ML12095A218

NRC Staff Pre-Filed Hearing Exhibit NRC00114B, Materials Reliability Program: PWR Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A). Part 2 of 6

ML12095A218
Person / Time
Site:	Indian Point, PROJ0669
Issue date:	01/09/2012
From:	Greenlee S Electric Power Research Institute, Entergy Nuclear Operations
To:	Stuchell S Office of Nuclear Reactor Regulation
Shared Package
ML12093A500	List: ML12093A502 ML12093A503 ML12095A218 ML12095A220
References
RAS 22205, 50-247-LR, 50-286-LR, ASLBP 07-858-03-LR-BD01, MRP 2011-036, MRP-227-A 1022863
Download: ML12095A218 (102)

Text

Aging Management Requirements The vent valves are contained in the core support shield assembly where the plenum assembly resides. These valves are check valves meant to relieve pressure in the interior of the core support assembly during a large break LOCA, preventing backpressure from reversing coolant flow through the core. These vent valves can be damaged due to mishandling when inserting and removing the plenum. The vent valve components listed above were identified as being susceptible to thermal aging embrittlement, which may lead to cracking. An existing program is in place at each of the B&W-designed units that requires testing and inspection of the vent valve assemblies each refueling outage. The aging management measures provided in these requirements include a provision to visually inspect the valve body and disc seating surfaces.

Continuation of the existing vent valve testing and inspection requirements will manage cracking of the vent valve component items that could cause loss of the vent valve function.

Primary (applicable to all plants):

Plenum cover weldment rib pads Plenum cover support flange CSS top flange There are no expansion items for these components.

The potential age-related degradation mechanism for the core clamp region is wear. The purpose of the clamping is to stabilize and significantly restrict rigid body pendulum motion of the core support assembly. Wear at these locations will progress from motions generated by fluid flow once the loss of core clamping is initiated. Note that a one-time physical measurement is to be performed prior to subsequent visual (VT-3) examination.

Primary:

Upper core barrel (UCB) bolt locking devices (applicable to all plants)

Expand to:

• Upper thermal shield (UTS) bolt locking devices (applicable to all plants)

• Lower thermal shield (LTS) bolt or stud/nut locking devices (applicable to all plants)

" Surveillance specimen holder tube (SSHT) bolt or stud/nut locking devices (Crystal River Unit 3 (CR-3) and Davis-Besse (DB) only)

• Lower grid shock pad bolt locking devices (TMI-1 only)

Lower core barrel (LCB) bolt locking devices (applicable to all plants)

Expand to:

" Upper thermal shield (UTS) bolt locking devices (applicable to all plants)

• Lower thermal shield (LTS) bolt or stud/nut locking devices (applicable to all plants)

• Surveillance specimen holder tube (SSHT) bolt or stud/nut locking devices (Crystal River Unit 3 (CR-3) and Davis-Besse (DB) only)

• Lower grid shock pad bolt locking devices (TMI-1 only) 4-9 NRC000114B Resubmitted: April 2, 2012

Aging Management Requirements Flow Distributor (FD) bolt locking devices (applicable to all plants)

Expand to:

• Upper thermal shield (UTS) bolt locking devices (applicable to all plants)

• Lower thermal shield (LTS) bolt or stud/nut locking devices (applicable to all plants)

" Surveillance specimen holder tube (SSHT) bolt or stud/nut locking devices (Crystal River Unit 3 (CR-3) and Davis-Besse (DB) only)

" Lower grid shock pad bolt locking devices (TMI-1 only)

Note that the bolts or stud/nuts associated with these locking devices are also examined by volumetric (UT) examination.

• Volumetric (UT) Examination Primary:

Upper core barrel (UCB) bolts (applicable to all plants)

Expand to:

• Upper thermal shield (UTS) bolts (applicable to all plants)

" Lower thermal shield (LTS) bolts or stud/nuts (applicable to all plants)

• Surveillance specimen holder tube (SSHT) bolts or stud/nuts (Crystal River Unit 3 (CR-3) and Davis-Besse (DB) only)

" Lower grid shock pad bolts (TMI-l only)

Lower core barrel (LCB) bolts (applicable to all plants)

Expand to:

• Upper thermal shield (UTS) bolts (applicable to all plants)

• Lower thermal shield (LTS) bolts or stud/nuts (applicable to all plants)

• Surveillance specimen holder tube (SSHT) bolts or stud/nuts (Crystal River Unit 3 (CR-3) and Davis-Besse (DB) only)

" Lower grid shock pad bolts (TMI-1 only)

Flow Distributor (FD) bolts (applicable to all plants)

Expand to:

" Upper thermal shield (UTS) bolts (applicable to all plants)

" Lower thermal shield (LTS) bolt or stud/nuts (applicable to all plants)

• Surveillance specimen holder tube (SSHT) bolt or stud/nuts (Crystal River Unit 3 (CR-3) and Davis-Besse (DB) only)

" Lower grid shock pad bolts (TMI-1 only) 4-10

Aging Management Requirements Note that the locking devices associated with these bolts or stud/nuts are also examined by visual (VT-3) examination.

The potential degradation mechanism for the high-strength bolting rings is stress corrosion cracking. For bolting or stud/nuts, this mechanism is best detected using ultrasonic examination techniques.

The upper core barrel bolts are accessible for ultrasonic examination while the core support shield assembly is in the reactor vessel and the plenum is removed. Ultrasonic examination of the upper core barrel bolts can be performed during a normal refueling outage. The lower core barrel bolts and flow distributor bolts are only accessible when the core support shield assembly is removed from the reactor vessel. Some lower core barrel bolts are more difficult to examine and are inaccessible for replacement due to the presence of the core guide blocks mounted on the side of the lower grid assembly.

Primary (applicable to all plants):

Baffle-to-former (FB) bolts Expand to:

• Baffle-to-baffle (BB) bolts

• Core barrel-to-former (CBF) bolts Note that the locking devices associated with these bolts are also examined by visual (VT-3) examination.

Note that even though the baffle-to-baffle (BB) bolts and core barrel-to-former (CBF) bolts are Expansion components, they require an evaluation and not an inspection.

0 Physical Measurement Primary (applicable to all plants):

Plenum cover weldment rib pads Plenum cover support flange CSS top flange There are no expansion items for these components.

Note: the measurement is performed to determine the differential height of top of the plenum rib pads to the reactor vessel seating surface with all three items inside the reactor vessel, but with the fuel assemblies removed.

Note that these components are subsequently examined by visual (VT-3) examination.

4.3.2 CE Components Tables 4-2 and 4-5 describe the examination requirements for the PWR internals Primary and Expansion components for CE plants.

The following is a list of the CE Primary and Expansion components by examination technique.

4-11

Aging Management Requirements Visual (VT-3) Examination Primary (applicable to all plants):

Core support column welds There are no expansion items for this component.

Primary (applicable to bolted plant designs):

Core shroud assembly (bolted)

There are no expansion items for this component.

Note that the core shroud assembly (bolted) is examined in order to detect void swelling effects as evidenced by abnormal interaction with fuel assemblies, gaps along high fluence shroud plate joints, vertical displacement of shroud plates near high fluence joint.

Primary (applicable to all plants with instrument guide tubes in the control element assembly (CEA) shroud assembly):

Instrument guide tubes (peripheral)

Expand to:

9 Remaining instrument guide tubes within the CEA shroud assemblies Visual (VT-1 and EVT-1) Examinations Primary (applicable to plant designs with core shrouds assembled in two vertical sections):

Core shroud assembly (welded)

There are no expansion items for this component.

Note that the core shroud assembly (welded) is examined in order to detect void swelling effects as evidenced by separation between the upper and lower core shroud segments.

Primary (applicable to plant designs with core shrouds assembled in two vertical sections):

Core shroud plate-former plate weld Expands to:

9 Remaining axial welds Primary (applicable to plant designs with core shrouds assembled with full-height shroud plates)

Shroud plates Expand to:

• Remaining axial welds

" Ribs and rings Primary (applicable to all plants):

Upper (core support barrel) flange weld 4-12

Aging Management Requirements Expands to:

" Lower core support beams

" Core support barrel assembly upper cylinder (including welds)

Primary (applicable to all plants):

Core support barrel assembly lower cylinder girth welds Expands to:

• Core support barrel assembly lower cylinder axial welds Note that the core support barrel lower cylinder axial welds are not included as Primary components since they are subject to lower stresses than the girth welds and are thus less susceptible to stress corrosion cracking (SCC or IASCC).

Primary (applicable to all plants with core shrouds assembled with full-height shroud plates):

Deep beams There are no expansion items for this component.

Primary (depends on time-limited aging analysis [TLAA]):

Core support barrel assembly lower flange weld (applicable to all plants)

Core support plate (applicable to all plants with a core support plate)

Fuel alignment plate (applicable to all plants with core shrouds assembled with full-height shroud plates)

There are no expansion items for these components.

0 Volumetric (UT) Examination Primary (applicable to bolted plant designs):

Core shroud bolts Expand to:

• Core support column bolts

• Barrel-shroud bolts 4.3.3 Westinghouse Components Tables 4-3 and 4-6 describe the examination requirements for the PWR internals Primary and Expansion components for Westinghouse plants.

The following is a list of the Westinghouse Primary and Expansion components by examination technique.

Visual (VT-3) Examination Primary:

Baffle-former assembly (applicable to all plants) 4-13

Aging Management Requirements Thermal shield flexures (applicable to all plants with thermal shields)

Guide plates (cards) (applicable to all plants)

There are no expansion items for these components.

Note that the baffle-former assembly is examined in order to detect void swelling effects as evidenced by abnormal interaction with fuel assemblies, gaps along high fluence baffle joint, vertical displacement of baffle plates near high fluence joint, or broken or damaged edge bolt locking systems along high fluence baffle joint. Also note that the PWROG is conducting a guide card wear project.

