NUREG/CR-6048

ORNU-TM-12371

Pressurized-Water Reactor

Internals Aging

Degradation Study

Phase I

Prepared by

K. H. Luk

Oak Ridge National Laboratory

Prepared for

U.S. Nuclear Regulatory Commission
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NUREG/CR-6048

ORNL-TM-12371

RV

Pressurized-Water Reactor

Internals Aging

Degradation Study

Phase I

Manuscript Completed: August 1993

Date Published: September 1993

Prepared by

K. H. Luk

Oak Ridge National Laboratory

Operated by

Martin Marietta Energy Systems, Inc.

Oak Ridge National Laboratory

Oak Ridge, TN 37831-6285

Prepared for

Division of Engineering

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

NRC FIN B0828

Under Contract No. DE-ACO5-84OR21400
Abstract

Ibis report is a summary of the results of a Phase 1 study

on the effects of aging degradations in pressurized-water

reactor (PWR) internal components. Westinghouse (WE),

Combustion Engineering (CE), and Babcock & Wilcox

(B&W) reactors are included in the study.

Stressors associated with the operating environment inside

the reactor pressure vessel provide conditions that are

favorable to the development of aging-related degradation

mechanisms. The dominant stressor is flow-induced oscillatory hydrodynamic forces generated by the reactor primary coolant flow. Results of a survey of the component

failure information identified three major aging-related

degradation mechanisms: fatigue, stress corrosion cracking, and mechanical wear.

between stressors and aging degradation mechanisms.

Flow-induced vibration problems are resolved by conventional engineering practices: by the elimination of excitation sources or by de-tuning the structure from input excitations. Uncertainties remaining in the assessment of aging

effects on PWR internals include long-term neutron irradiation effects and the influence of environmental factors on

high-cycle fatigue failures.

An effective plant in-service inspection (ISI) program will

ensure the structural integrity of reactor internals. Reactor

internals can be replaced if it is deemed necessary. Therefore, an inspection method with early failure detection

capability will further enhance the safety as well as the

efficiency of plant operations.

Strategies for controlling and managing aging degradations

are formulated based on the understanding of the linkage

iii NUREG/CR-6048
Contents

Page

Abstract ............................................................................... iii

List of Figures ........ ........................ ................................................ vii

List of Tables ........ ......................... ............................................... ix

Acknowledgments ................................................................................. xi

Summary ......... ........................... .................................................... xiii

1 Introduction ................................................................................ 1

Reference ........ ........................................................................ 2

2 PWR Internal Components ................................................................................ 3

2.1 Westinghouse (WE) Internals ............................................................................... 3

2.2 Babcox & Wilcox (B&W) Internals ............................................................................... 11

2.3 Combustion Engineering (CE) Internals ....................... ........................................................ 16

Reference ............................................................................... 22

3 Primary Stressors ......... ....................................................................... 25

3.1 Applied Loads .......... ..................................................................... 25

3.2 Eivironmental Stressors ................................................................................ 25

3.3 Manufacturing Stressors ............................................................................... 26

References............................................................................................................................................................... .. 27

4 Aging-Related Degradation Mechanisms ............................................................................... 29

4.1. Corrosion........................................................................................................................................................ 29

4.2 Fatigue ........ ....................................................................... 31

4.3 Erosion ........ ....................................................................... 32

4.4 Mechanical Wear ............................................................................... 32

4.5 Embrittlement ............................................................................... 32

4.6 Creep and Stress Relaxation ............................................................................... 33

4.7 PWR Internals and Potential Aging-Degradation Mechanisms .................................... I................................. 34

References ........ ..................................................... I. ................ 35

5 Survey of Aging-Related Failures ............................................................................... 37

5.1 ISI Program for Reactor Internals ............................................................................... 37

5.2 Failure Information Summary ............................................................................... 37

53 Failure Information Survey Results ............................................................................... 43

References............................................................................................................................................................... .. 45

6 ISI and Monitoring Programs ............................................................................... 47

6.1 LPMS ........ ....................................................................... 47

v V ~~~~~~~NUREG/CR
6.2 Vibration Monitoring and Trending Studies ................................................ .................................... 48

References ........ ............................................................................ 49

7 Discussion and Conclusions ..................................................................................... 51

7.1 FIV ........ ............................................................................. 51

7.2 SCC ..................................................................................... 52

7.3 Other Potential Aging-Related Degradation Mechanisms ............................................................................. 52

NUREG/CR-6048 vi
List of Figures

Figure Page

2.1 Westinghouse PWR internals ......................................................................... 4

2.2 Westinghouse PWR lower core support structure . ........................................................................ 5

2.3 Westinghouse PWR core barrel cross sections ......................... ............................................... 6

2.4 Westinghouse PWR radial key-keyway system .......................... .............................................. 7

2.5 Westinghouse PWR thermal shield flexure support system ........................................................................ 7

2.6 Westinghouse PWR secondary core support assembly . ......................................................................... 8

2.7 Westinghouse PWR upper core support structure ............................... .......................................... 9

2.8 Westinghouse PWR thimble and guide tube ..................... ................................................... 11

2.9 B&W PWR internals ........................................................................ 12

2.10 B&W PWR eral shield lower support ........................................................................ 14

2.11 B&W PWR internal vent valve ......................................................................... 5

2.12 CE PWR internals ........................................................................ 17

2.13 CE PWR core support assembly ........................................................................ 18

2.14 CE PWR core support barrel ........................................................................ 19

2.15 CE PWR core support snubber assembly ........................................................................ 20

2.16 CE PWR upper grid assembly ........................................................................ 21

2.17 CE PWR in-core instrumentation support structure ...................................... .................................. 23

5.1 Westinghouse PWR downward bypass flow scheme ............................................ ............................ 39

52 Westinghouse PWR baffle water-jet impingement patterns ........................................................................ 40

5.3 Westinghouse PWR guide tube support pins ...................... .................................................. 41

5A KWU-built PWR core baffle and bolting scheme ............................... ......................................... 42

5.5 KWU-built PWR core baffle bolt failures ........................................................................ 42

vii NUREGICR-6048
List of Tables

Table Page

4.1 PWR internals primary stressors and aging-related degradation mechanisms ...................... .............................. 35

5.1 Summary of WE internals failure information ............................................................................. 44

5.2 Summary of B&W internals failure information ............................................................................. 45

5.3 Summary of CE internals failure infonnation . ............................................................................ 45

ix NUREG/CR-6048
Acknowledgments

The guidance and direction provided by D. A. Casada,

W. S. Farmer, and R. D. Cheverton in dhe conduct of this

study are greatly appreciated.

The assistance of C. C. Soutbmayd and E. W. Carver in

preparing this report is gratefully acknowledged.

The author also wishes to hank A. E. Cross and W. P.

Poore m for their help in obtaining the component failure

data for PWR internals.

xi NUREG/CR6048
Summary

Reactor internals operate in the environment inside the

pressure vessel; this is favorable to the development of

time-dependent or aging degradations. The main objective

of this study is to assess the effects of aging degradations

on pressurized-water reactor (PWR) internal components.

The assessment Includes an evaluation of the effectiveness

of the plant in-service inspection (ISI) program in detecting

failures in nternals before they can affect the safety and

efficiency of plant operations. Westinghouse (WE),

Combustion Engineering (CE), and Babcock & Wilcox

(B&W) reactors are included in the study.

Reactor internals selected for the study serve three basic

functions: they support the core; they provide housings for

control rods, control rod drive mechanisms, and in-core

monitors; and they direct and guide the reactor coolant

flow through the reactor vessel. Most internals are made of

type 304 stainless steel. Although they are located inside

the reactor vessel, fuel assemblies, control rods, control rod

drive mechanisms, and in-core monitors are not included

as internals in the present study.

Stressors are conditions that can initiate and sustain the

growth of aging-related degradation mechanisms. Reactor

internals are subjected to stressors generated by applied

loadings (thermal and mechanical), by contacts with hightempemture flowing water, and by exposures to highenergy (E > I MeV) neutron fluxes. The applied loadings

of primary concern are flow-induced oscillatory hydrodynamic forces because they can excite structural components into vibrations. The reactor cooling water provides

an environment that may contain conditions that are considered as favorable to the development of stress corrosion

cracking (SCC). Neutron irradiation effects are important

stressors to internal components located in close proximity

of the core. Manufacturing processes may also impose

stressors on nternals.

Aging-related degradation mechanisms generally associated with the primary stressors for reactor internals are

corrosion (including SCC), fatigue, mechanical wear, erosion, embrittlement, creep, and stress relaxation. These

degradation mechanisms may develop at different rates,

and as a result they are not of equal mportance in the

design life expectancy of the reactor. However, aging

degradations, if they are not mitigated, will eventually lead

to a failure in the affected component. Reported failure

information and laboratory testing results can identify the

more significant aging-related degradation mechanisms.

SCC, fatigue, and mechanical wear are the major agingrelated degradation mechanisms for PWR internals.

Major reported failure cases for PWR internals include

bolting failures in core support structures caused by fatigue

and SCC, fuel assembly damages caused by high-speed

leakage flows through enlarged gaps in core baffle plates,

excessive thinning of flux thimble guide tubes caused by

flow-induced vibrations, and crevice-assisted SCC in control rod guide tube support pins. Some of these failures

have resulted In extended outages for repair work, but currently available information did not indicate that they have

compromised the safety-related functions of reactor

internals.

The plant in-service inspection ([SI) program calls for the

visual inspection of accessible areas of reactor internals

during refueling outages. Limitations of the visual Inspection method are well known. Reactors licensed since 1978

are equipped with loose parts monitoring systems

(LPMSs). The main objective of the LPMS is to alert plant

operators that loose parts are present in the reactor primary

system so that appropriate actions can be taken to limit further damages to other reactor components and systems.

New technologies have been developed, and they have the

capability of early failure detection in key core-support

internal components. They involve the use of neutron noise

vibration measurements and trending studies. These practices, while common in France and Germany, have not

been formally incorporated into the ISI program for U.S.

nuclear plants. They have been used on a voluntary basis.

Visual inspection, supplemented by ultrasonic and eddycurrent inspection methods, remain the major tools for

inspecting reactor internals in the U.S.

Most flow-induced vibration and SCC problems have been

resolved by conventional engineering practices such as

elimination of excitation sources, detuning the structure

from external excitations, design changes, and use of materials That will make the components less susceptible to corrosion attacks.

Small-amplitude flow-induced vibrations in reactor internals are difficult to eliminate, and high-cycle fatigue

remains as an active degradation mechanism. Internals

close to the core are exposed to high-energy neutron fluxes

and neutron irradiation effects, and related degradation

mechanisms may become more prominent in the remaining

life of the reactor. Thermal aging in cast austenitic stainless

steel (CASS) internal components is in the same category.

In the presence of active aging degradations, a vigilant

inspection program is needed to ensure the structural

integrity of reactor internals. The Incorporation of an effective early failure detection method will further enhance the

safety and efficiency of plant operations.

xiii NUREGICR-6048
1 Introduction

Systems, structures, and components of a commercial

nuclear power plant are subjected to time-dependent or

aging degradations during operations. Effects of aging

degradations, if they are not mitigated, will eventually lead

to failures that could adversely affect plant safety and perfornance. The Office of Nuclear Regulatory Research

(RES) has established the Nuclear Plant Aging Research

(NPAR) Programl to increase basic understanding of

aging-related degradations and their effects on reactor systems, structures, and components. The NPAR approach is

to perform in-depth studies on selected reactor systems,

structures, and components that are judged to be vulnerable

to aging degradations. Understanding the interrelationship

between stessors and aging-related degradation mechanisms is also the basis for the formulation of strategies for

controlling and managing aging effects. One of the reactor

systems selected for aging study is reactor internals, and

the study is assigned to the Oak Ridge National Laboratory

(ORNL). The effects of aging on boiling-water reactor

(BWR) internals has been addressed in a previous report.

This report will concentrate on the aging assessment of

pressurized-water r (PWR) internals. Operating histories of Babcock & Wilcox (B&W), Combustion

Engineering (CE), and Westinghouse (WE) reactors provide the majority of the information for the aging assessment process. Westinghouse has licensed Its PWR technology to European and Japanese vendors, and appropriate

aging-related failure information of overseas reactors will

also be included in the study.

The term "internal" is generally applied to reactor components that are located inside the reactor pressure vessel.

Fuel assemblies, control rods, control rod drive mechanisms, and in-core monitoring equipment are routinely

replaced and are excluded from this study. Housings for

these components are considered as internal components.

Reactor Internals perform many functions; the primary one

is to provide structural support and orientation to the core

and control rod assemblies (CRAs). Other internal components direct and guide the coolant flow through the core

region and provide shielding to the pressure vessel wall.

At the present time there are 73 PWRs licensed for commercial operation in the United States. Using the commercial operation starting date as the reference for counting

reactor ages, 6 reactors or about 8% of the total are over 20

years old; 42 reactors or 58% are between 10 and 20 years

old, and 25 reactors or 34% are less than 10 years old.

There is a total of 907 reactor-years of PWR operations,

and the accumulated operating histories of these reactors

1K H. Lk, EMcling Water Ram Iteals Aing Degadaion

Study-A Phas 1 Reor4" USNRC Report NUREGACR-S754

(ORNITM- 1 176), to be pubished.

provide the Information for studying aging effects in

selected reactor components. The plant ISI Program s the

major source of information on aging-related failure for

reactor systems.

The aging assessment is performed in a multiple-step process. The first step is the identification and description of

reactor internal components included in the study. The second step is to identify stressors that are presented in the

operating environment inside the pressure vesseL The third

step Is to establish linkage between stressors and agingrelated degradation mechanisms. The final step is the identification of the more significant aging-related degradation

mechanisms based on a review of the operating histories of

PWRs and reported component failure information. The

establishment of the proper linkage between an agingrelated degradation mechanism and the associated stressors

can be used as the basis for formulating strategies for controlling and managing aging effects.

Selected reactor internals are identified in Chap. 2 of the

report The information provided in Chap. 2 includes a

brief description of each selected component, the functions

it performs, and the material of construction. Primary

stressors Inside the reactor pressure vessel are discussed in

Chap. 3. Stressors generated by applied loads, environmental conditions, and manufacturing processes are included in

the discussion. Chapter 4 identifies aging-related degradation mechanisms associated with the primary stressors.

Potential aging degradation mechanisms include corrosion

(including SCC), fatigue, mechanical wear, erosion,

embrittlement (thermal and radiation induced), creep, and

stress relaxation. Chapter 5 is a summary of the more significant reported aging-related failures of PWR internals. It

includes discussions in the thermal shield support bolt failures in B&W, CE, and WE reactors; baffle plate waterjetting problems In WE units; WE control rod guide tube

split pins failures; and flux thimble tube thinning problems. The core baffle bolt failures in Kraftwerk Union

(KIWU)-built PWRs of the WE design are also included in

Chap. 5. The reported aging-related failure information

identifies three major aging-related degradation mechanisms: fatigue, SCC, and mechanical wear.

This study also addresses issues concerning the Inspection

and maintenance methods used to control and manage

aging effects in reactor internals. The effectiveness of the

visual inspection method is discussed in Chap. 6 of the

report, which also provides information on the development of new technologies in these areas, such as loose-part

monitoring, neutron noise vibration measurements, and

I NUREGICR-6048
Introduction

trending studies. Chapter 7 is a summary of important

results in this Phase 1 aging assessment of PWR internals.

