,.~R REGlI~;. . . 0CEA GUIDE TUBE EXTENSIONS (b) ALIGNMENT PLATE (a) Schematic illustration of a portion of the fuel alignment plate, and (b) Radial-view ... Typical Westinghouse control

��

United States Nuclear Regulatory Commission Official Hearing Exhibit

In the Matter of: Entergy Nuclear Operations, Inc. (Indian Point Nuclear Generating Units 2 and 3)

ASLBP #: 07-858-03-LR-BD01 Docket #: 05000247 | 05000286 Exhibit #: Identified: Admitted: Withdrawn: Rejected: Stricken:

Other:

NYS00307C-00-BD01 10/15/201210/15/2012

NYS00307C Submitted: December 22, 2011

c:..\,.~""R REGlI~;.

l~'~ • 0 · . · .. Aging Management Requirements ~ ~

.... ,,1- 0 .... ? ~

****il

Guide Lug

P late

Top Plate

Ring

Brace

Rib

Bottom Plate

Figure 4-13 CE welded core shroud with full height panels

4-47

Aging Management Requirements

Inspect for Separation of Upper and Lower Shroud Sections

Figure 4-14 Locations of potential separation between core shroud sections caused by swelling induced warping of thick flange plates in CE welded core shroud assembled in stacked sections

4-48

IPEC_PA_OAG0000104

Aging Mw!u.gement Requiremenis

Inspect for Separation of Upper and Lower Shroud Sections

Figure 4-14 Locations of potential separation between core shroud sections caused by swelling induced warping of thick flange plates in CE welded core shroud assembled in stacked sections

4-48


Flange Weld

Axial Weld

....... ~~~:.,.,__._._.~U~p~p er Core Barrel to

Figure 4-15 Typical CE core support barrel structure

Lower Core Barrel Circumferential Weld

Lower Barrel Axial Weld

Lower Barrel Circumferential Weld


Core Barrel to Support Plate Weld

4-49

IPEC_PA_OAG0000105

Figure 4-15 Typical CE core support barrel structure

Agillg Mallagemellt Reqlljrement.~

Flange Weld

Axial Weld

Upp er C ore Barre l to Lower C ore Barre l Circumferential Weld


Lower Barrel Circumfe rent ial Weld


Core Barrel to Suppo rt Plate Weld


i I I I I I Core B arrel --- I

I

Core .Support

I Column

r I Core .Support Plate ._ .... -

-\ v~ \/~ / j~ ~)' v~ lT ~

~ ~ !I' ~ 1 .Support Beam I y ).j r' I"'"

fr~

I

Flange and r ...._ ..___r' u::Jd' 'k.. ~ ;go; :::.. Flexure

"Winged" Core Support Columns

Core B arrel

Core .Support Plate ---- - -I- - ---

Core Support Columns without "Winged" Design

Figure 4-16 CE lower support structures for welded core shrouds: separate core barrel and lower support structure assembly with lower flange and core support plate

4-50

Core .Support Column

.Support Beam

IPEC_PA_OAG0000106


~ I COI~ Barr~l

I

I I

COle Support. Pl ate ~ I

"- V' \ V" ?~" V', V' " 7

_I Flange and Fl~xuu

"Winged" Core Support Columns

Core Barrel I I

, Core SupportPlate

~ - - - - - - -

1 U1 r' ,

I Flange and I ,

Flexllre • ,

Core Support Columns without "Winged" Design

Figure 4-16 CE lower support structures for welded core shrouds: separate core barrel and lower support structure assembly with lower flange and core support plate

4-50

COle Support. Column

Support. Beam

COle Support. Column

Support Beam


CEA GUIDE TUBES

Figure 4-17

(a)

E

CEA GUIDE TUBE EXTENSIONS

(b)

ALIGNMENT PLATE

(a) Schematic illustration of a portion of the fuel alignment plate, and (b) Radial-view schematic illustration of the guide tubes protruding through the plate in upper internals assembly of CE core shrouds with full-height shroud plates

4-51

IPEC_PA_OAG0000107

Aging Management Reqlliremenl.~

UGS SUPPORT PLATE

s

CEA GUIDE TUBES

Figure 4-17

(a)

CEA GUIDE TUBE EXTENSIONS

(b)

c

FUEL ALIGNMENT

PLATE

D

(a) Schematic illustration of a portion of the fuel alignment plate, and (b) Radial-view schematic illustration of the guide tubes protruding through the plate in upper internals assembly of CE core shrouds with full-height shroud plates

4-5 1


_c :::L rLl F== ~[ ::J rL :::L. rL P- rL ::::L

L \

I

I I I I I I I I I I !':\,

E7U'//// / /, I / '////LLL.J.J / ~ ~ ~ v v

I

v / I I I :; I ~ I :; v I I I

/ ~ :; v :; I I I / ~ v v v v ~ / ~ ~ v v

- - lL r--

I I I I I I I I I I I



I-- r-- - -

f1

n~ f' 1\ f1 (\ {1 (\ r 1\ I' 1\

{1 I~ (\

I I I I - - - l - l t - -

\ I H ~ H w+' ~ H }{- r- \ J ~ 1S<S<I llil IS<l<l v I v

Figure 4-18 CE control element assembly (CEA) shroud instrument tubes (circled in red) are shown, along with the welded supports attaching them to the CEA shroud tube, in this schematic illustration

4-52

IPEC_PA_OAG0000108

Aging Mmwgel1lelll Req/l;remelllJ

-"p... -"p... If 1'--

cP cp cp I ICP ICP I

I I I I f\ eLL", I ~

I I I I I I I I I I I I

r-




l-

I A I I :r I 1 1 t

V V

Figure 4-18 CE control element assembly (CEA) shroud instrument tubes (circled in red) are shown, along with the welded supports attaching them to the CEA shroud tube, in this schematic illustration

4-52


Illustrates the deep beam grid structure (number 3), as well as the fuel alignment pins

(numbers 1 and 2)

Figure 4-19 Isometric view of the lower support structure in the CE core shrouds with full-height shroud plates units. Fuel rests on alignment pins

4-53

IPEC_PA_OAG0000109

Agil/g Mal/agemelll Requirements

Illustrates the deep beam grid strllcture (number 3). as well as the fuel alignmen t pins (numbers 1 and 2)

Figure 4-19 Isometric view of the lower support structure in the CE core shrouds with full-height shroud plates units. Fuel rests on alignment pins

4-53


Wear Area

Figure 4-20 Typical Westinghouse control rod guide card (17x17 fuel assembly)

4-54

IPEC_PA_OAG0000110

Aging Management Requiremellts

Wear Area

Figure 4-20 Typical Westinghouse control rod guide card (17x17 fuel assembly)

4-54


Lower Flange Weld

Figure 4-21 Typical Westinghouse control rod guide tube assembly

4-55

IPEC_PA_OAG0000111

Agillg Managemelll Reqllirelllent.~

...

Lower F1unge Weld

Figure 4-21 Typical Westinghouse control rod guide tube assembly

4-55


Flange Weld

Upper Core Barrel to Lower Core Barrel Circumferential Weld

Lower Barrel Circumferential Weld


Figure 4-22 Major fabrication welds in typical Westinghouse core barrel

4-56

Axial Weld



IPEC_PA_OAG0000112

Aging Management Requiremenis

Flange Weld

Upper Core Barrel to ~r Core Barrel Cf!'Cumlerential Weld

lower Barre l Cirtumferential Weld


Figure 4-22

o

o o o

Major fabrication welds in typical Westinghouse core barrel

4-56

Axial Weld

l ower Barrel Axial Weld

Lower Barrel A;i(ial Weld


HAFFI..E P1.A TE EDGE BOLT

Figure 4-23

BAFFLE TO FORMER BOLT (LONG&: SHORT)

CORE BARREL TO FOP.f&:RBOLT

COP.NE:R EDGE BRACXET BAFFLE TO FORMER BOLT

Bolt locations in typical Westinghouse baffle-former-barrel structure. In CE plants with bolted shrouds, the core shroud bolts are equivalent to baffle-former bolts and barrelshroud bolts are equivalent to barrel-former bolts

4-57

IPEC_PA_OAG0000113

HAFfl.E PI..A 11i EDGEBOtT

Figure 4-23

BAFFLE TO FORMER BOLT(l.ONGA: SHORl)


~-COI""'REl. "TO R>R.Mr;:R.BOLT

COANER EDGE BRACXET BAFFLE TO FORMER. BOLT

Bolt locations in typical Westinghouse baffle-former-barrel structure. In CE plants with bolted shrouds, the core shroud bolts are equivalent to baffle-former bolts and barrelshroud bolts are equivalent to barrel-former bolts


/-~

0

00000 0

0 0 0 0 0

0 0 0 0 0

0 0 0

0 0 00000

0 0 0 0 0 0 0

0 0

0 0 0

0 e:' 0 00000 0 ~ 0 «o" 0 0

0 0

~0 # 0 0 0

0 0

~<() 0 0

0 00000 0

0 0 0 0 0 0 0

0 0 0

0 0

0 00000 0

0 0 0 0

0 0 0 0 0

0 0

0 0 00000

0 0 0 0 0 0 0 0 0

0 0 0

0

Figure 4-24 Baffle-edge bolt and baffle-former bolt locations at high fluence seams in bolted baffleformer assembly (note: equivalent baffle-former bolt locations in bolted CE shroud designs are core shroud bolts)

4-58

IPEC_PA_OAG0000114


Figure 4-24

~/ d.ooooo ••

o

.00000 o

.00000 o

• • 0

~ 000000

o .,'

o

~ «0' i ..... · 0

~;f'0 <Qd:-I:-j 000000 IV.,.

o

i 0000 000

000000 o • i 0000000

000000 o 0

o • • • 000000

o 0 o

o

00000

o

o ••••• ! V o

wi en

"0 w 2l Q) 0

,§ '" '"

Baffle-edge bolt and baffle-former bolt locations at high fluence seams in bolted baffleformer assembly (note: equivalent baffle-former bolt locations in bolted CE shroud designs are core shroud bolts)

4-58


High Fluence Seams

Figure 4-25 High fluence seam locations in Westinghouse baffle-former assembly

4-59

IPEC_PA_OAG0000115

Agillg Mallagemelll Reqlljrement.~

High Flucncc Scams

Figure 4-25 High fluence seam locations in Westinghouse baffle-former assembly

4-59


Q = ..;:> =

= c. =

= -=- = =

= ""=" <:;> =

= = .:. =

Potential Gaps at Baffle-Former Plate Levels

Figure 4-26

.,:,.. = .::;:..

= -= ..,.

.,::,. = ..;-

=- =

-=- ..::. ~

= =

Potential Bowing Along High Fluence Seam

- =

-..::.

=

Exaggerated view of void swelling induced distortion in Westinghouse baffle-former assembly. This figure also applies to bolted CE shroud designs

4-60

IPEC_PA_OAG0000116


Potential Caps at Baffle-Former Plate Levels

Figure 4-26

~

~ - -> =

- - = =

= - = I-= = = =

= = = <

= = =

~ ~'

= = = =

= = * =

= = = = ~

= = * = =~ = =

= = = - - *

- = = = = -= -= = = -= =

= = - -= =

- =

t Potential Bowing Along High Fluence Seam

Exaggerated view of void swelling induced distortion in Westinghouse baffle-former assembly. This figure also applies to bolted CE shroud designs

4-60


Vertical

Figure 4-27 Vertical displacement of Westinghouse baffle plates caused by void swelling. This figure also applies to bolted CE shroud designs

4-61

IPEC_PA_OAG0000117


o

Vertical

Figure 4-27 Vertical displacement of Westinghouse baffle plates caused by void swelling. This figure also applies to bolted CE shroud designs

4-61


\

Figure 4-28

VESSEl HEAD

/ TOP SUPPORT PLA lE

CORE BARRElL

Schematic cross-sections of the Westinghouse hold-down springs

4-62

IPEC_PA_OAG0000118

Aging Maul/gement Requirements

Figure 4-28

VESSEL HEAD

VESSEL

/ TOP SUPPORT PLATE

Hold-down spring

CORE BARREL

Schematic cross-sections of the Westinghouse hold-down springs

4-62

... 0 ....

Thermal Shield

~

I Core Support

Figure 4-29 Location of Westinghouse thermal shield flexures


Thermal Shield Flexure

4-63

IPEC_PA_OAG0000119

o

o o

. .....

. . . . . . ..

Thermal Shield . .. . . .

Core Support

Figure 4-29 Location of Westinghouse thermal shield flexures

, f

Agillg Mal/agell/el1l Requirements

Core Bane]

Thennal Shield Flexure

4-63


I Core Barrel l = I = I - I

= I = = = Core Support

I

= = Column I ~ I I 7

/ ..... ~ :~~ U-+-L J-L _1---l 'i~~ 1--I I I I

[ J Support Beam I I Core Support Plate ? r ; f'77'7777<

I

Figure 4-30 CE lower support structure assembly for plants with integrated core barrel and lower support structure with a core support plate (this design does not contain a lower core barrel flange)

a) b) c)

a) Early support column design

b) "Winged" support column design and plants with second generation core support assemblies

c) Later su ort column desi n used in lants with second eneration core su ort assemblies

Figure 4-31 CE core support columns

4-64

IPEC_PA_OAG0000120

Aging Management ReqlliremelllJ

I CO" B,nd LJ ~,

Core Support

, - Column

V i ! ~ :...-i! ,

I Core Support Plate li .~ Support Beam I r- ,- ;<.- i.--\ .... =--

Figure 4-30 CE lower support structure assembly for plants with integrated core barrel and lower support structure with a core support plate (this design does not contain a lower core barrel flange)

esp

a) b) c)

a) Early support colu mn deslgn b) "Winged" support colu mn design and plants with second generalion core support assemblies c) Later su art coili mn desi n used in lanls with second eneration core su ort assembli es

Figure 4-31 CE core support columns

4-64


Figure 4-32

Core Support Plate (Forging)

Core Plate

Lower Core Support Structure

Schematic indicating location of Westinghouse lower core support structure. Additional details shown in Figure 4-33

Figure 4-33

CORE~ SUPPORT COLUMN

LOWER CORE PLATE

Dl FFUSER PLATE

CORE SUPPORT PLATE/FORGING

\_BOTTOM MOUNTED INSTRUMENTATION COLUMN

Westinghouse lower core support structure and bottom mounted instrumentation columns. Core support column bolts fasten the core support columns to the lower core plate

4-65

IPEC_PA_OAG0000121

Agillg Mal/agemel1l Requirements

Figure 4-32

Core Suppcrt Plate (Forging)

Lm."c",~ Plate

LO\IIIt'tr Core i!W Su,.port SIN""'"

Schematic indicating location of Westinghouse lower core support structure. Additional details shown in Figure 4-33

Figure 4-33

CORE ':, SUPPORT COLUMN

LOWER CORE PLATE

DIFFUSER PLATE

CORE SUPPORT PLATE/FORGING

~ BOTTOM MOUNTED INSTRUMENTATION COLUMN

Westinghouse lower core support structure and bottom mounted instrumentation columns. Core support column bolts fasten the core support columns to the lower core plate

4-65


Figure 4-34 Typical Westinghouse core support column. Core support column bolts fasten the top of the support column to the lower core plate

Figure 4-35 Examples of Westinghouse bottom mounted instrumentation column designs

4-66

IPEC_PA_OAG0000122


Figure 4-34 Typical Westinghouse core support column. Core support column bolts fasten the top of the support column to the lower core plate

Figure 4-35 Examples of Westinghouse bottom mounted instrumentation column designs

4-66


Figure 4-36 Typical Westinghouse thermal shield flexure

4.4 Existing Programs Component Requirements

Existing Programs components are those PWR internals for which current aging management activities required to maintain functionality are being implemented. The continuation of these activities is credited within these guidelines for adequate aging management for specific components.

Included in the Existing Programs are PWR internals that are classified as removable core support structures. ASME Section XI, IWB-2500, Examination Category B-N-3 [2] does not list component specific examination requirements for removable core support structures. Accordingly, factors such as original design, licensing and code of construction variability could result in significant differences in an individual plant's current B-N-3 requirements. These guidelines credit specific components contained within the general B-N-3 classification for maintaining functionality.

These examination requirements, as applied to the components designated in Tables 4-7, 4-8, and 4-9, have been determined to provide sufficient aging management for these components.

Table 4-7 B&W plants Existing Programs components

No existing generic industry programs were considered sufficient for monitoring the aging effects addressed by these guidelines forB& W plants. Therefore, no components forB& W plants were placed into the Existing Programs group.

