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Internals.ppt 2

Reactor Internals Overview

Mechanisms: Cracking due to Irradiation

Assisted Stress Corrosion (IASCC) and Stress Corrosion

(SCC)

Reduction of Fracture Toughness due to Irradiation

Embrittlement (IE) and Thermal Embrittlement (TE)

Dimensional Changes due to Void Swelling (VS)

Loss of Mechanical Closure Integrity due to Stress

Relaxation (SR)

Synergistic Effects of These Mechanisms

2

Internals.ppt 3

B&W Typical Internals Layout

3

Core(not in scope)

PlenumAssembly

Plenum CoverAssembly

Vent Valves

Control RodGuide TubeAssembly

Upper GridAssembly

Thermal Shield

Core Barrel

Baffle Plates

Former Plates

Lower Grid

Flow DistributorHead

CoreSupportAssembly

CoreSupportShieldAssembly

Core BarrelAssembly

LowerInternalsAssembly

Incore GuideTubes

Internals.ppt 4

Typical CE Internals Layout

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Internals.ppt 5

Westinghouse Internals with

Forged Lower Support

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Internals.ppt 6

Westinghouse Internals with Cast

Lower Support

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Internals.ppt 7

Reactor Internals Lower Section

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Internals.ppt 8

PWR Core Internals

• The baffle-former

assembly or shroud

surrounds the core and

separates the coolant

down-flow from up-flow

• Constructed primarily

from annealed 304

stainless steel

• Can be of either bolted

or welded construction

• Baffle plates, especially

at reentrant corners, will

receive doses as large as

100 dpa over a 40 year

lifetime

Internals.ppt 9

Baffle-former-barrel Assembly in a

Typical Westinghouse PWR

core

barrel

bolt

former

baffle baffle

Internals.ppt 10

Typical Baffle Former Assembly

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Internals.ppt 11

Reactor Internals Upper Section

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Internals.ppt 12

Categorization

Aging

Management

Analysis

Component

List

Screening

Criteria

Initial

Screening

Cat. ANo Adverse

Effects

Category B Category C

Below Screening Above Screening

No Credible

Damage Issue

HighModerate

Functionality

Analysis

Probability &

Consequence

Analysis

Existing Subordinate Principal

Resolved by

Analysis

Aging Management Program

I&E GuidelinesMonitoring &

Trending

Aging Management

Strategy

Existing Guidelines New Recommendations

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Internals.ppt 13

Approach for Evaluating Functionality

Analysis for I&E Guidelines What?

Damage mechanisms of concern?

Metrics used to characterize a damage mechanism?

Observable effects/consequences on functionality?

Where?

Location of degradation?

When?

Estimate the likelihood and timing of future damage?

How?

Inspection, monitoring or trending technique

Task is to utilize representative plant results and

apply to entire fleet

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Internals.ppt 14

The Cracking Mechanisms

SCC

IASCC

Fatigue

Produce observable cracks

Most probable in regions of stress concentration

Expect to manage through an

integrated inspection program.

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Internals.ppt 15

Stress Corrosion Cracking (SCC)

Austenitic stainless steel

No experience with SCC in 300 series stainless steel under normal primary water conditions

○ No model to evaluate or rank potential for SCC

Large structural welds identified due to large potential residual stresses

X-750

Programs for guide tube support pins in place

Cast austenitic stainless steel

Verify that specifications meet minimum ferrite requirements

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Internals.ppt 16

Irradiation-Assisted SCC

(IASCC)

Stainless steel bolts (316 SS)

FEA intended to provide basis for ranking of

time to failure

Limited number of CE plants with bolted baffles

Stainless steel plate (304 SS)

CE shroud welds included in plate waterfall

Will identify locations with IASCC susceptibility

from FEA

Eliminated components associated with

Westinghouse lower core plate on basis of

completed analysis

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Internals.ppt 17

Parameters Influencing IASCC

Fluence

IASCC in PWRs occurs above

a threshold fluence of ~ 2 x 1021

n/cm2, E > 1 MeV

This fluence level is higher than

in BWRs by about an order of

magnitude

The threshold fluence level

does not correlate directly with

the onset or saturation of

radiation-induced materials

changes such as grain

boundary segregation or

hardening

Start of BWR IASCC

Start of PWR IASCC

BWR End of Life Dose

*Max PWR End of Life Dose

1020

1021

1022

1023

Neutron/cm2 (E 1 MeV)

0.1 1 10 100 dpa**

Start of Start of Saturation Possible Grain Boundary Ductility of Sensitization Start of Sensitization Loss and Ductility Swelling Loss

* For 32 EFPY ** Based on 15 dpa = 1022

n/cm² E 1 MeV

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Internals.ppt 18

Fatigue

Expect that fatigue evaluation will be

required to justify extended life

Real vs. assumed stress history

Realistic stress/strain amplitudes

Potential environmental effects

Two main groupings

Additional evaluation required

Addressed via SCC, IASCC, etc.