Reference

1. J. P. Vora, Nuclear Plant Aging Research (NPAR)

Program Plans, USNRC Report NUREG-1 144, Rev. 1,

September 1987.*

Available for purchwe fron National Technical Information Service,

Springfield. VA 22161.

NUREG/CR-6048 2
2 PWR Internal Components

lbree domestic PWR vendors in the United States are the

Westinghouse Electric Corporation (WE), the Combustion

Engineering Company (CE), and the Babcock & Wilcox

Company (B&W). Of the 73 domestic PWRs in commercial operation, 51 or about 70% of the total are WE, 15

(20%) are CE, and 7 (10%) are B&W reactors. The three

reactor designs share some common features, but the internals are sufficiently different that using "generic" components foraging studies is not feasible. The three reactor

internal systems will be treated separately.

The primary function of reactor internals is to provide

structural supports to the core and to properly position

CRAs under normal and accident operating conditions.

Reactor internals that perform such functions are components of the core support system. Other internal components direct and guide the coolant flow through the core

region and help to generate a uniform flow distribution to

enhance core heat transfer. A third type of core internals is

designed to provide gamma and neutron shielding to the

reactor pressure vessel. Housings for in-core instrumentalions are also considered as reactor internals. Even though

they are also located inside the reactor vessel, aging effects

in fuel assemblies, control rods, control rod drive mechanisms, and in-core monitors are not addressed in this study.

The majority of reactor internals are made of type 304

stainless steel. Unless it is specifically stated, It can be

assumed that the material of construction for an internal

component is type 304 stainless steel. Reactor internals are

designed in accordance with the requirements of Sect. m

of the American Society of Mechanical Engineers (ASME)

Boiler and Pressure Vessel (B&PV) Code.I For reactors

that were designed and built before the establishment of

Sect. HI of the ASME B&PV Code, analyses were performed to ensure that calculated stress values for reactor

components met the intent of the Code under specified

design conditions.

Information on internal components is obtained from various plant Final Safety Analysis Reports (FSARs) and technical reports published by the Electric Power Research

Institute (EPRI). The components included in the study are

those for a "typical" reactor made by the vendors. Reactor

design is an evolving process, and many design features

are plant specific. The report will attempt to identify major

design changes in internal components, but no attempt will

be made to account for all design changes. Reactor internals are identified by their names commonly used in plant

FSARs. They may be referred to by different names in

other reports, and confusions can be avoided by referring

to the detailed descriptions of the component in question.

2.1 Westinghouse (WE) Internals

WE PWR internals are divided into three structural units:

the lower core support structure, the upper core support

structure, and the in-core instrumentation support structure.

A simplified sketch of the arrangement of WE reactor

internals is shown in Fig. 2.1.

2.1.1 Lower Core Support Structure

The lower core support structure is the principal core support structure. It consists of the core barrel, the core baffle,

the lower core plate, the lower support columns, the bottom

support plate, the intermediate diffuser plate, the thermal

shield, and the secondary core support assembly. A sketch

of the lower core support structure Is shown in Fig. 2.2.

In addition to its core support functions, the lower core

support structure also directs and guides the coolant flow

through the core. The coolant enters the vessel through

inlet nozzles and flows down the annular region between

the core barrel and the vessel wall. The thermal shield,

when it is used, is located in the annular region. The main

coolant flow goes into a plenum at the bottom of the vessel

where it is turned around and then flows up into the core

through perforations of the bottom support plate, the diffuser plate, and the lower core plate. After passing through

the core, the coolant enters the upper core barrel region

occupied by the upper core support structure. The flow

then turns radially outward and leaves the core barrel

through outlet nozzles. The core barrel outlet nozzles direct

the coolant flow into the pressure vessel outlet nozzles.

In addition to the main coolant flow described above, there

are also secondary flows in the reactor coolant system. A

small amount of cooling water is diverted into the region

between the core baffle and the core barrel, and this bypass

flow provides additional cooling to the core barrel. Coolant

flow is also directed into the vessel head plenum for cooling purposes, and it exits through vessel outlet nozzles.

2.1.1.1 Core Barrel

The core barrel is the main structure of the core support

system. Other core support structures are attached to the

core barrel, which transmits the weight of the core to the

reactor vessel. It is also used to position the fuel

assemblies.

3 NUREG/CR-6048
PWR

ORNL-DW 93-2884 EM

* Control Pl dtove shat

Control rod cust (withdrawn)

Outlet noose

* Stun

D BafU radl suppon

Lower core plate

Fowimft plate

Core euppon columns

InsuISmentItln thumble quide"

Figure 2.1 Westinghouse PWR Internals

The barrel is a long, cylindrical, one-piece welded structure

divided into an upper and a lower region. The upper flange

of the core barrel rests on a ledge in the reactor vessel bead

flange, and the lower end of the barrel is welded to the

bottom support plate, which in turn is restrained by a radial

support system attached to the wall of the pressure vessel.

The outlet nozzles are made by welding forged rings to

openings in the upper part of the core barrel.

In some of the older WE reactors, the core barrel is a twopiece structure with the bottom of the core barrel connected

to the bottom support plate by tie rods. The upper and

lower barrel sections are connected by a bolted joint.

NUREG/CR-6048 4
PWR

ORNL-OWG 92-3460 ETD

Figure 2.2 Westinghouse PWR lower core support structure

2.1.1.2 Core Baffle

The core baffle forms the boundary of the core, and it also

directs and guides the coolant flow through the core region.

The core baffle is made of vertical baffle plates and horzontal former plates (also known as baffle radial support

plates). The horizontal former plates are bolted to the

inside surface of the lower part of the core barrel. The vertical plates are bolted to the inner edges of the horizontal

plates, forming the boundary of the core. The bolts are

made of type 316 stainless steel. Holes in the horizontal

former plates provide a flow path for the bypass cooling

flow in the region between the core barrel and the vertical

baffle plates. A cross section of the core showing the core

baffle and other components of the lower core support

structure is shown in Fig. 2.3.

2.1.1.3 Lower Core Plate

The. lower core plate positions the fuel assemblies and

transmits their weights to the core barrel. This perforated

plate is located at the bottom of the core below the core

baffle assembly. The plate thickness is -5.08 cm (2 in.). It

rests on a ledge on the Inside of the lower core barrel. The

holes distribute the coolant flow to the core. Fuel assembly

locating pins, two per assembly, are inserted into the lower

core plate. They provide proper alignment as well as lateral

support to the fuel assemblies.

2.1.1A Lower Support Columns

The weight of the fuel assemblies Is -113,276 kg

(250,000 lb), and to prevent excessive deformations of the

lower core plate, it is stiffened by the placement of short

columns between the lower core plate and the bottom

support plate (Fig. 2.2). The lower support columns are

also referred to as the core support columns. The top ends

of the columns are bolted to the under surface of the lower

core plate. The bolts are made of type 316 stainless steel.

The column lower ends are inserted into the bottom

support plate. A major portion of the fuel assemblies'

weight is then transmitted to the core barrel through the

bottom support plate.

5 NUREG/CR-6048
PWR

OANLWM ZN5 ETO

THERMAL

BAFFLE

ET NOZZLES

REACTOR VESSEL'

FUEL ASSEPABLY

Figure 2.3 Westinghouse PWR core barrel cross sections

2.1.1.5 Bottom Support Plate

The bottom support plate supports the lower core plate

through lower support columns. It transmits a major portion of the fuel assemblies weight to the bottom of the core

barrel.

The bottom support plate, also known as the bottom disk,

is a perforated plate welded to the lower end of the core

barrel. Typical plate thickness is -20.32 cm (8 in.). Lower

ends of the core support columns are inserted into the bottom plate. Perforations in the plate distribute coolant flow

to the core.

The bottom support plate is an integral part of the radial

support system at the lower end of the core barrel. The

support system consists of keys and keyways attached to

the bottom support plate and the reactor vessel wall.

Inconel clevis blocks are welded to the inside of the reactor

vessel at six uniformly spaced locations around the circumference. Six Inconel insert blocks are bolted to these clevises and act as keyways. Six keys, also uniformly spaced,

are welded to the outside surface of the bottom support

plate. When the core barrel is lowered into the pressure

vessel, the keys engage the keyways in the axial direction,

and lateral motions of the core barrel lower end are

restrained. A sketch of the radial key-keyway support system is shown in Fig. 2.4.

The bottom support plates in some of the older reactors are

made of Grade CF-8 cast austenitic stainless steel. When

they are casted, the bottom support plate is also referred to

as the bottom support casting. The bottom support plates

are also machined from type 304 stainless steel blocks.

2.1.1.6 Intermediate Diffuser Plate

The intermediate diffuser plate is a perforated plate located

between the lower core plate and the bottom support plate.

It is attached to the lower support columns. The intermediate diffuser plate, also called the flow mixer plate, generates a uniform coolant flow to the fuel assemblies. Not all

WE units are equipped with intermediate diffuser plates.

Perforations in the lower core plate and the bottom support

plate are used to provide the uniform flow distribution

when the diffuser plate is not used. Intermediate diffuser

plates in some of the older reactors are made of Grade

CF-8 cast austenitic stainless steel. Others are machined

from type 304 stainless steel plates.

2.1.1.7 Thermal Shield

The thermal shield protects the reactor vessel wall from

excessive gamma heating and radiation damages. It shields

the vessel wall from fast neutron fluxes and gamma radiations. It is located in the annular region between the core

barrel and the reactor vessel wall.

NUREGICR-6048 6
PWR

ORNL-DWG 03-288 ETM

LEVIS

LEVIS INSERT

ADIAL KEY

VESSEL

Figure 2.4 Westinghouse PWR radial key-keyway system

The early thermal shields are made of shell segments

assembled inside the reactor vessel. The shell segments

(typically three) were fastened together to form a cylindrical shell by the use of pins or bolts. The lower end of the

shield was keyed to support lugs attached to the bottom of

the vessel The top end was free. One WE PWR was

equipped with radial spacer pins at the top of the thermal

shield.

The intermediate thermal shield design is a one-piece

cylindrical structure and supported In one of two ways. In

one support system, the shield is rigidly bolted to the lower

edge of the core barrel, and a flexure support system is

used at the top. The flexure support is basically a structural

element with one end attached to the thermal shield and the

other end bolted to the core barrel. Six flexure supports are

spaced unifornly on the top of the shield. The flexure support system allows limited displacements at the top end of

the shield. A top-mount flexure support system is shown in

Fig. 2.5. Support conditions are reversed for the second

type of shield support system; the top has a rigid support,

and a flexure support system is used at the bottom edge.

Material surveillance samples are inserted into surveillance

specimen holder tubes that are bolted and pinned to the

outside of the thermal shield.

ORNL4.WG 9232? EMD

CORE

BARREL

FLEXURE

SUPPORT N

IN

THERMAL

SHIELD -

Figure 2.5 Westinghouse PWR thermal shield Ilexure

support system

7 NUREG/CR-6048
PWR

The latest thermal shield design uses shielding pads. Four

shielding pads, also called neutron shield pads, are bolted

and pinned to the outside of the core barrel. These pads are

located in areas with high neutron fluence levels to offer

the maximum protection to the reactor vessel wall. In this

design, material surveillance samples are located in specimen holder tubes bolted to the outside of the neutron shield

pads.

2.1.1.8 Secondary Core Support Assembly

The secondary core support assembly, often referred to as

the core catcher, is an energy-absorbing device used to

limit the vertical displacement of the core and to absorb

some of the impact energy in the event of a catastrophic

failure of the core support system. Figure 2.6 is a sketch of

a typical secondary core support assembly. The number of

energy absorbers required is plant specific, and it is determined by the condition that the maximum stress value in

reactor internals (except in the absorbers) is below the

yield stress of the material of construction for the affected

component during a postulated core drop. Yield stress values can be found in SecL III of the ASME B&PV Code.

The energy absorbers, cylindrical in shape, are attached to

a base plate that is contoured to fit the bottom surface of

the reactor vessel. The top ends of the absorbers are connected to columns that are bolted to the bottom support

plate. In the event of a core support failure, the energy

absorbers will limit the fall of the core as well as absorbing

some of the kinetic energy of the dropped core assembly.

This is accomplished through plastic deformations of a

volume of stainless steel, initially loaded in tension, in the

absorber. The plastic strain in the stainless steel piece is

limited to -15%, and then a positive stop is used to limit

the fall and transmit the additional loads to the bottom of

the reactor vessel.

2.1.2 Upper Core Support Structure

Tle upper core support structure is located in the upper

region of the core barrel. The assembly consists of the top

support plate, hold-down spring, deep beam sections, support columns, upper core plate, and guide tube assemblies.

ORNL-DWO 93-2887 Ero

CORE BARREL

8cTTON SUPPORT

PLATE

LP D

ENERGY ABSORBING DEVICE

Figure 2.6 Westinghouse PWR secondary core support assembly

NUREG/CR-6048 8
PWR

The upper core support assembly is defined by the top support plate on the top and the upper core plate in the bottom.

Proper spacing between the two plates is maintained by

support columns. Deep beam sections are attached to the

bottom of the top support plate to increase its stiffness and

to minimize deflections. Guide tube assemblies provide

housing for CRA drive shafts and rod control cluster

assemblies (RCCAs). The upper core support structure is

moved as a unit during refueling operations. A sketch of

the upper core support structure is shown in Fig. 2.7.

2.12.1 Top Support Plate

The top support plate, also called the upper support plate,

serves as the primary structure for the attachments of other

upper internal components. The plate transmits the weight

of these components to the reactor vessel.

The top support plate is a perforated flat plate that is stiffened by deep beam sections at its bottom surface. The edge

of the plate rests on top of the upper core barrel flange. The

plate is held in place by the hold-down spring, which is In

turn held down by the vessel head flange. Perforations on

the plate provide access for the RCCAs and in-core thermocouple conduits.

The top support plate is also the dividing boundary

between the upper plenum and the reactor vessel head

region. The coolant flow, which enters the upper plenum

through the upper core plate, is turned to a radial direction

TOP SUPPORT

PLATE

DEEP 

BEAM .

by the top support plate. The coolant flow exits through the

core barrel outlet nozzles.

In some WE units, the flat top support plate is replaced by

an upper support assembly. The upper support assembly is

in the form of a shallow top hat or an inverted shallow top

hat. In the top hat configuration, perforations on the top

surface provide access for the RCCA and in-core thermocouple conduits. The rim of the top hat serves as a support

flange, and it rests on the upper flange of the core barrel.

The undersurface of the top hat is stiffened by deep beam

sections. In the inverted top hat configuration, the bottom

is a thick plate, and perforations on the plate provide

access for the RCCA and in-core thermocouple conduits.

The rim of the inverted top hat rests on the upper flange of

the core barrel. The Inverted top hat upper support assembly is not stiffened by deep beam sections.

2.1.2.2 Hold-down Spring

The hold-down spring limits the axial movements of core

support structures. It is a large annular or ring spring. The

core barrel upper flange, the annular hold-down spring, and

the top support plate all rest on a ledge on the reactor vessel head flange. When the vessel head is installed on the

pressure vessel, the hold-down spring is compressed by the

tightening of the closure nuts, and the components resting

on the ledge form a rigid joint that would restrict axial

movement of the core support structures. The hold-down

spring is made of type 403 stainless steel.