4-67

IPEC_PA_OAG0000123

Agillg Mal/agell/el1l Requirements

Figure 4-36 Typical Westinghouse thermal shield flexure

4.4 Existing Programs Component Requirements

Ex isting Programs components are those PW"R inte rnals for which current aging management acti vities required 1,0 maintain functionality are be ing implemented. The continuatio n of these activities is credited within these guidelines for adequate aging management for specific components .

Included in the Existing Programs are PWR inlernals th ai are class i.fied as removab le core SUppOlt structW'es. ASME Section Xl, lWB-2500, Examination Category 8-N-3 [2] does not list component specific examination requirements for removable core support structures. Accordingly, fac tors such as orig inal design, licensing and code of construction variability could result in significant differences in an individual plant' s current 8 -N-3 requirements. These guidelines credit specifi c components contained within the general 8-N-3 class ification for maintaining functionality .

These examination requirements, as applied to !.he components designated in Tables 4-7, 4-8, and 4-9, have been determined to provide sufficient aging management for these components.

Table 4-7 B&W plants Existing Programs components

No existing generic industry programs were considered sufficienl, for monitoring the aging effects addressed by these guidelines for B&W plants. Therefore, no components for B&W plants were placed inlo the Existing Programs group.

4-67


Table 4-8 CE plants Existing Programs components

Item Applicability Effect

Primary Link Examination Method Examination Coverage (Mechanism)

Core Shroud Assembly All plants Loss of material ASME Code Visual (VT-3) First 1 0-year lSI after 40 (Wear) Section XI examination, general years of operation, and at

Guide lugs condition examination each subsequent

Guide lug inserts and for detection of inspection interval.

bolts excessive or asymmetrical wear.

Lower Support All plants with Cracking (SCC, ASME Code Visual (VT-3) Accessible surfaces at Structure core shrouds IASCC, Fatigue) Section XI examination to detect specified frequency.

assembled with severed fuel alignment Fuel alignment pins full-height pins, missing locking

shroud plates tabs, or excessive wear on the fuel alignment pin nose or flange.

Lower Support All plants with Loss of material ASME Code Visual (VT-3) Accessible surfaces at Structure core shrouds (Wear) Section XI examination. specified frequency.

assembled in Fuel alignment pins two vertical

sections

Core Barrel Assembly All plants Loss of material ASME Code Visual (VT-3) Area of the upper flange (Wear) Section XI examination. potentially susceptible to

Upper flange wear.

4-68

IPEC_PA_OAG0000124


Table 4-8 CE plants Existing Programs components


Primary link Examination Method Examination Coverage (Mechanism)

Core Shroud Assembly All plants Loss of material ASME Code Visual (VT-3) First 10-year lSI after 40 (Wear) Section XI examination, general years of operation, and at

Guide lugs condition examination each subsequent

Guide lug inserts and for detection of inspection interval.

bolts excessive or asymmetrical wear.

lower Support All plants with Cracking (SCC, ASME Code Visual (VT-3) Accessible surfaces at Structure core shrouds IASCC, Fatigue) Section XI examination to detect specified frequency.

assembled with severed fuel alignment Fuel alignment pins full-height pins, missing locking

shroud plates tabs, or excessive wear on the fuel alignment pin nose or flange.

lower Support All plants with Loss of material ASME Code Visual (VT-3) Accessible surfaces at Structure core shrouds (Wear) Section XI examination. specified frequency.

assembled in Fuel alignment pins two vertical

sections

Core Barrel Assembly All plants Loss of material ASME Code Visual (VT-3) Area of the upper flange (Wear) Section XI examination. potentially susceptible to

Upper flange wear.


Table 4-9 Westinghouse plants Existing Programs components

Item Applicability Effect Primary Link Examination Method Examination Coverage (Mechanism)

Core Barrel Assembly All plants Loss of material ASME Code Visual (VT-3) All accessible surfaces at Core barrel flange (Wear) Section XI examination to specified frequency.

determine general condition for excessive wear.

Upper Internals Assembly All plants Cracking (SCC, ASME Code Visual (VT-3) All accessible surfaces at Upper support ring or skirt Fatigue) Section XI examination. specified frequency. Lower Internals Assembly All plants Cracking (IASCC, ASME Code Visual (VT-3) All accessible surfaces at Lower core plate Fatigue) Section XI examination of the lower specified frequency. XL lower core plate (Note 1) core plates to detect

evidence of distortion and/or loss of bolt integrity.

Lower Internals Assembly All plants Loss of material ASME Code Visual (VT-3) All accessible surfaces at Lower core plate (Wear) Section XI examination. specified frequency. XL lower core plate (Note 1) Bottom Mounted All plants Loss of material NUREG-1801 Surface (ET) Eddy current surface Instrumentation System (Wear) Rev. 1 examination. examination as defined in Flux thimble tubes plant response to IEB 88-

09. Alignment and Interfacing All plants Loss of material ASME Code Visual (VT-3) All accessible surfaces at Components (Wear) Section XI examination. specified frequency. Clevis insert bolts

(Note 2) Alignment and Interfacing All plants Loss of material ASME Code Visual (VT-3) All accessible surfaces at Components (Wear) Section XI examination. specified frequency. Upper core plate alignment pins

Notes:

1. XL= "Extra Long" referring to Westinghouse plants with 14-foot cores.

2. Bolt was screened in because of stress relaxation and associated cracking; however, wear of the clevis/insert is the issue.

4-69

IPEC_PA_OAG0000125

Agillg Mallagemellt Requiremellts

Table 4-9 Westinghouse plants Existing Programs components


Primary Link Examination Method Examination Coverage (Mechanism)

Core Barrel Assembly All plants Loss of material ASME Code Visual (VT 3) All accessible surtaces at Core barrel flange (Wear) Section XI examination to specified frequency.

determine general condition for excessive wear.

Upper Internals Assembly All plants Cracking (SCC, ASME Code Visual (VT-3) All accessible surtaces at Upper support ring or skirt Fatiguef Section XI examination. specified frequency. Lower Internals Assembly All plants Cracking (IASCC, ASME Code Visual (VT-3) All accessible surtaces at Lower core plate Fatigue) Section XI examination of the lower specified frequency. XL lower core plate (Note 1) core plates to detect

evidence of distortion and/or loss of bolt inteqritv.

Lower Internals Assembly All plants Loss of material ASME Code Visual (VT-3) All accessible surtaces at Lower core plate (Wear) Section XI examination. specified frequency. XL lower core plate (Note 1) Bottom Mounted All plants Loss of material NUREG· 1801 Surtace (ET) Eddy current surtace Instrumentation System (Wear) Rev. 1 examination. examination as defined in Flux thimble tubes plant response to IEB 88-

09. Alignment and Interfacing All plants Loss of material ASME Code Visual (VT-3) All accessible surtaces at Components (Wear) Section XI examination. specified frequency. Clevis insert bolts

(Note 2) Alignment and Interfacing All plants Loss of material ASME Code Visual (VT-3) All accessible surtaces at Components (Wear) Section XI examination. specified frequency. Upper core plate alignment pins

Notes:

1. XL = "Extra Long· referring to Westinghouse plants with 14-foot cores.

2. Bolt was screened in because of stress relaxation and associated cracking; however, wear of the clevis/insert is the issue.


Also included in Existing Programs are those components for which existing guidance has been issued (e.g., from the nuclear steam supply system (NSSS) vendors or Owners Groups) to address degradation that manifested itself during the current operational life of the PWR fleet. The continued implementation of this guidance has been determined to adequately manage the aging effects for these components.

4.4. 1 B& W Components

Table 4-7 describes the PWR internals in the Existing Programs for B& W plants.

No existing generic industry programs contain the specificity considered sufficient for monitoring the aging effects addressed by these guidelines for B&W plants. Therefore, no components forB& W plants were placed into the Existing Programs group.

4.4.2 CE Components

Table 4-8 describes the PWR internals in the Existing Programs forCE plants.

The following is a list of the CE Existing Programs Components.

• ASME Section XI

Existing:

Guide lugs and guide lug inserts and bolts (applicable to all plants)

- Fuel alignment pins (applicable to all plants with core shrouds assembled with fullheight shroud plates and all plants with core shrouds assembled in two vertical sections)

- Upper flange (applicable to all plants)

These component items may be considered core support structures listings that are typically examined during the 10-year inservice inspection per ASME Code Section XI Table IWB-2510, B-N-3 [2]. For these component items, the requirements of B-N-3 (visual VT-3) are considered sufficient to monitor for the aging effects addressed by these guidelines.

• Plant-specific

The guidance for ICI thimble tubes and thermal shield positioning pins is limited to plant specific recommendations and thus have no generic reference, nor are they included in Table 4-8. The owner should review their specific design, upgrade status, and plant commitments for CE ICI thimble tubes.

4-70

IPEC_PA_OAG0000126


Also included in Existing Progmms are those components for which existing guidance has been issued (e .g., from the nuclear steam supply system (NSSS) vendors or Owners Groups) to address degradation that manifested itself during the current operational life of the PWR fleet. The continued implementation of thi s guidance has been detennined to adequately manage the aging effects for these components.

4.4.1 B& W Components

Table 4-7 describes the PWR intemals in the Existing Progmms for B&W plants.

No existing generic industry programs contain the specificity considered suffic ient for monitoring the ag ing effects addressed by these guidelines for B&W plants. Therefore, no components for B& W plants were placed into the Existing Programs group.

4.4.2 CE Components

Table 4-8 describes the PWR internals in the Existing Progmms for CE plants.

The fo llowing is a li st of the CE Existing Programs Components.

• ASME Section XI

Existing:

Guide lugs and guide lug inserts and bolts (applicable to aJi plants)

Fuel alignment pins (applicable to all plants with core shrouds assembled with fu llheight shroud plates and all plants with core shrouds assembled in two vertical sections)

Upper flange (applicable to all plants)

These component items may be considered core SUppOl1 stllletures li stings that are typically examined during the 10-year inserviee inspection per ASME Code Section XI Table IW8-25 10, 8-N-3 ['2 ]. For these component items, the requirements of 8-N-3 (v isual VT-3) arc considered suffic ient to monitor for the aging effects addressed by these guidelines.

• Plant-specific

The guidance for ICI thimble tubes and thermaJ shield positioning pins is limited to plant specific recommendations and thus have no generic reference, nor are they included in Table 4-8. The owner should rev iew their specific des ign, upgrade status, and plant commitments fo r CE ICI thimble tubes.

4-70


4.4.3 Westinghouse Components

Table 4-9 describes the PWR internals in the Existing Programs for Westinghouse plants.

The following is a list of the Westinghouse Existing Programs Components.

• ASME Section XI

Existing:

- Core barrel flange (applicable to all plants)

- Upper support ring or skirt (applicable to all plants)

- Lower core plate and XL lower core plate (applicable to all plants)

- Clevis insert bolts (applicable to all plants)

- Upper core plate alignment pins (applicable to all plants)

These component items are considered core support structures that are typically examined during the 10-year inservice inspection per ASME Code Section XI Table IWB-2510, B-N-3 [2]. For these component items, the requirements of B-N-3 (visual VT-3) are considered sufficient to monitor for the aging effects addressed by these guidelines.

• Plant-specific

The guidance for flux thimble tubes is included in Table 4-9 and is based on owner commitments.

The guidance for guide tube support pins (split pins) is limited to plant specific recommendations and thus have no generic reference. Subsequent performance monitoring should follow the supplier recommendations. They thus are not included in Table 4-9. The owner should review their specific design, upgrade status, and asset management plans for Westinghouse guide tube support pins (split pins).

4.5 No Additional Measures Components

It has been determined that no additional aging management is necessary for components in this group. In no case does this determination relieve utilities of the ASME Code Section XI [2] IWB Examination Category B-N-3 inservice inspection requirements for components from this group classified as core support structures unless specific relief is granted as allowed by 10CFR50.55a [ 4].

4-71

IPEC_PA_OAG0000127


4.4.3 Westinghouse Components

Table 4-9 describes the PWR inte l11 als in the Existing Programs for Westinghouse plants.

The following is a li st of the Westinghouse Existing Programs Components.

• ASM E Section XI

Existing:

Core barrel flange (applicable to all plants)

Upper support ring or skiJ1 (applicable to all plants)

Lower core plate and XL lower core plate (applicable to all plants)

Clevis insert bolts (applicable to all plants)

Upper core plate alignment pins (applicable to all plants)

These component items are considered core support structu res that are typically examined during the IO-year inservice inspection per ASME Code Section XI Table rWB-25 10, 8-N-3 [21. For these component items, the requirements of 8-N-3 (v isual VT-3) are considered suffic ient to monitor fo r the aging effects addressed by these guidelines.

• Plant-specific

The guidance for flux thimble tubes is included in Table 4-9 and is based on owner commitments.

The guidance for guide tube support pins (split pins) is limited to plant specific recommendations and thus have no generic reference. Subsequent perfonnance monitoring should follow the supplier recommendations. They thus are not included in Table 4-9. The owner should review their specific design, upgrade status, and asset management plans for Westinghouse guide tube support pins (split pins).

4.5 No Additional Measures Components

It has been determined that no additional ag ing management is necessary for components in thi s group. In no case does thi s determination relieve utilities ofthe ASME Code Section XI [2] JWB Examination Category 8 -N-3 inservice inspection requirements fo r components from this group classified as core support structures unless specif ic relief is granted as allowed by IOCFR50. 55a [41.

4·7 1

IPEC_PA_OAG0000128

5 EXAMINATION ACCEPTANCE CRITERIA AND EXPANSION CRITERIA

The purpose of this section is to provide both examination acceptance criteria for conditions detected as a result of the examination requirements in Section 4, Tables 4-1 through 4-6, as well as criteria for expanding examinations to the Expansion components when warranted by the level of degradation detected in the Primary components.

Examination acceptance criteria identify the visual examination relevant condition(s) or signalbased level or relevance of an indication that requires formal disposition for acceptability. Based on the identified condition, and supplemental examinations if required, the disposition process results in an evaluation and determination of whether to accept the condition until the next examination or repair or replace the item. An acceptable disposition process is described in Section 6. Section 5.1 provides a discussion of relevant conditions applicable to the visual examination methods and of relevant indications applicable to the volumetric examinations employed in the guidelines. Section 5.2 provides examination acceptance criteria for physical measurements. These criteria are contained in Tables 5-1, 5-2, and 5-3 for B&W, CE, and Westinghouse plants, respectively.

Additionally, Tables 5-1, 5-2, and 5-3 contain expansion criteria for B&W, CE, and Westinghouse plants, respectively. Expansion criteria are intended to form the basis for decisions about expanding the set of components selected for examination or other aging management activity, in order to determine whether the level of degradation represented by the detected conditions has extended to other components judged to be less affected by the degradation.

5-1

IPEC_PA_OAG0000129

5 EXAMINATION ACCEPTANCE CRITERIA AND EXPANSION CRITERIA

The purpose of thi s section is to prov ide both examination acceptance criteri a fo r conditions detected as a result of the examination requirements in Section 4, Tables 4~ J through 4-6, as well as criteri a fo r expanding examinations to the Expansion components when warranted by the level of degradation detected in the Primary components.

Examination acceptance criteria identify the visual examination relevant condition(s) or signalbased level or re levance of an indication that requires fonnal di spos ition for acceptability. Based on the identified condition, and supplemental examinations if required, the di spos ition process results in an evaluation and determination of whether to accept the condition until the nex t examination or repair or replace the item. An acceptable di spos ition process is described in Section 6. Section 5. 1 prov ides a di scussion of relevant conditions applicable to the visuaJ examination methods and of relevant indications applicable to the vo lumetric examinations employed in the guide lines. Section 5.2 provides examination acce ptance criteri a for phys ical measurements. These crite ria are contained in Tables 5~ I, 5~2, and 5~3 for 8&W, CE, and Westinghouse plants, respectively.

Additionally, Tables 5- 1, 5~2, and 5·3 contain expansion criteria for 8 &W, CE, and Westinghouse plants, respectively. Expansion criteri a are intended to fOlm the basis fo r decis ions about expanding the set of components selected for examination or other ag ing management acti vity, in order to determine whether the leve l of degmdation represented by the detected conditions has ex tended to other components judged to be less affected by the degmdation.

5· 1

Examination Acceptance Criteria and Expansion Criteria

Table 5-1 B&W plants examination acceptance and expansion criteria

Item Applicability Examination Acceptance Expansion

Expansion Criteria Additional Examination

Criteria (Note 1) Link(s) Acceptance Criteria

Plenum Cover All plants One-time physical None N/A N/A Assembly & Core measurement. In addition, a Support Shield visual (VT-3) examination is Assembly conducted for these items.