Primary:

Baffle-edge bolts (applicable to all plants with baffle-edge bolts)

There are no expansion items for these components.

Note that the baffle-edge bolts are examined in order to detect lost or broken locking devices, failed or missing bolts, or protrusion of bolt heads.

0 Visual (VT-1 and EVT-1) Examinations Primary (applicable to all plants):

Upper core barrel flange weld Expands to:

• Lower support column bodies (non cast)

Primary (applicable to all plants):

Lower core barrel flange weld (alternatively designated as core barrel-to-support plate weld in some plant designs)

Upper and lower core barrel girth welds Expand to:

• Upper and lower core barrel axial welds Note that the upper and lower core barrel axial welds are not included as Primary components since they are subject to lower stresses than the girth welds and are thus less susceptible to stress corrosion cracking (SCC or IASCC).

Primary (applicable to all plants):

Control rod guide tube (CRGT) assembly lower flange welds Expand to:

o Bottom-mounted instrumentation (BMI) column bodies (these components receive a visual (VT-3) examination)

• Lower support column bodies (cast) o Upper core plate 4-14

Aging Management Requirements 0

Lower support forging or casting Note that the examination coverage is 100% of outer (accessible) CRGT lower flange weld surfaces and adjacent base metal.

Volumetric (UT) Examination Primary (applicable to all plants):

Baffle-former bolts Expand to:

" Lower support column bolts

" Barrel-former bolts

• Physical Measurement Primary (applicable to all plants with 304 stainless steel hold down springs):

Internals hold down spring There are no expansion items for this component.

4-15

Aging Management Requirements Table 4-1 B&W plants Primary components Item Applicability Effect (Mechanism)

Expansion Examination Examination Link (Note 2)

MethodlFrequency (Note 2)

Coverage Plenum Cover Assembly &

All plants Loss of material and None One-time physical Determination of Core Support Shield associated loss of measurement no later than differential height of top Assembly core clamping two refueling outages from of plenum rib pads to Plenum cover weldment rib pre-load (Wear) the beginning of the license reactor vessel seating pads renewal period, surface, with plenum in Plenum cover support flange reactor vessel.

CSS top flange Perform subsequent visual (VT-3) examination on the See Figure 4-1.

10-year ISI interval.

Control Rod Guide Tube All plants Cracking (TE),

None Visual (VT-3) examination Accessible surfaces at Assembly including the detection during the next 10-year ISI.

each of the 4 screw CRGT spacer castings of fractured spacers or Subsequent examinations on locations (at every 900) missing screws the 10-year ISI interval, of 100% of the CRGT spacer castings (limited accessibility).

See Figure 4-5.

Core Support Shield All plants Cracking (TE),

None Visual (VT-3) examination 100% of accessible Assembly including the during the next 10-year ISI.

surfaces CSS vent valve top retaining detection of surface (see BAW-2248A, page ring irregularities, such as Subsequent examinations on 4.3 and Table 4-1).

CSS vent valve bottom damaged, fractured the 10-year ISI interval.

retaining ring material, or missing See Figure 4-11.

(Note 1) items 4-16

Aging Management Requirements Table 4-1 B&W plants Primary components (continued)

Effect Examination Examination Item Applicability (Mechanism)

Expansion Link (Note 2)

MethodlFrequency (Note 2)

Coverage Core Support Shield All plants Bolts: Cracking UTS bolts and LTS Volumetric examination (UT) 100% of accessible Assembly (SCC) studs/nuts or bolts and of the bolts within two bolts and their Upper core barrel Locking their locking devices refueling outages from locking devices.

(UCB) bolts and their Devices: Loss 1/1/2006 or next 10-year ISI (Note 3) locking devices of material, SSHT studs/nuts or bolts interval, whichever is first.

damaged, and their locking devices See Figure 4-7.

distorted or (CR-3 and DB only)

Subsequent examination on missing the 10-year ISI interval unless locking Lower grid shock pad an evaluation of the baseline devices (Wear bolts and their locking results submitted for NRC or Fatigue devices (TMI-1 only) staff approval justifies a damage by longer interval between failed bolts).

examinations.

Visual (VT-3) examination of bolt locking devices on the 10-year ISI interval.

Core Barrel All plants Bolt: Cracking UTS bolts and LTS Volumetric examination (UT) 100% of accessible Assembly (SCC) studs/nuts or bolts and of the bolts during the next bolts and their Lower core barrel Locking their locking devices 10-year ISI interval from locking devices (LCB) bolts and their Devices: Loss 1/1/2006.

(Note 3) locking devices of material, SSHT studs/nuts or bolts

damaged, and their locking devices Subsequent examination on See Figure 4-8.

distorted or (CR-3 and DB only) the 10-year ISI interval unless missing an evaluation of the baseline locking Lower grid shock pad results submitted for NRC devices (Wear bolts and their locking staff approval justifies a or Fatigue devices (TMI-1 only) longer interval between damage by examinations.

failed bolts).

Visual (VT-3) examination of bolt locking devices on the 10-year ISI interval.

4-17

Aging Management Requirements Table 4-1 B&W plants Primary components (continued)

Examination Examination Item Applicability Effect (Mechanism)

Expansion Link (Note 2)

Method/Frequency (Note 2)

Coverage Core Barrel All plants Cracking (IASCC, IE, Baffle-to-baffle bolts, Baseline volumetric 100% of accessible Assembly Overload)

Core barrel-to-former bolts examination (UT) no later than bolts. (Note 3)

Baffle-to-former (Note 4) two refueling outages from the bolts beginning of the license See Figure 4-2.

renewal period with subsequent examination after 10 additional years.

Core Barrel All plants Cracking (IE),

Core barrel cylinder Visual (VT-3) examination 100% of the Assembly including the (including vertical and during the next 10-year ISI.

accessible surface Baffle plates detection of readily circumferential seam within 1 inch around detectable cracking welds),

Subsequent examinations on each flow and bolt in the baffle plates Former plates the 10-year ISI interval, hole.

See Figure 4-2.

Core Barrel All plants Cracking (IASCC, IE, Locking devices, including Visual (VT-3) examination 100% of accessible Assembly Overload), including locking welds, for the during the next 10-year ISI.

baffle-to-former and Locking devices, the detection of external baffle-to-baffle internal baffle-to-including locking missing, non-bolts and Core barrel-to-Subsequent examinations on baffle bolt locking welds, of baffle-to-functional, or former bolts the 10-year ISI interval, devices. (Note 3) former bolts and removed locking internal baffle-to-devices or welds See Figure 4-2.

baffle bolts 4-18

Aging Management Requirements Table 4-1 B&W plants Primary components (continued)

Effect Examination Examination Item Applicability (Mechanism)

Expansion Link (Note 2)

Method/Frequency (Note 2)

Coverage Flow Distributor All plants Bolt: Cracking UTS bolts and LTS Volumetric examination (UT) of 100% of Assembly (SCC) studs/nuts or bolts and their the bolts during the next 10-accessible bolts Flow distributor Locking Devices:

locking devices.

year ISI interval from 1/1/2006.

and their (FD) bolts and their Loss of material, SSHT studs/nuts or bolts Subsequent examination on locking devices.

locking devices damaged or and their locking devices the 10-year ISI interval unless (Note 3) distorted or missing (CR-3 and DB only) an evaluation of the baseline locking devices Lower grid shock pad bolts results, submitted for NRC staff See Figure 4-8.

(Wear or Fatigue and their locking devices approval, justifies a longer damage by failed (TMI-1 only) interval between examinations.

bolts).

Visual (VT-3) examination of bolt locking devices on the 10-year ISI interval.

Lower Grid All plants Cracking (SCC),

Alloy X-750 dowel locking Initial visual (VT-3)

Accessible Assembly including the welds to the upper and examination no later than two surfaces of Alloy X-750 dowel-detection of lower grid fuel assembly refueling outages from the 100% of the 24 to-guide block separated or support pads.

beginning of the license dowel-to-guide welds missing locking renewal period, block welds.

welds, or missing dowels Subsequent examinations on See Figure 4-4.

the 10-year ISI interval.

Incore Monitoring All plants Cracking (TE/IE),

Lower grid fuel assembly Initial visual (VT-3) 100% of top Instrumentation including the support pad items: pad, examination no later than two surfaces of 52 (IMI) Guide Tube detection of pad-to-rib section welds, refueling outages from the spider castings Assembly fractured or missing Alloy X-750 dowel, cap beginning of the license and welds to IMI guide tube spider arms or, screw, and their locking renewal period, the adjacent spiders Cracking (IE),

welds lower grid rib IMI guide tube including separation (Note: the pads, dowels, Subsequent examinations on section.

spider-to-lower grid of spider arms from and cap screws are the 10-year ISI interval.

rib section welds the lower grid rib included because of IE of See Figures 4-3 section at the weld the welds) and 4-6.

4-19

Aging Management Requirements Notes to Table 4-1:

1.

A verfication of the operation of each vent valve shall also be performed through manual actuation of the valve. Verify that the valves are not stuck in the open position and that no abnormal degradation has occurred. Examine the valves for evidence of scratches, pitting, embedded particles, leakage of the seating surfaces, cracking of lock welds and locking cups, jack screws for proper position, and wear. The frequency is defined in each unit's technical specifications or in their pump and valve inservice test programs (see BAW-2248A, page 4-3 and Table 4-1[18]).

2.

Examination acceptance criteria and expansion criteria for the B&W components are in Table 5-1.

3.

A minimum of 75% of the total population (examined + unexamined), including coverage consistent with the Expansion criteria in Table 5-1, must be examined for inspection credit.

4.