ORNL-DWG 92S461 ETD

SUPPORT

COLMNS

GUIDE

TUBE

PLATE

Figure 2.7 Westinghouse PWR upper core support structure

9 NUREGICR-6048
PWR

2.1.2.3 Deep Beam Sections

Deep beam sections are welded to the bottom surfaces of

the top support plate and the upper support assembly with a

top hat configuration. The beam sections increase the plate

stiffness, and loadings acting on the plate are transmitted to

the reactor vessel without excessive deformations. Deep

beam sections are not used in the upper support assembly

with an inverted top hat configuration.

2.1.2.4 Upper Support Columns

Upper support columns are tubular structures bolted to the

top support plate and the upper core plate. The bolts are

made of type 316 stainless steel.

The upper support columns maintain the proper spacing

between the top support plate and the upper core plate.

They transmit loadings from the upper core plate to the

stiffened top support plate that, in turn, transmits the loadings to the reactor vessel. Some of the support columns

also provide structural supports to in-core thermocouple

conduits.

2.1.2.5 Upper Core Plate

The primary function of the upper core plate is to establish

proper alignment for the upper core support structure, the

lower core support structure, fuel assemblies, and control

rods. The upper core plate is a perforated plate bolted to

the lower ends of the upper support columns. The bolts are

made of type 316 stainless steel.

The upper core support structure is positioned with respect

to the lower core support structure by a pin-slot alignment

scheme. Four unifornly spaced flat-sided pins are welded

to the inside surface of the core barrel at the elevation of

the upper core plate. Slots are milled into the upper core

plate at the corresponding positions. When the upper core

support structure is lowered into the core barrel, the slots

engage the flat-sided pins in the axial direction, and proper

alignment is achieved. The pin-slot alignment scheme will

also restrict lateral displacements of the upper core support

structure.

The top ends of fuel assemblies are aligned by locating

pins protruding from the bottom of the upper core plate.

The pins engage the fuel assemblies when the upper core

support structure is lowered into place.

2.1.2.6 Guide Tube Assemblies

Guide tubes provide housings for control rod drive shafts

and RCCAs. They shield these components from effects of

cross-flows in the upper plenum region of the reactor pressure vessel. Guide plates are used to maintain proper spacing of the RCCA rodlets in the guide tubes.

The guide tube assembly is a tubular structure and is

divided into two parts or assemblies, with the upper support plate serving as the dividing line. The upper assembly,

often referred to as the control rod shroud tube, is fastened

to the top of the upper support plate. RCCAs enter the

upper plenum through openings in the reactor vessel head,

and they are inserted into the top openings of the control

rod shroud tube. The lower part is called the control rod

guide tube. The top end of the control rod guide tube is fastened to the bottom of the top support plate, and the lower

end is held in place by split pins inserted into the upper

core plate. The split pins are bolted to the guide tube bottom; the support will allow limited axial movements of the

lower end of the guide tube, but lateral displacement will

be restrained.

2.1.3 In-core Instrumentation Support

Structures

In-core instrumentation support structures are stainless

steel tubular structures that provide housing and support to

in-core instrumentation such as in-core thermocouples and

thimbles of the reactor flux-mapping system.

Thermocouple housings penetrate the vessel through the

vessel head, while the flux thimble guide tubes enter the

vessel through the bottom. In a few of the older reactors,

all in-core instrumentations penetrate the vessel from the

vessel head.

2.13.1 In-core Thermocouple Housing

In-core thermocouple housings house and guide in-core

thermocouples before their insertion into the core. The

housings start as port columns that penetrate the vessel

from the vessel head. They are also referred to as upper incore instrumentation port columns. These port columns are

slip-connected to in-line columns fastened to the top support plate. Thermocouples are inserted through these port

columns and the top support plate to positions above their

sensor locations in the core.

2.13.2 Flux Thimble Guide Tubes

Flux thimbles are components of the reactor in-core neutron monitoring or flux-mapping system. They are inserted

into the reactor pressure vessel through the bottom. Flux

thimbles are retracted during refueling and maintenance

operations. Guide tubes provide housing and support to

these thimbles along most of their lengths. Because they

NUREG/CR-6048 10
enter the reactor vessel through the bottom, they are also

referred to as lower in-core instrumentation.

A thimble guide tube is divided into two parts. The upper

part is located inside the fuel assembly, while the lower

portion is located between the reactor vessel bottom and

the lower core plate. The segment of the flux thimble

between the bottom of the fuel assembly and the top of the

lower core plate is not protected by the guide tube, and it is

exposed to the reactor coolant flows.

When they leave the reactor vessel, flux thimbles are

located inside high-pressure conduits, which penetrate the

pressure vessel from the bottom. The high-pressure conduits, containing the flux thimbles, continue on to the seal

table. A simplified sketch of the flux thimble and guide

tube arrangement is shown in Fig. 2.8.

PWR

by a mechanicil seal at the seal table. The flux thimble

tube and the high-pressure conduit are considered as a part

of the reactor primary pressure boundary.

2.2 Babcock & Wilcox (B&W)

Internals

B&W reactor internals are grouped into two main structural assemblies: the core support assembly and the plenum

assembly. In-core instrumentation guide tubes are a part of

the core support assembly, and CRA guide tubes belong to

the plenum assembly. A sketch of the arrangement of

B&W reactor internals is shown in Fig. 2.9.

The core support assembly is the primary core support

structure. The plenum assembly is used to maintain a

proper alignment of the control rod guide tubes.

2.2.1 Core Support Assembly

The thimbles are sealed at the reactor end. The region

between the flux thimble and the high-pressure conduit is

maintained at the reactor coolant pressure, and it is sealed

The core support assembly is the major core support structure, and it consists of the core support shield, the core

ORL-MG 93-28 ETD

con

SIALTABLE

UNGWIDD THIMILE lUBE

Figure 2.8 Westinghouse PWR thimble and guide tube

11 NUREGtCR-6048
PWR

ORNLDWC3 93-2889 ErD

PLENUM ASSEMBLY -

OUTLET NOZZL -

CORE BARRELLOWER GR'

FLOW OISTRUUTOR

Figure 2.9 B&W PWR Internals

barrel, the lower grid assembly, the flow distributor, the

thermal shield, surveillance specimen holder tube, in-core

instrumentation guide tubes, and internal vent valves. The

core support shield and the core barrel are the major components of the core support assembly. They provide structural support and attachment points to other internal

components.

In addition to its core support functions, the core support

assembly also directs and guides the coolant flow in the

core region. The flow enters the vessel through inlet nozzles, and the core support shield turns the coolant flow

downward in the annular region between the core barrel

and the vessel wall. The coolant flow is turned upward at

the bottom of the vessel, and it goes up to the core through

the flow distributor and the bottom grid.

NUREG/CR-6048 12
The core barrel guides the coolant flow through the core.

The main flow is upward through the fuel assemblies outlined by the core baffle. A small quantity of the coolant is

also diverted to flow upward in the region between the core

baffle and the core barrel. The coolant pressure in the core

baffle-core barrel region is maintained at a lower value

than the main coolant pressure in the core. The resulting

pressure differential helps to prevent the formation of tension stresses in the bolts attaching the vertical baffle plates

to the horizontal former plates. In the event of flow leakages through gaps in the baffle plate joints, the pressure

differential will force the coolant to move outward and

away from the fuel assemblies. This bypass flow scheme

can prevent fuel assembly damages caused by baffle plate

water-Jetting problems.

2.2.1.1 Core Support Shield

The core support shield provides structure supports to other

internals and transmits loadings to the reactor vessel. The

shield is a cylindrical structure with flanges at the ends.

The top flange is forged, and it rests on a circumferential

ledge in the reactor vessel closure flange. The core support

shield lower flange is bolted to the top flange of the core

barrel. The plenum assembly is located inside the core

support shield. The position and orientation of the plenum

assembly with respect to the core support shield are maintained by fixtures located on the shield inside surface. In

some B&W units the lower end of the core support shield

is welded to the top of the core barrel.

The core support shield wall contains two types of openings. The first type consists of outlet nozzles used for reactor coolant flow. The exact number of outlet nozzles is

plant specific. The flow nozzles are formed by welding

forged rings to the wall openings. The rings mate with

internal projections of the reactor vessel outlet nozzles. The

ring seal surfaces are sized and finished so that a clearance

gap is maintained at cold condition to facilitate the installation and removal of components of the core support

assembly. When the reactor is heated to the operating temperature, the differential thermal expansions of the stainless steel core support shield and the carbon steel reactor

vessel will close the gap and form the necessary seal surface. The cold gap is sized so that when contact is made,

maximum stresses in the reactor vessel and internal components remain below allowable values as stipulated in

Sect. M of the ASME BPV Code.

The second types of opening in the shield wall are internal

vent valve openings. Mounting rings are welded to these

openings for the installation of the internal vent valve

assemblies. The exact number of vent valve openings is

plant specific, but most B&W plants are equipped with

four to ight internal vent valves.

PWR

2.2.1.2 Core Barrel

The core barrel supports the fuel assemblies, the lower grid

assembly, flow distributor, and in-core instrumentation

guide tubes. It guides the primary coolant flow through the

core.

The core barrel is a cylindrical structure with flanged ends.

The upper flange is bolted to the lower flange of the core

support shield, and the lower flange is bolted to the lower

grid assembly. In some B&W units, the top end of the core

barrel is welded to the lower end of the core support shield.

The core baffle is considered as an integral part of the core

barrel. The baffle is formed by bolting a series of horizontal former plates to the inside surface of the core barrel.

Vertical baffle plates are then bolted to the inside edges of

the horizontal former plates. The vertical baffle plates form

the boundary of the fuel assemblies.

2.2.1.3 Lower Grid Assembly

The lower grid assembly provides structural supports to the

fuel assemblies, the thermal shield, and the flow distnbutor. Fixtures attached to the lower grid assembly also align

the in-core instrumentation guide tubes with fuel assemblies' instrument tubes.

The assembly consists of two grid structures connected by

short tubular columns and surrounded by a forged cylinder

with flanges at the ends. The upper grid structure is a perforated plate attached to the top flange of the forged cylinder. Pads bolted to the perforated plate are used to align the

fuel assemblies and in-core instrumentation guide tubes.

The lower grid structure is formed by welded intersecting

plates. In some units, the lower grid structure is a machined

forging.

The lower and upper grid structures are connected by tubular columns. Also a perforated plate located midway

between the two grid structures is used to generate a uniformly distributed coolant flow to the core. The lower grid

structure rests on and is also bolted to the lower flange of

the core support shield. The top flange of the forged cylinder is bolted to the lower flange of the core barreL

22.1.4 Flow Distributor

The flow distributor is a perforated dished head with an

external flange that is bolted to the bottom of the lower

grid assembly.

13 NUREGICR-6048
PWR

Perforations in the flow distributor produce a uniform

coolant flow to the core entrance. The flow distributor also

provides support to the in-core instrument guide tubes.

2.2.1.5 Thermal Shield

The thermal shield protects the reactor vessel wall from

excessive gamma beating and radiation damages. It shields

the vessel wall from fast neutron fluxes and gamma radiations. The shield is located in the annular region between

the core barrel and the reactor vessel wall.

The thermal shield is a cylindrical structure. The upper end

of the shield is restricted against radial vibratory motions

by restraints bolted to the core barrel cylinder. The lower

end of the shield is shrunk-fit onto the upper flange of the

forged cylinder in the lower grid assembly. As an added

assurance, 96 high-strength bolts are also used to secure

the thermal shield to the upper flange of the forged cylinder. The bolts are made of grade A 286 stainless steel. A

sketch of the thermal shield lower support is shown in

Fig. 2.10.

2.2.1.6 Surveillance Specimen Holder Tube (SSHT)

SSHTs are cylindrical tubes mounted on the outside surface of the thermal shield. Each tube can hold two surveillance capsules, and the exact number of tubes in a reactor

is plant specific. There is an off-set for the tubes, and they

are located at a short distance from the thermal shield outside surface. The span of the tube typically extends from

the top of the core support shield to the lower end of the

thermal shield. There are usually three support mounts for

each holder tube.

2.2.1.7 In-core Instrumentation Guide Tubes

In-core instrumentation guide tubes provide housing and

supports to in-core instruments from penetrations in the

reactor vessel bottom head to instrument tubes in the fuel

assemblies. Guide tubes are segmented tubular structures

with different diameters for the various sections. The lower

end of the tubes are connected to instrument penetrations at

the vessel bottom head. The tubes then go through openings in the flow distributor and the lower grid assembly

before being connected to instrument tubes in the fuel

assemblies.

2.2.1.8 Internal Vent Valves

The use of internal vent valves is a unique design feature

of B&W PWRs. The primary function of the internal vent

valve is to release the pressure buildup in the reactor upper

plenum region following a cold-leg pipe break accident.

The internal vent valve assemblies are attached to mounting rings welded to openings in the wall of the core support

ORNL-DWG 92-3328 ET

THERMAL n oEL 

0.00 TO 0.060

INTERFERENCE

FIT

LOWER GRID

FORGING

96-0.8' BOLTS

* LOCXING

CLIPS

Figure 2.10 B&W PWR thermal shield lower support

NUREG/CR-6048 14
shield. The mounting rings also contain devices and features that are used to align and position the valve assembly

for proper operation of the valve disk. The internal vent

valve assembly consists of a hinge assembly, the valve disk

with sealing surfaces, split-retaining rings, and fasteners.

The design of the internal vent valve is similar to that for

swing check valves. The vent valves are uniformly spaced

around the circumference of the core support shield wall,

and the exact number of valves required is plant specific. A

sketch of an internal vent valve assembly is shown in

Fig. 2.1 1.

Under normal reactor operating conditions, the coolant

pressure in the annular region between the core support

shield and the reactor vessel wall (downcomer) is higher

than the coolant pressure in the core. The magnitude of this

pressure differential during normal operations is

-0.29 MPa (42 psi). It will generate a large force that will

press the valve disk seal face tightly against the tilted valve

body seal face. Thus internal vent valves are closed during

normal operations.

In the event of a cold-leg pipe break accident, the sudden

loss of pressure would lead to rapid phase changes in the

primary reactor coolant. The coolant becomes a two-phase

mixture of steam and water. The pressure in the core region

will drop from the operating pressure of 15.5 MPa

(2250 psi) to -10.6 MPa (1541 psi), which is the saturated

pressure at the operating temperature of 316C (6000F).

The large thennal inertia of the core would prevent a rapid

change in the fluid temperature in the core region. A large

pressure differential now exists between the core region

and the pipe break location, which is at the ambient pressure. The pressure differential will drive the reactor coolant

from the core region to the pipe break location through two

PWR

potential flow paths. The first flow path is through the hot

leg and the steam generator. The second flow path is

through the core and the cold leg. The flow rate through

each path is determined by hydraulic resistances in each

flow path. In any event, reactor coolant is transported from

the core region to the pipe break location. If the process is

not mitigated, the water level in the core will drop, and a

portion of the core may be uncovered. Note that the process is dynamic, and the event could take place in a very

short time interval.

