow-accelerated
corrosion
Proposed Draft NUREG-1801
A plant-specific aging management Yes, plant
program is to be evaluated. As
specific
noted in Combustion Engineering
(CE) Information Notice (IN) 90-04
and NRC IN 91-19 and LER 50362/90-05-01, this form of
degradation has been detected
only in certain CE System 80
steam generators.
D2. STEAM GENERATOR (ONCE–THROUGH)
D2.1
Pressure Boundary and Structural
D2.1.1
D2.1.2
D2.1.3
D2.1.4
D2.1.5
D2.1.6
D2.1.7
D2.1.8
D2.1.9
D2.1.10
D2.2
Upper and Lower Heads
Tube Sheets
Primary Nozzles
Shell Assembly
Feed Water and Auxiliary Feed Water Nozzles and Safe Ends
Steam Nozzles and Safe Ends
Primary Side Drain Nozzles
Secondary Side Nozzles (Vent, Drain, and Instrumentation)
Primary Manways (Cover and Bolting)
Secondary Manways and Handholes (Cover and Bolting)
Tube Bundle
D2.2.1
D2.2.2
September 2004
Tubes and Sleeves
Tube Plugs
IV D2-1
Proposed Draft NUREG-1801
D2. STEAM GENERATOR (ONCE-THROUGH)
Systems, Structures, and Components
This section consists of 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.
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 (VIII.D1), the
steam generator blowdown system (VIII.F), and the auxiliary feedwater system (VIII.G).
Proposed Draft NUREG-1801
IV D2-2
September 2004
September 2004
IV
D2
Item
R-17
R-01
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Steam Generator (Once-Through)
Structure
and/or
Component
IV.D2.1-j External
IV.D2.1-b surfaces
Material
Environment
Aging Effect/
Mechanism
Steel
Air with borated
water leakage
Loss of material/
Chapter XI.M10, “Boric Acid
boric acid corrosion Corrosion”
IV.D2.1- Instrument
penetrations
h
Nickel alloy
Reactor coolant
Cracking/ primary Chapter XI.M1, “ASME Section XI Yes, plant
water stress
Inservice Inspection, Subsections specific
corrosion cracking IWB, IWC, and IWD,” for Class 1
components and Chapter XI.M2,
“Water Chemistry,” for PWR
primary water in EPRI TR-105714
and plant specific AMP consistent
with applicant commitments to
NRC Orders, Bulletins and Generic
Letters associated with nickel
alloys.
IV.D2.1- Piping, piping
components,
c
Steel,
Reactor coolant
stainless
steel, cast
austenitic
stainless
steel, carbon
steel with
nickel-alloy
or stainless
steel
cladding,
nickel-alloy
Link
and primary
side nozzles
IV D2-3
R-04
and piping
elements
Aging Management Program
(AMP)
Further
Evaluation
No
Proposed Draft NUREG-1801
Cumulative fatigue Fatigue is a time-limited aging
Yes,
damage/ fatigue
analysis (TLAA) to be performed TLAA
for the period of extended
operation, and, for Class 1
components, environmental effects
on fatigue are to be addressed.
See the Standard Review Plan,
Section 4.3 “Metal Fatigue,” for
acceptable methods for meeting
the requirements of 10 CFR
54.21(c)(1)(i) and (ii), and for
addressing environmental effects
on fatigue.
See Chapter X.M1 of this report for
meeting the requirements of
10 CFR 54.21(c)(1)(iii).
Proposed Draft NUREG-1801
IV
D2
Item
R-38
R-31
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Steam Generator (Once-Through)
Structure
and/or
Material
Component
Steel
IV.D2.1- Pressure
boundary and
f
structural
FW and AFW
nozzles and
safe ends
Steam nozzles
and safe ends
Link
Aging Management Program
(AMP)
Further
Evaluation
Secondary
Wall thinning/ flow- Chapter XI.M17, “Flow-Accelerated No
Corrosion”
feedwater/steam accelerated
corrosion
IV D2-4
Steel
Air with leaking
secondary-side
water and/or
steam
Loss of material/
erosion
Chapter XI.M1, “ASME Section XI No
Inservice Inspection, Subsections
IWB, IWC, and IWD,” for Class 2
components
IV.D2.1- Steam
generator
k
Steel
System
Loss of preload/
temperature up to stress relaxation
340°C (644°F)
Chapter XI.M18, “Bolting Integrity” No
IV.D2.1-g Steam
IV.D2.1-d generator
components
Steel
Secondary
Cumulative fatigue Fatigue is a time-limited aging
Yes,
feedwater/steam damage/ fatigue
analysis (TLAA) to be evaluated for TLAA
the period of extended operation.
See the Standard Review Plan,
Section 4.3, “Metal Fatigue” for
acceptable methods for meeting
the requirements of 10 CFR
54.21(c).
closure bolting
R-33
Aging Effect/
Mechanism
IV.D2.1-l Secondary
manways and
handholes
(cover only)
R-32
Environment
September 2004
September 2004
IV
D2
Item
R-36
R-35
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Steam Generator (Once-Through)
Structure
and/or
Material
Component
Nickel alloy
IV.D2.1-i Steam
generator
components
Such as,
secondary side
nozzles (vent,
drain, and
instrumentation
)
Link
IV.D2.1- Steam
generator
a
IV D2-5
components
Upper and
lower heads
Tube sheets
Steel with
stainless
steel or
nickel-alloy
cladding
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
Secondary
Cracking/ stress
A plant-specific aging management Yes, plant
feedwater/steam corrosion cracking program is to be evaluated.