The primary aging degradation mechanisms for loss of joint tightness for this item are IC and ISR. Fatigue and Wear, which can also lead to cracking, are secondary aging degradation mechanisms after significant stress relaxation and loss of preload has occurred due to IC/ISR. Bolt stress relaxation cannot readily be inspected by NDE. Only bolt cracking is inspected by UT inspection. The effect of loss of joint tightness on the functionality will be addressed by analysis of the core barrel assembly, which will be performed to address Applicant/Licensee action item 6 in the SE [27].

4-20

Aging Management Requirements Table 4-2 CE plants Primary components Effect Expansion Examination Examination Item Applicability (Mechanism)

Link (Note 1)

MethodlFrequency (Note 1)

Coverage Core Shroud Assembly Bolted plant Cracking (IASCC, Core support Baseline volumetric (UT) 100% of accessible (Bolted) designs Fatigue) column bolts, examination between 25 and bolts (see Note 3).

Core shroud bolts Aging Management Barrel-shroud 35 EFPY, with subsequent Heads are (IE and ISR) bolts examination on a ten-year accessible from the (Note 2)

interval, core side. UT accessibility may be affected by complexity of head and locking device designs.

See Figure 4-24.

Core Shroud Assembly Plant designs Cracking (IASCC)

Remaining Enhanced visual (EVT-1)

Axial and horizontal (Welded) with core Aging Management axial welds examination no later than 2 weld seams at the Core shroud plate-former shrouds (lE) refueling outages from the core shroud re-plate weld assembled in (Note 2) beginning of the license entrant corners as two vertical renewal period and visible from the core sections subsequent examination on a side of the shroud, ten-year interval, within six inches of central flange and horizontal stiffeners.

See Figures 4-12 and 4-14.

Core Shroud Assembly Plant designs Cracking (IASCC)

Remaining Enhanced visual (EVT-1)

Axial weld seams at (Welded) with core Aging Management axial welds, examination no later than 2 the core shroud re-Shroud plates shrouds (IE)

Ribs and refueling outages from the entrant corners, at assembled (Note 2) rings beginning of the license the core mid-plane with full-height renewal period and

(+/- three feet in shroud plates subsequent examination on a height) as visible ten-year interval, from the core side of the shroud.

See Figure 4-13.

4-21

Aging Management Requirements Table 4-2 CE plants Primary components (continued)

Effect Expansion Examination Examination Item Applicability (Mechanism)

Link (Note 1)

Method/Frequency (Note 1)

Coverage Core Shroud Assembly Bolted plant Distortion None Visual (VT-3) examination no Core side surfaces as (Bolted) designs (Void Swelling),

later than 2 refueling outages indicated.

Assembly including:

from the beginning of the

• Abnormal license renewal period.

See Figures 4-25 and interaction with Subsequent examinations on 4-26.

fuel assemblies a ten-year interval.

• Gaps along high fluence shroud plate joints

• Vertical displacement of shroud plates near high fluence joint Aging Management (IE)

Core Shroud Assembly Plant designs Distortion None Visual (VT-1) examination no If a gap exists, make (Welded) with core (Void Swelling), as later than 2 refueling outages three to five Assembly shrouds evidenced by from the beginning of the measurements of gap assembled in separation license renewal period, opening from the core two vertical between the upper Subsequent examinations on side at the core shroud sections and lower core a ten-year interval, re-entrant corners.

shroud segments Then, evaluate the Aging Management swelling on a plant-(IE) specific basis to determine frequency and method for additional examinations.

See Figures 4-12 and 4-14.

4-22

Aging Management Requirements Table 4-2 CE plants Primary components (continued)

Effect Expansion Examination Examination Item Applicability (Mechanism)

Link (Note 1)

Method/Frequency (Note 1)

Coverage Core Support Barrel All plants Cracking (SCC)

Lower core Enhanced visual (EVT-1) 100% of the Assembly support examination no later than 2 accessible surfaces Upper (core support barrel) beams refueling outages from the of the upper flange flange weld Core support beginning of the license weld (Note 4).

barrel renewal period. Subsequent assembly examinations on a ten-year See Figure 4-15.

upper interval.

cylinder Upper core barrel flange Core Support Barrel All plants Cracking (SCC, Lower Enhanced visual (EVT-1) 100% of the Assembly IASCC) cylinder axial examination no later than 2 accessible surfaces Lower cylinder girth welds Aging Management welds refueling outages from the of the lower cylinder (IE) beginning of the license welds (Note 4).

renewal period. Subsequent examinations on a ten-year See Figure 4-15 interval.

Lower Support Structure All plants Cracking (SCC, None Visual (VT-3) examination no 100% of the Core support column welds IASCC, Fatigue later than 2 refueling outages accessible surfaces including damaged from the beginning of the of the core support or fractured license renewal period, column welds (Note material)

Subsequent examinations on 5).

Aging Management a ten-year interval.

See Figures 4-16 (IE, TE) and 4-31 4-23

Aging Management Requirements Table 4-2 CE plants Primary components (continued)

Effect Expansion Examination Examination Item Applicability (Mechanism)

Link (Note 1)

MethodlFrequency (Note 1)

Coverage Core Support Barrel All plants Cracking (Fatigue)

None If fatigue life cannot be Examination coverage Assembly demonstrated by time-limited to be defined by Lower flange weld aging analysis (TLAA),

evaluation to determine enhanced visual (EVT-1) the potential location examination, no later than 2 and extent of fatigue refueling outages from the cracking.

beginning of the license renewal period. Subsequent See Figures 4-15 and examination on a ten-year 4-16.

interval.

Lower Support Structure All plants with Cracking (Fatigue)

None If fatigue life cannot be Examination coverage Core support plate a core support Aging Management demonstrated by time-limited to be defined by plate (IE) aging analysis (TLAA),

evaluation to determine enhanced visual (EVT-1) the potential location examination, no later than 2 and extent of fatigue refueling outages from the cracking.

beginning of the license renewal period. Subsequent See Figure 4-16.

examination on a ten-year interval.

Upper Internals Assembly All plants with Cracking (Fatigue)

None If fatigue life cannot be Examination coverage Fuel alignment plate core shrouds demonstrated by time-limited to be defined by assembled aging analysis (TLAA),

evaluation to determine with full-height enhanced visual (EVT-1) the potential location shroud plates examination, no later than 2 and extent of fatigue refueling outages from the cracking.

beginning of the license renewal period. Subsequent See Figure 4-17.

examination on a ten-year interval.

4-24

Aging Management Requirements Table 4-2 CE plants Primary components (continued)

Effect Expansion Examination Examination Item Applicability (Mechanism)

Link (Note 1)

Method/Frequency (Note 1)

Coverage Control Element Assembly All plants with Cracking (SCC, Remaining Visual (VT-3) examination, 100% of tubes in Instrument guide tubes instrument Fatigue) that instrument no later than 2 refueling peripheral CEA guide tubes in results in missing guide tubes outages from the beginning shroud assemblies the CEA supports or within the of the license renewal period.

(i.e., those adjacent shroud separation at the CEA shroud Subsequent examination on to the perimeter of assembly welded joint assemblies a ten-year interval, the fuel alignment between the tubes plate).

and supports Plant-specific component integrity assessments may See Figure 4-18.

be required if degradation is detected and remedial action is needed.

Lower Support Structure All plants with Cracking (Fatigue)

None Enhanced visual (EVT-1)

Examine beam-to-Deep beams core shrouds that results in a examination, no later than 2 beam welds, in the assembled detectable surface-refueling outages from the axial elevation from with full-height breaking indication beginning of the license the beam top shroud plates in the welds or renewal period. Subsequent surface to four beams examination on a ten-year inches below.

Aging Management interval, if adequacy of (IE) remaining fatigue life cannot See Figure 4-19.

be demonstrated.

Note to Table 4-2:

1.

Examination acceptance criteria and expansion criteria for the CE components are in Table 5-2.

2.

Void swelling effects on this component is managed through management of void swelling on the entire core shroud assembly.

3.

A minimum of 75% of the total population (examined + unexamined), including coverage consistent with the Expansion criteria in Table 5-2, must be examined for inspection credit.

4.

A minimum of 75% of the total weld length (examined + unexamined), including coverage consistent with the Expansion criteria in Table 5-2, must be examined from either the inner or outer diameter for inspection credit.

5.

A minimum of 75% of the total population of core support column welds.

4-25

Aging Management Requirements Table 4-3 Westinghouse plants Primary components Item Applicability Effect Expansion Link Examination (Mechanism)

(Note 1)

Method/Frequency (Note 1)

Examination Coverage Control Rod Guide Tube All plants Loss of Material None Visual (VT-3) examination no 20% examination of the Assembly (Wear) later than 2 refueling outages number of CRGT Guide plates (cards) from the beginning of the assemblies, with all guide license renewal period, and cards within each no earlier than two refueling selected CRGT assembly outages prior to the start of examined.

the license renewal period.

Subsequent examinations See Figure 4-20 are required on a ten-year interval.

Control Rod Guide Tube All plants Cracking (SCC, Bottom-mounted Enhanced visual (EVT-1) 100% of outer Assembly Fatigue) instrumentation examination to determine the (accessible) CRGT lower Lower flange welds Aging (BMI) column bodies, presence of crack-like flange weld surfaces and Management (IE Lower support surface flaws in flange welds adjacent base metal on and TE) column bodies (cast) no later than 2 refueling the individual periphery Upper core plate outages from the beginning CRGT assemblies.

Lower support of the license renewal period (Note 2) forging/casting and subsequent examination on a ten-year interval.

See Figure 4-21.