The internal vent valve is designed to achieve a quick pressure equalization between the core region and the cold-leg

pipe break location and to prevent the lowering of the water level in the vessel below the top of the core. After the

pipe break occurs, decompression waves propagate from

the cold-leg pipe break location into the reactor vessel. The

fluid pressure of the core region becomes higher than the

pressure in the annular region between the core support

shield and the vessel wall. The direction of the resulting

pressure force will now oppose the valve closing force

generated by the hinge assembly. The valve closing force

is overcome when the magnitude of the reversed pressure

differential attains a value of 1034 Pa (0.15 psi). The valve

will be fully opened with a pressure differential of no more

than 2069 Pa (03 psi). When the internal vent valves are

opened, the major reactor coolant flow path, the one with

the least flow resistances, will be from the upper plenum of

the core region to the cold-leg pipe break location. The

opening of internal vent valves together with the activation

of the emergency core cooling system can prevent the

uncovering of the core in a cold-leg pipe break accident.

Because of their important safety function, internal vent

valves are inspected and tested during refueling outages.

ORNL4-WG 929 ETD

rI-Z

-EXERCISE WG

SECTION Z-Z

- REACTOR

VESSEL

WALL

L-Z

Figure 2.11 B&W PWR Internal vent valve

1S NUREG/CR-6048
-

PWR

The valves become accessible to visual inspections and

mechanical testings after the vessel head and the plenum

assembly are removed. A hook tool is used to engage the

exercise lug, and the freedom of the movement of the

hinged valve disk is tested. When the valve disk is raised,

a remote visual inspection is also performed on the valve

body and the sealing faces.

2.2.2 Plenum Assembly

The plenum assembly positions and supports the CRA

guide tubes. It also provides a path for the reactor coolant

to flow to the reactor outlet nozzles.

The assembly is made up of the plenum cylinder, the

plenum cover, the upper grid, and CRA guide tubes. It is

located inside the core support shield and directly above

the core. The structural unit is formed by attaching the

plenum cover and the upper grid to the upper and lower

flanges of the plenum cylinder. CRA guide tubes connect

the plenum cover and the upper grid. Lifting lugs are

welded to the plenum cover, and the plenum assembly is

removed as a single unit during refueling operations.

The plenum assembly is aligned with the reactor vessel

closure head, control rod drive penetrations, and the core

support assembly. When the bolts of the vessel closure

head are tightened, a large clamping force is produced

between the reactor vessel, vessel closure head, the core

support shield upper flange, and the plenum cover assembly upper flange. The clamping force helps to form a rigid

joint between the components involved, and it will limit

movements of these components. Proper positioning is

attained by the locking of keyways in the plenum assembly

cover flange to keys in the reactor vessel flange. The bottom of the assembly, which is the upper grid, is restrained

by the inside surface of the core support shield.

2.2.2.1 Plenum Cylinder

The plenum cylinder positions and supports CRA guide

tubes and guides the reactor coolant flow to the outlet nozzles. It is a cylindrical structure with flanges at the two

ends. The side wall contains holes as outlets for the reactor

coolant flow. The plenum cover is attached to the top

flange, and the upper grid is connected to the lower flange

of the plenum cylinder.

2.2.2.2 Plenum Cover

The plenum cover aligns the upper ends of CRA guide

tubes and provides support points for the lifting of the

plenum assembly. The cover is formed by welding parallel

intersecting plates to form square lattices, and the lattice

structure is covered by a perforated flat plate with an integral flange at the periphery. The inner edge of the integral

flange is attached to the upper flange of the plenum cylinder, while the outer edge of the flange is attached to the

upper flange of the core support shield. Holes in the perforated plate are positioned to match the upper ends of the

CRA guide tubes. Lifting lugs are welded to the plenum

cover.

2.2.2.3 Upper Grid

The upper grid aligns the lower end of CRA guide tubes to

the upper ends of fuel assemblies. The grid is a perforated

plate bolted to the lower flange of the plenum cylinder, and

it is also guided by the inside surface of the lower flange of

the core support shield. Perforations in the plate position

the lower ends of CRA guide tubes to the upper ends of

fuel assemblies.

2.2.2.4 CRA Guide Tubes

CRA guide tubes provide housing and support to the

CRAs. The tubes also shield the CRA from the effects of

cross-flows in the plenum assembly. Guide tubes are also

considered as structural components connecting the

plenum cover to the upper grid.

A CRA guide tube is a tubular structure with a mounting

flange welded to one end. The mounting flange is bolted to

an opening of the upper grid, while the top end of the guide

tube is welded to a perforation in the plenum cover plate.

CRAs are located inside the guide tubes.

2.3 Combustion Engineering (CE)

Internals

CE internals are divided into four structural units: the core

support assembly, the upper guide assembly, the flow skirt,

and the in-core instrumentation support system. A sketch

of the arrangement of CE reactor internals is given in

Fig. 2.12.

2.3.1 Core Support Assembly

The major core support structure is the core support

assembly. The assembly consists of the core support barrel,

the core support plate, the lower support structure, the core

shroud, thermal shield, and the core support barrel to pressure vessel snubbers. Because of flow-induced vibration

problems, thermal shields have been removed from several

CE reactors. A typical core support assembly is shown in

Fig. 2.13.

NUREG/CR-6048 16
PWR

ORNo 38W ED

UN-CORE INSTRUMENTATION UPPORT PLATE

Figure 2.12 CE PWR Internals

The core support assembly also directs and guides the

reactor coolant flow through the core region. The coolant

enters the pressure vessel through inlet nozzles and is

turned to flow downward in the annular region between

the core support barrel and the reactor vessel wall. The

coolant flow is turned upward in the lower plenum. It

flows through the flow skirt, the bottom of the core support

barrel, and the lower support structure before entering the

core through flow distribution holes in the core support

plate. At the cow exit, the coolant flow is turned to a radial

direction in the upper region of the core support barrel and

leaves the bawrel through outlet nozzles. A small portion of

the main coolant flow Is diverted into the gap region

between the core shroud and the core support barrel. This

secondary flow is used to provide a more effective cooling

of the core shroud. Other minor bypass flows are leakage

flows through key-keyway alignment systems, instrumentation guide tubes, and nozzle clearances.

23.1.1 Core Support Barrel

The core support barrel transmits the weight of the fuel

assemblies to the reactor vessel. A typical total weight

of the fuel assemblies and claddings for a CE unit is

-133,333 kg (-300,000 lb). The barrel also provides proper

alignment for the core support assembly, the upper guide

structure assembly, the reactor vessel, and the closure head.

The core support barrel is a long cylindrical structure with

an external ring flange at the top and an internal ring flange

at the bottom. The top flange rests on a ledge in the reactor

pressure vessel. Four equally spaced alignment keys are

press-fitted into the top flange of the core support barrel.

The reactor vessel, closure head, and the upper guide structure assembly flange are slotted to fit into the alignment

key locations. The keys are used to align the core support

assembly and the upper guide assembly with respect to the

reactor pressure vessel. A cut-away view of the core support barrel is shown in Fig. 2.14.

Two outlet nozzles in the upper section of the core support

barrel are fitted to internal projections of the reactor vessel

exit nozzles. Four equally spaced guide pins are attached to

the inside of the core support banmel just below the outlet

17 NURECR-6048
PWR

ORNL-OW 93oa EMD

OUTLET

I II I I

Figure 2.13 CE PWR core support assembly

nozzles, and the pins are used to align and to limit the

movement of the upper guide assembly.

The core support plate rests on a ledge in the inside surface

of the core support barrel. The ledge is located near the

bottom of the barrel. The lower support structure rests on

the internal ring flange at the bottom of the barrel.

The lower end of the core support barrel is restrained by

snubbers located on the outside surface of the barrel. The

snubbers allow radial and axial expansions of the core support barrel, but they will restrict lateral or circumferential

barrel displacements.

2.3.1.2 Core Support Plate

The core support plate positions and supports the fuel

assemblies. It is a perforated flat plate, and the perforations

distribute coolant flow to the fuel assemblies. The fuel

assemblies rest on the core support plate and are positioned

by locating pins (four for each assembly) shrunk-fit to the

plate. The core support plate rests on a ledge located in the

inside surface of the core support barrel near the bottom

end. The periphery of the plate is pinned, bolted, and lock

welded to the ledge. The core shroud is also attached to the

core support plate.

NUREGICR-6048 18
PWR

ORNUWM 034M Io

Figure 2.14 CE PWR core support barrel

bolts. The gap in the region between the core boundary and

the core support barrel is maintained by seven tiers of horzontal centering plates. These centering plates are bolted to

vertical core sMiud plates and centered during assembly

by adjusting bushings located in the core support barrel.

All bolted joints in the core shroud are lock-welded. The

location of the core shroud in the core support barrel is

shown in Fig. 2.14.

Most of the coolant flows through the core. Holes are

drilled in the horizontal plates to provide some coolant

flow through the region between the core shroud and the

core support barrel. The flow in the gap region will eliminate stagnation pockets and provide a more effective cooling to the shroud. Improved shroud cooling would reduce

temperature gradients and minimize thermal stresses in the

shroud and the core support barrel.

2.3.15 Core Support Barrel to Reactor Vessel

Snubbers

The upper end of the core support barrel is clamped

between the pressure vessel flange and the closure head.

From a structural analysis standpoint, the core support

barrel has a cantilevered support at the upper end, and no

additional support would be needed. However, the core

support barrel is submerged in a turbulent flow and is susceptible to flow-induced vibrations. The amplitudes of

such flow-induced vibrations would be larger at the lower

end of the barrel when it is not restrained. Snubbers are

installed at the lower end of the core support barrel to limit

and reduce the amplitude of potential flow-induced vibrations. The locations of the snubbers on the core support

barrel are shown in Fig. 2.14.

The snubber system consists of six equally spaced lugs

welded to the core support barrel outside surface. The core

support barrel lugs act as the groove of a tongue-andgroove assembly. Mating lugs attached to the inside surface of the pressure vessel wall serve as tongues. An

exploded view of the snubber assembly is shown in

Fig. 2.15.

When the core support barrel is lowered into the pressure

vessel, the pressure vessel lugs slide into the grooves of the

core support barrel lugs. Shims are bolted to the side surfaces of the vessel lugs, and the corresponding surfaces of

the barrel lugs are hard faced to minimize wear. There is

no direct connection between the lugs. The gap clearance

between mating surfaces will impose a limit on the barrel

displacements. The snubber system will not impede thermal expansions in the axial and radial directions. It wil,

however, restrict lateral or circumferential expansions. The

23.13 Lower Support Structure

The lower support structure provides additional stiffness to

the core support plate and also transmits a part of the core

assembly weight to the bottom of the core support barrel.

The lower support structure is a welded assembly that

consists of a cylinder, columns, support beams, and a bottom plate. The core support plate is supported by columns,

and the bases of these columns are welded to support

beams. The bottoms of the support beams are welded to the

bottom plate, which is perforated for coolant flow passage.

The ends of the support beams are welded to the cylinder.

The lower end of the cylinder is welded to the perforated

bottom plate. The whole assembly rests on the internal ring

flange of the core support barrel. The outer edge of the

core support plate is also supported by the top end of the

cylinder. The cylinder guides the main reactor coolant flow

to the core and limits the core shroud bypass flow by

means of holes located near the bottom of the cylinder.

23.1A Core Shroud

The core shroud forms the boundary of the core and controls the coolant flow through the core region. The shroud

is formed by bolting vertical shroud plates to horizontal

centering plates. The vertical plates, of varying widths,

become the core boundary. The bottom of the vertical

plates are attached to the core support plate by anchor

19 NUREGR-6048
-

PWR

ORNL 2S3 EMD

CORE SUPPORT BARREL

.1ra PIN14 REQ'D PER ASSEMBLY I

SOLT 4 REQ'D PER ASEMBLY )

HIM I 2 REQ'D PER ASSEMBLY 

Figure 2.1S CE PWR core support snubber assembly

snubber system is not a fixed support system, and the core

suppoit barrel is not immune from small-amplitude

vibrations.

23.1.6 Thermal Shield

The thennal shield is a cylindrical structure located in the

annular region between the core support barrel and the

reactor vessel wall. The upper end of the shield is supported by nine uniformly spaced lugs attached to the outside of the core support barrel. The lugs restrict the axial

and tangential movement of the shield. A preloaded positioning pin under each lug is threaded radially through the

shield and butts against the core barrel. The lower end of

the shield is held in place in a similar manner by 17 radial

positioning pins.

Because of flow-induced vibration problems, thermal

shields had been removed from several CE reactors.

Analysis results indicated that it is not necessary to replace

these shields in the remaining design life of these reactors.

Thermal shields are not used in the newer CE units.

2.3.2 Upper Guide Assembly

The upper guide assembly consists of the upper guide

structure support plate assembly, the fuel assembly alignment plate, control element assembly (CEA) shrouds, and a

hold-down ring. The assembly is handled as a single unit

during refueling operations. A sketch of the upper grid

assembly is shown in Fig. 2.16.

The upper guide structure (UGS) support plate assembly

provides support and alignment to CEA shrouds, which

shield the control rods from cross-flow effects in the upper

plenum region. The CEA shrouds also provide support to

in-core instrumentation. The fuel alignment plate positions

the upper ends of the fuel assemblies. The hold-down

spring holds down the fuel assemblies during normal operation and prevents the fuel assemblies from being lifted out

of the core during accidents.

23.2.1 UGS Support Plate Assembly

The UGS support plate assembly is a welded structure. The

assembly is constructed by welding a support flange to the

top of a cylinder, and a support plate is welded to the

cylinder inside surface at the cylinder midsection. In some

CE units the support plate is located near the top of the

cylinder, just below the support flange. The top of a grid

array structure, made of welding intersecting deep beams,

is welded to the bottom of the support plate. The ends of

the deep beams are welded to the cylinder inside wall. The

support plate and the deep beam grid array structure position and support the upper ends of CEA shrouds. Four

NUREGCR-6048 20
PWR

ORNL-DWG 03-2894 EM

MEL AsSEMBLY -

ALIGNMENT PLATE

Figure 2.16 CE PWR upper grid assembly

emay spaced keyways are machined into the top support

flange, and they engage the core support barrel alignment

keys. The key-keyway alignment scheme ensures a proper

alignment of the core with respect to the reactor closure

head and CEA drive mechanisms.

machined Into the support barrel ring flange and the holddown ring, and they will engage core support barrel alignment keys. The bottom plate of the support barrel provides

support and alignment to CEA shrouds.

2.3.2.2 Fuel Assembly Alignment Plate

In the new CE umits, the UGS support plate assembly Is

replaced by a UGS support barrel assembly. The UGS support barrel assembly consists of a ring flange welded to the

top of a circular cylinder. A circular plate is welded to he

bottom of the cylinder. The ring flange rests on the holddown ring that, in tn, sits on the core support barrel

upper flange. Four uniformly spaced keyways are

The fuel assembly alignment plate positions the upper ends

of the fuel assemblies and also provides support to the

lower ends of CEA shrouds. Locating holes are machined

Into the plate to engage posts on the fuel assembly upper

end fittings. Four equally spaced slots or keyways are also

machined into the outer edge of the fuel assembly alignment plate, and they engage pins protruding from the core

21 NUREG/CR-6048
PWR

support barrel. The pin-keyway arrangement restricts lateral displacements of the upper guide assembly during

operations. The fuel assembly alignment plate presses

down on the fuel assembly hold-down ring, and the upward

reaction forces are transmitted via the alignment plate and

CEA shrouds to the flanged UGS support plate.