specific
Reactor coolant
Cracking/ stress
Chapter XI.M1, “ASME Section XI No
corrosion cracking Inservice Inspection, Subsections
IWB, IWC, and IWD,” for Class 1
components and
Chapter XI.M2, “Water Chemistry,”
for PWR primary water in EPRI
TR-105714
Proposed Draft NUREG-1801
Proposed Draft NUREG-1801
IV
D2
Item
R-34
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Steam Generator (Once-Through)
Structure
and/or
Material
Component
Steel
IV.D2.1- Steam
generator shell
e
assembly (for
OTSG), upper
and lower shell,
and transition
cone (for
recirculating
steam
generator)
Link
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Secondary
Loss of material/
Chapter XI.M1, “ASME Section XI
feedwater/steam general, pitting and Inservice Inspection, Subsections
crevice corrosion
IWB, IWC, and IWD,” for Class 2
components and
Further
Evaluation
Yes,
detection of
aging effects is
to be
evaluated
Chapter XI.M2, “Water Chemistry,”
for PWR secondary water in EPRI
TR-102134
IV D2-6
R-40
IV.D2.2-f Tube plugs
IV.D2.2-g
Nickel alloy
Reactor coolant
R-44
IV.D2.2- Tubes and
sleeves
a
Nickel alloy
Reactor coolant
As noted in NRC Information
Notice IN 90-04, general and
pitting corrosion of the shell exists,
the AMP guidelines in Chapter
XI.M1 may not be sufficient to
detect general and pitting
corrosion, and additional
inspection procedures are to be
developed, if required.
Cracking/ primary Chapter XI.M19, “Steam Generator No
water stress
Tubing Integrity” and Chapter
corrosion cracking XI.M2, “Water Chemistry,” for PWR
primary water in EPRI TR-105714
September 2004
Cracking/ primary Chapter XI.M19, “Steam Generator No
water stress
Tubing Integrity” and
corrosion cracking
Chapter XI.M2, “Water Chemistry,”
for PWR primary water in EPRI
TR-105714
.
September 2004
IV
D2
Item
R-46
R-48
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Steam Generator (Once-Through)
Structure
and/or
Component
IV.D2.2- Tubes and
sleeves
e
Link
IV.D2.2- Tubes and
sleeves
c
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
Material
Environment
Nickel alloy
Reactor coolant Cumulative fatigue Fatigue is a time-limited aging
Yes,
and secondary
damage/ fatigue
analysis (TLAA) to be performed TLAA
feedwater/steam
for the period of license renewal.
See the Standard Review Plan,
Section 4.3 “Metal Fatigue,” for
acceptable methods for meeting
the requirements of 10 CFR
54.21(c).
Secondary
Cracking/
Chapter XI.M19, “Steam Generator no
feedwater/steam intergranular attack Tubing Integrity” and
Nickel alloy
IV D2-7
Chapter XI.M2, “Water Chemistry,”
for PWR secondary water in EPRI
TR-102134
Proposed Draft NUREG-1801
R-47
IV.D2.2- Tubes and
sleeves
b
Nickel alloy
Secondary
Cracking/ outer
Chapter XI.M19, “Steam Generator no
feedwater/steam diameter stress
Tubing Integrity” and
corrosion cracking
Chapter XI.M2, “Water Chemistry,”
for PWR secondary water in EPRI
TR-102134
R-49
IV.D2.2- Tubes and
sleeves
d
Nickel alloy
Secondary
Loss of material/
feedwater/steam fretting and wear
Chapter XI.M19, “Steam Generator No
Tubing Integrity” and
Chapter XI.M2, “Water Chemistry,”
for PWR secondary water in EPRI
TR-102134
This Page Intentionally Left Blank
September 2004
IV D2-8
Proposed Draft NUREG-1801
F.
COMMON MISCELLANEOUS MATERIAL ENVIRONMENT COMBINATIONS
Systems, Structures, and Components
This section includes 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,
there are no aging effects which are expected to degrade the ability of the structure or
component from performing 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 PWRs and BWRs. (For example, see System Interfaces in V.A to V.D2 for details.)
September 2004
IV F1-1
Proposed Draft NUREG-1801
Proposed Draft NUREG-1801
IV
E
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Common Miscellaneous Material Environment Combinations
Structure
and/or
Component
Piping, piping
components,
and piping
elements
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
Air – indoor
uncontrolled
(External)
None
None
No
Nickel alloy
Air – indoor
uncontrolled
(External)
None
None
No
Piping, piping
components,
and piping
elements
Stainless
steel
Air – indoor
uncontrolled
(External)
None
None
No
RP-05
Piping, piping
components,
and piping
elements
Stainless
steel
Air with borated
water leakage
None
None
No
RP-06
RP-06
Piping, piping
components,
and piping
elements
Stainless
steel
Concrete
None
None
No
RP-07
RP-07
Piping, piping
components,
and piping
elements
Stainless
steel
Gas
None
None
No
RP-08
RP-08
Piping, piping
components,
and piping
elements
Stainless
steel
Treated borated None
water
None
No
Item
Link
Environment
RP-02
RP-02
Cast
austenitic
stainless
steel
RP-03
RP-03
Piping, piping
components,
and piping
elements
RP-04
RP-04
RP-05
V F1-2
Material
September 2004
September 2004
IV
E
REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM
Common Miscellaneous Material Environment Combinations
Item
Link
RP-01
RP-01
Structure
and/or
Component
Piping, piping
components,
and piping
elements
Material
Environment
Aging Effect/
Mechanism
Aging Management Program
(AMP)
Further
Evaluation
Steel
Concrete
None
None
No
V F-3
Proposed Draft NUREG-1801
This Page Intentionally Left Blank
Proposed Draft NUREG-1801
IV E-2
September 2004