Core Barrel Assembly All plants Cracking (SCC)

Lower support Periodic enhanced visual 100% of one side of the Upper core barrel flange column bodies (non (EVT-1) examination, no later accessible surfaces of weld cast) than 2 refueling outages from the selected weld and Core barrel outlet the beginning of the license adjacent base metal nozzle welds renewal period and (Note 4).

subsequent examination on a ten-year interval.

See Figure 4-22.

Core Barrel Assembly All plants Cracking (SCC, Upper and lower Periodic enhanced visual 100% of one side of the Upper and lower core IASCC, Fatigue) core barrel cylinder (EVT-1) examination, no later accessible surfaces of barrel cylinder girth welds axial welds than 2 refueling outages from the selected weld and the beginning of the license adjacent base metal renewal period and (Note 4).

subsequent examination on a I ten-year interval.

I See Figure 4-22 4-26

Aging Management Requirements Table 4-3 Westinghouse plants Primary components (continued)

Item Applicability Effect Expansion Link Examination Examination ItemApplicability_(Mechanism)

(Note 1)

MethodlFrequency (Note 1)

Coverage Core Barrel Assembly All plants Cracking (SCC, None Periodic enhanced visual 100% of one side of Lower core barrel flange Fatigue)

(EVT-1) examination, no later the accessible weld (Note 5) than 2 refueling outages from surfaces of the the beginning of the license selected weld and renewal period and adjacent base metal subsequent examination on a (Note 4).

ten-year interval.

Baffle-Former Assembly All plants with Cracking (IASCC, None Visual (VT-3) examination, Bolts and locking Baffle-edge bolts baffle-edge Fatigue) that with baseline examination devices on high bolts results in between 20 and 40 EFPY fluence seams. 100%

• Lost or broken and subsequent of components locking devices examinations on a ten-year accessible from core

• Failed or

interval, side missing bolts (Note 3).

• Protrusion of See Figure 4-23.

bolt heads Aging Management (IE and ISR)

(Note 6)

Baffle-Former Assembly All plants Cracking (IASCC, Lower support Baseline volumetric (UT) 100% of accessible Baffle-former bolts Fatigue) column bolts, examination between 25 and bolts (Note 3). Heads Aging Barrel-former bolts 35 EFPY, with subsequent accessible from the Management (IE examination on a ten-year core side. UT and ISR)

interval, accessibility may be (Note 6) affected by complexity of head and locking device designs.

See Figures 4-23 and 1

1_

14-24.

4-27

Aging Management Requirements Table 4-3 Westinghouse plants Primary components (continued)

Expansion Examination Item Applicability Effect (Mechanism)

Link (Note 1) Method/Frequency (Note 1)

Examination Coverage Baffle-Former Assembly All plants Distortion (Void None Visual (VT-3) examination to Core side surface as Assembly Swelling), or check for evidence of indicated.

(Includes: Baffle plates, baffle Cracking (IASCC) distortion, with baseline edge bolts and indirect effects that results in examination between 20 and See Figures 4-24, 4-25, of void swelling in former

• Abnormal 40 EFPY and subsequent 4-26 and 4-27.

plates) interaction with fuel examinations on a ten-year assemblies interval.

eGaps along high fluence baffle joint eVertical displacement of baffle plates near high fluence joint eBroken or damaged edge bolt locking systems along high fluence baffle joint Alignment and Interfacing All plants with Distortion (Loss of None Direct measurement of spring Measurements should be Components 304 stainless Load) height within three cycles of taken at several points Internals hold down spring steel hold down the beginning of the license around the circumference springs Note: This renewal period. If the first set of the spring, with a mechanism was not of measurements is not statistically adequate strictly identified in sufficient to determine life, number of measurements the original list of spring height measurements at each point to minimize age-related must be taken during the next uncertainty.

degradation two outages, in order to mechanisms [7].

extrapolate the expected See Figure 4-28.

1 spring height to 60 years.

I

_I 4-28

Aging Management Requirements Table 4-3 Westinghouse plants Primary components (continued)

Effect Expansion Link Examination Item Applicability (Mechanism)

(Note 1)

Method/Frequency (Note 1)

Examination Coverage Thermal Shield Assembly All plants with Cracking None Visual (VT-3) no later than 2 100% of thermal shield Thermal shield flexures thermal shields (Fatigue) refueling outages from the flexures.

or Loss of beginning of the license Material (Wear) renewal period. Subsequent See Figures 4-29 and 4-that results in examinations on a ten-year 36.

thermal shield interval.

flexures excessive wear, fracture, or complete I separation I

Notes to Table 4-3:

1.

Examination acceptance criteria and expansion criteria for the Westinghouse components are in Table 5-3.

2.

A minimum of 75% of the total identified sample population must be examined.

3.

A minimum of 75% of the total population (examined + unexamined), including coverage consistent with the Expansion criteria in Table 5-3, must be examined for inspection credit.

4.

A minimum of 75% of the total weld length (examined + unexamined), including coverage consistent with the Expansion criteria in Table 5-3, must be examined from either the inner or outer diameter for inspection credit.

5.

The lower core barrel flange weld may be alternatively designated as the core barrel-to-support plate weld in some Westinghouse plant designs.

6.

Void swelling effects on this component is managed through management of void swelling on the entire baffle-former assembly.

4-29

Aging Management Requirements Table 4-4 B&W plants Expansion components Examination Effect Primary LinkExmnto Item Applicability (Mechanism)

(Note 1)

Method/Frequency Examination Coverage (Note 1)

Upper Grid Assembly All plants Cracking (SCC),

Alloy X-750 Visual (VT-3) examination.

Accessible surfaces of Alloy X-750 dowel-to-upper (except DB) including the dowel-to-Subsequent examinations on 100% of the dowel grid fuel assembly support detection of guide block the 10-year ISI interval unless locking welds.

pad welds separated or welds an applicant/licensee provides missing locking an evaluation for NRC staff See Figure 4-6 welds, or missing approval that justifies a longer dowels interval between inspections.

(i.e., these are similar to the lower grid fuel assembly support pads).

Core Barrel Assembly All plants Bolt or Stud/Nut:

UCB, LCB or Bolt or Stud/Nut: Volumetric 100% of accessible bolts Upper thermal shield (UTS)

Cracking (SCC)

FD bolts and examination (UT).

or studs/nuts and their bolts and their locking devices Locking Devices:

their locking Locking Devices: Visual (VT-3) locking devices (Note 2).

Loss of material, devices examination Core Barrel Assembly CR-3, DB

damaged, Subsequent examinations on See Figure 4-7.

Surveillance specimen holder distorted or the 10-year ISI interval unless tube (SSHT) studs/nuts (CR-missing locking an applicant/licensee provides

3) or bolts (DB) and their devices (Wear or an evaluation for NRC staff locking devices Fatigue damage approval that justifies a longer by failed bolts).

interval between inspections.

Core Barrel Assembly All plants Cracking (IE),

Baffle plates No examination requirements.

Inaccessible.

Core barrel cylinder (including including readily Justify by evaluation or by vertical and circumferential detectable replacement.

See Figure 4-2.

seam welds) cracking Former plates 4-30

Aging Management Requirements Table 4-4 B&W plants Expansion components (continued)

Effect Primary Link Examination Item Applicability MethodlFrequency Examination Coverage y

(Mechanism)

(Note 1)

(Note 1)

Core Barrel Assembly All plants Cracking (IASCC, Baffle-to-former Internal baffle-to-baffle An acceptable Baffle-to-baffle bolts IE, Overload) bolts bolts:

examination technique Core barrel-to-former bolts (Note 3)

No examination currently not available.

requirements, Justify by evaluation or See Figure 4-2.

by replacement.

External baffle-to-baffle Inaccessible.

bolts, core barrel-to-former bolts: No See Figure 4-2.

examination requirements.

Justify by evaluation or by replacement.

Core Barrel Assembly All plants Cracking (IASCC, Locking devices, No examination Inaccessible.

Locking devices, including IE) including locking requirements.

locking welds, for the external welds, of baffle-Justify by evaluation or See Figure 4-2.

baffle-to-baffle bolts and core to-former bolts or by replacement.

barrel-to-former bolts internal baffle-to-baffle bolts Lower Grid Assembly All plants Cracking (IE),

IMI guide tube Visual (VT-3)

Accessible surfaces of Lower grid fuel assembly including the spiders and examination, the pads, dowels, and support pad items: pad, pad-detection of spider-to-lower Subsequent cap screws, and to-rib section welds, Alloy X-separated or grid rib section examinations on the 10-associated welds in 750 dowel, cap screw, and missing welds, welds year ISI interval unless 100% of the lower grid their locking welds missing support an applicant/licensee fuel assembly support (Note: the pads, dowels and pads, dowels, cap provides an evaluation for pads.

cap scres pare dowuelsd screws and NRC staff approval that cap screws are included locking welds, or justifies a longer interval See Figure 4-6.

because of IE of the welds) misalignment of between inspections.

the support pads 4-31

Aging Management Requirements Table 4-4 B&W plants Expansion components (continued)

Effect Primary Link Examination Item Applicability Method/Frequency Examination Coverage y (Mechanism)

(Note 1)

(Note 1)

Lower Grid Assembly All plants Cracking (SCC),

Alloy X-750 Visual (VT-3) examination.

Accessible surfaces of Alloy X-750 dowel-to-lower including the dowel-to-guide Subsequent examinations 100% of the support pad grid fuel assembly support detection of block welds on the 10-year ISI interval dowel locking welds.

pad welds separated or unless an applicant/licensee missing locking provides an evaluation for See Figure 4-6.

welds, or missing NRC staff approval that dowels justifies a longer interval between inspections.