23.2.3 CEA Shrouds

CEA shrouds in most CE reactors are tubular structures.

They extend from the fuel assembly alignment plate to an

elevation above the support plate of the UGS assembly.

The shrouds protect CEAs from cross-flow effects in the

upper plenum.

The majority of CEA shrouds are the five-element type,

and they are made by welding a cylindrical section to a

base; the base is bolted and lock-welded to the fuel assembly alignment plate. Flow channel inlets are machined into

the cylindrical section at the base, and they serve as a passageway for the coolant flow through the fuel assembly

alignment plate. The upper ends of the shrouds are

connected to the UGS support plate by spanner nuts, which

would allow the shrouds to expand in an axial direction.

Four-element shrouds are located at the periphery of the

UGS support plate, and they consist of a cylindrical section

welded to a base; the base is bolted and lock-welded to the

fuel assembly alignment plate. The upper section of the

shroud is welded directly to the UGS support plate.

In the older reactors, the CEA shrouds have a cruciform

configuration, and they extend from the fuel assembly

alignment plate to an elevation just above the reactor vessel

flange. The shroud is fabricated by welding four formed

plates to four end bars to complete a cruciform-shaped

structure. The shroud ends are fitted with support pads.

The bottom ends are bolted and lock-welded to the fuel

assembly alignment plate and the top ends to the UGS support plate. CEAs located inside these crucifonn shrouds

also shield them from cross-flow effects.

2.3.2.4 Hold-down Ring

The hold-down ring restricts axial displacements of internal components. Differential thermal expansions, fuel

growth, and rotations of the closure head during bolt tightening and pressurization are major causes of internal component axial displacements. The hold-down ring is also

referred to as the expansion compensation ring.

ring segment contains plungers supported by Belleville

washers, and the compression of these devices will result

in an axial hold-down force acting on the upper guide

assembly. The ring segments are fabricated from type 403

stainless steel.

A shim plate is inserted into the space between the UGS

and the core support barrel flange to accommodate internal

components' axial expansions.

2.3.3 Flow Skirt

The flow skirt is a perforated right circular cylindrical

structure with stiffening rings at its top and bottom. The

skirt is supported by nine equally spaced machined sections welded to the bottom head of the reactor pressure

vessel. There is no connection between the flow skirt and

other internal components. The skirt is made of Inconel.

The flow holes are designed to provide a uniform inlet

flow to the core.

2.3.4 In-core Instrumentation Support

Structure

The in-core instrumentation support structure is a part of

the in-core neutron flux monitoring system. It consists of

an instrumentation support plate that fits in the recess section of the UGS assembly and is supported by four bearing

pins. CEA shrouds extend through perforations in the

instrumentation plate. The in-core instruments are guided

and protected by in-core instrumentation guide tubes that

route the instruments to various locations in the core. The

guide tubes are bent and grouped together to form cluster

assemblies above the instrumentation plate. The clusters

are supported by frame-type structures bolted to the instrumentation plate, and they extend into the reactor vessel

head instrumentation nozzles. A sketch of an in-core

instrumentation support structure is shown in Fig. 2.17.

Reference

1. American Society of Mechanical Engineers, ASME

Boiler and Pressure Vessel Code, Section III, Nuclear

Power Plant Components, Div. 1.*

Available from American National Standards Institute, 1430 Broadway,

New York, NY lOOS, copyrighted.

The hold-down device is a segmented circular frame, and

each ring segment is bolted to the flange of the UGS. Each

NUREG/CR-6048 22
PWR

ORNL4OWG 932895 EM

4- WGUIDE TUBE

Figure 2.17 CE PWR In-core Instrumentation support structure

23 N2R3G/CR 6048
3 Primary Stressors

In the context of aging studies, stressors are conditions that

will promote the development and sustain the growth of

aging-related degradation mechanisms. Applied loads,

environmental conditions, and manufacturing processes

can Impose stressors on reactor internals.

Applied loading is an important stressor for internal components. The majority of the internals are not parts of the

reactor coolant primary pressure boundary and are not

subjected to static pressure differential loadings. Flowgenerated oscillatory hydrodynamic forces and preloads in

bolts are the applied loadings of concern to reactor

internals.

The operating environment inside the pressure vessel also

imposes many stressors on internal components. Reactor

internals are submerged in the reactor primary coolant

low. The coolant temperature in the core region, where

most internals are located, is -3160C (6000F), and the

average coolant flow speed is -49 ms (16 fWs). The nominal system pressure in the reactor vessel is -15.5 MPa

(2250 psia). These conditions can generate many agingrelated stressors. Normal reactor operating conditions for

the three types of PWRs may vary, but In general they do

not deviate significantly from conditions mentioned above.

Because of their proximity to the core, some internals are

exposed to stressors associated with fast neutron fluxes and

the heating of gamma radiation.

Accidents can impose much more severe thermal and

mechanical loadings on reactor components. However, the

emphasis of the present study is on effects of normal plant

operations, which constitute the great majority of the operating history of the reactor.

3.1 Applied Loads

Thernal and mechanical loads are the major applied loads

acting on reactor internals during normal steady-state and

transient (startup or shutdown) operations. Thermal loads

are produced by temperature gradients in a component, by

thermal expansions of different materials, and by restricted

thermal expansions. Mechanical loads are generated by

static pressure differentials, preloads in bolts, and fluidflow-generated cyclic forces.

Reactor internals are designed to accommodate thermal

expansions; as a result, constraint-induced thermal loads

are kept at a low level. The existence of temperature gradients and rapidly changing temperatures in a component are

important thermal stressors. They can lead to thermal

cycling and fatigue crack initiation.

The two major sources of static applied loads are differential pressure loads and preloads in bolts. Most internals are

not a part of the reactor primary pressure boundary and are

not subjected to large static differential pressure loads.

Preloads in bolts can produce tensile stresses that may

reach the yield stress value of the bolt material. Tensile

stresses of such a magnitude are considered as an essential

factor to the development of stress corrosion cracking

(SCC).1

The applied loads of primary concern to reactor internals

are flow-induced oscillatory hydrodynamic forces. The

volume flow rate for PWRs ranges from 250 to 380 Us

(4000 to 6000 gal/min), and the coolant is forced through

the reactor vessel by reactor coolant pumps. Major sources

of flow-related excitations are the pump-generated pressure

pulsations at the pump rotational speed, blade passing

frequencies, and their harmonics. Pressure pulsations can

act as periodic forcing functions on reactor internals, and

their effects are most pronounced at the entrance regions of

the vessel inlet nozzles.

Two types of hydrodynamic forces act on a blunt object

when it is placed in a flow stream. The first type is a static

load and is usually referred to as a drag force. The second

type is a time-dependent force induced by flow separations. The weights of the components and structural supports are sufficient to counterbalance the drag forces. The

cyclic or periodic flow-generated hydrodynamic forces

may force a structure to vibrate. Flow-induced vibrations

can lead to fatigue failures and mechanical wear.

A third source of flow-related stressors is the highly turbulent flow generated by the forced flow through gaps and

other small openings. A structure located in the wake of

such high-intensity turbulent flows can undergo vibrations

caused by the time-dependent hydrodynamic forces. Baffle

plate water-jetting is an example of such flow-induced

vibration problems.

3.2 Environmental Stressors

The primary environmental stressors for reactor internals

are related to the operating environment inside the reactor

pressure vessel. They include contact with the primary

25 NURBGtCR-60-48
-

Pimary

reactor coolant flow and exposure to fast neutron fluxes.

Long-term contacts with a high-temperatute fluid medium

and exposures to fast neutron fluxes may lead to physical

changes and deterioration in some of the materials of construction for reactor internals. Neutron irradiation effects

may cause embrittlement and irradiation-assisted SCC in

stainless steel components. Parts made from cast austenitic

stainless steels (CASSs) are susceptible to thermal aging

effects.2

The coolant is borated water of high purity. The corrosiveness of the coolant is determined by the quantity of dissolved oxygen content, concentrations of impurities, and

boric acid present in the flow stream. The dissolved oxygen is a product of the radiolytic reactions in the core, and

the process is generic to reactor operations. The hydrogen

overpressure system in the volume control tank of the

chemical and volume control system adds hydrogen gases

to the flow stream, which act as scavengers and remove

most of the dissolved oxygen. However, it is possible that

locally high concentrations of dissolved oxygen and other

impurities may exist in crevices in some internal components. PWR internals are susceptible to corrosion attacks.

Chlorides and fluorides are the two impurity components

in the reactor primary cooling water that are of concern to

reactor internals. They can be introduced into the flow system by condenser leakage and as impurities in the reactor

make-up water. Impurities may be trapped in crevices, and

their concentrations may reach such a level that corrosion

cracks can be initiated in the affected components.

Boron, as boric acid dissolved in the cooling water, is the

preferred nuclear poison used for reactivity control in

PWRs. Some reactor components, usually made of highnickel alloys, are susceptible to corrosion attacks in boric

acid solutions under a stagnant condition. Reactor internals

are submerged in a flowing fluid medium, and they are

usually not sensitive to boric acid corrosion attacks.

An important environmental stressor for some internals is

the exposure to fast neutron (E > I MeV) fluxes. The neutron irradiation effects are most pronounced for components located in the immediate vicinity of the core.

Prolonged bombardments by neutrons can change the

mechanical and physical properties of the materials.

Specifically, they will increase yield and ultimate strengths

and reduce the uniform elongation to fracture and fracture

toughness. Irradiation effects can also lower the temperature at which creep can become a significant deformation

mechanism. Exposure to neutron fluxes can also lead to a

lowering of the threshold stress level that is considered as

necessary for the development of SCC.3 These changes

NUREG/CR-6048

could have potentially adverse effects in the structural

integrity of reactor internals.

3.3 Manufacturing Stressors

The processes used to fabricate reactor components may

introduce stressors to the finished parts. Welding, bolting,

cold working, and casting are four common processes used

in the making of reactor internals. Stressors are associated

with each of these processes.

Welding is a common method for attaching components

together to form an integral structural unit. Austenitic

stainless steel such as type 304 may be sensitized in a

welding process. The chromium depletion sensitization

process can make the finished products susceptible to SCC.

Residual stresses in weldments, if not properly heat

relieved, may also contribute to the development of SCC.

Many reactor core support structures are joined together by

bolts. Gaps and crevices in bolted joints can create a local

environment that is conducive to the development of corrosion attacks. Preloads in bolts and the resulting tensile

stresses are stressors that can aid the SCC process.

Cold working is used in the making of some internal components. The component is formed to a predetermined

shape by a bending operation. Plastic strain accumulation

and surface flaws are the stressors associated with a cold

working operation. They can lead to crack initiations and

accelerated crack growths.

Some stainless steel internals are cast in one piece. While

casting may eliminate many of the stressors associated

with welding, bolting, and cold working, it is not a

stressor-free manufacturing operation. CASS components

are susceptible to thermal aging effects.

Stressors imposed on reactor internals by applied loads,

environmental conditions, and manufacturing processes are

the basic ingredients for the development of aging-related

degradation mechanisms. Stressors that are present in a

system, whether acting independently or in conjunction

with others, will initiate the aging degradation process.

Degradation mechanisms develop at different rates, and,

unless correction or preventive measures are taken, will

eventually lead to failures in the affected components. A

review of the failure history of reactor internals can provide useful information on the relative development rates

of potential aging-related degradation mechanisms.

Understanding of the synergistic relationship between

26
stressors and aging degradations is also the basis for the

formulation of strategies for managing aging effects.

References

1. R. L. Cowan and C. S. Tedmon, -Intragranular Corrosion of Iron-Nickel-Chromium Alloys," Advances in

Corrosion Science and Technology, Vol. 3, M. S.

Fontana and R. W. Staehle, Eds., Plenum Press, New

York, 1973.

2. 0. K. Chopra and H. M. Chung, "Long-Term Aging of

Cast Stainless Steel: Mechanisms and Resulting

Primary

Properties," in Proceedings of the Fifteenth Water

Reactor Safety Research lqfornation Meeting,

Gaithersburg, MD, USNRC Conference Proceedings

NUREG/CP-0090, October 1987*

3. A. J. Jacobs and G. P. Wozaldo, "Irradiation Assisted

Stress Corrosion Cracing as a Factor in Nuclear

Power Plant Aging:' J. Material Engineering 9(4)

(1988).t

Avaiale for purchase from he National Technical Information Service,

Springfield. VA 22161.

tAvailable in public tecnical libraries.

27 NUREG/CR-6048
4 Aging-Related Degradation Mechanisms

The development of an aging-related degradation mechanism is a time-dependent process. The process is initiated

when the necessary stressor or stressors are present in the

operating environment of the affected components. Once

started, an aging degradation mechanism will remain

active, and the deteriorating effects are cumulative. The

affected components will eventually fail unless mitigating

methods are used to remove the responsible stressors from

the operating environment When it is not feasible to eliminate the stressors, then it is essential to replace the

degraded components before failure. One of the objectives

of an aging study is to Identify potential aging-related

degradation mechanisms associated with the primary stressors of the system.

Aging-related degradation mechanisms generally associated with the operating environment of reactor internals are

corrosion, fatigue, erosion, mechanical wear, embrittlement, creep, and stress relaxation. Corrosion includes general wastage and SCCs. Erosion is the loss of material

caused by the flow of an abrasive and/or high-velocity

fluid medium. Fatigue is vibration induced, and the responsible oscillatory forces can be either mechanical or thermal

in nature. Vibration can also lead to mechanical wear and

fretting. Embrittlement decreases the fracture toughness of

a material and is caused by thermal aging and/or irradiation

effects. Creep and stress relaxation may lead to changes in

material properties and structural deformation mechanisms.

They are usually caused by prolonged exposure to high

temperatures and irradiation effects. The operating environment inside a reactor pressure vessel contains stressors

that can activate all these potential aging-related degradation mechanisms.

4.1 Corrosion

Corrosion is a term applied to a class of aging effects in

which the structural integrity of a component is weakened

as a result of material deterioration caused by electrochemical reactions with the surrounding medium. The

effects can be highly localized, or they can cover a large

portion of the structure. The localized effects usually take

the form of crack initiation and development, while the

more global effects are general corrosion and wastage.

Operating conditions, corrodents, and alloy compositions

would determine the dominant corrosion mechanisms in a

particular situation.

4.1.1 General Corrosion and Wastage

When a structure is submerged in a corrosive medium, the

surface of the structure can become oxidized, and corrosion

products can be removed by the fluid. The corrosion process occurs more or less uniformly over the entire contact

surface. A moving fluid medium can increase the corrosion

product removal rate. Austenitic stainless steels, specifically type 304, have good resistance to general corrosion

and wastage.1 General corrosion and wastage are not considered as significant aging-related degradation mechanisms for reactor internals made of stainless steels.

4.1.2 Corrosion Cracking

Materials that provide good resistance to general corrosion

and wastage, such as stainless steels, are often susceptible

to corrosion cracking. These failures are localized, and

they have the appearance of microscopic brittle fractures.