Plenum cover weldment rib pads The measured differential Plenum cover support height from the top of the flange plenum rib pads to the

CSS top flange vessel seating surface shall average less than 0.004 inches compared to the as-built condition.

The specific relevant condition for these items is wear that may lead to a loss of function.

Core Support Shield ONS-3, DB Visual (VT-3) examination. CRGT Confirmed evidence of relevant The specific relevant Assembly spacer conditions for a single CSS cast condition is evidence of

CSS cast outlet nozzles castings outlet nozzle shall require that the fractured spacers or

The specific relevant VT-3 examination be expanded to missing screws. condition is evidence of include 100% of the accessible surface irregularities, such surfaces at the 4 screw locations as damaged or fractured (at every 90°) of the CRGT spacer nozzle material. castings by the completion of the

next refueling outage.

5-2

IPEC_PA_OAG0000130

E.xamillatioll Acceptance Criteria alld Ex pallsioll Criteria

Table 5-1 B&W plants examination acceptance and expansion criteria

Hem Applicability Examination Acceptance Expansion



Plenum Cover All plants One-time physical None N/A N/A Assembly & Core measurement. In addition, a Support Shield visual (VT-3) examination is Assembly conducted for these items.

Plenum cover weldment rib pads The measured differential Plenum cover support height from the top of the flange plenum rib pads to the

CSS top flange vessel seating surface shall average less than 0.004 inches compared to the as-bu ill condition.

The specific relevant condition for these items is wear that may lead to a loss of function.

Core Support Shield ONS-3. DB Visual (VT-3) examination. CRGT Confirmed evidence of relevant The specific re levant Assembly spacer conditions for a single CSS cast condition is evidence of

CSS cast oullet nozzles castings oullet nozzle shall requ ire that the fractu red spacers or

The specific relevant VT-3 examination be expanded to missing screws. condition is evidence of include 100% of the accessible surface irregularities, such surfaces at the 4 screw locations as damaged or fractured (at every 90°) of the CRGT spacer nozzle material. castings by the completion of the

next refueling outage.


Table 5-1 B&W plants examination acceptance and expansion criteria (continued)




Core Support Shield All plants Visual (VT-3) examination. CRGT Confirmed evidence of relevant The specific relevant Assembly spacer conditions in two or more CSS condition is evidence of

CSS vent valve discs castings vent valve discs shall require that fractured spacers or

The specific relevant the VT-3 examination be missing screws. condition is evidence of expanded to include 100% of the surface irregularities, such accessible surfaces at the 4 screw as damaged or fractured locations (at every 90°) of the disc material. CRGT spacer castings by the

completion of the next refueling outage.

Core Support Shield All plants Visual (VT-3) examination. None N/A N/A Assembly

CSS vent valve top The specific relevant retaining ring condition is evidence of CSS vent valve bottom damaged or fractured retaining ring material, and missing items.

CSS vent valve disc shaft or hinge pin

5-3

IPEC_PA_OAG0000131

/:,).:amination Acceptance Criteria and E..'7Jansion Criteria





Core Support Shield All plants Visual (VT-3) examination. CRGT Confirmed evidence of relevant The specific relevant Assembly spacer conditions in two or more CSS condition is evidence of

CSS vent valve discs castings vent valve discs shall require that fractu red spacers or

The specific relevant the VT-3 examination be missing screws. condition is evidence of expanded to include 100% of the surface irregularities, such accessible surfaces at the 4 screw as damaged or fractured locations (at every 90·) of the disc material. CRGT spacer castings by the

completion of the next refueling outage.

Core Support Shield All plants Visual (VT-3) examination. None NiA NiA Assembly

CSS vent valve top The specific relevant retaining ring condition is evidence of CSS vent valve bottom damaged or fractured retaining ring material , and missing items.

CSS vent valve disc shall or hinge pin

5-3



Item Applicability Examination Acceptance Expansion Expansion Criteria Additional Examination Criteria (Note 1) Link(s) Acceptance Criteria

Core Support Shield All plants 1) Volumetric (UT) LCB bolts 1) Confirmed unacceptable 1) The examination Assembly examination of the UCB (Note 2) indications exceeding 10% of the acceptance criteria for Upper core barrel (UCB) bolts. UCB bolts shall require that the UT the UT of the expansion bolts and their locking UTS, LTS, examination be expanded by the bolting shall be devices The examination acceptance and FD bolts completion of the next refueling established as part of the

criteria for the UT of the outage to include: examination technical UCB bolts shall be SSHT bolts For all plants justification. established as part of the (CR-3 and 100% of the accessible UTS, L TS, examination technical DB only) and FD bolts, 2) The specific relevant justification. Additionally for TMI-1 condition for the

Lower grid UT examination to include 100% of expansion of the VT-3 2) Visual (VT-3) examination shock pad the accessible lower grid shock locking devices is of the UCB bolt locking bolts (TMI-1 pad bolts, evidence of broken or devices. only) Additionally for CR-3 and DB missing bolt locking

UT examination to include 100% of devices. The specific relevant the accessible SSHT bolts. condition for the VT -3 of the UCB bolt locking devices is 2) Confirmed evidence of relevant evidence of broken or conditions exceeding 10% of the missing bolt locking devices. UCB bolt locking devices shall

require that the VT-3 examination be expanded by the completion of the next refueling outage to include: For all plants 100% of the accessible UTS, L TS, and FD bolt locking devices, Additionally for TMI-1 1 00% of the accessible lower grid shock pad bolt locking devices, Additionally for CR-3 and DB 100% of the accessible SSHT bolt locking devices.

5-4

IPEC_PA_OAG0000132






Core Support Shield All plants 1) Volumetric (UT) LCB bolts 1) Confirmed unacceptable 1) The examination Assembly examination of the UCB (Note 2) indications exceeding 10% of the acceptance criteria for Upper core barrel (UCB) bolts. UCB bolts shall requi re that the UT the UT of the expansion bolts and their locking UTS. LTS. examination be expanded by the bolting shall be devices The examination acceptance and FD bolts completion of the next refueling established as part of the

criteria for the UT of the outage to include: examination technical UCB bolts shall be SSHT bolts For all plants justification. established as part of the (CR-3 and 100% of the accessible UTS, LTS, examination technical DB only) and FD bolts, 2) The specific relevant justification. Additionally for TMf-1 condition for the

Lower grid UT examination to include 100% of expansion of the VT-3 2) Visual (VT-3) examination shock pad the accessible lower grid shock locking devices is of the UCB bolt locking bolls (TM 1-1 pad bolts, evidence of broken or devices. only) Additionally for CR-3 and DB missing bolt locking

UT examination to include 100% of devices. The specific relevant the accessible SSHT bolts. condition for the VT -3 of the UCB bolt locking devices is 2) Confirmed evidence of relevant evidence of broken or conditions exceeding 10% of the missing bolt locking devices. UCB bolt locking devices shall

require that the VT-3 examination be expanded by the completion of the next refueling outage to include: For all plants 100% of the accessible UTS, LTS, and FD bolt locking devices, Additionally for TMf-1 100% of the accessible lower grid shock pad bolt locking devices, Additionally for CR-3 and DB 100% of the accessible SSHT bolt locking devices.





Criteria (Note 1) Link(s) Acceptance Criteria Core Barrel Assembly All plants 1) Volumetric (UT) UTS, LTS, 1) Confirmed unacceptable 1) The examination Lower core barrel (LCB) examination of the LCB and FD bolts indications exceeding 10% of the acceptance criteria for bolts and their locking bolts. LCB bolts shall require that the UT the UT of the expansion devices SSHT bolts examination be expanded by the bolting shall be

The examination acceptance (CR-3 and completion of the next refueling established as part of the criteria for the UT of the LCB DB only) outage to include: examination technical bolts shall be established as For all plants justification. part of the examination Lower grid 100% of the accessible UTS, L TS, technical justification. shock pad and FD bolts,

bolts (TMI-1 Additionally for TMI-1 2) The specific relevant 2) Visual (VT-3) examination only) 1 00% of the accessible lower grid condition for the of the LCB bolt locking shock pad bolts, expansion of the VT -3 of devices. Additionally for CR-3 and DB the locking devices is

100% of the accessible SSHT evidence of broken or The specific relevant bolts. missing bolt locking condition for the VT -3 of the devices. LCB bolt locking devices is 2) Confirmed evidence of relevant evidence of broken or conditions exceeding 10% of the missing bolt locking devices. LCB bolt locking devices shall

require that the VT-3 examination be expanded by the completion of the next refueling outage to include: For all plants 100% of the accessible UTS, L TS, and FD bolt locking devices, Additionally for TMI-1 1 00% of the accessible lower grid shock pad bolt locking devices, Additionally for CR-3 and DB, 100% of the accessible SSHT bolt locking devices.

5-5

IPEC_PA_OAG0000133





Criteria (Note '1) Link(s) Acceptance Criteria Core Barrel Assembly All plants 1) Volumetric (UT) UTS. LTS. 1) Confirmed unacceptable 1) The examination Lower core barrel (LCB) examination of the LCB and FD bolls indications exceeding 10% of the acceptance criteria for bolls and their locking bolls. LCB bolls shall require that the UT the UT of the expansion devices SSHT bolls examination be expanded by the boiling shall be

The examination acceptance (CR-3 and completion of the next refueling established as part of the criteria for the UT of the LCB DB only) outage to include: examination technical bolts shall be established as For all plants justification. part of the examination Lower grid 100% of the accessible UTS, LTS, technical justification. shock pad and FD bolls,

bolts (TMI-1 Additionally for TMI-1 2) The specific relevant 2) Visual (VT-3) examination only) 100% of the accessible lower grid condition for the of the LCB boll locking shock pad bolls, expansion of the VT -3 01 devices. Additionally for CR-3 and DB the locking devices is

100% of the accessible SSHT evidence of broken or The specific relevant bolls. missing boll locking condition for the VT -3 of the devices. LCB boll locking devices is 2) Confirmed evidence of relevant evidence of broken or conditions exceeding 10% of the missing bolt locking devices. LCB bolt locking devices shall

require that the VT-3 examination be expanded by the completion of the next refueling outage to include: For all plants 100% of the accessible UTS, LTS, and FD bolt locking devices, Additionally for TMI-1 100% of the accessible lower grid shock pad boll locking devices, Additionally for CR-3 and DB, 100% of the accessible SSHT boll locking devices.

5·5






Core Barrel Assembly All plants Baseline volumetric (UT) Baffle-to-baffle Confirmed unacceptable N/A Baffle-to-former bolts examination of the baffle-to- bolts, indications in greater than or

former bolts. Core barrel-to- equal to 5% (or 43) of the baffle-former bolts to-former bolts, provided that

The examination acceptance none of the unacceptable bolts criteria for the UT of the are on former elevations 3, 4, baffle-to-former bolts shall and 5, or greater than 25% of the be established as part of the bolts on a single former plate, examination technical shall require an evaluation of the justification. internal baffle-to-baffle bolts for

the purpose of determining whether to examine or replace the internal baffle-to-baffle bolts. The evaluation may include external baffle-to-baffle bolts and core barrel-to-former bolts for the purpose of determining whether to replace them.

Core Barrel Assembly All plants Visual (VT-3) examination. a. Former a and b. Confirmed cracking in a and b. N/A Baffle plates plates multiple (2 or more) locations in

The specific relevant the baffle plates shall require condition is readily b. Core barrel expansion, with continued detectable cracking in the cylinder operation of former plates and baffle plates. (including the core barrel cylinder justified

vertical and by evaluation or by replacement circumferential by the completion of the next seam welds) refueling outage.

5-6

IPEC_PA_OAG0000134






Core Barrel Assembly All plants Baseline volumetric (UT) Baffle-to-baffle Confi rmed unacceptable NiA Baffle-to-former bolls examination of the baffle-to- bolls, indications in greater than or

former bolls. Core barrel-to- equal to 5% (or 43) of the baffle-former bolls to-former bolts, provided that

The examination acceptance none of the unacceptable bolls criteria for the UT of the are on former elevations 3, 4, baffl e-to-former bolts shall and 5, or greater than 25% of the be established as part of the bolts on a single former plate, examination technical shall requ ire an evaluation of the justification. internal baffle-to-baffle bolls for

the purpose of determining whether to examine or replace the internal baffle-to-baffle bolls. The evaluation may include external baffle-to-baffl e bolls and core barrel-to-former bolls for the purpose of determining whether to replace them.

Core Barrel Assembly All plants Visual (VT-3) examination. a. Former a and b. Confirmed cracking in a and b. N/A Baffle plates plates multiple (2 or more) locations in

The specific relevant the baffle plates shall require condition is readily b. Core barrel expansion, with continued detectable cracking in the cylinder operation of former plates and baffle plates. (including the core barrel cylinder justified

vertical and by evaluation or by replacement circumferential by the completion of the next seam welds) refueling outage.






Core Barrel Assembly All plants Visual (VT-3) examination. Locking Confirmed relevant conditions in N/A Locking devices, devices for greater than or equal to 1% (or 11) including locking welds, The specific relevant the external of the baffle-to-former or internal of baffle-to-former bolts condition is missing, non- baffle-to- baffle-to-baffle bolt locking devices and internal baffle-to- functional, or removed baffle bolts shall require an evaluation of the baffle bolts locking devices. and barrel- external baffle-to-baffle and core

to-former barrel-to-former bolt locking bolts devices for the purpose of

determining continued operation or replacement.

Lower Grid Assembly All plants Initial visual (VT-3) Alloy X-750 Confirmed evidence of relevant The specific relevant Alloy X-750 dowel-to- examination. dowel conditions at two or more locations condition is separated or guide block welds locking shall require that the VT-3 missing locking weld, or

The specific relevant welds to the examination be expanded to missing dowel. condition is separated or upper and include the Alloy X-750 dowel missing locking weld, or lower fuel locking welds to the upper and missing dowel. assembly lower fuel assembly support pads

support pads by the completion of the next refueling outage.

5-7

IPEC_PA_OAG0000135






Core Barrel Assembly All plants Visual (VT-3) examination. Locking Confi rmed relevant conditions in NiA Locking devices, devices for greater than or equal to 1 % (or 11) including locking welds, The specific relevant the external of the baffle-to-former or internal of baffle-Io-former bolls condition is missing, non- baffle-to- baffle-to-baffle bolt locking devices and internal baffle-Io- functional, or removed baffle bolts shall require an evaluation of the baffle bolts locking devices. and barrel - external baffle-to-baffle and core

to-former barrel-to-former bolt locking bolts devices for the purpose of

determining continued operation or replacement.

Lower Grid Assembly All plants Initial visual (VT-3) Alloy X-750 Confirmed evidence of relevant The specific relevant Alloy X-750 dowel-to- examination. dowel conditions at two or more locations condition is separated or guide block welds locking shall require that the VT-3 missing locking weld, or

The specific relevant welds to the examination be expanded to missing dowel. condition is separated or upper and include the Alloy X-7S0 dowel missing locking weld, or lower fuel locking welds to the upper and missing dowel. assembly lower fuel assembly support pads

support pads by the completion of the next refueling outage.

5·7






lncore Monitoring All plants Initial visual (VT-3) a. CRGT a. Confirmed evidence of relevant a. For the CRGT spacer Instrumentation (IMI) examination. spacer conditions for two or more IMI castings, the specific Guide Tube Assembly castings guide tube spider locations shall relevant conditions are IMI guide tube spiders The specific relevant require that the VT-3 examination fractured spacers or IMI guide tube spider-to- conditions for the IMI guide b. Lower fuel be expanded to include 100% of missing screws. lower grid rib section tube spiders are fractured or assembly the accessible surfaces at the 4 welds missing spider arms. support pad screw locations (at every 90°) of b. For the lower fuel

items: pad, the CRGT spacer castings by the assembly support pad The specific relevant pad-to-rib completion of the next refueling items (pads, pad-to-rib conditions for the IMI spider- section outage. section welds, Alloy X-to-lower grid rib section welds, Alloy 750 dowels, cap screws, welds are separated or X-750 b. Confirmed evidence of relevant and their locking welds), missing welds. dowel, cap conditions at two or more IMI the specific relevant

screw, and guide tube spider locations or IMI conditions are separated their locking guide tube spider-to-lower grid rib or missing welds, welds section welds shall require that the missing support pads,

VT-3 examination be expanded to dowels, cap screws and include lower fuel assembly locking welds, or support pad items by the misalignment of the completion of the next refueling support pads. outage.