Lower Grid Assembly TMI-1 Bolts: Cracking UCB, LCB or FD Bolt: Volumetric examination 100% of accessible bolts Lower grid shock pad bolts (SCC) bolts and their (UT).

and their locking and their locking devices Locking Devices:

locking devices Locking Devices: Visual (VT-devices. (Note 2)

Loss of material,

3) examination.

damaged, Subsequent examinations See Figure 4-4.

distorted or on the 10-year ISI interval missing locking unless an applicant/licensee devices (Wear or provides an evaluation for Fatigue damage NRC staff approval that by failed bolts).

justifies a longer interval between inspections.

Lower Grid Assembly All plants Bolts or UCB,LCB or FD Bolt or Stud/Nut: Volumetric 100% of accessible bolts Lower thermal shield (LTS)

Studs/Nuts:

bolts and their examination (UT).

and their locking bolts (ANO-1, DB and TMI-1)

Cracking (SCC) locking devices Locking Devices: Visual (VT-devices. (Note 2) or studs/nuts (ONS, CR-3)

Locking Devices:

3) examination.

and their locking devices Loss of material, Subsequent examinations See Figure 4-8.

damaged, on the 10-year ISI interval distorted or unless an applicant/licensee missing locking provides an evaluation for devices (Wear or NRC staff approval that Fatigue damage justifies a longer interval by failed bolts).

between inspections.

4-32

Aging Management Requirements Notes to Table 4-4:

1.

Examination acceptance criteria and expansion criteria for the B&W components are in Table 5-1.

2.

A minimum of 75% of the total population (examined + unexamined) must be examined for inspection credit.

3.

The primary aging degradation mechanisms for loss of joint tightness for these items are IC and ISR. Fatigue and Wear, which can also lead to cracking, are secondary aging degradation mechanisms after significant stress relaxation and loss of preload has occurred due to ICIISR. Bolt stress relaxation cannot readily be inspected by NDE. Only bolt cracking could be inspected by UT inspection if it were possible for these bolts. Therefore, the effects of loss of joint tightness and/or cracking on the functionality of these bolts relative to the entire core barrel assembly will be addressed by analysis of the core barrel assembly, which will be performed to address Applicant/Licensee action item 6 of the SE [27].

4-33

Aging Management Requirements Table 4-5 CE plants Expansion components Effect Primary Link Examination Item Applicability (Mechanism)

(Note 1)

Method/Frequency Examination Coverage (Note 1)

Core Shroud Assembly Bolted plant Cracking Core shroud Volumetric (UT) 100% (or as supported (Bolted) designs (IASCC, bolts examination, by plant-specific Barrel-shroud bolts Fatigue)

Re-inspection every 10 justification; Note 2) of Aging years following initial barrel-shroud and guide Management (IE inspection, lug insert bolts with and ISR) neutron fluence exposures > 3 displacements per atom (dpa).

See Figure 4-23.

Core Support Barrel All plants Cracking (SCC, Upper (core Enhanced visual (EVT-1) 100% of accessible Assembly Fatigue) support barrel) examination, welds and adjacent Lower core barrel flange flange weld Re-inspection every 10 base metal (Note 2).

years following initial inspection.

See Figure 4-15.

Core Support Barrel All plants Cracking (SCC)

Upper (core Enhanced visual (EVT-1) 100% of accessible Assembly Aging support barrel) examination.

surfaces of the welds Upper cylinder (including Management flange weld Re-inspection every 10 and adjacent base welds)

(IE) years following initial metal (Note 2).

inspection.

See Figure 4-15.

Core Support Barrel All plants Cracking (SCC)

Upper (core Enhanced visual (EVT-1) 100% of accessible Assembly support barrel) examination.

bottom surface of the Upper core barrel flange flange weld Re-inspection every 10 flange (Note 2).

years following initial inspection.

See Figure 4-15.

4-34

Aging Management Requirements Table 4-5 CE plants Expansion components (continued)

Effect Primary Link Examination Item Applicability (Mechanism)

(Note 1)

MethodlFrequency Examination Coverage (Note 1)

Core Support Barrel All plants Cracking (SCC)

Core barrel Enhanced visual (EVT-1) 100% of one side of the Assembly assembly girth examination, with initial accessible weld and Core barrel assembly axial welds and subsequent adjacent base metal welds examinations dependent surfaces for the weld on the results of core with the highest barrel assembly girth calculated operating weld examinations, stress.

See Figure 4-15.

Lower Support Structure All plants Cracking (SCC, Upper (core Visual (EVT-1) 100% of accessible Lower core support beams except those Fatigue) support barrel) examination, surfaces (Note 2).

with core including flange weld Re-inspection every 10 shrouds damaged or years following initial See Figures 4-16 and assembled fractured inspection.

4-31.

with full-height material shroud plates Core Shroud Assembly Bolted plant Cracking Core shroud Ultrasonic (UT) 100% (or as supported (Bolted) designs (IASCC, bolts examination, by plant-specific Core support column bolts Fatigue)

Re-inspection every 10 analysis) of core support Aging years following initial column bolts with Management inspection, neutron fluence (IE) exposures > 3 dpa (Note 2).

See Figures 4-16 and 4-33.

4-35

Aging Management Requirements Table 4-5 CE plants Expansion components (continued)

Effect Primary Link Examination Item Applicability (Mechanism)

(Note 1)

Method/Frequency Examination Coverage (Note 1)

Core Shroud Assembly Plant designs Cracking Shroud plates of Enhanced visual (EVT-I)

Axial weld seams other (Welded) with core (IASCC) welded core examination, than the core shroud re-Remaining axial welds, shrouds Aging shroud Re-inspection every 10 entrant corner welds at assembled Management assemblies years following initial the core mid-plane, plus Ribs and rings with full-height (lE) inspection, ribs and rings.

shroud plates See Figure 4-13.

Control Element Assembly All plants with Cracking (SCC, Peripheral Visual (VT-3) examination.

100% of tubes in CEA Remaining instrument guide instrument Fatigue) that instrument guide Re-inspection every 10 shroud assemblies tubes guide tubes in results in tubes within the years following initial (Note 2).

the CEA missing supports CEA shroud inspection.

shroud or separation at assemblies inspection.

assembly the welded joint See Figure 4-18.

between the tubes and supports.

Notes to Table 4-5:

1.

Examination acceptance criteria and expansion criteria for the CE components are in Table 5-2.

2.

A minimum of 75% coverage of the entire examination area or volume, or a minimum sample size of 75% of the total population of like components of the examination is required (including both the accessible and inaccessible portions).

4-36

Aging Management Requirements Table 4-6 Westinghouse plants Expansion components Effect Primary Link Examination Item Applicability MethodlFrequency Examination Coverage y

(Mechanism)

(Note 1)

(Note 1)

Upper Internals Assembly All plants Cracking CRGT lower Enhanced visual (EVT-100% of accessible Upper core plate (Fatigue, Wear) flange weld

1) examination, surfaces (Note 2).

Re-inspection every 10 years following initial inspection.

Lower Internals Assembly All plants Cracking CRGT lower Enhanced visual (EVT-100% of accessible Lower support forging or Aging flange weld

1) examination, surfaces (Note 2).

castings Management (TE Re-inspection every 10 in Casting) years following initial See Figure 4-33.

inspection.

Core Barrel Assembly All plants Cracking Baffle-former Volumetric (UT) 100% of accessible Barrel-former bolts (IASCC, Fatigue) bolts examination, bolts. Accessibility may Aging Re-inspection every 10 be limited by presence Management (IE, years following initial of thermal shields or Void Swelling inspection, neutron pads (Note 2).

and ISR)

See Figure 4-23.

Lower Support Assembly All plants Cracking Baffle-former Volumetric (UT) 100% of accessible Lower support column bolts (IASCC, Fatigue) bolts examination, bolts or as supported by Aging Re-inspection every 10 plant-specific Management (IE years following initial justification (Note 2).

and ISR) inspection.

See Figures 4-32 and 4-33.

4-37

Aging Management Requirements Table 4-6 Westinghouse plants Expansion components (continued)

Effect Primary Link Examination Item Applicability (Mechanism)

(Note 1)

Method/Frequency Examination Coverage (Note 1)

Core Barrel Assembly All plants Cracking (SCC, Upper core Enhanced visual (EVT-1) 100% of one side of the Core barrel outlet nozzle Fatigue) barrel flange examination, accessible surfaces of welds Aging weld Re-inspection every 10 the selected weld and Management (IE years following initial adjacent base metal of lower inspection.

(Note 2).

sections)

See Figure 4-22.

Core Barrel Assembly All plants Cracking (SCC, Upper and Enhanced visual (EVT-1) 100% of one side of the Upper and lower core barrel IASCC) lower core examination, accessible surfaces of cylinder axial welds Aging barrel cylinder Re-inspection every 10 the selected weld and Management girth welds years following initial adjacent base metal (IE) inspection.

(Note 2).

See Figure 4-22.

Lower Support Assembly All plants Cracking Upper core Enhanced visual (EVT-1) 100% of accessible Lower support column bodies (IASCC) barrel flange examination, surfaces (Note 2).

(non cast)

Aging weld Re-inspection every 10 Management years following initial See Figure 4-34.

(IE) inspection.

Lower Support Assembly All plants Cracking Control rod Visual (EVT-1) examination.

100% of accessible Lower support column bodies (IASCC) guide tube Re-inspection every 10 support columns (Note (cast) including the (CRGT) lower years following initial 2).

detection of flanges inspection.

fractured See Figure 4-34.

support columns Aging Management (IE) 4-38

Aging Management Requirements Table 4-6 Westinghouse plants Expansion components (continued)

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

(Note 1)

(Note 1)

Bottom Mounted All plants Cracking Control rod Visual (VT-3) 100% of BMI column Instrumentation System (Fatigue) guide tube examination of BMI bodies for which Bottom-mounted including the (CRGT) lower column bodies as difficulty is detected instrumentation (BMI) column detection of flanges indicated by difficulty of during flux thimble bodies completely insertion/withdrawal of insertion/withdrawal.

fractured column flux thimbles.

bodies Re-inspection every 10 See Figure 4-35.