There is no visual indication of the presence of corrosion

products, and if the cracks are not detected, failures can

occur with little or no advance warnings. Most SCC mechanisms require the presence of tensile stresses in the structure as well as a corrosive medium. Other types of corrosion crackings can develop without tensile stresses. As a

general rule, the presence of tensile stresses may accelerate

the crack growth rate.

SCC is a major aging-related degradation mechanism for

reactor internals. Crevices and Irradiation effects can assist

the SCC process.

4.1.2.1 SCC

It is generally accepted that the simultaneous presence of

three conditions are necessary for the development of SCC:

a susceptible material, a corrosive environment, and tensile

stresses. Elimination of any one of the three conditions will

stop the SCC process. Depending on the alloy compositions and corrodents involved, cracks can develop along

boundaries between grains; such failures are known as

intergranular stress corrosion crackings (IGSCC). In other

cases, cracks propagate along certain crystallographic slip

planes within the grains. These failures are referred to as

transgranular stress corrosion crackings (TGSCC). More

information on the fundamentals of SCCs can be found in

texts by Logan2 and Romanov. 3

Austenitic stainless steels, such as type 304, have good

resistance to general corrosion and wastage, but they can

be made susceptible to SCCs by a sensitization process. 4

When a component made of type 304 stainless steel is

heated or cooled slowly through the temperature range of

482 to 8160C (900 to 15000F), carbon in the steel will precipitate out as chromium carbide along grain boundaries.

Chromium, which is a key alloying element for providing

29 NUREGICR-6048
Aging-Related

corrosion resistance, is depleted in regions adjacent to

grain boundaries, and the chromium-depleted regions are

susceptible to corrosion attacks. Slow cooling after annealings, prolonged stress-relieving operations, and welding

can provide the temperature condition that is needed to

sensitize stainless steels. For reactor internals made of type

304 stainless steel, welding is a likely cause of the sensitization process, and weld heat-affected zones (HAZ) are

common locations for SCCs.

It should be emphasized that a susceptible material by

itself, such as a sensitized stainless steel, cannot cause

SCC. Other conditions must also be present in the system

before SCC can occur.

The second requirement is a corrosive environment. Under

normal PWR operating conditions, the primary reactor

cooling water contains small quantities of dissolved oxygen, chlorides, fluorides, and other impurities. The impurities may contribute to the development of SCC when they

are trapped in crevices in internal components.

Dissolved oxygen is a product of radiolytic reactions in the

core. In steady-state PWR operations, the NRC Standard

Technical Specification stipulates that the dissolved oxygen content in the reactor primary cooling water be kept at

<100 ppb. The plant primary cooling water chemistry

requirements, as stated in the Final Safety Analysis Report

(FSAR), meet and exceed the Standard Technical Specification requirements for dissolved oxygen. The reactor

chemical and volume control system, using hydrogen gas

as a scavenger for the dissolved oxygen, can maintain a

dissolved oxygen content at <5 ppb in the bulk of the primary cooling water system during steady-state power generation. At this concentration level, dissolved oxygen is not

a factor in the development of SCC. However, pockets of

high concentrations of dissolved oxygen may exist in

crevices and can contribute to the development of SCC. In

the presence of impurities such as chlorides, SCC may be

initiated when the dissolved oxygen content is >40 ppb.5

SCC in an oxygenated water environment is intergranular.

The Standard Technical Specification for PWR primary

water chemistry also requires that the chloride and fluoride

contents be kept at <150 ppb. Chlorides are of special concern. Under favorable conditions, the presence of chlorides

and dissolved oxygen in the reactor primary cooling water

may promote the development of SCC in austenitic stainless steel components. 6 In general, the quantity of chlorides required for the development of SCC decreases with

increasing dissolved oxygen content. Experimental results

with water in the range of 204 to 3160C (400 to 6000F)

indicated that in the absence of dissolved oxygen, SCC will

NUREG/CR-6048

not occur even when the chloride content attains a level as

high as 20,000 ppm. During steady-state PWR operations,

the dissolved oxygen content in the bulk of the reactor

cooling water is <5 ppb, and it is not likely that SCC can

develop under such a low level of dissolved oxygen.

Crevices, which can trap and create high local impurity

concentrations, are needed to initiate SCC in PWR

internals.

Sulfide is another conrodent that can cause SCC. It is not

present in the reactor cooling water. However, molybdenum disulfide (MoS2 ), often used as a thread lubricant,

can create locally high sulfide concentrations in reactor

internals with crevice conditions such as those that existed

in bolted joints.7 MoS2 reacts with borated water, and one

of the reaction products is hydrogen sulphide (H2S), which

is highly corrosive. H2S can cause SCC in components

made of stainless steel. Reactor internals that use MoS2 as

a lubricant are susceptible to sulfide-induced SCC.

Small quantities of fluoride are present in the reactor primary cooling water, and they may induce SCC in stainless

steel components. However, information on fluorideinduced SCC in PWR conditions is very limited.

The development of SCC in reactor internals is only one of

many concerns that need to be taken into consideration in

establishing an optimum primary cooling water chemistry

program for PWRs. Cracking problems in steam generator

tubings and out-of-core radiation control are equally

important factors. More detailed discussions concerning

PWR primary water chemistry control can be found in

Refs. 8 and 9.

The third requirement for the development of SCC is the

presence of tensile stresses in a structural component. In

the absence of significant irradiation effects, the magnitude

of the tensile stresses must exceed a threshold value before

SCC can occur. A generally accepted value for the threshold stress is the yield stress of the material of construction.

Bolt preloads and weld residual stresses can generate the

necessary tensile stresses in a reactor internal component.

Based on this information, it is reasonable to suggest that

crevices in PWR internals are likely locations for the

development of SCC. A "crevice condition" is a general

term that is used to describe a small region in which high

concentrations of corrodents may be trapped. Crevices are

created by small holes, surface deposits, narrow gaps in

gasket surfaces, lap, and bolted joints. The small region is

usually filled with a stagnant liquid, and a differential aeration cell is established within the stagnation region. The

30
differential aeration cell can create locally high concentrations of anionic species such as chlorides in the crevice.

The presence of bolt tensile stresses and dissolved oxygen

would supply the necessary conditions for the development

of SCC. Most crevice-assisted SCC is intergranular.

In addition to the crevice condition, exposures to fast neutron fluxes can also assist the SCC process. SCC has been

observed in reactor components made of nonsensitized

stainless steels. In most cases, stresses in these components

are not high, and they are much lower than the yield stress

of the material. A common factor in these nonsensitized

stainless steel components is the exposure to high energy

or fast neutron fluxes. These observations seem to suggest

a new mechanism for the development of SCC. This form

of SCC is known as irradiation-assisted SCC (IASCC);

IASCC is intergranular.

Basic understanding of IASCC is not complete. Many

theories have been proposed, but no one single theory can

satisfactorily account for the iradiation effects on SCC.

General agreement is that a threshold neutron fluence level

exists below which IASCC is not likely to occur. The best

estimate of the threshold neutron fluence levels for stainless steel is -5 x 102° neutrons/cm 2 (E > 1 MeV). The

expected lifetime neutron fluence levels for internals

located in close proximity to the core, such as the core baffle, in-core monitor housings, and the core support plate,

can exceed the threshold value, and such components are

susceptible to IASCC.

One of the proposed theories suggested that the material

may be weakened by the formation of bubbles in the solid

caused by a transmutation process. Austenitic stainless

steels contain trace quantities of boron and can react with

thermal neutrons to form lithium and helium.9 Hydrogen

gases are also produced by transmutation reactions involving fast neutrons and elements such as nitrogen, nickel,

iron, and chromium.8 These insoluble gases precipate from

the solid and have a tendency to migrate to dislocations or

to grain boundaries.9 Gas bubbles are discontinuities, and

they can weaken the structural material. The solid may disintegrate if the bubbles are fused together, even when no

significant stresses are acting on the system. Basic understandings on bubble sizes, their weakening effects, and the

driving forces behind bubble movements are not complete.

The presence of impurities such as phosphorus (P) and silicone (Si) increases the susceptibility of unirradiated stainless steels to IGSCC. This seems to suggest that grain

boundary segregation of impurities could be the responsible mechanism for IASCC. Therefore, high-purity stainless

steels, with lower impurity contents, may be less susceptiAging-Related

ble to IASCC. However, there is not sufficient information

to correlate grain boundary impurity concentrations with

neutron fluence level and IASCC susceptibility. Testing

results are inconclusive; some high-purity stainless steels

are more resistant to IASCC under PWR operating conditions while others (including type 304) performed no better

than commercial-grade stainless steels.

There is no agreement on the primary mechanism for

IASCC, and there are many active research works in the

field. It is generally accepted that IASCC would require the

excedence of a threshold neutron fluence level. Also, the

threshold tensile stress value for IASCC is much lower

than the yield stress of the material. Stresses incurred during manufacturing and handling operations may be sufficient for the development of IASCC. It is also accepted

that the critical flaw size decreases with increasing neutron

fluence level. The reduction of the critical flaw size can

accelerate crack growth rates.

The operating environment inside the pressure vessel of a

PWR can produce conditions that are favorable to the

development of SCC. SCC, in one form or another, is

expected to be a major aging-related degradation mechanism for reactor internals.

4.12.2 Pitting

Pitting corrosion is a localized phenomenon. Holes or pits

are etched into the metal surfaces and are filled with corrosion products. The pits are created by crevices or other fabrication flaws in the form of small cuts and nicks. They act

as differential aeration cells in which the pit is an anode

and the surrounding surface is a cathode.I The anodic reaction will create locally high concentrations of anionic

species such as chlorides and sulfides in the pits. The corrosion process is the result of high concentrations of

aggressive corrodents. Pitting can occur without the presence of significant stresses in the cornponent. A stagnant

condition will promote the growth of pitting corrosion.

Pitting is a concern for reactor internals when they are in

storage during extended plant outages. Unless the reactor

has a history of long outages, pitting is not considered as a

primary aging-related degradation mechanism for reactor

internals.

4.2 Fatigue

In addition to static loads, reactor internals are subjected to

dynamic or time-dependent forces. A structure will

undergo some form of vibratory motions as a response to

dynamic loads. Vibrations can lead to crack initiation and

31 NUREGICR-6048
Aging-Related

subsequent crack growth. Structural failures caused by

vibrations are classified as fatigue failures. Fatigue failures

are further divided into two types: low- and high-cycle

fatigues. The determining factor is the number of vibration

cycles that a component experienced before the development of a crack. The counting process begins with the

onset of vibrations and is terminated with the initiation of a

fatigue crack in the structure. There is not a precise dividing line separating low- and high-cycle fatigue failures.

Generally speaking, when a crack is initiated between 103

and 104 cycles, the failure is considered as a low-cycle

fatigue failure. Plastic strain accumulations are associated

with low-cycle fatigue failures. High-cycle fatigue failures

are characterized by large numbers of cycles and elastic

stresses. There are usually little or no plastic strain accumulations in a high-cycle fatigue failure.

Fatigue failures are environment dependent Understanding

of the interactions between fatigue and environmental factors such as corrosion is limited. Most fatigue analyses are

performed without inputs from environmental conditions.

Fatigue design curves, such as those provided by the ASME

B&PV Code,10 are developed for dry air where the environmental corrosion effects are insignificant.

Some dynamic loads, such as those caused by hydraulic

transients and seismic excitations, are treated as parts of

the design loadings for the reactor system. Stresses, strains,

and the dominant frequencies associated with the structural

response to these events can be calculated, and their contributions are added to the fatigue life estimation for the

affected components.

The primary cause of fatigue failures in reactor internals is

flow-induced vibrations. The amplitude of the vibrations is

determined by the amount of damping in the system as

well as the closeness between he structural natural frequency and the dominant excitation frequency. When these

frequencies are well separated, the amplitude of the vibrations is small. Large-amplitude resonant vibrations may

develop when the input excitation frequency is close to the

fundamental natural frequency of the structure. Damping

can reduce the amplitude of such vibrations. The development of large-amplitude resonant vibrations in a reactor

component will quickly lead to a low-cycle fatigue failure.

When they are detected during reactor preoperational testings, corrective actions will be implemented to eliminate

such large-amplitude vibrations. Therefore, in an aging

assessment study, high-cycle fatigue failures caused by

small-amplitude vibrations is the more important and relevant aging-related degradation mechanism.

4.3 Erosion

Erosion is the removal of materials from a metal surface

when it is submerged in a moving fluid medium.I The process is best illustrated by the sandblasting operation in

which a gas stream laden with fine solid particles literally

grinds away a metal surface. Reactor cooling water is filtered, and the quantity of suspended solid particles in the

flow stream is kept at a very low level (<1 ppm). Erosion

caused by the actions of suspended solid particles is not

expected to be a major aging-related degradation mechanism for PWR internals.

Cavitation can also cause erosion on a metal surface submerged in a liquid flow stream. In regions where the flow

speed is high, the local static pressure may fall below the

vapor pressure of the liquid, and a phase change takes

place. Bubbles are formed and can become attached to the

metal surface causing erosion problems. There is no phase

change in the primary reactor coolant in normal PWR

operations, and the flow speed in the core region is not sufficient to produce cavitations. Cavitation-induced erosion

is not expected to be an aging issue for PWR internals.

4.4 Mechanical Wear

Mechanical wear is the loss of material as a result of relative motions between two contact surfaces. The presence of

a corrosive medium may accelerate the material loss rate.

Most mechanical wear problems in reactor internals are

vibration-induced.

Mechanical wear can also develop in components that are

held fixed initially. Examples are wears observed in

flanges and bolted joints. Preloads are applied to these

parts to prevent slippage between adjacent contact surfaces. Vibrations and creep can reduce the forces holding

these parts together, and slippage can occur. The resulting

relative motions can cause mechanical wear.

Mechanical wear can be minimized by hardening the contact surfaces or by the use of special wear pads. Periodic

tightening of bolts in bolted joints can prevent mechanical

wear caused by slippage.

4.5 Embrittlement

Embrittlement is the loss of ductility and fracture toughness in a material. It is usually accompanied by increases

NUREG/CR-6048 32
in the material yield and ultimate strengths. The process

can transform a normally ductile material to one that is

susceptible to brittle fractures. Prolonged exposures to fast

neutron fluxes and thermal aging can cause embrittlement

in reactor components made of stainless steels.

4.5.1 Radiation Embrittlement

Effects of radiation embrittlement are controlled by several

parameters; the more important ones are neutron fluence,

irradiation temperature, and material compositions. The

experimental data base is obtained from testings using

surveillance specimens from commercial power reactors

and other test reactor experiments. Reductions in uniform

elongations are used as indications of the decrease in ductility in the material after exposure to fast neutron fluxes.

Changes in fracture toughness can be deduced from results

of Charpy V-notch impact testings as well as other types of

fracture-toughness testings.

At a temperature of -316°C (6000F), testing results1 1 with

type 304 stainless steel indicate that the effects of radiation

embrittlement are becoming noticeable at a neutron fluence

level of -S x 1020 neutrons/cm2. The estimated neutron

fluence for reactor internals is determined by the distance

of the component from the core. In 40 years of power

operations, for those components located in close proximity to the core, the expected maximum neutron fluencet2 is

-I x 1022 neutrons/cm 2. Such reactor internal components

are susceptible to the effects of radiation embrittlement.