Notes:

1. The examination acceptance criterion for visual examination is the absence of the specified relevant condition(s).

2. Expansion to LCB applies if the required Primary examination of LCB has not been performed as scheduled in Table 4-1.

5-8

IPEC_PA_OAG0000136






Incore Monitoring All plants Initial visual (VT-3) a. CRGT a. Confi rmed evidence of relevant a. For the CRGT spacer Instrumentation (IMI) examination. spacer conditions for two or more IMI castings, the specific Guide Tube Assembly castings guide tube spider locations shall relevant conditions are IMI guide tube spiders The specific relevant require that the VT-3 examination fractu red spacers or IMI guide tube spider-to- conditions for the IMI guide b. Lower fuel be expanded to include 100% of missing screws. lower grid rib section tube spiders are fractured or assembly the accessible surtaces at the 4 welds missing spider arms. support pad screw locations (at every 90°) of b. For the lower fuel

items: pad, the CRGT spacer castings by the assembly support pad The specific relevant pad-to-rib completion of the next refueling items (pads, pad-to-rib conditions for the IMI spider- section outage. section welds, Alloy X-to-lower grid rib section welds, Alloy 750 dowels, cap screws, welds are separated or X-7S0 b. Confi rmed evidence of relevant and thei r locking welds), missing welds. dowel, cap conditions at two or more IMI the specific relevant

screw, and guide tube spider locations or IMI conditions are separated their locking guide tube spider-to-Iower grid rib or missing welds, welds section welds shall require that the missing support pads,

VT-3 examination be expanded to dowels, cap screws and include lower fuel assembly locking welds, or support pad items by the misalignment of the completion of the next refueling support pads. outage.

Notes:

1. The examination acceptance criterion for visual examination is the absence of the specified relevant condition{s).

2. Expansion to LCB applies if the required Primary examination of LCB has not been pertormed as scheduled in Table 4- t .


Table 5-2 CE plants examination acceptance and expansion criteria

Examination Additional Examination

Item Applicability Acceptance Criteria Expansion Link(s) Expansion Criteria Acceptance Criteria

(Note 1)

Core Shroud Assembly Bolted plant Volumetric (UT) a. Core support a. Confirmation that >5% of the a and b. The (Bolted) designs examination. column bolts core shroud bolts in the four examination acceptance

Core shroud bolts b. Barrel-shroud plates at the largest distance criteria for the UT of the

bolts from the core contain core support column

The examination unacceptable indications shall bolts and barrel-shroud acceptance criteria for require UT examination of the bolts shall be established the UT of the core lower support column bolts as part of the shroud bolts shall be barrel within the next 3 refueling examination technical established as part of the cycles. justification. examination technical justification.

b. Confirmation that >5% of the core support column bolts contain unacceptable indications shall require UT examination of the barrel-shroud bolts within the next 3 refueling cycles.

Core Shroud Assembly Plant designs Visual (EVT -1) Remaining axial Confirmation that a surface- The specific relevant (Welded) with core examination. welds breaking indication> 2 inches in condition is a detectable

Core shroud plate-former shrouds length has been detected and crack-like surface assembled in sized in the core shroud plate- indication.

plate weld two vertical The specific relevant former plate weld at the core sections condition is a detectable shroud re-entrant corners (as

crack-like surface visible from the core side of the indication. shroud), within 6 inches of the

central flange and horizontal stiffeners, shall require EVT -1 examination of all remaining axial welds by the completion of the next refueling outage.

5-9

IPEC_PA_OAG0000137


Table 5-2 CE plants examinat ion acceptance and expansion criteria



(Note 1)

Core Shroud Assembly Bolted plant Volumetric (UT) a. Core support a. Confirmation that >5% of the a and b. The (Bolted) designs examination. column bolls core shroud bolls in the four examination acceptance

Core shroud bolts b. Barrel-shroud plates at the largest distance criteria for the UT of the

bolts from the core contain core support column

The examination unacceptable indications shall bolts and barrel-shroud acceptance criteria for require UT examination of the bolts shall be established the UT of the core lower support column bolts as part of the shroud bolls shall be barrel within the next 3 refueling examination technical established as part of the cycles. justification. examination technical justification.

b. Confirmation that >5% of the core support column bolts contain unacceptable indications shall require UT examination of the barrel-shroud bolts within the next 3 refueling cycles.

Core Shroud Assembly Plant designs Visual (EVT-1 ) Remaining axial Confi rmation that a surface- The specific relevant (Welded) with core examination. welds breaking indication> 2 inches in condition is a detectable

Core shroud plate-former shrouds length has been detected and crack-like surface assembled in sized in the core shroud plate- indication.

plate weld two vertical The specific relevant former plate weld at the core sections condition is a detectable shroud re-entrant corners (as

crack-like surface visible from the core side of the indication. shroud), within 6 inches of the

central flange and horizontal stiffeners, shall require EVT-1 examination of all remaining axial welds by the completion of the next refueling outage.

5-9


Table 5-2 CE plants examination acceptance and expansion criteria (continued)



(Note 1)

Core Shroud Assembly Plant designs Visual (EVT -1) a. Remaining axial a. Confirmation that a surface- The specific relevant (Welded) with core examination. welds breaking indication> 2 inches in condition is a detectable

Shroud plates shrouds b. Ribs and rings length has been detected and crack-like surface assembled sized in the axial weld seams at indication. with full- The specific relevant the core shroud re-entrant height shroud condition is a detectable corners at the core mid-plane plates crack-like surface shall require EVT-1 or UT

indication. examination of all remaining axial welds by the completion of the next refueling outage.

b. If extensive cracking is detected in the remaining axial welds, an EVT-1 examination shall be required of all accessible rib and ring welds by the completion of the next refueling outage.

5-10

IPEC_PA_OAG0000138




Item Applicability Acceptance Criteria Expansion link(s) Expansion Criteria Acceptance Criteria

(Note 1)

Core Shroud Assembly Plant designs Visual (EVT-1) a. Remaining axial a. Confirmation that a surface- The specific relevant (Welded) with core examination. welds breaking indication> 2 inches in condition is a detectable

Shroud plates shrouds b. Ribs and rings length has been detected and crack-like surface assembled sized in the axial weld seams at indication. with full- The specific relevant the core shroud re-entrant height shroud condition is a detectable corners at the core mid-plane plates crack-like surface shall require EVT-1 or UT

indication. examination of all remaining axial welds by the completion of the next refueling outage.

b. If extensive cracking is detected in the remaining axial welds, an EVT-1 examination shall be required of all accessible rib and ri ng welds by the completion of the next refueling outage.



Examination Additional Examination Item Applicability Acceptance Criteria Expansion Link(s) Expansion Criteria

Acceptance Criteria (Note 1)

Core Shroud Assembly Bolted plant Visual (VT-3) None N/A N/A (Bolted) designs examination.

Assembly

The specific relevant conditions are evidence of abnormal interaction with fuel assemblies, gaps along high fluence shroud plate joints, and vertical displacement of shroud plates near high fluence joints.

Core Shroud Assembly Plant designs Visual (VT -1) None N/A N/A (Welded) with core examination.

Assembly shrouds assembled in two vertical The specific relevant sections condition is evidence of

physical separation between the upper and lower core shroud sections.

5-11

IPEC_PA_OAG0000139





(Note 1)

Core Shroud Assembly Bolted plant Visual (VT -3) None N/A N/A (Bolted) designs examination.

Assembly

The specific relevant conditions are evidence of abnormal interaction with fuel assemblies, gaps along high fluence shroud plate joints, and vertical displacement of shroud plates near high fluence joints.

Core Shroud Assembly Plant designs Visual (VT·1 ) None N/A N/A (Welded) with core examination.

Assembly shrouds assembled in two vertical The specific relevant sections condition is evidence of

physical separation between the upper and lower core shroud sections.





(Note 1)

Core Support Barrel All plants Visual (EVT-1) Remaining core a. Confirmation that a surface- a and b. The specific Assembly examination. barrel assembly breaking indication >2 inches in relevant condition is a

Upper (core support welds beginning length has been detected and detectable crack-like with: sized in the upper flange weld surface indication. barrel) flange weld The specific relevant shall require that an EVT-1

condition is a detectable a. lower flange examination of the lower flange

crack-like surface weld, weld be performed by the c. The specific relevant

indication. followed by: completion of the next refueling condition is damaged or

b. remaining outage. fractured material of the

accessible core cast core support column

barrel assembly welds.

welds, and b. Confirmation that a surface-breaking indication >2 inches in

c. core support length has been detected and column welds sized in the lower flange weld (cast) shall require an EVT-1

examination of all remaining accessible core barrel assembly welds by the completion of the next refueling outage.

c. Confirmation of cracking in any of the remaining accessible core barrel assembly welds shall require a VT-3 examination of cast core support column welds, taking into account the general compressive loading of these columns and the potential for thermal aging embrittlement of the castings.

5-12

IPEC_PA_OAG0000140

E.xamillatioll Acceptance Criteria alld Expallsioll Criteria

Table 5-2 CE plants examinat ion acceptance and expansion criteria (continued)



(Note 1)

Core Support Barrel All plants Visual (EVT-1) Remaining core a. Confirmation that a surface- a and b. The specific Assembly examination. barrel assembly breaking indication >2 inches in relevant condition is a

Upper (core support welds beginning length has been detected and detectable crack-like with: sized in the upper flange weld surlace indication. barrel) flange weld The specific relevant shall require that an EVT-1

condition is a detectable a. lower flange examination of the lower flange

crack-like surlace weld, weld be perlormed by the c. The specific relevant

indication. followed by: completion of the next refueling condition is damaged or

b. remaining outage. fractured material of the

accessible core cast core support column

barrel assembly welds.

welds, and b. Confirmation that a surface-breaking indication >2 inches in

c. core support length has been detected and column welds sized in the lower flange weld (cast) shall require an EVT-1

examination of all remaining accessible core barrel assembly welds by the completion of the next refueling outage.

c. Confirmation of cracking in any of the remaining accessible core barrel assembly welds shall requi re a VT-3 examination of cast core support column welds, taking into account the general compressive loading of these columns and the potential for thermal aging embrittlement of the castings.





(Note 1)

Core Support Barrel All plants Visual (EVT-1) None N/A N/A Assembly examination.

Lower flange weld The specific relevant condition is a detectable crack-like indication.

Lower Support All plants Visual (EVT -1) None N/A N/A Structure with a core examination.

Core support plate support plate

The specific relevant condition is a detectable crack-like surface indication.

Upper Internals All plants Visual (EVT -1) None N/A N/A Assembly with core examination.

Fuel alignment plate shrouds assembled The specific relevant with full- condition is a detectable height shroud crack-like surface plates indication.

5-13

IPEC_PA_OAG0000141





(Note 1)

Core Support Barrel All plants Visual (EVT-1) None N/A N/A Assembly examination. Lower flange weld

The specific relevant condition is a detectable crack-like indication.

Lower Support All plants Visual (EVT -1) None N/A N/A Structure with a core examination.

Core support plate support plate

The specific relevant condition is a detectable crack-like surface indication.

Upper Internals All plants Visual (EVT-1) None N/A N/A Assembly with core examination. Fuel alignment plate shrouds

assembled The specific relevant with full - condition is a detectable height shroud crack-like surtace plates indication.





(Note 1)

Control Element All plants Visual (VT-3) Remaining Confirmed evidence of missing The specific relevant Assembly with examination. instrument tubes supports or separation at the conditions are missing

Instrument guide tubes instruments within the CEA welded joint between the tubes supports and separation tubes in the shroud assemblies and supports shall require the at the welded joint CEA shroud The specific relevant visual (VT-3) examination to be between the tubes and assembly conditions are missing expanded to the remaining the supports.

supports and separation instrument tubes within the CEA at the welded joint shroud assemblies by between the tubes and completion of the next refueling the supports. outage.

Lower Support All plants Visual (EVT -1) None N/A N/A Structure with core examination.

Deep beams shrouds assembled with full- The specific relevant

height shroud condition is a detectable

plates crack-like indication.

Notes: 1. The examination acceptance criterion for visual examination is the absence of the specified relevant condition(s).

5-14

IPEC_PA_OAG0000142





(Note 1)

Control Element All plants Visual (VT -3) Remaining Confirmed evidence of missing The specific relevant Assembly with examination. instrument tubes supports or separation at the conditions are missing

Instrument gu ide tubes instruments within the CEA welded joint between the tubes supports and separation tubes in the shroud assemblies and supports shall require the at the welded joint CEA shroud The specific relevant visual (VT-3) examination to be between the tubes and assembly conditions are missing expanded to the remaining the supports.

supports and separation instrument tubes within the CEA at the welded joint shroud assemblies by between the tubes and completion of the next refueling the supports. outage.

Lower Support All plants Visual (EVT-1 ) None N/A N/A Structure with core examination.

Deep beams shrouds assembled with full- The specific relevant

height shroud condition is a detectable

plates crack-like indication.



Table 5-3 Westinghouse plants examination acceptance and expansion criteria



Control Rod Guide All plants Visual (VT-3) None N/A N/A Tube Assembly examination.

Guide plates (cards)

The specific relevant condition is wear that could lead to loss of control rod alignment and impede control assembly insertion.

Control Rod Guide All plants Enhanced visual (EVT-1) a. Bottom-mounted a. Confirmation of surface- a. For BMI column Tube Assembly examination. instrumentation breaking indications in two or bodies, the specific

(BMI) column bodies more CRGT lower flange relevant condition for the Lower flange welds welds, combined with flux VT-3 examination is

The specific relevant thimble insertion/withdrawal completely fractured

condition is a detectable b. Lower support difficulty, shall require visual column bodies.

crack-like surface column bodies (cast) (VT-3) examination of BMI

indication. column bodies by the completion of the next b. For cast lower support refueling outage. column bodies, the

specific relevant condition is a detectable

b. Confirmation of surface- crack-like surface breaking indications in two or indication. more CRGT lower flange welds shall require EVT-1 examination of cast lower support column bodies within three fuel cycles following the initial observation.

5-15

IPEC_PA_OAG0000143


Table 5-3 Westinghouse plants examinat ion acceptance and expansion c riteria


Item Applicability Acceptance Criter ia Expansion Link(s) Expansion Criteria Acceptance Criteria

(Note 1)

Control Rod Guide All plants Visual (VT -3) None NiA NiA Tube Assembly examination.

Guide plates (cards)

The specific relevant condition is wear that could lead to loss of control rod alignment and impede control assembly insertion.

Control Rod Guide All plants Enhanced visual (EVT-1 ) a. Bottom-mounted a. Confirmation of surface- a. For BMI column Tube Assembly examination. instrumentation breaking indications in two or bodies, the specific

(BMI) column bodies more CRGT lower flange relevant condition for the Lower flange welds welds, combined with flux VT-3 examination is

The specific relevant thimble insertion/withdrawal completely fractured

condition is a detectable b. l ower support difficulty, shall require visual column bodies.

crack-like surface column bodies (cast) (VT-3) examination of BMI

indication. column bodies by the completion of the next b. For cast lower support refueling outage. column bodies, the

specific relevant condition is a detectable

b. Confirmation of surface- crack-like surface breaking indications in two or indication. more CRGT lower flange welds shall requ ire EVT-1 examination of cast lower support column bod ies within three fuel cycles following the initial observation.


Table 5-3 Westinghouse plants examination acceptance and expansion criteria (continued)



Core Barrel Assembly All plants Periodic enhanced visual a. Remaining core a. The confirmed detection and a and b. The specific (EVT-1) examination. barrel welds sizing of a surface-breaking relevant condition is a

Upper core barrel flange indication with a length greater detectable crack-like weld than two inches in the upper surface indication.

The specific relevant b. Lower support core barrel flange weld shall

condition is a detectable column bodies (non require that the EVT-1

crack-like surface cast) examination, and any

indication. supplementary UT examination, be expanded to include the core barrel-to-support plate weld by the completion of the next refueling outage. If extensive confirmed indications in the core barrel-to-support plate weld are detected, further expansion of the EVT-1 examination shall include the remaining core barrel assembly welds.

b. If extensive cracking in the remaining core barrel welds is detected, EVT-1 examination shall be expanded to include the upper six inches of the accessible surfaces of the non-cast lower support column bodies within three fuel cycles following the initial observation.

5-16

IPEC_PA_OAG0000144


Table 5-3 Westinghouse plants examinat ion acceptance and expansion c riteria (continued)



(Note 1)

Core Barrel Assembly All plants Periodic enhanced visual a. Remaining core a. The confirmed detection and a and b. The specific (EVT-1) examination. barrel welds sizing of a surface-breaking relevant condition is a

Upper core barrel flange indication with a length greater detectable crack-like weld than two inches in the upper surlace indication.