Aging years following initial Management inspection.

(IE)

Flux thimble insertion/withdrawal to be monitored at each inspection interval.

Notes to Table 4-6:

1.

Examination acceptance criteria and expansion criteria for the Westinghouse components are in Table 5-3.

2.

A minimum of 75% coverage of the entire examination area or volume, or a minimum sample size of 75% of the total population of like components of the examination is required (including both the accessible and inaccessible portions).

4-39

Aging Management Requirements Figure 4-1 Typical upper internals arrangement for B&W-designed PWRs 4-40

Aging Management Requirements Core Barrel-to-Former Bolt and Baffle-to-Former Bolt Locations Figure 4-2 Typical internals core barrel assembly for B&W-designed PWRs 4-41

Aging Management Requirements Figure 4-3 Typical lower internals arrangement for B&W-designed PWRs 4-42

Aging Management Requirements Figure 4-4 Typical guide block and shock pad locations for B&W-designed PWRs 4-43

Aging Management Requirements Figure 4-5 Typical control rod guide tube (CRGT) for B&W-designed PWRs (one of 69 CRGTs shown) 4-44

Aging Management Requirements Figure 4-6 Typical lower grid assembly and fuel assembly support pads for B&W-designed PWRs 4-45

Aging Management Requirements Figure 4-7 Typical upper thermal shield bolts and upper core barrel bolts for B&W-designed PWRs 4-46

Aging Management Requirements Figure 4-8 Typical lower thermal shield studslnuts, lower core barrel bolts, and flow distributor bolts for the B&W-designed PWRs 4-47

Aging Management Requirements Figure 4-9 Typical core support shield (CSS) outlet nozzle for the B&W-designed PWRs 4-48

Aging Management Requirements CSS Vent Valve Journal Receptacle Welded Cover Plate Figure 4-10 Typical core support shield (CSS) vent valve - outside view - for the B&W-designed PWRs 4-49

Aging Management Requirements Figure 4-11 Typical core support shield (CSS) vent valve - inside view - for the B&W-designed PWRs 4-50

Aging Management Requirements Weld locations potentially affected by swelling in horizontal stiffeners Core shroud plate-former plate weld locations with stresses potentially above IASCC threshold.

Weld locations potentially affected by swelling in horizontal stiffeners Figure 4-12 Potential crack locations for CE welded core shroud assembled in stacked sections 4-51

Aging Management Requirements Guide Lug Top Plate Ring Brace Rib Bottom Plate Figure 4-13 CE welded core shroud with full height panels 4-52

Aging Management Requirements 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 4-53

Aging Management Requirements

.Flange Weld

ý0 Axial Weld Upper Core Barrel to Lower Core Barrel Circumferential Weld Lower Barrel Axial Weld Lower Barrel Circumferential Weld Lower Barrel Axial Weld Core Barrel to Support Plate Weld Figure 4-15 Typical CE core support barrel structure 4-54

Aging Management Requirements 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 4-55

Aging Management Requirements (a)

UGS SUPPORT PLATE A

FD CEA GUIDE TUBES E

CEA GUIDE TUBE ALIGNMENT EXTENSIONS PLATE (b)

Figure 4-17 (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 4-56

Aging Management Requirements r// I I II-A.

E

-E I

I I

I I

I I

I 44W I I

I I

I I

I I

I I'1 II

-C

-TI r

AIP I

I, iit*ciu I *i~~~~IL*J I

I I

I I

I I

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 4-57

Aging Management Requirements Illustrates the deep beam grid structure (number 3), as well as the fuel alignment pins (numbers I 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 4-58

Aging Management Requirements Wear Area Figure 4-20 Typical Westinghouse control rod guide card (17x17 fuel assembly) 4-59

Aging Management Requirements Lower Flange Welds Figure 4-21 Typical Westinghouse control rod guide tube assembly 4-60

Aging Management Requirements Flange Weld Upper Core Barrel to Lower Core Barrel Circumferential Weld Lower Barrel Circumferential Weld Core Barrel to Support Plate Weld Figure 4-22 Major fabrication welds in typical Westinghouse core barrel 4-61

Aging Management Requirements SAPFR.E 10 FoRMER BOLT(LOSS&.103 COAHER EIDE BRACKET DAFFLE TO FORMER BOLT Figure 4-23 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 barrel-shroud bolts are equivalent to barrel-former bolts 4-62

Aging Management Requirements L1~

00000 V.,00000 0000 0

0000 0000 00000 0000 0

0,0 0'

000 Figur 4-2 Bafl-eg bltad afl-frmr ot octon a ig fuec sam n ole bffe fome asebl noe:eqivletbafl-frmr ot octinsinbltd E hru desins ae coe shoud 0olts 4-63

Aging Management Requirements High Fluence Seams Figure 4-25 High fluence seam locations in Westinghouse baffle-former assembly (full axial length of each of the re-entrant baffle plate corners) 4-64

Aging Management Requirements Potential Gaps at Bafflo-Formcr Plate LecVCS Figure 4-26 Exaggerated view of void swelling induced distortion in Westinghouse baffle-former assembly. This figure also applies to bolted CE shroud designs 4-65

Aging Management Requirements Vertical Displacement Figure 4-27 Vertical displacement of Westinghouse baffle plates caused by void swelling. This figure also applies to bolted CE shroud designs 4-66

Aging Management Requirements TOP SUPPORT PLATE CORE BARREL Figure 4-28 Schematic cross-sections of the Westinghouse hold-down springs 4-67

Aging Management Requirements Core Barrel Thermal Shield Flexure Thermal Shield Core Support Figure 4-29 Location of Westinghouse thermal shield flexures 4-68

Aging Management Requirements IICore Support 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)

-J a) b)

c) a)

Early support column design b)

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

Later support column design used in plants with second generation core support assemblies Figure 4-31 CE core support columns 4-69

Aging Management Requirements er Core Plate T

M Lower Core Support Structure Core Supporti o

Plate (Forging)

Figure 4-32 Schematic indicating location of Westinghouse lower core support structure. Additional details shown in Figure 4-33 LOWER CORE PLATE DIFFUSER PLATE CORE SUPPORT PLATE/FORGING CORE

• SUPPORT COLUMN BOTTOM MOUNTED INSTRUMENTATION COLUMN Figure 4-33 Westinghouse lower core support structure and bottom mounted instrumentation columns. Core support column bolts fasten the core support columns to the lower core plate 4-70

Aging Management Requirements 9

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 4-71

Aging Management Requirements 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 for B&W plants. Therefore, no components for B&W plants were placed into the Existing Programs group.

4-72

Aging Management Requirements Table 4-8 CE plants Existing Programs components Effect Item Applicability (Mechanism)

Reference Examination Method Examination Coverage Core Shroud Assembly All plants Loss of material ASME Code Visual (VT-3)

First 10-year ISl after 40 Guide lugs (Wear)

Section Xl examination, general years of operation, and at Aging condition examination each subsequent Guide lug inserts and Management for detection of inspection interval.

bolts (ISR) excessive or Accessible surfaces at asymmetrical wear.

specified frequency.

Lower Support All plants with Cracking (SCC, ASME Code Visual (VT-3)

Accessible surfaces at Structure core shrouds IASCC, Fatigue)

Section Xl examination to detect specified frequency.

Fuel alignment pins assembled with Aging severed fuel alignment full-height Management (IE pins, missing locking shroud plates and ISR) 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 Xl examination, specified frequency.

Fuel alignment pins assembled in Aging two vertical Management (IE sections and ISR)

Core Barrel Assembly All plants Loss of material ASME Code Visual (VT-3)

Area of the upper flange Upper flange (Wear)

Section Xl examination, potentially susceptible to wear.

4-73

Aging Management Requirements Table 4-9 Westinghouse plants Existing Programs components Effect Item Applicability (Mechanism)

Reference Examination Method Examination Coverage Core Barrel Assembly All plants Loss of material ASME Code Visual (VT-3)

All accessible surfaces at Core barrel flange (Wear)

Section Xl 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)

Aging core plates to detect Management (IE) 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 to Table 4-9:

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.

4-74

Aging Management Requirements 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 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 Programs for CE 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 fuill-height 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-25 10, 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.

4-75

Aging Management Requirements 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-25 10, 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].

4-76

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 signal-based 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 and in Reference 26. 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

Examination Acceptance Criteria and Expansion Criteria Table 5-1 B&W plants examination acceptance and expansion criteria Examination Acceptance Expansion Additional Examination Item Applicability Criteria (Note 1)

Link(s)

Expansion Criteria 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 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 retaining ring material, and missing items.

Control Rod Guide Tube All plants The specific relevant None N/A N/A Assembly condition for the VT-3 of the CRGT spacer castings CRGT spacer castings is evidence of fractured spacers or missing screws.

5-2

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

Item Applicability Examination Acceptance Expansion Expansion Criteria dditional Examination Item AppI cabIity Criteria (Note 1) 1 Link(s)

ExpansionICriteria Acceptance Criteria Core Support Shield Assembly Upper core barrel (UCB) bolts and their locking devices All plants

1) Volumetric (UT) examination of the UCB bolts.

The examination acceptance criteria for the UT of the UCB bolts shall be established as part of the examination technical justification.

2) Visual (VT-3) examination of the UCB bolt locking devices.

The specific relevant condition for the VT-3 of the UCB bolt locking devices is evidence of broken or missing bolt locking devices.