At specified neutron fluence levels and using annealed type

304 stainless steel, experimental resultst3 showed that the

material hardens and there is a corresponding decrease in

ductility, as indicated by a reduction in uniform elongation.

The experiments were conducted at temperatures ranging

frm room temperature to -7600C (1400'F') and at neutron

fluence levels (E > 1 MeV) up to 6 x 1021 neutrons/cm2.

At 299°C (5700F) the unirradiated uniform elongation has

a value of -38%. The uniform elongation at a neutron fluence level of 1.5 x 1020 neutrons/cm 2 is -22% at the same

temperature. At -1.5 x 1021 neutrons/cm2, the uniform

elongation dropped to -0.5%. The decrease seemed to

level off at higher neutron fluence levels. Assuming these

results are representative values for 304 stainless steels, it

can be speculated that when the expected neutron fluence

level does not exceed 1.5 x 1020 neutrons/cm 2, reactor

internals should retain sufficient ductility that brittle fracture can be prevented. When the neutron fluence level for

an internal component approaches the maximum value of

I x 1022 neutrons/cm2, the decrease in ductility would

make a component susceptible to brittle fractures (reasonable combination of critical fracture toughness, stress, and

flaw size). The core baffle, in-core instrumentation guide

Aging-Related

tubes, and possibly the core support plates are internal

components that can experience neutron fluence levels

close to the maximum value,12 and they are vulnerable to

radiation embrittlement effects.

4.5.2 Thermal Embrittlement

Thermal embrittlement is an aging-related degradation

mechanism for reactor components made of cast austenitic

stainless steels14 (CASS). CASS is a two-phase alloy consisting of austenite and ferrite. The austenite is ductile, and

its ductility is not affected by thermal exposure. The ferrite

can become embrittled by thermal exposures, and its content is the controlling factor for thermal aging effects in

CASS. When the ferrite content is <20%, the thermal

embrittlement of CASS components is low.

When the ferrite content is >25%, CASS components can

become embrittled when they are aged in the PWR operating temperature range of 280 to 3200C (535 to 6100F).

Thermal-aging effects under these conditions can lead to a

brittle cleavage fracture of the ferrite. Failures can also

take the form of a separation of the ferrite/austenite phase

boundary, which provides a convient crack propagation

path. In either case, the fracture toughness of the material

is reduced, and CASS components are susceptible to brittle

fractures.

4.6 Creep and Stress Relaxation

Creep and stress relaxations can cause a load-bearing

structure to lose its structural integrity when it is exposed

to high temperatures for an extended period of time. Exposures to fast neutron fluxes can lower the temperature in

which creep and stress relaxation can become significant

deformation mechanisms.

Creep is the progressive deformation of a structure under a

constant internal stress. Stress relaxation is the reduction of

internal stresses in a structure with a constant deformation.

Creep and stress relaxation are caused by the same mechanism, but they differ in the constraints imposed on the

structure. Creep can lead to brittle fractures in structural

components that can undergo deformations. Stress relaxation is a concern for constrained structures such as bolted

joints, where deformations are held fixed. The decrease in

stresses in a bolted joint can cause the joint to lose its initial tightness, and leakage and slippage may occur. The

development of creep and stress relaxation is influenced by

the operating temperature and neutron fluence level.

33 33 ~~~~~NUREG/CR
Aging-Related

Creep and stress relaxation are considered as significant

structural deformation mechanisms when the operating

temperature is higher than half of the melting-point temperature of the material. The melting points for austenitic

stainless steels are -14270 C (26000F). The normal operating temperature range for PWR internals is between 280

and 3200C (535 and 6100F), and it is much below 714 0C

(13000F), which is half the melting-point temperature for

stainless steels. The creep analysis temperature limit for

austenitic stainless steels, as specified in the ASME B&PV

Code, is 4270C (8000F). Creep analyses are not required

for reactor components operating at the normal PWR operating temperature. When the neutron fluence level is low,

thermally induced creep and stress relaxation are not considered as major aging-related degradation mechanisms for

PWR internals.

In-pile experimental results at -288 0C (550F) indicated

that stress relaxation can occur in bolts made of type 304

stainless steelsl at a neutron fluence level (E > 1 MeV) of

-6 x 1019 neutrons/cm2. At the temperature range from 60

to 3160C (140 to 6000F), significant stress relaxation has

been observed in type 304 stainless steels at neutron fluence levels >5 x 1020 neutrons/cm2. The expected lifetime

neutron fluence levels for reactor internals located near the

core can exceed this value. The core baffle, in-core instrumentation guide tubes, and core support plates are susceptible to the aging effects of irradiation-assisted creep and

stress relaxation.

4.7 PWR Internals and Potential

Aging-Degradation Mechanisms

A reactor internal component is exposed to many stressors,

and it is subjected to the effects of more than one agingrelated degradation mechanism. Aging-related degradation

mechanisms develop at different rates. In most cases, a

dominant degradation mechanism emerges and would

eventually cause a failure in the affected component.

Actual operating conditions may favor the development of

certain aging-related degradation mechanisms. As an

example, it is well understood that impurity contents and

tensile stresses have a strong influence in the development

of SCC in stainless steel components. In PWR operating

conditions, crevices are needed to trap high-impurity

concentrations. Preloads in bolts can produce internal

tensile stresses that are close to the yield stress of the bolt

material. The combination of crevice conditions and high

bolt tensile stresses indicates that PWR internals with

bolted joints and other tight-fit connections are probable

locations for the development of SCC.

The location of an internal component can have some

influence in the development of a dominant aging-related

degradation mechanism. A major stressor for reactor internals is flow-generated excitations, which may take the

form of pump-generated pressure pulsations, vortex sheddings, and high-intensity turbulent flows. Pump-generated

pressure pulsations are at their strongest level in regions

around inlet nozzles. Reactor internals located in these

regions, such as core support shields, thermal shields, surveillance specimen holder tubes, and core barrels, are susceptible to flow-induced pressure pulse excitations. As the

coolant flow travels deeper into the core region, fluid

damping and friction losses reduce the intensity of the

pressure pulsations, and they would become less of a factor

in exciting internal components.

Reactor internals located close to tube-bank-like structures,

such as in-core instrumentation guide tubes and shrouds,

are susceptible to the effects of cross-flow generated vortex

sheddings. Flow-induced vibration is a major stressor for

these components.

The effects of fast neutron fluxes are most pronounced in

regions around the core. Reactor internals such as core baffle, core barrel, thermal shield, surveillance specimen

holder tubes, core support plates, and in-core instrumentation guide tubes are susceptible to irradiation-assisted SCC

and radiation-induced embrittlement.

Reported aging-related failure information for PWR internals is needed to assess the relative importance of the different aging-related degradation mechanisms associated

with the many stressors that may exist inside a reactor

pressure vessel.

Primary stressors and associated aging-related degradation

mechanisms for PWR internals are summarized in

Table 4.1.

NUREG/CR-6048 334
Aging-Related

Table 4.1 PWR Internals primary stressors and aging-related degradation rnechanisms

Stressor Source Aging-related degradation mechanisms

Oscillatory hydrodynamic forces Reactor coolant flow Fatigue and mechanical wear

(dynamic stresses)

Preloads in bolts (static stresses) External applied loads Contributing factor to SCCa

(torquing during installation)

Residual stresses (static stresses) Welding Contributing factor to SCC

Thermal stresses (static and dynamic) Reactor coolant and gamma heating Fatigue, contributing factor to SCC

Dissolved oxygen and other impurities Reactor coolant Contributing factor to SCC

Local concentration of corrodents Crevice condition in reactor coolant Contributing factor to SCC

flow

Chromium depletion in grain Welding Enhance potential for SCC

boundaries (sensitization)

High operating temperature Reactor coolant Thermal embrittlement in CASS

[-3160C (600*F)J components

Exposure to fast neutron fluxes Nuclear reaction in the core Irradiation-assisted SCC and

embrittlement, irradiation-enhanced

creep and stress relaxationb

4

SCC is most likely t occur as a result of high tensile stresses, local concentration of orrodents (crevices), and a susceptible material.

bradition effects can make creep and ses relaxation major degradation mechanisms at reactor eperating temperature 3160C (6000F)].

References Type 304 Stainless Steel:' Corrosion 35, 288

(197 9).t

1. "Component Life Estimation: LWR Structural Materials Degradation Mechanisms:' EPRI NP-5461, an

interim report prepared by Structural Integrity Associates, Inc., for the Electric Power Research Institute,

Palo Alto, Calif, September 1987.*

2. H. L Logan, The Stress Corrosion of Metals, John

Wlley & Sons, New York, 1966.

3. V. V. Romanov, "Stress Corrosion Cracing of

Metals," the Israel Program for Scientific Translations and the National Science Foundation,

Washington, D.C., 1961.

4. R. L. Cowan and C. S. Tedmon, "Intergranular Corrosion of Iron-Nickel-Chromium Alloys," Advances

in Corrosion Science and Technology, Vol. 3, M. S.

Fontana and R. W. Stachle, Eds., Plenum Press, New

York, 1973.

5. M. E. Indig and A. R. Mcllree, "High Temperature

Electrochemical Studies of the Sress Corrosion of

6. B. M. Gordon, "The Effects of Chloride and Oxygen

on the Stress Corrosion Cracking of Stainless Steels:

Review of Literature," NACE 36(8) (1980).1

7. "Degradation and Failure of Bolting in Nuclear

Power Plants" Vol. 1, EPRI NP-5769, a final report

prepared by Applied Science & Technology for the

Electric Power Research Institute, Palo Alto, Calif,

1988.

8. A. J. Jacobs and 0. P. Wozadlo, "Irradiation-Assisted

Stress Corrosion Cradcng as a Factor in Nuclear

Power Plant Aging," . Materials Engineerig 9(4)

(1988).1

9. 1. Gittus, Irradiation Effects In Crystalline Solids,

Applied Science Publishers, Ltd., London, 1978.

10. American Society of Mechanical Engineers, ASME

Boier and Pressure Vessel Code, Section I, Nuclear

Power Plant Components, Div. 1. 

35 NUREGCR-6D48
Aging-Related

11. R. E. Robbins, J. J. Holmes, and J. E. Irvin, "Post

Irradiation Tensile Properties of Annealed and ColdWorked AISI-304 Stainless Steel," Trans. American

Nuclear Society, November 1967.t

12. V. S. Shah and P. E. MacDonald, Eds., "Residual

Life Assessment of Major Light Water Reactor

Components-Overview:' USNRC Report NUREG/

CR-4731, November 1989.*

13. 0. K. Chopra and H. M. Chung, "Long-Term Aging of

Cast Stainless Steel: Mechanisms and Resulting Properties," in Proc. of the Fifteenth Water Reactor Safety

Research Information Meeting, Gaithersburg, MiL,

USNRC Conference Proceeding NUREG/CR-0090,

October 1987.*

14. 0. K. Chopra, Argonne National Laboratory, "Estimation of Fracture Toughness of Cast Stainless Steels

During Thermal Aging in LWR Systems," USNRC

Report NUREGICR-4513, June 199.t

Available fr purchase from National Technical Infornation Service,

Springfield, VA 22161.

tAvailable in public technical libraries.

NUREG/CR-6048 36
S Survey of Aging-Related Failures

Reactor Internals are subjected to stressors that can activate

many potential aging-related degradation mechanisms.

Aging-related component failure information is needed to

properly identify the more Important aging-related degradation mechanisms. The identification of major agingrelated degradation mechanisms may provide information

that can be used to formulate strategies for controlling and

managing aging effects.

The reactor component failure information is compiled

from results of the plant in-service inspection (ISI) programs. The program specifies requirements and procedures

for inspecting reactor systems or components to ensure

safe plant operations. The requirements and procedures

may include orders, rules, criteria, and guidelines established by regulatory or licensing agencies as well as industry standards developed by technical organizations. ISI

programs for U.S. commercial nuclear power plants are

established under the rules and regulations of Sect. XI of

the ASME Boiler and Pressure Vessel (B&PV) Code.I The

inspection of some reactor internals is included as a part of

the plant ISI program.

5.1 ISI Program for Reactor Internals

The plant ISI program requires visual inspections for accessible areas of reactor internals. A complete inspection

cycle is 10 years, and selected internal components are inspected at refueling outages.

Section XI of the ASME B&PV Code specifies thre

classes of visual inspections: VT-I, VT-2, and VT-3. A

VT-i visual examination is conducted to determine the

condition of the part, component, or surface examined, including such conditions as cracks, wear, corrosion, erosion,

or physical damage on the surface of the part or component. The examination can be performed either directly or

remotely. A VT-2 examination is used to detect leakage

from pressure-retaining components. Most reactor internals

are not pressure-retaining components, and VT-2 inspections are seldom used on internals. A VT-3 examination is

conducted to determine the general mechanical and structural conditions of components and their supports, such as

verification of clearances, settings, physical displacements,

loose or missing parts, debris, corrosion, wear, erosion, or

the loss of integrity at bolted or welded connections. VT-3

inspections can be performed either directly or remotely.

During a refueling outage, some internal components are

removed from the pressure vessel and stored in a pool.

VT-3 examinations are performed on the accessible areas

of these components by remote television cameras under

proper lighting. Internals that remain in the pressure vessel

are also inspected by renote television cameras. Note that

the ISI plan only calls for the inspection of accessible areas

of the internals. Details of the inspecting procedures can be

found in SecL XI of the ASME B&PV Code.

When a failure is detected, information concerning the

failure such as its location and suspected causes is

recorded. The information is sent to NRC for inclusion in

the Licensee Event Report (LER). The LER may also contain information on corrective actions taken to repair the

failure or damages. For these reasons, LERs are considered

as reliable sources for reactor systems or components failure information.

5.2 Failure Information Summary

The majority of PWRs in the United States started commercial operations in the 1970s, and since that time there

were many reported aging-related failures of reactor internals. Most of the reported failures can be attributed to three

aging-related degradation mechanisms: (I) fatigue,

(2) SCC, and (3) mechanical wear.

Most of the reported aging-related failure cases involved

domestic PWRs. Many overseas PWRs of the

Westinghouse design have experienced aging-related failures similar to those of the domestic units. However, the

only overseas case discussed in details in this report is the

SCC failures in core baffle bolts detected in some German

and Swiss PWRs. This failure case is included because of

its safety significance.

The following is a brief summary of the reported failure

cases for each of the three major aging-related degradation

mechanisms.

5.2.1 Fatigue

Most fatigue failures in PWR internals are caused by flowinduced vibrations (FIVs). Important excitation sources

include pump-generated pressure pulsations, cross-flow

vortex sheddings, and highly turbulent flows. There is usually a dominant frequency associated with these disturbances. Large-amplitude vibrations can develop when the

structural natural frequency matches one of these excitation

frequencies. The component will be subjected to smallamplitude vibrations when the frequencies are well separated.

37 NUREG/CR-6048
Survey

5.2.1.1 WE Thermal Shield

The failure of thermal shields in older WE PWRs have

been attributed to FIV.2 Thermal shields are located near

inlet nozzles and are subjected to pump-generated pressure

pulsations. Most of these failed thermal shields were of the

segmented-shell design with shell segments (usually three)

keyed to support lugs at the bottom of the reactor vessel.