The specific relevant b. l ower support core barrel flange weld shall

condition is a detectable column bodies (non requ ire that the EVT-1

crack-like surlace cast) examination, and any

indication. supplementary UT examination, be expanded to include the core barrel-to-support plate weld by the completion of the next refueling outage. If extensive confirmed ind ications in the core barrel-to-support plate weld are detected, further expansion of the EVT-1 examination shall include the remaining core barrel assembly welds.

b. If extensive cracking in the remaining core barrel welds is detected, EVT-1 examination shall be expanded to include the upper six inches of the accessible surlaces of the non-cast lower support column bodies within three fuel cycles following the initial observation.





Baffle-Former All plants Visual (VT-3) None N/A N/A Assembly with baffle- examination.

Baffle-edge bolts edge bolts

The specific relevant conditions are missing or broken locking devices, failed or missing bolts, and protrusion of bolt heads.

Baffle-Former All plants Volumetric (UT) a. Lower support a. Confirmation that more than a and b. The Assembly examination. column bolts 5% of the baffle-former bolts examination acceptance

actually examined on the four criteria for the UT of the Baffle-former bolts baffle plates at the largest lower support column

The examination b. Barrel-former bolts distance from the core bolts and the barrel-

acceptance criteria for (presumed to be the lowest former bolts shall be

the UT of the baffle- dose locations) contain established as part of the

former bolts shall be unacceptable indications shall examination technical

established as part of require UT examination of the justification.

the examination lower support column bolts

technical justification. within the next three fuel cycles.

b. Confirmation that more than 5% of the lower support column bolts actually examined contain unacceptable indications shall require UT examination of the barrel-former bolts.

5-17

IPEC_PA_OAG0000145


Table 5-3 Westinghouse plants examinat ion acceptance and expansion c riteria (continued)



(Note 1)

Baffle-Former All plants Visual (VT -3) None NiA NiA Assembly with baffle- examination.

Baffle-edge bolls edge bolts

The specific relevant conditions are missing or broken locking devices, failed or missing bolts, and protrusion of bolt heads.

Baffle-Former All plants Volumetric (UT) a. Lower support a. Confi rmation that more than a and b. The Assembly examination. !column bolts 5% of the baffle-former bolts examination acceptance

actually examined on the four criteria for the UT of the Baffle-former bolts baffle plates at the largest lower support column

The examination b. Barrel-former bolts distance from the core bolts and the barrel-

acceptance criteria for (presumed to be the lowest former bolts shall be

the UT of the baffle- dose locations) contain established as part of the

former bolts shall be unacceptable indications shall examination technical

established as part of requi re UT examination of the justification.

the examination lower support column bolts

technical justification. within the next three fuel cycles.

b. Confi rmation that more than 5% of the lower support column bolts actually examined contain unacceptable indications shall require UT examination of the barrel-former bolts.





(Note 1)

Baffle-Former All plants Visual (VT-3) None N/A N/A Assembly examination.

Assembly

The specific relevant conditions are evidence of abnormal interaction with fuel assemblies, gaps along high fluence shroud plate joints, vertical displacement of shroud plates near high fluence joints, and broken or damaged edge bolt locking systems along high fluence baffle plate joints.

Alignment and All plants Direct physical None N/A N/A Interfacing Components with 304 measurement of spring

Internals hold down stainless height. steel hold

spring down springs The examination acceptance criterion for this measurement is that the remaining compressible height of the spring shall provide hold-down forces within the plant-specific design tolerance.

5-18

IPEC_PA_OAG0000146


Table 5-3 Westinghouse plants examination acceptance and expansion c riteria (continued)



(Note 1)

Baffle-Former All plants Visual (VT -3) None N/A N/A Assembly examination.

Assembly

The specific relevant conditions are evidence of abnormal interaction with fuel assemblies, gaps along high fluence shroud plate joints, vertical displacement of shroud plates near high flue nee joints, and broken or damaged edge bolt locking systems along high flue nee baffle plate joints.

Alignment and All plants Direct physical None N/A N/A Interfacing Components with 304 measurement of spring

Internals hold down stainless height. steel hold

spring down springs The examination acceptance criterion for this measurement is that the remaining compressible height of the spring shall provide hold-down forces within the plant-specific design tolerance.





Thermal Shield All plants Visual (VT-3) None N/A N/A Assembly with thermal examination.

shields Thermal shield flexures

The specific relevant conditions for thermal shield flexures are excessive wear, fracture, or complete separation.


5-19

IPEC_PA_OAG0000147





(Note 1)

Thermal Shield All plants Visual (VT -3) None NiA NiA Assembly with thermal examination.

shields Thermal shield fl exures

The specific relevant conditions for thermal shield flexures are excessive wear, fracture, or complete separation.



5.1 Examination Acceptance Criteria

5.1.1 Visual {VT-3) Examination

Visual (VT-3) examination has been determined to be an appropriate NDE method for the detection of general degradation conditions in many of the susceptible components. The ASME Code Section XI, Examination Category B-N-3 [2], provides a set of relevant conditions for the visual (VT-3) examination of removable core support structures in IWB-3520.2. These are:

1. structural distortion or displacement of parts to the extent that component function may be impaired;

2. loose, missing, cracked, or fractured parts, bolting, or fasteners;

3. corrosion or erosion that reduces the nominal section thickness by more than 5%;

4. wear of mating surfaces that may lead to loss of function; and

5. structural degradation of interior attachments such that the original cross-sectional area is reduced more than 5%.

For components in the Existing Programs group, these general relevant conditions are sufficient. However, for components where visual (VT-3) is specified in the Primary or the Expansion group, more specific descriptions of the relevant conditions are provided in Tables 5-1, 5-2, and 5-3 for the benefit of the examiners. Typical examples are "fractured material" and "completely separated material." One or more of these specific relevant condition descriptions may be applicable to the Primary and Expansion components listed in Tables 5-1, 5-2, and 5-3.

The examination acceptance criteria for components requiring visual (VT-3) examination is thus the absence of the relevant condition(s) specified in Tables 5-1, 5-2, and 5-3.

The disposition can include a supplementary examination to further characterize the relevant condition, an engineering evaluation to show that the component is capable of continued operation with a known relevant condition, or repair/replacement to remediate the relevant condition.

5.1.2 Visual {VT-1) Examination

Visual (VT-1) examination is defined in the ASME Code Section XI [2] as an examination "conducted to detect discontinuities and imperfections on the surface of components, including such conditions as cracks, wear, corrosion, or erosion." For these guidelines VT -1 has only been selected to detect distortion as evidenced by small gaps between the upper-to-lower mating surfaces of CE welded core shrouds assembled in two vertical sections.

The examination acceptance criterion is thus the absence of the relevant condition of gaps that would be indicative of distortion from void swelling.

5-20

IPEC_PA_OAG0000148

Examination AccelJtance Criteria and ExpallSion Criteria

5.1 Examination Acceptance Criteria

5.1.1 Visual (VT-3) Examination

Visual (VT-3) examination has been determined to be an appropriate NDE method for the detec tion of general degradation conditions in many of the susceptible components. The ASME Code Sect ion XI, Examination Category B-N-3 [2], provides a set of relevant conditions for the visual (VT-3) examination of removable core support structures in IWB-3520.2. These arc;

I. structural distortion or di splacement of parts to the extent that component function may be impaired;

2. loose, miss ing, cracked, or fractured parts, bolting, or fastene rs;

3. corrosion or erosion that reduces the nominal section thickness by more than 5%;

4. wear of mating sUlfaces that may lead to loss of function; and

5. structural degradation of interior attachments such that the original c ross-sectional area is reduced more than 5%.

For components in the Existing Programs group, these general relevant conditions are sufficient. However, for components where visual (VT-3) is specified in the Primary or the Expansion group, more specific descriptions of the relevant conditions are provided in Tables 5-1, 5-2, and 5-3 for the benefit of the examiners. Typical examples are "fractured material" and "completely separated material." One or more ofthese specific relevant condition descriptions may be applicable to the Primary and Expans ion components listed in Tables 5-1 , 5-2, and 5-3.

The examination acceptance criteria for components requiring visual (VT -3) examination is thus the absence of the relevant condition(s) specified in Tables 5- 1, 5-2, and 5-3.

The disposition can include a supplementary examination to fUlther characte rize the relevant condition , an engineering evaluat ion to show that the component is capable of continued operation with a known relevant condition, or repair/replacement to remediate the relevant condition.

5.1.2 Visual (VT·I) Examination

Visual (VT-I) examination is defined in the ASME Code Section XI 121 as an examination "conducted to detect di scontinuities and imperfections on the sUlface of components, including such conditions as cracks, wear, corrosion, or erosion." For these gu idelines VT - 1 has only been selected to detect di stortion as evidenced by small gaps between the upper-to-Iower mating sUlfaces of CE welded core shrouds assembled in two vertical sections.

The examination acceptance criterion is thus the absence of the relevant condition of gaps that would be indicative of di stortion from vo id swelling.

5-20


5.1.3 Enhanced Visual (EVT-1) Examination

Enhanced visual (EVT -1) examination has the same requirements as the ASME Code Section XI [2] visual (VT-1) examination, with additional requirements given in the Inspection Standard [3]. These enhancements are intended to improve the detection and characterization of discontinuities taking into account the remote visual aspect of reactor internals examinations. As a result, EVT -1 examinations are capable of detecting small surface breaking cracks and surface crack length sizing when used in conjunction with sizing aids (e.g. landmarks, ruler, and tape measure). EVT-1 examination has been selected to be the appropriate NDE method for detection of cracking in plates or their welded joints. Thus the relevant condition applied for EVT -1 examination is the same as found for cracking in Reference 2 which is crack-like surface breaking indications.

Therefore, until such time as generic engineering studies develop the basis by which a quantitative amount of degradation can be shown to be tolerable for the specific component, any relevant condition is to be dispositioned. In the interim, the examination acceptance criterion is thus the absence of any detectable surface breaking indication.

5.1.4 Surface Examination

Surface ET (eddy current) examination is specified as an alternative or as a supplement to visual examinations. No specific acceptance criteria for surface (ET) examination of PWR internals locations are provided in the ASME Code Section XI [2]. Since surface ET is employed as a signal-based examination, a technical justification per the Inspection Standard [3] provides the basis for detection and length sizing of surface-breaking or near-surface cracks. The signal-based relevant indication for surface (ET) is thus the same as the relevant condition for enhanced visual (EVT -1) examination. The acceptance criteria for enhanced visual (EVT -1) examinations in 5.1.3 (and accompanying entries in Tables 5-1, 5-2, and 5-3) are therefore applied when this method is used as an alternative or supplement to visual examination.

5. 1.5 Volumetric Examination

The intent of volumetric examinations specified for bolts in Section 4.3 of these I&E guidelines is to detect planar defects. No flaw sizing measurements are recorded or assumed in the acceptance or rejection of individual bolts or pins. Individual bolts or pins are accepted based on the detection of relevant indications established as part of the examination technical justification. When a relevant indication is detected in the cross-sectional area of the bolt or pin, it is assumed to be non-functional and the indication is recorded. A bolt or pin that passes the criterion of the examination is assumed to be functional.

Because of this pass/fail acceptance of individual bolts or pins, the examination acceptance criterion for volumetric (UT) examination of bolts and pins is based on a reliable detection of indications as established by the individual technical justification for the proposed examination. This is in keeping with current industry practice. For example, planar flaws on the order of 30% of the cross-sectional area have been demonstrated to be reliably detectable in previous bolt NDE technical justifications for baffle-former bolting.

5-21

IPEC_PA_OAG0000149


5.1.3 Enhanced Visual (EVT-l) Examination

Enhanced visual (EVT-I ) examination has the same requirements as the ASME Code Section XI [2 [ visual (YT-I ) examination, with additional requirements given in the Inspection Standard [3 [. These enhancements are intended to improve the detection and characterization of di scontinuities taking into account the remote visual aspect of reactor inte rnals examinations. As a result, EVT-I examinations are capable of detecting small surface breaking cmcks and sUiface crack length siz ing when used in conjunction with sizing aids (e.g. landmarks, ruler, and tape measure). EVTI examination has been se lected to be the appropriate NDE method for detection of cracking in plates or their welded joints. Thus the relevant condition applied for EVT- I examination is the same as found for cracking in Reference 2 which is crack-like sUiface breaking indications.

Therefore, until such time as generic engineering studies develop the basis by which a quantitative amount of degradation can be shown to be tole rable for the specific component, any relevant condition is to be di spositioned. In the interim, the examination acceptance criterion is thus the absence of any detectable surface breaking indication.

5.1.4 Surface Examination

5U1face ET (eddy curre nt) examination is specified as an alternative or as a supplement to visual examinations. No specific acceptance criteri a for surface (ET) examination of PWR intemals locations are provided in the ASME Code Sec tion XI [2]. Since surface ET is employed as a signal-based examination , a technical justification per the Inspection Standard [3] provides the basis for detection and length sizing of surface-breaking or near-surface cracks. The signal -based relevant indication for surface (ET) is thus the same as the relevant condition fo r enhanced visual (EVT- l) examination. The acceptance criteri a for enhanced visual (EVT- I) examinations in 5. 1.3 (and accompanying entries in Tables 5-1 , 5-2, and 5-3) are therefore applied when thi s method is used as an alternative or supplement to visual examination.

5.1.5 Volumetric Examination

The intent of volumetric examinations spec ified for bolts in Section 4.3 of these I&E guidelines is to detect planar defec ts. No flaw sizing measurements are recorded or assumed in the acceptance or rejection of individual bolts or pins. Individual bolts or pins are accepted based on the detec tion of relevant indications established as part of the examination technical justification. When a relevant indication is detected in the cross-sectional area of the bolt or pin, it is assumed to be non-functional and the indication is recorded. A bolt or pin that passes the criterion of the examination is assumed to be functional.

Because of thi s pass/fail acceptance of individual bolts or pins, the examination acceptance criterion for volumetric (UT) examination of bolts and pins is based on a reliable detection of indications as established by the individual technical justification for the proposed examination. This is in keeping with curre nt industry pmctice. For example, planar flaws on the order of 30% of the cross-sectional area have been demonstrated to be reliably detectable in previous bolt NDE technical justifications for baffle-fonner bolting.

5-21


Bolted and pinned assemblies are evaluated for acceptance based on meeting a specified number and distribution of functional bolts and pins. As discussed in Section 6.4, criteria for this evaluation can be: 1) found in previous Owners Group reports, 2) developed for use by the PWROG or 3) developed on a plant-specific basis by the applicable NSSS vendor.

5.2 Physical Measurements Examination Acceptance Criteria

Continued functionality can be confirmed by physical measurements where, for example, loss of material caused by wear, loss of pre-load of clamping force caused by various degradation mechanisms, or distortion/deflection caused by void swelling may occur. Where appropriate, these physical measurements are described in Section 4.3, with limits applicable to the various designs. ForB& W designs, the acceptable tolerance for the measured differential height from the top of the plenum rib pads to the vessel seating surface has been generically established and is provided in Table 5-1. For Westinghouse designs, tolerances are available on a design or plantspecific basis and thus are not provided generically in these guidelines. ForCE designs, no physical measurements are specified.

5.3 Expansion Criteria

The criteria for expanding the scope of examination from the Primary components to their linked Expansion components is contained in Tables 5-1, 5-2, and 5-3 for B&W, CE, and Westinghouse plants, respectively. The logic and basis for the levels of degradation warranting expansion is documented in an MRP letter [15].

5-22

IPEC_PA_OAG0000150

Examination AccelJtance Criteria and ExpallSion Criteria

Bolted and pinned assemblies are evaluated for acceptance based on meeting a specified number and di stribution of functional bolts and pins. As di scussed in Section 6.4, criteri a for thi s evaluation can be: I) found in previous Owners Group repol1s, 2) developed for use by the PWROG or 3) developed on a plant-specific basis by the applicable NSSS vendor.

5.2 Physical Measurements Examination Acceptance Criteria

Continued functionality can be confirmed by physical measurements where, for example, loss of materi al caused by wear, loss of pre-load of clamping force caused by various degradation mechanisms, or di stortion/deflection caused by void swelling may occur. Where appropriate, these phys ical measurements arc described in Section 4.3, with limits applicable to the various des igns. For 8 & W des igns, the acceptable tolerance for the measured differential height from the top ofthe plenum rib pads to the vessel seatjng sUlface has been generi cally established and is provided in Table 5- 1. For Westinghouse des igns, tolerances are available on a des ign or plalltspecific basis and thus are not provided generically in these guidelines. For CE des igns, no phys ical measurements are specified.

5.3 Expansion Criteria

The criteri a fo r expanding the scope of examination from the Primary components to their linked Expansion components is contained in Tables 5-1 , 5-2, and 5-3 for 8 & W, CE, and Westinghouse plants, respectively. The logic and basis for the levels of degradation warranting expansion is documented in an MRP letter ["1 51.