UTS bolts and LTS bolts or studs/nuts and their locking devices SSHT studs/nuts or bolts and their locking devices (CR-3 and DB only)

Lower grid shock pad bolts and their locking devices (TMI-1 only)

1) Confirmed unacceptable indications exceeding 10% of the UCB bolts shall require that the UT examination be expanded by the completion of the next refueling outage to include:

For all plants 100% of the accessible UTS bolts and 100% of the accessible LTS bolts or studs/nuts, Additionally for TMI-1 100% of the accessible lower grid shock pad bolts, Additionally for CR-3 and DB 100% of the accessible SSHT studs/nuts or bolts.

2) Confirmed evidence of relevant conditions exceeding 10% of the 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 bolt and 100% of the accessible LTS bolt or stud/nut locking devices, Additionally for TMI-I 100% of the accessible lower grid shock pad bolt locking devices, Additionally for CR-3 and DB 100% of the accessible SSHT bolt or stud/nut locking devices.

1) The examination acceptance criteria for the UT of the expansion bolting shall be established as part of the examination technical justification.

2) The specific relevant condition for the VT-3 of the expansion locking devices is evidence of broken or missing bolt locking devices.

5-3

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

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

Link(s)

Expansion Criteria Acceptance Criteria Core Barrel Assembly All plants

1) Volumetric (UT)

UTS bolts

1) Confirmed unacceptable

1) The examination Lower core barrel (LCB) examination of the LCB and LTS indications exceeding 10% of the acceptance criteria for bolts and their locking bolts.

bolts or LCB bolts shall require that the UT the LIT of the expansion devices studs/nuts examination be expanded by the bolting shall be The examination acceptance and their completion of the next refueling established as part of the criteria for the UT of the LCB locking outage to include:

examination technical bolts shall be established as devices For all plants justification.

part of the examination 100% of the accessible UTS bolts technical justification.

SSHT and 100% of the accessible LTS studs/nuts or bolts or studs/nuts

2) The specific relevant

2) Visual (VT-3) examination bolts and Additionally for TMI-1 condition for the VT-3 of of the LCB bolt locking their locking 100% of the accessible lower grid the expansion locking

devices, devices (CR-shock pad bolts, devices is evidence of 3 and DB Additionally for CR-3 and DB broken or missing bolt The specific relevant only) 100% of the accessible SSHT locking devices.

condition for the VT-3 of the studs/nuts or bolts.

LCB bolt locking devices is Lower grid evidence of broken or shock pad

2) Confirmed evidence of relevant missing bolt locking devices, bolts and conditions exceeding 10% of the their locking LCB bolt locking devices shall devices require that the VT-3 examination (TMI-1 only) be expanded by the completion of the next refueling outage to include:

For all plants 100% of the accessible UTS bolts and 100% of the accessible LTS bolt or stud/nut locking devices, Additionally for TMI-1 100% of the accessible lower grid shock pad bolt locking devices, Additionally for CR-3 and DB, 100% of the accessible SSHT stud/nut or bolt locking devices.

5-4

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

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

Link(s)

Expansion Criteria Acceptance Criteria 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 baffle 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 I seam welds) refueling outage.

5-5

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

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

Link(s)

Expansion Criteria Acceptance Criteria Core Barrel All plants Visual (VT-3) examination.

Locking Confirmed relevant conditions in greater N/A Assembly

devices, than or equal to 1% (or 11) of the baffle-Locking devices, The specific relevant condition including to-former or internal baffle-to-baffle bolt including locking is missing, non-functional, or locking welds, locking devices, including locking welds, welds, of baffle-to-removed locking devices, for the shall require an evaluation of the external former bolts and including locking welds.

external baffle-to-baffle and core barrel-to-former internal baffle-to-baffle-to-baffle bolt locking devices for the purpose of baffle bolts bolts and core determining continued operation or barrel-to-replacement.

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

upper and expanded to include the Alloy X-750 VT-3 of the lower grid fuel dowel locking welds to the upper and expansion dowel assembly lower grid fuel assembly support pads by locking weld is support pads the completion of the next refueling separated or outage.

missing locking weld, or missing

_dowel.

5-6

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

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

Link(s)

Expansion Criteria Acceptance Criteria Flow Distributor All plants

1) Volumetric (UT)

UTS bolts

1) Confirmed unacceptable

1) The examination Assembly examination of the FD bolts.

and LTS indications exceeding 10% of the acceptance criteria for Flow distributor bolts or FD bolts shall require that the UT the UT of the expansion (FD) bolts and The examination acceptance studs/nuts examination be expanded by the bolting shall be their locking criteria for the UT of the FD and their completion of the next refueling established as part of devices bolts shall be established as locking outage to include:

the examination part of the examination devices For all plants technical justification.

technical justification.

100% of the accessible UTS bolts SSHT and 100% of the accessible LTS

2) Visual (VT-3) examination studs/nuts bolts or studs/nuts

2) The specific relevant of the FD bolt locking or bolts and Additionally for TMI-I condition for the VT-3 of

devices, their locking 100% of the accessible lower grid the expansion locking devices shock pad bolts, devices is evidence of The specific relevant (CR-3 and Additionally for CR-3 and DB broken or missing bolt condition for the VT-3 of the DB only) 100% of the accessible SSHT locking devices.

FD bolt locking devices is studs/nuts or bolts.

evidence of broken or Lower grid missing bolt locking devices, shock pad

2) Confirmed evidence of relevant bolts and conditions exceeding 10% of the their locking FD bolt locking devices shall devices require that the VT-3 examination (TMI-1 only) be expanded by the completion of the next refueling outage to include:

For all plants 100% of the accessible UTS bolts and 100% of the accessible LTS bolt or stud/nut locking devices, Additionally for TMI-I 100% of the accessible lower grid shock pad bolt locking devices, Additionally for CR-3 and DB.

100% of the accessible SSHT stud/nut or bolt locking devices.

5-7

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

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

Link(s)

Expansion Criteria Acceptance Criteria Incore Monitoring All plants Initial visual (VT-3)

Lower fuel Confirmed evidence of relevant The specific relevant Instrumentation (IMI) examination, grid conditions at two or more IMI conditions for the VT-3 of Guide Tube Assembly assembly guide tube spider locations or IMI the lower grid fuel IMI guide tube spiders The specific relevant support pad guide tube spider-to-lower grid rib assembly support pad IMI guide tube spider-to-conditions for the IMI guide items: pad, section welds shall require that the items (pads, pad-to-rib lower grid rib section tube spiders are fractured or pad-to-rib VT-3 examination be expanded to section welds, Alloy X-welds missing spider arms.

section include lower fuel assembly 750 dowels, cap screws, welds, Alloy support pad items by the and their locking welds)

The specific relevant X-750 completion of the next refueling are separated or missing conditions for the IMI spider-dowel, cap outage.

welds, missing support to-lower grid rib section screw, and pads, dowels, cap welds are separated or their locking screws and locking missing welds.

welds welds, or misalignment I

I_

I of the support pads.

Notes to Table 5-1:

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

5-8

Examination Acceptance Criteria and Expansion Criteria 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 The examination bolts from the core contain core support column acceptance criteria for unacceptable indications shall bolts and barrel-shroud the UT of the core require UT examination of the bolts shall be established shroud bolts shall be lower support column bolts as part of the estishrud b

s shall bhe 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 plate weld assembled in sized in the core shroud plate-indication.

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

Examination Acceptance Criteria and Expansion Criteria Table 5-2 CE plants examination acceptance and expansion criteria (continued)

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

Expansion Criteria Accetan Critia (Note 1)

Acceptance Criteria 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 The specific relevant sized in the axial weld seams at indication.

with full-condition is a detectable the core shroud re-entrant height shroud crack-like surface corners at the core mid-plane plates indication, shall require EVT-1 or UT 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

Examination Acceptance Criteria and Expansion Criteria Table 5-2 CE plants examination acceptance and expansion criteria (continued)

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

Expansion Criteria Additinal Eaitio (Note 1)

Acceptance Criteria 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 twsembled in The specific relevant two vertical condition is evidence of sections physical separation between the upper and lower core shroud sections.

5-11

Examination Acceptance Criteria and Expansion Criteria Table 5-2 CE plants examination acceptance and expansion criteria (continued)

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

Expansion Criteria Acceptance Criteria (Note 1)

AcceptanceCriteria Core Support Barrel All plants Visual (EVT-1)

Confirmation that a surface-The specific relevant Assembly examination.

Lower core support breaking indication >2 inches in condition is a detectable Upper (core support beams length has been detected and crack-like surface barrel) flange weld The specific relevant Upper core barrel sized in the upper flange weld indication.

condition is a detectable cylinder (including shall require that an EVT-1 crack-like surface welds) examination of the lower core indicracike surfaels coreba support beams, upper core indication.

Upper core barrel barrel cylinder and upper core flange barrel flange be performed by the completion of the next refueling outage.

Core Support Barrel All plants Visual (EVT-1)

Lower cylinder Confirmation that a surface-The specific relevant Assembly examination, axial welds breaking indication >2 inches in condition for the Lower cylinder girth welds length has been detected and expansion lower cylinder The specific relevant sized in the lower cylinder girth axial welds is a condition is a detectable weld shall require an EVT-1 detectable crack-like crack-like surface examination of all accessible surface indication.

indication, lower cylinder axial welds by the completion of the next refueling outage.

Lower Support All plants Visual (VT-3)

None None Structure examination.

Core support column welds The specific relevant condition is missing or separated welds.

5-12

Examination Acceptance Criteria and Expansion Criteria Table 5-2 CE plants examination acceptance and expansion criteria (continued)

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

Expansion Criteria Additinal Eaitio (Note 1)

Acceptance Criteria 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

Examination Acceptance Criteria and Expansion Criteria Table 5-2 CE plants examination acceptance and expansion criteria (continued)

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

Expansion Criteria Acceptance Criteria (Note 1)

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

at the welded joint instrument tubes within the CEA between the tubes and shroud assemblies by the supports.

completion of the next refueling outage.