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Internals.ppt 19

The Embrittlement Mechanisms

Irradiation embrittlement

Thermal embrittlement

Changes in material properties

○ Strength (increase)

○ Ductility (decrease)

○ Toughness (decrease)

Expect to manage through an industry

trending program

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Internals.ppt 20

Irradiation Embrittlement

Industry trend curves for strength and ductility are embedded in computer codes Westinghouse lower core plate

Westinghouse baffle-former-barrel

CE core shroud

Extrapolate to remaining components based on fluence and temperature

Fracture toughness estimates required for components with active cracking mechanisms

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Internals.ppt 21

Thermal Embrittlement

Evaluate composition and

temperature to determine

susceptibility to thermal

embrittlement

Fracture toughness estimate required

if there is an active cracking

mechanism

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Internals.ppt 22

Dimensional Stability

Mechanisms

Void swelling

Irradiation induced stress

relaxation/creep

Component distortion

Modify stress/strain distribution

○ Affects SCC, IASCC and fatigue

Expect to manage through industry trending

and inspecting for distortion

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Internals.ppt 23

Void Swelling

FEA analysis provide ranking based on swelling model in computer codes

Westinghouse baffle-former-barrel

Westinghouse lower core plate

CE core shroud

Components not included in FEA that can be easily compared to analyzed components

Westinghouse lower core support structure

CE baffle bolts

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Internals.ppt 24

Irradiation Induced Stress Relaxation/

Creep

FEA model incorporates stress

relaxation and creep effects (can rank

effect)

Stress relaxation may have significant

impact on other stress related mechanisms

(e.g., IASCC)

Loss of bolt preload must be considered as

contributing to wear and fatigue waterfalls

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Internals.ppt 25

Wear Mechanism

Difficult to compare or rank wear

potential in identified components

Match inspection/trending monitoring program

to component requirements

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Internals.ppt 26

Wear

Existing wear management programs Westinghouse flux thimbles and tubes

CE thermal shield positioning pins

CE In-core Instrumentation thimble tubes

Monitored through control rod drop

times

Inspect & monitor neutron noise

Inspection requirements combined

with integrated crack monitoring

programs

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Internals.ppt 27

What is a Reactor Internals Aging

Management Program (AMP)?

A document (procedure, instruction,

specification) that describes a plant’s program

to ensure the long-term integrity and safe

operation of PWR internal components

Why is it required? Previously required only for plants applying for license

renewal

With publication of MRP-227A, now required for all plants (Mandatory requirement under NEI 03-08)

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Internals.ppt 28

Contents of an AMP

What are the required contents of

an AMP?

MRP-227A, Appendix A defines the 10

elements which constitute an acceptable AMP

These elements are from NUREG-1801

(Generic Aging Lessons Learned [GALL] Report)

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Internals.ppt 29

Requirements for Reactor

Internals AMPs

GALL report NUREG-1801, “PWR Vessel Internals” identified 10 Attributes/Elements necessary for the Evaluation and Technical Basis

1. Scope of Program

2. Preventive Actions

3. Parameters Monitored/Inspected

4. Detection of Aging Effects

5. Monitoring and Trending

6. Acceptance Criteria

7. Corrective Actions

8. Confirmation Process

9. Administrative Controls

10. Operating Experience

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Internals.ppt 30

First MRP-227 Inspections

Started in 2011 First plant was Ginna in May 2011

Baffle-former bolts

Baffle-former assembly

Baffle former edge bolts

Lower guide tube flange weld

Upper core barrel flange to shell weld

Thermal shield flexures

Control rod guide cards

Type X-750 split pins replaced using cold-

worked 316 stainless steel

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Internals.ppt 31

Ginna Inspection

Alignment and interfacing for internals

hold-down spring; at Ginna spring

material is 410 stainless steel, therefore,

only EVT-1 scan on the external surface

was needed

In addition to MRP-227 requirements,

ASME Code Section XI ISI examinations

require visual exam of lower core plate

and fuel pins

Only problem area was baffle-former

bolts 31

Internals.ppt 32

Ginna Baffle-Former Bolt

Inspections Proactive inspection in 1999 found 59

bolts with indications (347 SS) – 56

replaced (316 SS) and 5 broke when

extracted

Original plan for 2011 was to replace

126 old bolts

Big problem – bolts could not easily be

removed and replaced (25 of 28 successful)

Shifted to UT inspection approach of 98

bolts

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Internals.ppt 33

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License Renewal Reactor Internals MRP 227-A Inspections

Internals.ppt 34

Reactor Vessel Internals Aging

Managing Program

Pure Water Stress

Corrosion Cracking

PWSCC

Irradiated Assisted

Stress Corrosion

Cracking

IASCC

Stress Corrosion

Cracking

SCC

Irradiation Enhanced

Stress Relaxation

SR

CASS TE/IE

CASS

Irradiation

Embrittlement

IE

Irradiated Induced

Void Swelling

VSWearFatigue

Use of Material Data, Fluence and Generic Analyses to Screen and Select Most-Affected Internals Components

Baffle-BoltsWelds

Thermal Shield & Core Barrel Bolts

Bolted Joints

CASS Components (Synergistic

Effect)

Baffle-Former Plates and Bolts

Internals Components

(Increased Strength & Loss

of Ductility)

Guide Cards

Guide Lugs

Use Plant-Specific Analyses to Determine Critical Locations, Critical Crack Sizes and Flaw Tolerance for Actual Stresses, Fluences and Temperatures

UT InspectionsEnhance VT-1

Inspections

MRP 227-A Inspection GuidelinesEmbrittlement

Analysis with Stress and Deflection Due to LOCA and SSE

Dimensional Changes and Functionality

1. Scope 6. Acceptance Criteria2. Preventive Actions 7. Corrective Action3. Parameters Monitored 8. Confirmation Process4. Detection of Aging Effects 9. Administrative Controls5. Monitoring and Trending 10. Operation Experience

Meet Current License Basis

1. Scope 6. Acceptance Criteria2. Preventive Actions 7. Corrective Action3. Parameters Monitored 8. Confirmation Process4. Detection of Aging Effects 9. Administrative Controls5. Monitoring and Trending 10. Operation Experience

Meet Current License Basis Meet Current License Basis

Improved Inspection Techniques Proposed After Functionality

Assessment.

· Integrated Baffle Bolt Microstructural Evaluation

· Possible Lead Irradiation of SS at Bounding Temp?