Y·22

6 EVALUATION METHODOLOGIES

There are various options that are available for the disposition of conditions detected during examinations (Section 4) that are unable to satisfy the examination acceptance criteria (Section 5). These options include, but are not limited to: (1) supplemental examinations, such as a surface examination, to supplement a visual (VT -1) or an enhanced visual (EVT -1) examination, to further characterize and potentially dispose of a detected condition; (2) engineering evaluation that demonstrates the acceptability of a detected condition; (3) repair, in order to restore a component with a detected condition to acceptable status; or ( 4) replacement of a component with an unacceptable detected condition.

The first option involves the re-examination of a component with an unacceptable detected condition with an alternative examination method that has the potential capability to further define or confirm with greater precision the component physical condition. This additional characterization may enable the more precise character of that detected condition to be found acceptable for continued service. An example would be the volumetric (UT) examination to depth size a surface-breaking flaw detected by either visual (VT -1) or enhanced visual (EVT -1) examination.

Section 6 concentrates on the second option, evaluation methodologies that can be used for evaluating flaws detected during the examinations described in Section 4 that exceed the examination acceptance criteria described in Section 5. The evaluation process depends upon the loading applied to the component, assembly, or system. Typical loading information to be considered is provided in Section 6.1 and evaluation methodology options are described in subsequent sections. These methodologies range from the satisfaction of limit load requirements for the internals assembly or component cross section to the satisfaction of flaw stability requirements using either linear elastic fracture mechanics (LEFM) or elastic-plastic fracture mechanics (EPFM), depending upon applicability. In addition, recommendations for flaw depth assumptions, in the absence of flaw depth sizing during examination, and flaw growth assumptions for subsequent operation until the next examination, are described. Justification for flaw evaluation fracture toughness limits is also provided. Design-specific or fleet-specific flaw handbooks may be used as an engineering evaluation tool.

6.1 Loading Conditions

The purpose of this section is to describe the typical loading conditions that govern the evaluation of flaws exceeding the examination acceptance criteria of Section 5.

6-1

IPEC_PA_OAG0000151

6 EVALUATION METHODOLOGIES

There are various options that are avaj lable for the disposition of conditions detected during examinations (Section 4) that are unable to satisfy the examination acceptance criteri a (Section 5). These options include, but arc not limited to: (I) supplemental examinations, such as a sUifacc examination, to supplement a visual (VT-I) or an enhanced visual (EVT-l) examination, to further characteri ze and potentially dispose of a detected condition; (2) engineering evaluation that dcmonstmtcs the acceptability of a detected condition; (3) repair, in order to restore a component with a detected condition to acceptable status; or (4) replacement of a component with an unacceptable detected condition.

The first option involves the re+examination of a component with an unacceptable detected condition with an alternative examination method that has the potential capability to fUl1her define or confu'm with greater precision the component physical condition. This additional characteri zation may enable the more precise character of that detected condition to be found acceptable for continued service. An example would be the volumetric (UT) examination to depth size a surface+breaking flaw detected by either visual (VT+ I) or enhanced visual (EVT -1) examination.

Section 6 concentrates on the second option, evaluation methodologies that can be used for evaluating fl aws detected during the examinations described in Section 4 that exceed the examination acceptance criteria described in Section 5. The evaluation process depends upon the loading applied to the component, assembly, or system. Typical loading information to be considered is provided in Section 6.1 and evaluation methodology options are described in subsequent sections. These methodologies range from the sati sfaction of limit load requirements for the internals assembly or component cross section to the sati sfaction of flaw stability requirements using either linear elastic fracture mechanics (LEFM) or elastic-plastic fracture mechanics (EPFM), depending upon applicability. In addition, recommendations for flaw depth assumptions, in the absence of flaw depth sizing during examination, and flaw growth assumptions for subsequent operation until the next examination, are described. Justification for flaw evaluation fracture toughness limits is also provided. Design-specific or fleet-specific flaw handbooks may be used as an engineering evaluation tool.

6.1 Loading Conditions

The purpose of thi s section is to describe the typical loading conditions that govern the evaluation of flaws exceeding the examination acceptance criteria of Section 5.

6-1

Evaluation Methodologies

Core support structures are designed to a set of defined loading conditions that typically include deadweight, such as the weight of the structure itself and an assigned portion of the weight of the fuel assemblies; mechanical loads, such as fuel assembly spring forces and control rod actuation loads; hydraulic loads; loadings caused by flow-induced vibration; loss-of-coolant accident (LOCA) loads; thermal loads, such as those from both normal operation thermal transients and upset condition thermal transients, as well as gamma heating; operating basis earthquake (OBE) and safe shutdown earthquake (SSE) seismic loads; handling loads that might occur during refueling and internals removal for inservice examinations; and interference conditions, friction forces, and dynamic insertion loads. Confirmation of required loading and combination requirements on an individual plant basis is essential prior to conducting any assessment.

For the case of many bolts and pins, the defined loading conditions include interference conditions, friction forces due to differential thermal growth, and dynamic insertion loads, in addition to dead weight, seismic, and vibration loadings.

The loading conditions for internal structures that are not core support structures are less well documented publicly. However, should an engineering evaluation be required for any internals structure (both core support structures and other internals), the original design basis should be examined, in order to determine the availability of actual or potential loading conditions.

6.2 Evaluation Requirements

The evaluation of component conditions that do not satisfy the examination acceptance criteria of Section 5 must be performed for a future state that corresponds to the next required examination or later. This future state should be determined based on the observed condition and a projection of future condition based on progressing degradation. The progressing degradation estimate should be based on a combination of operating experience (bolt failure histories), applicable testing data (crack growth rates in plate material), and available analytical results for that component. Uncertainties in predictive measures should be considered where applicable. Options for performing evaluations are contained in the following sub-sections.

6.2. 1 Limit Load Evaluation

Evaluation Requirement

An assembly or component that cannot meet the examination acceptance criteria of Section 5 of these I&E guidelines may be subject to limit load requirements as an evaluation disposition option, in order to continue in service in the existing condition. For PWR internals, the threshold for limit load requirements only is based on the accumulated neutron fluence exposure identified in BWRVIP-100-A [19]. This requirement states that, for accumulated neutron fluence less than 3xl020 n/cm2 (E > 1 MeV), or approximately 0.5 dpa, only a limit load evaluation requirement must be met for continued service of the internals assembly or individual component. A discussion and explanation of this requirement is contained in the following paragraphs.

6-2

IPEC_PA_OAG0000152

Em/l/atiOIl Methodologies

Core support structures are des igned to a set of defined loading conditions that typically include deadweight, such as the we ight of the structure itself and an assigned portion of the we ight of the fuel assemblies; mechanical loads, such as fuel assembly spring forces and control rod actuation loads; hydraulic loads; loadings caused by flow- induced vibration; loss-of-coolant acc ident (LOCA) loads; thennalloads, such as those from both normal operation thermal transients and upset condition thermal tra nsients, as well as gamma heating; operating basis earthquake (OBE) and safe shutdown earthquake (SSE) seismic loads; handling loads that might occur during refueling and internal s remova l for inserv ice examinations; and interference conditions, friction forces, and dynamic insertion loads. Confirmation of required loading and combination requirements on an individual plant basis is essential prior to conducting any assessment.

For the case of many bolts and pins, the defined loading conditions include interfe rence conditions, friction forces due to differential thermal growth, and dynamic insertion loads, in addition to dead weight, se ismic, and vibration load ings.

The loading conditions for inte rnal structures that are not core support structures are less well documented publicly. However, should an engineering evaluation be required for any internal s structure (both core SUppOl1 structures and other intemals), the original design bas is should be examined, in order to determine the avai lability of actual or potential loading conditions.

6.2 Evaluation Requirements

The evaluation of component conditions that do not satisfy the examination acceptance criteri a of Section 5 must be perfonned for a future state that corresponds to the nex t required examination or late r. This future state should be determined based on the observed condition and a projection of future condition based on progress ing degradation. The progress ing degradation estimate should be based on a combination of operating experience (bolt failure hi stories), applicable testing data (crack growth rates in plate material), and available analytical results fo r that component. Uncertainties in predictive measures should be considered where applicable. Options for perfolming evaluations are contained in the fo llowing sub-sections.

6.2. 1 Limit Load Evaluation

Evaluation Requirement

An assembly or component that cannot meet the examination acceptance criteri a of Section 5 of these I&E guidelines may be subject to limit load requirements as an evaluation di sposition option, in order to continue in serv ice in the ex isting condition. For PWR internals, the threshold for limit load requirements only is based on the accumulated neutron fluence exposure identified in BWRVlP-I OO-A (191 . This requirement states that, for accumulated neutron fluence less than 3x I 020 n/cm! (E > I MeV), or approximately 0.5 dpa , onl y a limit load evaluation requirement must be met fo r continued serv ice of the internals assembly or individual component. A di scuss ion and explanation of thi s requirement is contained in the following paragraphs.

6-2


Discussion and Explanation

Irrespective of the level of neutron irradiation exposure, limit load requirements can be satisfied for the affected assembly or component, in order to continue service until the end of the current inservice inspection interval. Therefore, the affected assembly or component can be shown to satisfy limit load requirements which may follow procedures similar to those given in the ASME Code Section XI, Appendix C [20]. The limit load calculation is carried out to find the critical degree of degradation within the elements of the assembly, or the progress of flaw parameters (location of the remaining cross section neutral axis and the effective flaw length) that cause the cross section to reach its limit load. For austenitic stainless steel, the stress limits for primary loading may be based on the irradiated mechanical strength properties for the minimum estimated fluence accumulated at the loaded section.

A safety factor of 2.77 on the limit load for expected loadings (ASME Service Loadings A and B) and a safety factor of 1.39 on the limit load for unexpected loadings (ASME Service Loadings C and D) must be met for the applied load on the assembly, or on the membrane and bending stresses in the component. The component analysis must demonstrate that a plastic hinge does not form in the remaining ligament of the cross section. For sections that have relatively uniform loss of material, and for unflawed sections that experience increased loading due to failure in other sections, the limiting primary stress and deflections for ASME Level C and D combinations should meet the plant design basis, or alternatively, meet the requirements of ASME Section Ill, Appendix F [21].

If the neutron fluence exposure is less than 3x1020 n/cm2 (E > 1 MeV), or approximately 0.5 dpa, this is the only evaluation that needs to be met for acceptance of the PWR internals assembly or individual component. No fracture toughness requirements need to be met for neutron fluence exposures less than this value.

6.2.2 Fracture Mechanics Evaluation

For neutron fluence levels exceeding 0.5 dpa, either an elastic-plastic fracture mechanics (EPFM) evaluation or a linear elastic fracture mechanics (LEFM) evaluation must be performed to assure continued structural integrity in the presence of detected flaws that exceed the examination acceptance criteria of Section 5. For neutron fluence above 0.5 dpa and below 5 dpa, EPFM is the preferred method. For neutron fluence above 5 dpa, LEFM should be utilized. Non-mandatory Appendix C of the ASME Code Section XI [20] provides general guidance which may be followed for performing such evaluations. Although the appendix strictly applies to austenitic stainless steel piping, the discussion of flaw growth due to fatigue, or due to stress corrosion cracking (SCC), or due to a combination of the two is relevant. Note, however, that fatigue crack growth rates in Article C-8000 are limited to air environments only, and that fatigue crack growth in water environments and SCC crack growth rates are not available yet.

For the case of IASCC, considerable research has been conducted on the effects of various levels of irradiation exposure on crack growth resistance, primarily by the Boiling Water Reactor Vessel & Internals Project (BWRVIP) [19]. Reference 19 also provides the technical basis for the recommendation of either LEFM or EPFM. Figure 6-1, reproduced from Reference 19, shows the data that were used to produce a set of conservative J-R curves (crack growth

6-3

IPEC_PA_OAG0000153

Emlua/ion Me/lwdologie.~

Discussion and Explanation

Inespective of the level of neutron irradiation exposure, limit load requirements can be sati sfied for the affected assembly or component, in order to continue service until the end of the current inse rvice inspection intervaL Therefore, the affec ted assembly or component can be shown to satisfy limit load requirements which may fo llow procedures similar to those given in the ASME Code Section XI, Appendix C [20] . The limit load calculation is carried out to find the critical degree of degradation within the elements of the assembly, or the progress of flaw parameters (location ofthe remaining cross sec tion neutral ax is and the effec ti ve fl aw length) that cause the cross sec tion to reach its limit load. For austenitic stainless steel, the stress limits for primary loading may be based on the inadiated mechanical strength properties fo r the minimum estimated fluence accumulated at the loaded section.

A safety factor of 2.77 on the limit load for expected loadings (ASME Service Loadings A and B) and a safety factor of 1.39 on the limit load for unexpected loadings (ASME Service Loadings C and D) must be met fo r the applied load on the assembly, or on the membrane and bending stresses in the component. The component analys is must demonstrate that a plastic hinge does not fo rm in the remaining ligament of the cross section. For sections that have relati vely unifonn loss of materi al, and for unflawed sections that experience increased loading due to fa ilure in other sections, the limiting primary stress and deflec tions for ASME Level C and 0 combinations should meet the plant des ign basis, or alternati vely, meet the requirements of ASME Section ILl , Appendix F 1211 .

If the neutron fluence exposure is less than 3x 1020 n/c m2 (E > 1 MeV), or approx imately 0.5 dpa, thi s is the only evaluation that needs to be met for acceptance of the PWR internals assembly or individual component. No fracture toughness requirements need to be met for neutron fluence exposures less than thi s value.

6.2.2 Fracture Mechanics Evaluation

For neutron fluence levels exceeding 0.5 dpa, either an elastic-plastic frac ture mechanics (EPFM) evaluation or a linear elastic fracture mechanics (LEFM) evaluation must be performed to assure continued stlUctura l integrity in the presence of detected flaws that exceed the examination acceptance crite ria of Section 5. For neutron fluence above 0.5 dpa and below 5 dpa, EPFM is the preferred method. For neutron fluence above 5 dpa, LEFM should be utili zed. Non-mandatory Appendix C of the ASME Code Section XI [20.1 prov ides general guidance which may be fo llowed for performing such evaluations. Although the appendix strictl y applies to austenitic stainless steel piping, the discuss ion of fl aw growth due to fa tigue, or due to stress corrosion cracking (SCC), or due to a combination of the two is relevant. Note, however, that fatigue crack growth rates in Article C-8000 are limited to air environments onl y, and that fatigue crack growth in water environments and SCC crack growth rates are not available yet.

For the case of lASee, considerable research has been conducted on the effects of various levels of irradiation exposure on crack growth res istance, primaril y by the Boiling Water Reactor Vessel & Internals Project (BWRVIP) 1"1 91. Reference 19 also provides the technical basis for the recommendation of e ither LEFM or EPFM. Figure 6-1 , reproduced from Reference 19, shows the data that were used to produce a set of conservative J-R curves (crack growth

6-3


resistance curves) for various exposure levels. Figures 6-2 and 6-3, also reproduced from Reference 19, show the lower bound for the power law parameter, C, and the upper bound for the power law parameter, n, in the curve fit to the crack growth resistance curve data given by

Jmat = C (~a)" Equation 6-1

where J and Care in KJ/m2 and ~a is in mm.

The lower bound expression for power law parameter Cis given by

C = (1217.9*6.697*1010 + 0.3908*F05563)/(6.697*1010 + F05563) Equation 6-2

The upper bound expression for power law parameter n is given by

n = 1/(4.962- 0.02439*p0°9976) Equation 6-3

The term F in the above expressions is the neutron fluence. At accumulated fluence values of approximately 1 dpa, the material has relatively high elastic-plastic crack growth resistance. For example, at 1 dpa, the upper bound power law parameter C equals 177 and the lower bound power law parameter n equals 0.492. Then, the crack growth resistance at 1.5 mm (0.059") of crack growth is 216 KJ/mm2

, or 1,609 in-lb/in2• Elastic-plastic behavior would be expected at

such a low fluence level.

At an accumulated fluence value of 10 dpa, C equals 55.2 and n equals 0.7833. Then, the crack growth resistance at 1.5 mm (0.059") of crack growth is 75.8 KJ/mm2

, or 565 in-lb/in2• If the

tangent to the crack growth resistance curve at 1.5 mm (0.059") is projected back to zero crack growth and converted to K1 through the expression

Equation 6-4

where E is the elastic modulus, then K1c equals 84 MPav'm, or about 76 ksiv'in. This value of fracture toughness is in the range that would suggest that LEFM is perhaps more suitable than EPFM, even though some amount of plastic response remains.