Lower Support All plants Visual (EVT-1)

None N/A N/A Structure with core examination.

Deep beams shrouds assembled The specific relevant with full-condition is a detectable height shroud crack-like indication.

plates I

Notes to Table 5-2:

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

5-14

Examination Acceptance Criteria and Expansion Criteria Table 5-3 Westinghouse plants examination acceptance and expansion criteria Examination Additional Examination Item Applicability Acceptance Criteria Expansion Expansion Criteria Additinal Eaitio (Note 1)

Link(s)

Acceptance 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-

a. Bottom-

a. Confirmation of surface-

a. For BMI column Tube Assembly

1) examination, mounted breaking indications in two or bodies, the specific Lower flange welds instrumentation more CRGT lower flange relevant condition for the (BMI) column welds, combined with flux VT-3 examination is The specific relevant bodies thimble insertion/withdrawal completely fractured condition is a difficulty, shall require visual column bodies.

detectable crack-like (VT-3) examination of BMI surface indication.

b. Lower support column bodies by the column bodies completion of the next

b. For cast lower support (cast), upper core refueling outage.

column bodies, upper plate and lower core plate and lower support forging or support forging/castings, casting

b. Confirmation of surface-the specific relevant breaking indications in two or condition is a detectable more CRGT lower flange crack-like surface welds shall require EVT-1 indication.

examination of cast lower support column bodies, upper core plate and lower support forging/castings within three fuel cycles following the initial observation.

5-15

Examination Acceptance Criteria and Expansion Criteria Table 5-3 Westinghouse plants examination acceptance and expansion criteria (continued)

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

Expansion Criteria Acceptance Criteria (Note 1)

Core Barrel Assembly All plants Periodic enhanced visual a. Core barrel outlet a. The confirmed detection and a and b. The specific Upper core barrel flange (EVT-1) examination, nozzle welds sizing of a surface-breaking relevant condition for the weld

b. Lower support indication with a length greater expansion core barrel The specific relevant column bodies (non than two inches in the upper outlet nozzle weld and Thenspeific arelevnta core barrel flange weld shall lower support column condition is a detectable cast) require that the EVT-1 body examination is a crack-like surface examination be expanded to detectable crack-like indication.

include the core barrel outlet surface indication.

nozzle welds by the completion of the next refueling outage.

b. If extensive cracking in the core barrel outlet nozzle 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.

Core Barrel Assembly All plants Periodic enhanced visual None None None Lower core barrel flange (EVT-1) examination.

weld (Note 2)

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

5-16 I

Examination Acceptance Criteria and Expansion Criteria Table 5-3 Westinghouse plants examination acceptance and expansion criteria (continued)

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

Expansion Criteria Acceptance Criteria (Note 1)

Core Barrel All plants Periodic enhanced Upper core barrel The confirmed detection and The specific relevant Assembly visual (EVT-1) cylinder axial welds sizing of a surface-breaking condition for the Upper core barrel examination, indication with a length greater expansion upper core cylinder girth welds than two inches in the upper barrel cylinder axial weld core barrel cylinder girth welds examination is a The specific relevant shall require that the EVT-1 detectable crack-like condition is a examination be expanded to surface indication.

detectable crack-like include the upper core barrel surface indication, cylinder axial welds by the completion of the next refueling outage.

Core Barrel All plants Periodic enhanced Lower core barrel The confirmed detection and The specific relevant Assembly visual (EVT-1) cylinder axial welds sizing of a surface-breaking condition for the Lower core barrel examination, indication with a length greater expansion lower core cylinder girth welds than two inches in the lower barrel cylinder axial weld core barrel cylinder girth welds examination is a The specific relevant shall require that the EVT-1 detectable crack-like condition is a examination be expanded to surface indication.

detectable crack-like include the lower core barrel surface indication, cylinder axial welds by the completion of the next refueling outage.

5-17

Examination Acceptance Criteria and Expansion Criteria Table 5-3 Westinghouse plants examination acceptance and expansion criteria (continued)

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

Expansion Criteria Acceptance Criteria (Note 1)

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 Baffle-former bolts actually examined on the four criteria for the UT of the 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.

5-18

Examination Acceptance Criteria and Expansion Criteria Table 5-3 Westinghouse plants examination acceptance and expansion criteria (continued)

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

Expansion Criteria Additinal Eaitio (Note 1)

Acceptance Criteria 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.

spring steel hold 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.

5-19

Examination Acceptance Criteria and Expansion Criteria Table 5-3 Westinghouse plants examination acceptance and expansion criteria (continued)

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

Expansion Criteria Acceptance Criteria (Note 1)

Thermal Shield All plants Visual (VT-3)

None N/A N/A Assembly with thermal examination.

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

Notes to Table 5-3:

1.

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

2.

The lower core barrel flange weld may alternatively be designated as the core barrel-to-support plate weld in some Westinghouse plant designs.

5-20

Examination Acceptance Criteria and Expansion 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 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-I 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.

5.1.3 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 5-21

Examination Acceptance Criteria and Expansion Criteria 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 or pins 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.

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, 5-22

Examination Acceptance Criteria and Expansion Criteria these physical measurements are described in Section 4.3, with limits applicable to the various designs. For B&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 plant-specific basis and thus are not provided generically in these guidelines. For CE 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].

5-23

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 guidance provided in this section is general; Reference 26 should be consulted for more detailed guidance.

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

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 3x102" 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

Evaluation Methodologies 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 III, 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 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 6-3

Evaluation Methodologies Jmat = C (Aa)n Equation 6-1 where J and C are in KJ/m2 and Aa is in mm.

The lower bound expression for power law parameter C is given by C = (1217.9*6.697*1010 + 0.3908*Fo5563)/(6.697* 100 + F°0 5563)

Equation 6-2 The upper bound expression for power law parameter n is given by n = 1/(4.962 - 0.02439*F°0 °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 of crack growth is 216 KJ/m 2. 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 of crack growth is 75.8 KJ/m 2. If the tangent to the crack growth resistance curve at 1.5 mm is projected back to zero crack growth and converted to K, through the expression J~c = (K1c) 2/E Equation 6-4 where E is the elastic modulus, then Kc equals 100 MPa'lm. 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 of crack growth is only 64 KJ/m 2. Extrapolating the tangent of the crack growth resistance curve back to zero crack growth and converting gives Kc = 92 MPa'hm. 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 MPa*m 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 Kc = 55 MPa/m for exposure levels between 5 and 15 dpa, and with a limiting fracture toughness Kc = 38 MPa'Im for exposure levels greater than 15 dpa.

6-4

Evaluation Methodologies 80D 7DO 600 eoo-30D 200 IO0 0

0.00 0.50 1.00 1.50 2.00 Crack Extension, Aa, mm 2.50 Figure 6-1 Experimental Jmaterial versus crack extension curves for stainless steel materials at various fluence levels [19]

800 700 000 500

.1 400 300

.200 9L 100 a

• Ba" Metal X HAZ 0

-C.

EqjaUWo 2-2

• -T 1.DE+20 1.OE+21 1.OE+22 Neutron Fluence. nicm2 Figure 6-2 J-R curve power law parameter C as a function of neutron fluence for stainless steel, applicable for fluence less than 3x10 21 nlcm 2[19]

6-5

Evaluation Methodologies 1

0.9 0.8

£ 0.7 1 0.0

.10.5 0.4 0.3 0.2 0.1 o Wuld Mefal x HAZ

-- n, Equain 24 19FU UEM

• 4 0 r__ _ _ L...........L LJ..L i.......J.....L.....

i 1.DE+20 1.DE+21 1.0E+22 Neutron Fluence, n/cm' Figure 6-3 J-R curve power law parameter n as a function of neutron fluence for stainless steel, applicable for fluence less than 3x10 21 nlcm 2 [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 1.1 xl 10' 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 of BWRVIP-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.

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 6-6

Evaluation Methodologies 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 2x 10-6 to 5.25x 10.5 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 lxl0-7 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 daldt = 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.

1.OOE-03 0

5 10 15 20 25 30 K ( ksi-inl/2)

Figure 6-4 Proposed BWR hydrogen water chemistry crack growth curves for stainless steel irradiated between 5x1 020 to 3x1 021 nlcm' [24]

6-7

Evaluation Methodologies

• 304 - 2880C 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 3 316 - 280'C to 289°C - BWR HWC - 2 to 2,9 dpa

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

• 316Ti - 2880C - PWR - 25 dpa

• 316Ti - 3200C to 340°C-PWR - 25 dpa

• 347-280°C - BWR HWC - 3,2 dpa

• 347-320°C - PWR -13,5 to 17 dpa 1.00E-04 00 1.00E-05 0°A 0

1.00E-06 0U A

0 a~

1.OOE-07

-AU o

A OW

~1.00E-08 0

A A

1.00E-09 o

1.OOE-10 0

5 10 15 20 25 30 35 40 45 50 55 60 Stress Intensity K ( MPa'Im)

Figure 6-5 Effect of stress intensity on IASCC crack growth rate [25]

6.3 Evaluation of Flaws in Bolts and Pins 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-finctional 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 where the assembly is found to be deficient are most often driven by loose parts or reassembly 6-8

Evaluation Methodologies 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.

6.5 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.

6-9

Evaluation Methodologies 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 J-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 J-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 crack at the inside or outside surface 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; 6-10

Evaluation Methodologies

" 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 comer crack in a rectangular bar that can represent: (i) a quarter-circular crack in the comer of baffle plates; or (ii) a quarter-circular crack at the inside comer 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