However, at 15 dpa, C equals 44.54 and n equals 0.889, so that the crack growth resistance at 1.5 mm (0.059") of crack growth is only 64 KJ/mm2

, or 476 in-lb/in2• Extrapolating the

tangent of the crack growth resistance curve back to zero crack growth and converting gives K1c =55 MPav'm, or 50 ksiv'in. Further analysis of more recent fracture toughness data at higher irradiation exposures for irradiated stainless steels has determined [25] that an appropriately conservative value for the fracture toughness of 38 MPav'm (34.6 ksiv'in) should be used for high neutron fluence exposure.

Therefore, for fluence levels below 5 dpa, the elastic-plastic crack growth resistance curves based on Equations 6-1 to 6-3 should be used. For neutron fluence greater than 5 dpa, LEFM analyses should be used with a limiting fracture toughness K1c =55 MPav'm (50 ksiv'in) for exposure levels between 5 and 15 dpa, and with a limiting fracture toughness K1c = 38 MPav'm (34.6 ksiv'in) for exposure levels greater than 15 dpa.

6-4

IPEC_PA_OAG0000154

Em/llation Methodologies

res istance curves) for various exposure levels. Figures 6-2 and 6-3, also reproduced from Reference 19, show the lower bound for the power law parameter, C, and the upper bound for the power law parameter, n, in the curve fit to the crack growth res istance curve data given by

Equation 6-1

where J and C are in KJ/ml and Lla is in mm.

The lower bound express ion for power law parameter C is given by

C = (121 7.9*6.697* 10'" + 0. 3908*F'''~')I(6.697* 10'" + F" '~) Equation 6-2

The upper bound express ion for power law parameter n is given by

n = 1/(4.962 - 0.02439* F" ~") Equation 6-3

The term F in the above express ions is the neutron fluence. At accumulated fluence values of approximately I dpa, the materi al has relatively high elastic-plastic crack growth resistance. For example, at I dpa, the upper bound power law parameter C equals 177 and the lower bound power law parameter n equals 0.492. Then, the crack growth resistance at 1.5 mm (0.059") of crack growth is 2 16 KJ/mm2

, or 1,609 in-Iblin1• Elastic-plastic behavior would be expected at

such a low fluence leve l.

At an accumulated fluence value of 10 dpa , C equals 55.2 and n equals 0.7833. Then, the crack growth resistance at 1.5 mm (0.059") of crack growth is 75.8 KJ/mm2

, or 565 in-Ib/ in2• If the

tangent to the crack growth resistance curve at 1. 5 mm (0.059") is projected back to zero crack growth and converted to K( through the express ion

Equation 6-4

where E is the elastic modulus, then KI( equals 84 M Pa--im, or about 76 ksi--iin . This va lue of fracture toughness is in the range that would suggest that LEFM is perhaps more suitable than EPFM, even though some amount of plastic response remains.

However, at 15 dpa, C equals 44.54 and n equals 0.889, so that the crack growth resistance at 1.5 mm (0.059") of crack growth is onl y 64 KJ/mm2

, or 476 in-Ib/in2. Extrapolating the

tangent of the crack growth resistance curve back to ze ro crack growth and converting gives K,C = 55 MPa--im, or 50 ksi--iin. Further analysis of more recent fracture toughness data at higher irrad iation exposures fo r irrad iated stainless steels has determined [25 J that an appropriately conservative value for the fracture toughness of 38 MPa--im (34.6 ksi--iin) should be used for high neutron fiuence exposure.

Therefore, for fiuence levels below 5 dpa, the elastic-plastic crack growth res istance curves based on Equations 6-1 to 6-3 should be used. For neutron fiuence greater than 5 dpa, LEFM analyses should be used with a limiting fracture toughness KI( = 55 MPa--im (50 ksi--iin) for exposure levels between 5 and 15 dpa, and with a limiting fracture toughness KI( = 38 MPa--im (34.6 ksi..Jin) for exposure levels greate r than 15 dpa.

6-4


8[]0

7[}0

5[]0

!:;; 5(}0 ""'j; :.:::

~ 4(}0

~.

·~ 3(]0 .,

2[}0

[}0

D

0.010 0 .5

Figure 6-1

.DO 1.5

Oraclk Extens i on , Aa, mm

---.- 1 .5E21J 8[6

---.- 1 ~E :!JJ 8[6

- 1.5E' J B[ii

----tr- :3E20 B[5

---4--- 3.5E:!JJ H B]

__.....,____ .!IE:!O W [21

- .... - J .:iE;-21 6[3· _ .. _ 3.5E+-21 8 [3

_ _, _ 3.5Ei-21 8[3

- -o.- aE+21 5[2

- .a - aE+21 3[2

- -o- 8E+21 fl[2

--:>: - 8E+21 W[2J:

- a - 8E+2 1 W[2J:

- . - 8E·+2 1 W [2.'

2.00 2.~

Experimental Jm.,.,,., versus crack extension curves for stainless steel materials at various fluence levels [19]

Figure 6-2

so

0

. 1.30

~

u~ w -.!! "' o::: 40 ~· 0 0 .!::: 30 LL.. ~

9-'

~ 20 a:_

10

.DE+2 CI

.. ~

• ~

'-......

' ~

• E-a.s~ Me1~

D 1/i.l_ d Metill

H,O.Z

C. EquO!liDfl 2 -2

• ~ - I~

1-

(

r--.. t 1'-- .... i ---- -

.OE+21

Neutron f i'Uen ce. rlic m?

~ 1.0E+ 2

J-R curve power law parameter Cas a function of neutron fluence for stainless steel, applicable for fluence less than 3x102

' n/cm2 [19]

6-5

IPEC_PA_OAG0000155

Emlua/ioll M ellwdologie.f

900

700

'''' , EO'" ~ ~

1 : 4C!'{) ;; .. ~ JOO ~

' 00

100

• ODO a.50

Figure 6-1

\ .00 ,.51)

Cr.lck Ext@onsion , "'iI , mm

---r- 1 51::29 1115]

- 15E:tI515l _ , 5E211 516)

----n- 3!lO 5[~J

~}~::;>o,...:el

----""- ':UO ' \O{2;-----O-- .:~:Il W12! - X'-'; .3f:t'lIl[23 ..... · 6,(£ZO 11(2)

•••• -BelO 11[11

·- . ·· ~~20!l[3i

•• ..,. . ' 9.:.1.1 SI;)

·· 0> ··20:.215[5j

" 9 " '!+2111[Jj

" O " 'E+~1 '3[~1

,, ''' ·· )!:HI S{Jj

- -. - :;~nl liP! - .. - 3 :1:: 4 1 1 SDl

-.- 3-510 · 21 5 [l1 - -fl - 3 ~~2 \ Jlf3)

- -0 - e! ;'21 sr.l - .(1_ S!.'"HI S[2!

- -c- ~~t21I1I2i

- "0:- lI!: t~ ' '1'1[2]

- .. -$10+21 wm

:?oo 2 50

Experimental J .... 1 ..... versus crack extension curves for stainless steel materials at various fluence levels (19]

Figure 6-2

. •

' DO ,.,

! .200 ~

'00

o 'f ()Et20

~ x

i',

• Sa", I~HJI

0 WNlI>t~1

X "'" - c. EqIR~ 2·2

I "~ ,~

-

r--.. •

1.0: .. 21

N~utron HUl!'nce. nlem:

II' 1.01:+22

J-R curve power law parameter C as a function of neutron fluence for stainless steel , applicable for fluence less than 3x10Zl n/cm2 (19]

6-5


• Ba:;r, MEr _g -

D lflleld Melal

_8 - It:: HI'.Z

- n, Equalfor• 2-J ~ ~FJ~ <: _7 ...;· <:: 'II _13 <:: 0 0. ,.;;

_5

/

-----~ f!- f-,... ·---- ~ • • (

t •

Neu ro n Fl 11enc e, nJcm~

Figure 6-3 J-R curve power law parameter nasa function of neutron fluence for stainless steel, applicable for fluence less than 3x1 021 n/cm2 [19]

6.2.3 Flaw Depth Assumptions

If the flaw depth has been determined by either the primary examination or by a supplementary examination method, that flaw depth should be used in any subsequent flaw evaluation. If only the flaw length has been determined by the examination, the evaluation should be based on the assumption that the flaw extends completely through the cross section of the component. The evaluation may be based on an assumption of depth if justified by a sufficiently robust technical demonstration.

6.2.4 Crack Growth Assumptions

Prior to the limit load and fracture mechanics calculations, the cyclic and time-dependent flaw growth from the current time to the next examination must be calculated. For example, if the inservice inspection interval is ten years, the flaw growth must be calculated for a ten-year period. If the examination is a one-time examination only, the growth of the flaw to the end of component life must be calculated and shown to satisfy acceptable limits. If the end-of-period flaw exceeds limits, the inservice inspection interval should be adjusted and a subsequent inspection performed prior to exceeding the flaw limit.

In the absence of sufficient information on crack growth in relevant PWR environments, data from BWR hydrogen water chemistry (HWC) environments is the most electrochemically appropriate and readily available source. A crack growth rate of l.lxl0 5 inches per hour (2.5 mm/year) in the depth direction has been accepted by the NRC staff for BWR HWC environments in their safety evaluation ofBWRVIP-14 [23]. This assumed flaw growth rate may be too conservative for a PWR water environment; therefore, the technical basis for reduced flaw growth rates is discussed in the following paragraphs.

6-6

IPEC_PA_OAG0000156


Figure 6-3

= O_i • ! o.e ~ w o_~

0.2

-

~

o 1.0E+20

• •

• en i! IJ~:.JI

0 W~d M~la l

• HAl - no ~q~ J~~r. 2·3 ~ ~'"

~

x • • , •

l.OE+2 1 l.OE+22

N~utron Flu~ne •• nJem2

J-R curve power law parameter n as a function of neutron fluence for stainless steel, applicable for fluence less than 3x1021 nfcm' [19]

6.2.3 Flaw Depth Assumptions

If the flaw depth has been determined by either the primary examination or by a supplementary examination method, that flaw depth should be used in any subsequent flaw evaluation. If onl y the fl aw length has been detennined by the examination, the evaluation should be based on the assumption that the flaw extends completely through the c ross sec tion of the component. The evaluation may be based on an assumption of depth if justified by a sufficiently robust technical demonstration.

6.2.4 Crack Growth Assumptions

Prior to the li mit load and frac ture mechanics ca lculations, the cycl ic and time-dependent flaw growth from the current time to the next examination must be calculated. For example, if the inselv ice inspection interval is ten years, the fl aw growth must be calculated for a ten-year period. If the examination is a one-time examination only, the growth of the flaw to the end of component life must be calculated and shown to sati sfy acceptable limits. If the end-of-period flaw exceeds limits, the inservice inspection interval should be adjusted and a subsequent inspection pelformed prior to exceeding the flaw limit.

In the absence of sufficient information on crack growth in relevant PWR environments, data from BWR hydrogen water chemistry (HWC) environments is the most electrochemically appropriate and read ily avaj lable source. A crack growth rate of 1.1 x 10-5 inches per hour (2.5 mm/year) in the depth direction has been accepted by the NRC staff fo r BWR HWC environments in their safety evaluation of BWRVIP- 14 [23]. Thjs assumed fl aw growth rate may be too conservative for a PWR water environment; therefore, the technical bas is for reduced fl aw growth rates is di scussed in the fo llowing paragraphs.

6-6


The most recent information on flaw growth rates for irradiated austenitic stainless steels in BWR environments is provided in BWRVIP-99 [24]. The information in BWRVIP-99 is based on both laboratory data and on field measurements of crack growth rates in BWR core shroud beltline welds, as measured by ultrasonic testing. The data are considered proprietary. The major findings were that field-measured crack growth rates varied from 2x10 6 to 5.25x10 5 inches per hour (about 0.5 mm to 11 mm per year), with the crack growth rate as a function of depth much lower than the crack growth rate as a function of length. Laboratory crack growth rates depended upon electro-chemical potential (ECP), with the growth rates substantially lower in a HWC environment that is more typical of a PWR environment. The HWC crack growth rates varied from 1x10 7 to 4x10 5 inches per hour (0.02 mm to 9 mm per year). The nominal reduction in crack growth rate for the HWC environment was found to be approximately 20 times lower than the corresponding crack growth rates in nominal BWR environments. However, the scatter in the data is very large.

For HWC environments, the recommended curve is given by

da/dt = 2.72 X 10 8 (K)25 Equation 6-5

Figure 6-4 shows that this curve approximates an upper bound to the relevant laboratory HWC data.

The BWR HWC curve is seen to be representative for PWR water environments, compared to limited crack growth rate data in PWR environment shown in Figure 6-5 [25]. Therefore, the HWC curve may be used for all PWR IASCC and SCC analyses until generic curves are established for IASCC and SCC in PWR environment. The use of alternative crack growth rate correlations in any analysis must be accompanied by an appropriate technical justification.

... .s:: -;:

1.00E-03

1.00E-04

:.:. 1.00E-05 '6 (ij '0

~ ,.

/

-~

........ ~--"""

~-"'" • Ill

-- da/dt=2.72E-8 KA2.5

1.00E-06

-- -.. .. 1.00E-07

0 5 10 15 20 25

K ( ksi-in1/2)

Figure 6-4 Proposed BWR hydrogen water chemistry crack growth curves for stainless steel irradiated between 5x1020 to 3x1021 n/cm2 [24]

30

6-7

IPEC_PA_OAG0000157

Evaluatioll Metlwdologie.f

The most recent information on flaw growth rates for irradiated austenitic stainless steels in BWR environments is provided in BWRVfP+99 124]. The information in BWRVIP~99 is based on both laboratory data and on field measurements of crack growth rates in BWR core shroud beltline welds, as measured by ultrason ic testing. The data arc considered proprietary. The major findings were that field-measured crack growth rates varied from 2x 10·tJ to 5.25x 10 S inches per hour (about 0.5 mm to 11 mm per year), with the c rack growth rate as a function of depth much lower than the crack growth rate as a function of length. Laboratory crack growth rates depended upon electro-chemical pOI.ential (ECP), with the growth rates substantially lower in a HWC environment that is more typical of a PWR environment. The HWC crack growth rates varied from J x 10.1 to 4x I O ~ inches per hour (0.02 mm to 9 mm per year). The nominal reduction in crack growth rate for the HWC environment was found to be approximately 20 times lower t.h an the cOITesponding crack growth rates in nominal BWR environments. However, the scatter in the data is very large.

For HWe environments, t.he recommended curve is given by

daJdt = 2.72 x 10 ' (K)" Equation 6-5

Figure 6-4 shows that this curve approx imates an upper bound to the relevant laboratory HWe data.

The BWR HWC curve is seen to be representative for PWR water environments, compared to limited crack growth rate data in PWR envi.ronment shown in Figure 6-5 [25] . Therefore, the HWC curve may be used for al l PWR IASCe and sec analyses until generic curves are. established for IASee and sec in PWR environment. The use of ahernative cmck growth rate correlations ill any analysis must be accompanied by an appropriare technical justification.

1.0DE'()3

1.DOE-04

i-'" • ~ ~ :::. 1 00E-05 ¥ . "

da!dt==;?72E·8 K" ;l .5

1.00E_06

1.00E-07 o 5 15 20

K { ksi-in1 l2J

Figure 6-4 Proposed BWR hydrogen water chemistry crack growth curves for stainless steel irradiated between 5x1020 to 3x1021 n/cm2 [24}

30

6-7


I:!. 304- 288°C to 315°C- PWR- 1,4 to 6,3 dpa

• 304- 316°C to 340°C- PWR- 6,3 to 32,9 dpa

o 304L- 280°C to 288°C- BWR HWC - 5,5 to 13,7 dpa

1.00E-04

1.00E-05

-~ e e 1.00E-06 -.f:l ~ ~ ..= 1.00E-07 ..... :::

~ -~

0 ;.... 0 •• ..::a=: 1.00E-08

"" ~ ;.... u

1.00E-09

1.00E-1 0

0 5

Figure 6-5

o 316- 280°C to 289°C- BWR HWC- 2 to 2,9 dpa

• 316- 320°C to 340°C- PWR- 17 to 25 dpa

~ 316Ti- 288°C- PWR- 25 dpa

•:• 316Ti- 320°C to 340°C- PWR- 25 dpa

<> 347- 280°C- BWR HWC- 3,2 dpa

• 347- 320°C- PWR -13,5 to 17 dpa

~:~

~:~

•: ~:~

~:~ )i(i

0 • 0 •

<> • <> • <> <> . • ( •

• ... . ~ • • oJF -• 0

• c. ll,c. 0

j&o 0

o• • I:!.

0

10 15 20 25 30 35 40

Stress Intensity K ( MPavlm)

45

Effect of stress intensity on IASCC crack growth rate [25]

6.3 Evaluation of Flaws in Bolts and Pins

50

•

55 60

For bolts and pins, no evaluation of individual items is required. Individual bolts or pins that are found to be unacceptable during the UT examination should be assumed to be non-functional, and the acceptance criterion for continued operation of the assembly that contains one or more non-functional bolts or pins are based on the functioning of the assembly, not the individual bolt or pin. In addition, no evaluation of individual items is required where visual examinations are the basis for determining functionality of bolts, pins or locking devices. Assessments in cases

6-8

IPEC_PA_OAG0000158

Eva/Ul/lio/l MelllOd%gies

1.00E-04

1.00E-05

-~ e 1.00E-06 ~ ., -" ~ .c 1.00E-07

~ 0 ..

<..'l '. -" 1.00E-08

" " .. U

1.00E-09

1.00E-10

a 5

Figure 6-5

6 304 - 288°C to 315°C - PWR - 1,4 to 6,3 dpa

... 304 - 316"C to 340"C - PWR - 6,3 to 32,9 dpa

o 304L - 280"C to 288°C - BWR HWC - 5,5 to 13,7 dpa

0 316 - 280"C to 289"C - BWR HWC - 2 to 2,9 dpa

. 316 - 320· C to 340·C - PWR -17 to 25 dpa

)( 316Ti - 288·C - PWR - 25 dpa

.:, 316Ti - 320°C to 340·C- PWR - 25 dpa

() 347- 280· C - BWR HWC - 3,2 dpa

. 347- 320· C - PWR - 13,5 to 17 dpa

.:.

x o • o •

o

o o • ~ o •

• • • · . ~ • • • o EO • o

o

o

10

• c .. •

• •

o

15 20 25 30 35 40 45

Stress Intensity K (MPa" m)

50

Effect of stress intensity on IASCC crack growth rate [25)

6.3 Evaluation of Flaws in Bolts and Pins

•

55 60

For bolts and pins. no evaluarion of individual itcms is required. Individual bolts or pins that are found to be unacceptable during the UT examination should be assumed to be non-functional , and the acceptance criterion for continued operation of the assembly that contains one or more non-functional bolts or pins are based on the functioning of the assembly. not the individual bolt or pin. [n addjt ion, no evaluat.ion of individual items is requ ired where visual examinations are the basis for dercrmin ing functionality of bolls, pins or locking devices. Assessments in cases

6-8


where the assembly is found to be deficient are most often driven by loose parts or reassembly interference evaluations that may be resolved using standard processes to support continued operation. Typically these are part of existing plant corrective action programs and as such should be sufficient to disposition.

6.4 Assembly Level Evaluations

As indicated in Sections 5.1.5, bolts are not accepted or rejected based on flaw sizing but on flaw detection. Thus the bolted assembly must be evaluated based on the number of rejected bolts, the minimum number required for functionality and an assumed failure rate until the next examination. Assemblies that satisfy an evaluation criterion that has been established by the NSSS vendor may be dispositioned. Alternatively, an assembly level evaluation may be performed to ensure that required functionality is maintained through the period until the next examination. Essential features of this type of evaluation are described below.

A process that can be followed for those system level evaluations is provided in the following paragraphs. The process builds on the vendor functionality evaluations [11, 12]. Other approaches can also be used. The finite element models to be used for the system level evaluation could take advantage of geometric and loading symmetry. Examples of such models have been demonstrated for the B&W-designed and Westinghouse-designed baffle-former assemblies, the CE-designed core shroud assembly, and bottom core plate assemblies for different vendor designs. The bolts and pins that are elements of the assembly should be modeled in sufficient detail to capture the essential structural behavior needed to demonstrate function or the lack thereof. For example, the assumption that a particular bolt, pin, or fastener has failed can be accounted for by modeling the bolt or pin as a one-dimensional finite element with no axial or shear strength. If a particular bolt or pin is assumed to maintain at least some or most of its preload, then the representation of material strength must be appropriate. That material strength should account conservatively for the local fluence and temperature for particular bolts or pins. The geometric modeling of the bolts and pins for system level evaluations does not require the level of detail that would be needed to predict localized failure in a bolt or pin.

The number of bolts or pins that are assumed to be non-functional should bound the estimated number and pattern of non-functional bolts or pins at the end of the evaluation interval. The estimation process is beyond the scope of this document. A conservative pattern that differs from the actual observed pattern of non-functional bolts or pins may be used. The loads referred to in Section 6.1 should be applied to this assembly model, and the structural response determined. This structural response should then be compared to assembly functional requirements, and a determination should be made about the capability to continue to operate the assembly through the remainder of the inspection interval.

The precise functionality criteria for each assembly are beyond the scope of this document. Reference should be made to vendor-recommended criteria.

6-9

IPEC_PA_OAG0000159


where the assembly is found to be defic ient are most often driven by loose parts or reassembly intelference evaluations that may be resolved using standard processes to suppOl1 continued operation. Typically these are part of existing plant correcti ve action programs and as such should be sufficient to di sposition.

6.4 Assembly Level Evaluations

As indicated in Sections 5. 1. 5, bolts are not accepted or rejected based on fl aw sizing but on flaw detection. Thus the bolted assembly must be evaluated based on the number of rejected bolts, the minimum number required for functionality and an assumed failure rate until the next examination. Assemblies that sati sfy an evaluation criterion that has been establi shed by the NSSS vendor may be dispos itioned. Alte rnatively, an assembly level evaluation may be perfonned to ensure that required functionality is maintained through the period until the nex t examinatjon. Essential features of this type of evaluation are described below.

A process that can be followed for those system level evaluations is provided in the following paragraphs. The process builds on the vendor functjonality evaluations [II , 12]. Other approaches can also be used. The finite e lement models to be used for the system level evaluation could take advantage of geometric and loading symmetry. Examples of such models have been demonstrated for the 8 &W-designed and Westinghouse-des igned baffle-former assemblies, the CE-designed core shroud assembly, and bottom core plate assemblies for different vendor designs. The bolts and pins that are elements of the assembly should be modeled in suffic ient detail to capture the essential structural behavior needed to demonstrate function or the lack thereof. For example, the assumption that a particular bolt, pin, or fas tener has failed can be accounted for by modeling the bolt or pin as a one-dimensional finite element with no ax ial or shear strength . If a partk ular bolt or pin is assumed to maintain at least some or most of its preload, then the representatjon of material strength must be appropriate. That material strength should account conservati vely for the local fluence and temperature for pal1icular bolts or pins. The geometric modeUng of the bolts and pins for system level evaluations does not require the level of detail that would be needed to predict locali zed failure in a bolt or pin.

The number of bolts or pins that are assumed to be non-functional should bound the estimated number and pattern of non-functional bolts or pins at the end of the evaluation interval. The estimation process is beyond the scope of thi s document. A conservati ve pattern that diffe rs from the actual observed patte rn of non-functional bolts or pins may be used. The loads referred to in Section 6.1 should be applied to thi s assembly model , and the structural response determined. This structural response should then be compared to assembly functional requirements, and a determinatjon should be made about the capability to continue to operate the assembly through the remainder of the inspection intervaL

The precise functionality criteria for each assembly are beyond the scope of thi s document. Reference should be made to vendor-recommended criteria .

6-9


6.5 Evaluation of Flaws in Other Internals Structures

Reference 22 describes a methodology to be used to evaluate detected and sized flaws found in PWR internals - other than bolts or pins - that exceed the examination acceptance criteria in Section 5.1. This methodology is summarized in the following steps.

First, the neutron fluence for the component is calculated or derived from existing calculations.

Second, the applied stresses are found from either existing stress analyses or from a new stress analysis of the assembly containing the affected component location.

Third, the detected and sized flaw from the examination is applied to a representation of the geometry of interest. Reference 22 has provided a number of representative PWR internal core support geometries of interest.

Fourth, the growth of the flaw over the period of time until the next examination, or until the end of component life, as applicable, is calculated. The flaw growth calculation will depend on the active mechanism driving the flaw extension (i.e. IASCC, SCC, or fatigue). Reference 22 assumed that negligible flaw growth occurred prior to application of nominal, design-basis, and bounding loads.

Fifth, load evaluation requirements (for example, limit load) for the flawed geometry after flaw growth, subject to both expected and unexpected loads, should be met.

Sixth, applied fracture mechanics stress intensities or applied }-integrals are calculated from the combination of the stresses and the grown flaws for the representative core support geometry of interest, as applicable. LEFM solutions may be obtained from the literature, with a conversion to an elastic-plastic crack driving force valid for localized plasticity at the crack tip.

Finally, the applied fracture mechanics stress intensities or the applied }-integrals must be shown to meet the limits of Section 6.2.2. For LEFM calculations, the applied fracture mechanics stress intensity must be shown to be less than the material fracture toughness. For EPFM calculation, the evaluation procedure specified in ASME Section XI, non-mandatory Appendix K, Article K-4000, K-4220 [2], can be used to demonstrate flaw stability. Specifically, Paragraph K-4220 provides a flaw stability criterion that limits the elastic-plastic crack driving force to less than the material elastic-plastic crack growth resistance at a crack extension of 0.1 inches. The safety margin that is demonstrated in meeting the limits of Section 6.2.2 should be identified and justified for the classes of loading considered.

The methodology outlined above has been demonstrated in Reference 22, where five simple geometries were analyzed with assumed dimensions that represented a wide variety of PWR internals locations. Because of the uncertainty in the applied stresses and the conservatism of the bounding material fracture toughness, no safety margins were applied to the critical flaw size calculations. The five simple geometries analyzed are described below:

• A semi-elliptical surface crack in a flat plate that can represent: (i) a semi-elliptical surface crack at the inside or outside flat surface of baffle plates; (ii) a semi-elliptical surface crack at the inside or outside flat surface of a core support barrel; or (iii) a semi-elliptical surface

6-10

IPEC_PA_OAG0000160


6.5 Evaluation of Flaws in Other Internals Structures

Reference 22 describes a methodology to be used to evaluate detected and sized flaws found in PWR internals - other than bolts or pins - that exceed the examination acceptance criteri a in Section 5.1. This methodology is summarized in the following steps.

First, the neutron fluence for the component is calculated or deri ved from ex isting calculations.

Second, the applied stresses are found from either existing stress analyses or from a new stress analys is of the assembly containing the affected component location.

Third, the detected and sized fl aw from the examination is applied to a representation of the geometry of interest. Reference 22 has provided a number of representati ve PWR internal core support geometries of interest.

Fourth, the growth of the flaw over the period of time until the next examination, or until the end of component life, as applicable, is calculated. The fl aw growth calculation will depend on the active mechanism driving the fl aw extension (i.e. IASee, sec, or fatigue). Reference 22 assumed that negligible flaw growth occurred prior to application of nominal, des ign-basis, and bounding loads.

Fifth, load evaluation requi re ments (for example, limit load) for the flawed geometry after fl aw growth, subject to both expected and unexpected loads, should be met.

Sixth, applied fracture mechanics stress intensities or applied l-integrals are calculated from the combination of the stresses and the grown fl aws for the representative core support geometry of interest, as applicable. LEFM solutions may be obtained from the literature, with a convers ion to an elastic-plastic crack driving force valid for localized plastic ity at the crack tip .

Finally, the applied frac ture mechanics stress intensities or the applied J-integral s must be shown to meet the limits of Section 6.2.2. For LEFM calculat ions, the applied fracture mechanics stress intensity must be shown to be less than the material fracture toughness. For EPFM calculation, the evaluation procedure specified in ASME Section XI , non-mandatory Appendix K, Article K-4000, K-4220 [2], can be used to demonstrate fl aw stability. Specifically, Paragra ph K-4220 provides a flaw stability criterion that limits the elastic-plastic crack driving force to less than the material elastic-plastic crack growth resistance at a crack ex tension of 0.1 inches. The safety margin that is demonstrated in meeting the limits of Section 6.2.2 should be identified and justified for the classes of load ing considered.

The methodology outlined above has been demonstrated in Reference 22, where five simple geometries were analyzed with assumed dimensions that represented a wide variety of PWR internals locations. Because of the uncertainty in the applied stresses and the conservati sm of the bounding materi al fracture toughness, no safety marg ins were applied to the critical fl aw size calculations. The five simple geometries analyzed are described below:

• A semi-elliptica l surface crack in a fl at plate that can represent: ( i) a semi-elliptical surface crack at the inside or outside flat surface of baffle plates; (ii) a semi-elliptical sUlface crack at the inside or outside fl at surface of a core support barrel; or (i ii) a semi-elliptical surface

6· 10


crack at the inside or outside smface of a core barrel. The flaw can be either circumferential (e.g., in the circumferential weld seam of the core barrel) or longitudinal (e.g., in the vertical weld seam). A flat plate solution is adequate for these cylinders when the radius to thickness ratio (R/t) is greater than 36 and loading level is fairly low;

• A through-wall crack in the center of a plate that can represent: (i) a through-wall crack in baffle plates; (ii) a through-wall crack in the flat surface of a core support barrel; (iii) a circumferential through-wall crack (e.g. in the circumferential weld seam) in a core barrel; or (iv) a longitudinal through-wall crack (e.g. in the vertical weld seam) in a core barrel;

• A through-wall edge crack in a flat plate that can represent: (i) a through-wall crack emanating from the side edges of baffle plates; or (ii) a through-wall crack emanating from the edge of former plates;

• A through-wall edge crack emanating from a 1 and 3/8-inch diameter hole that can represent: (i) two through-wall edge cracks emanating from baffle-to-former bolt holes or cooling holes; or (ii) two through-wall edge cracks emanating from holes in former plates; and

• A quarter-circular corner crack in a rectangular bar that can represent: (i) a quarter-circular crack in the corner of baffle plates; or (ii) a quarter-circular crack at the inside corner of a core support barrel.

Although no detailed loading/stress information was available for the various geometries, limited information was used to estimate the maximum normal operating stress (2.5 ksi) and the maximum LOCA stress (10 ksi) in highly irradiated components. For completeness, however, remote tensile stress levels up to 50 ksi were analyzed.

For the three types of postulated through-wall flaws, the analyses showed that the critical flaw is more limiting for a through-wall edge crack or a through-wall edge crack emanating from a hole than for a through-wall centered crack. For a medium-width baffle plate (26-inch), the critical flaw length for a through-wall crack is 22.8 inches at 2.5 ksi and 7.62 inches at 10 ksi. For the same baffle plate, the critical flaw length for a through-wall edge crack is 11.3 inches at 2.5 ksi and 2.65 inches at 10 ksi.

6-11

IPEC_PA_OAG0000161


crack at the inside or outside sUiface of a core barrel. The fl aw can be either c ircumferential (e.g., in the circumferential weld scam of the core barrel) or longitudinal (e.g., in the vertical weld scam). A flat plate solution is adequate for these cylinders when the radius to thickness ratio (Rlt) is greater than 36 and loading level is fairly low;

• A through-wall crack in the center of a plate that can represent: ( i) a through-wall crack in baffle plates; ( ii ) a through-wall crack in the flat surface of a core SUppOl1 barrel; (iii) a circumferential through-wall crack (e.g. in the circumferential weld seam) in a core barrel ; or (iv) a longitudinal through-wall crack (e.g. in the vel1ical weld seam) in a core ban·eI;

• A through-wall edge crack in a flat plate that can represent: (i) a through-wall c rack emanating from the side edges of baffle plates; or (ii) a through-wa ll crack emanating from the edge of former plates;

• A through-wall edge crack emanating from a I and 3/S-inch diameter hole that can represent: ( i) two through-wall edge cracks emanating from baffle-to-fonner bolt holes or cooling holes; or (ii) two through-wall edge cracks emanating from holes in fonner plates; and

• A quarter-circular corner crack in a rectangular bar that can represent: (i) a quarter-circular crack in the corner of baffle plates; or (ii) a quarter-circular crack at the inside corner of a core support barrel.

Although no detailed loading/stress infonnation was available for the various geometries, limited information was used to estimate the max imum normal operating stress (2.5 ksi) and the maximum LOCA stress (10 ksi) in highly irradiated components. For completeness, however, remote tensile stress levels up to 50 ksi were analyzed.

For the three types of postulated through-wall flaws, the analyses showed that the critical flaw is more limiting for a through-wall edge crack or a through-wall edge crack emanating from a hole than for a through-wall centered crack. For a medium-width baffle plate (26-inch), the critical flaw length for a through-wall crack is 22.S inches at 2.5 ksi and 7.62 inches at 10 ksi. For the same baffle plate, the critical flaw length for a through-wall edge crack is I J.3 inches at 2.5 ksi and 2.65 inches at 10 ksi.

6-11

IPEC_PA_OAG0000162

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