The top end was either free or was equipped with radial

spacer pins. The support condition can be characterized as

a cantilevered support with limited displacements at the

free end. The shell segments were fastened together by

vertical pins at the intersections. Coolant flow-induced excitations caused the assembled shield to vibrate in a shell

mode. During vibrations, some of the shell segments came

into contact with the core barrel. In these older reactors, the

core barrels were bolted together, and the repeated impact

loadings caused failures in the core barrel support bolts.

The impact loadings also damaged the thermal shield.

One of the reported failure cases involved a one-piece

cylindrical thermal shield and occurred during the reactor

functional hot testing. The shield was clamped to the core

barrel at the bottom. The top was free except for the presence of radial limiter pins, which fit into a keyway in the

core barrel. The pins were shrunk-fit into the thermal

shield and lock welded in place by light fillet welds. FIV

caused the pins to come into contact with the sides of the

keyway in the core barrel, and repeated impacts eventually

led to the cracking of the fillet welds that locked the pins to

the thermal shield. After the failure was detected, the decision was made to replace the pin-keyway system by a

flexure support system.

There was also one reported case of failure involving a

one-piece cylindrical thermal shield with a flexure support

system. The top-mounted flexure support system failed.

The cause was attributed to high-cycle fatigue caused by

small-amplitude FIV of the thermal shield.

There are no reported failures in thermal shields using the

neutron-shield pad design.

5.2.1.2 CE Thermal Shield

CE thermal shields are located near inlet nozzles where effects of pump-generated pressure pulsations are strong.

Two CE units equipped with thermal shields reported

problems with their support system.2 ISIs during refueling

outages revealed missing support and positioning pins. All

remaining pins showed signs of excessive wear or damage.

Lugs welded to the core barrel to support the top end of the

thermal shield were also damaged. In one unit the damaged

NUREG/CR-6048

lugs caused a through-the-wall crack in the core barreL No

damage to the reactor vessel was observed.

The damaged thermal shields were removed. Analysis indicated that removal of the thermal shield did not lead to

any significant changes in the core thermal-hydraulic operating conditions. The removal of the thermal shields is not

expected to have any undesirable effect on the reactors during their remaining design life.

5.2.1.3 CE Hold-Down Ring

During inspection of a maintenance outage, excessive mechanical wear was observed in the hold-down ring of a CE

unit.2 Failure was attributed to FIV caused by insufficient

hold-down spring force. The hold-down ring was replaced.

The new ring was fabricated with type 403 stainless steel

instead of type 304, which was used to make the old holddown ring. Additional hold-down spring force was applied

during the installation of the new hold-down ring. The internal hold-down spring force was also increased for other

CE reactors. There were no other reported hold-down ring

problems.

5.2.1.4 WE BaMe Plate Water-Jetting

Fuel rod damage caused by baffle plate water-jetting has

been reported in a number of WE reactors. 2'

3 The core

baffle outlines the boundary of the core. A bypass flow is

established in the region between the core baffle and the

core barrel, and it is used to provide more effective cooling

to the core barrel. In some of the older WE reactors, the

bypass flow is introduced into the region between the core

baffle and the core barrel by holes located in the upper core

barrel. The bypass flow moves in a downward direction

through holes in the horizontal former plates, and it is

turned around at the bottom of the core barrel and then

merged with the main flow going through the core. When

the bypass flow is in a downward direction, as illustrated in

Fig. 5.1, the bypass flow pressure is higher than that of the

main coolant flow in the core. A pressure differential is established between the bypass flow and the core, and it will

push the coolant into the core if gaps exist between the

vertical baffle plates. The jetlike leakage flow will impinge

on fuel rods in the vicinity of the gaps and set the rods into

whirling motions and vibrations. Excessive fuel rod motions will eventually lead to cladding degradations and

failures.

Two types of baffle plate water-jet impingement patterns

have been observed and are illustrated in Fig. 5.2. The jet

from a center-injection joint impinges directly on a fuel

rod, while that from a corner-injection joint will have a

more sideways, impact. They can both set the affected fuel

rods into vibrations and whirling motions.

38
Survey

ORNL-OWG 93-2896 ETM

CORE PLATE

CORE PLATE

Figure 5.1 Westinghouse PWR downward bypass flow scheme

Several remedies have been tried to reduce the severity of

the baffle plate water-jetting impingement problem. One

remedy that has been tried is the peening of an entire joint

to reduce the gap width between the vertical plates.

Subsequent inspections revealed that peening the joints

was not effective because damaged fuel rods were found

around the peened Joint. There was also evidence to suggest that peening a center-injection joint will have the effect of enlarging the gap width in nearby coner-injection

joints. A more effective remedy is to replace the fuel rods

in the water-jet impingement region with solid stainless

steel rods of the same diameter. The insertion of partial

grids, which will serve as midspan supports for fuel rods in

the impingement region, has also been effective in reducing fuel rod damage. However, uses of solid stainless steel

rods and partial grids are not considered as the permanent

solution to the baffle plate water-jeting problems. A more

effective solution Is to reduce or eliminate the driving force

behind the water-jetting flow.

The bypass flow rate is small when compared with the

main coolant flow rate through the core, and the coolant

pressure distribution in the core can be regarded as constant during normal reactor operations. The variable that

can be changed to affect the pressure differential between

the bypass flow region and the core is the pressure of the

bypass flow. A decrease in the bypass flow pressure reduces the pressure differential and the driving force behind

the water-jetting flow. The bypass flow pressure can be

lowered by increasing the pressure loss in the main coolant

flow before the diversion of a small portion of the main

flow into the bypass flow region. This is accomplished by

plugging the inlet holes near the top of the lower core barrel. The pressure in the bypass flow is reduced because of

the added pressure loss in the flow down the annular

downcomer region between the core barrel and the vessel

wall. The bypass flow is then diverted into the core bafflecore barrel region through the lower core plate. In the

modified flow scheme, the bypass flow is in an upward

direction. Ideally, pressures in the core and the bypass flow

region should be balanced, and there will be no driving

force to produce jetting flow through gaps. In practice, a

balanced pressure system is difficult to attain, and the

severity of the water-jetting problem is reduced by lowering the pressure differential between the bypass flow and

39 NUREGICR-6048
Survey

ORNL-DM 93-2897 ETM

COOLANT FLOW THROUGH

JOINT ENLARGED GAPI

Pi

P3 0000

00

pi), Pa FUEL RO 

Figure 5.2 Westinghouse PWR baffle water-jet Impingement patterns

the core. The effectiveness of the modified flow scheme

can be further improved by designing the flow system in

such a manner that the core flow pressure is higher than the

bypass flow pressure. In this situation, potential water-jetting flows will be from the core into the bypass flow region

and away from the fuel rods.

One WE unit that has a history of baffle plate water-jetting

problems has modified its downward bypass flow scheme

into an upward bypass flow scheme. Other plants with a

downward bypass flow have also developed plans to convert to an upward bypass flow scheme.

In the new WE reactor design, the bypass flow enters the

core baffle-core barrel region from the bottom region of

the core barrel, and the bypass flow travels upward through

holes in the horizontal fonner plates.

Discussions of the baffle plate water-jet impingement

problem in WE reactors are based on information provided

by NRC E Information Notice No. 82-27.3 There are no

reported baffle plate waler-jetting impingement problems

in CE and B&W PWRs.

5.2.2 SCC

A crevice condition is a common feature in the development of SCC in PWR internals. Bolted joints and tightly

fitted connections are likely sites for SCC attacks. Most

observed SCC is intergranular. Internals with crevice

conditions that are also exposed to high-energy neutron

fluxes may also be susceptible to irradiation-assisted SCC

5.2.2.1 B&W Thermal Shield Support Bolts

The B&W thermal shield is a cylindrical structure with its

upper end bolted to the upper flange of the core barrel. The

lower end of the shield is shrunk-fit into the upper flange

of the forged cylinder in the lower grid assembly and secured by 96 high-strength bolts. B&W thermal shield support bolts were made of a nickel alloy stainless steel (grade

A-286).

ISIs in several units revealed missing bolts from the thermal shield lower end support joint. 2 The majority (-80%)

of the remaining bolts were loose, and several bolt locking

cups were also missing. Cracks were also detected in bolts

at the shield upper end and in the SSHT mounted on the

outside wall of the thermal shield. The more serious failures were bolts at the lower end of the shield.

Failures were attributed to IGSCC at the bolt-head-to-boltshank transition. The replacement bolts were designed to

reduce the tensile stress level in the bolt, and this was accomplished by redesigning the shank region, peening the

surface of the bolt, and reducing the preload used to install

the bolts. The material of construction also was changed

from A-286 stainless steel to Inconel X-750.

NUREG/CR-6048 40
A review of LERs did not show any reported failure of the

new thermal shield support bolts.

5.22.2 B&W Core Barrel-to-Core Support Shield

Bolts

The core barrel and core support shield are flanged cylindrical structures. The upper flange of the core support

shield rests on a circumferential ledge of the vessel closure

flange. The lower flange of the core support shield is

bolted to the top flange of the core barrel. The bolts were

made of a nickel alloy stainless steel (grade A-286).

Ultrasonic (L inspections in two B&W plants showed

indications of cracking in a number of the bolts joining the

core barrel to the core support shield. The results were

verified when the bolt heads became separated from the

bolt shanks when the locking clips were removed. Failures

were attributed to IGSCC. Visual inspections failed to detect these failures.

The cracked bolts were replaced by bolts made of the same

material, grade A-286 stainless steel. The new bolts were

made by machining, while the old bolts were made by a

hot-headed operation. The torque applied to the new bolts

will be significantly reduced. A review of LERs showed no

reported failure of the new core barrel-to-core support

shield bolt.

522.3 WE Control Rod Guide Tube Support Pins

The guide tube assembly is a part of the upper core support

structure, and it houses the control rod drive shaft and the

RCCAs. The assembly is a tubular structure divided into

Survey

two parts: the upper part, called the control rod shroud

tube, and the lower part, the control rod guide tube. The

upper support plate is the dividing boundary. The top of

the control rod guide tube is fastened to the bottom of the

upper support plate, and the lower end of the tube is held in

place by split pins inserted into the lower core plate.

Several WE units had experienced steam generator failures

caused by loose parts in the reactor primary cooling system. The loose parts were identified as parts from failed

split pins in control rod guide tubes. The cause of the failures was identified as IGSCC.

The split pins are made from a nickel alloy (Inconel X-750)

and are bolted to the bottom of the guide tube column. The

support pins are then inserted into the upper core plate. The

pins support the guide tube against hydrodynamic forces

and also align the tubes with respect to the upper core plate

and the fuel assemblies. A sketch of the guide tube support

pin is shown in Fig. 5.3.

Crevice conditions are created when the pins are inserted

into the upper core plate and the pins are exposed to a locally corrosive fluid medium. Improper heat treatment and

overtorquing of the nuts during installation of the pins may

have contributed to the development of IGSCC in the bottotn region of the shank. As a remedy to the IGSCC problems, WE now recommends a solution heat treatment at a

higher temperature, increasing the size of the pins, peening

the nuts, and reducing the preloads during installation. The

objective is to reduce the susceptibility of the material to

the corrosive environment and to reduce tensile stresses

that are necessary to the development of IGSCC.

ORNUO 3G- ETD

m 1 a ,, -U

GUICE TUBE

OT T0M

GurDE TUBE

BOTTOM

PLATE

*. STRAIGHT PIN b. TAPERED PIN

Figure 5.3 Westinghouse PWR guide tube support pins

41 NUREGICR-6048
Survey

The guide tube split pin IGSCC problems at PWR plants

were addressed in a NRC IE Information Notice No.

82-29.4 Many plants have initiated programs to replace the

existing guide tube split pins with pins of the new design.

5.2.2.4 Core Baffle Bolt Failures In KWU-Built PWRs

The Kraftwerk Union (KWU) AG has built many PWR

plants of the Westinghouse design in Germany and

Switzerland. Five of these plants have reported SCC problems in the core baffle bolts.

The core baffle maintains the geometry of the fuel assemblies and controls the coolant flow through the core. It is a

bolted structure composed of horizontal former plates and

vertical baffle plates. The outer edges of the former plates

are bolted to the inside of the core barrel, and vertical baffle plates are bolted to inner edges of the former plates. In

reactors built by KWU before 1979, the core baffle bolts

were made of a nickel-alloy (Inconel X750). After they

were installed, the bolts were secured by tack welds. After

1979, the core baffle is a welded structure in KWU-built

units. A simplified view of the core baffle and its bolting

scheme are shown in Fig. 5A.

ORNL-DWG 92-3331 ETD

IGSCC X

I

i

I

I

I

i

IGSCC

FAILURE

IW

w

I

Figure S5 KCWU-built PWR core afte bolt failures

Routine UT inspections detected signs of cracks in core

baffle bolts in five KWU-built PWRs. A typical bolt failure

pattern is shown in Fig. 5.5. The problem was first reported

in 1987.5 Failures were attributed to IGSCC. The suspected cause was the sensitization of the nickel-alloy bolts

ORNL-DWG 92-3330 ETD

by the welding process, coupled with crevice conditions

and high tensile stresses in the bolts.

The nickel-alloy bolts were replaced by 1.4571 austenitic

steel bolts, and the replacement bolts were secured by a

mechanical scheme. There is no reported failure of the replacement core baffle bolts.

The core baffle in domestic WE PWRs is a bolted structure. The bolts are made of type 316 stainless steel. A

search of the LER data base did not locate any reported

failure of the core baffle bolts in U.S. reactors.

5.2.3 Mechanical Wear

The primary cause of mechanical wear in reactor internals

is FIV. Parts that are in close proximity to each other may

come into contact during vibrations, and the rubbing would

lead to excessive thinning of the contact areas. Bolted

joints, in-core instrument lines protected by tightly fitted

shields, and guide tubes are probable sites for mechanical

wear problems.

5.2.3.1 WE Reactor Flux Thimble and Guide Tube

Thinning

In-core neutron monitors, which are parts of the reactor

flux-mapping system, are located inside retractable thimble

Figure 5.4 KWU-built PWR core baffle and bolting

scheme

NUREG/CR-6048 42
tubes. The thimble tubes extend from selected locations inside the core, through the bottom of the pressure vessel,

high-pressure conduits, and seal tables to a ten-path transfer device. The thimble tubes, in turn, are supported by

guide tubes in the lower region of the reactor pressure vessel. The guide tube is divided Into two parts: the upper

part, located inside the fuel assembly, and the lower part,

extending from lower core plate to the reactor vessel bottom. The portion of the thimble tube between the fuel assembly and the lower core plate Is unprotected and exposed

to reactor coolant flows. The thimble tubes are susceptible

to FIV in the unshielded region. Contacts between thimble

tubes and guide tubes can lead to mechanical wear

problems in both structures. A sketch of the flux thimble

and guide tube arrangement is shown In Fig. 2.7.

Excessive thimble and guide tube thinnings and leakages

were reported in a U.S. reactor in 1981.6 Fretting-induced

failures of guide tubes have been reported in foreign WE

units. The consequence of a thimble tube leakage is serious

because it can result in a breach of the reactor coolant primary pressure boundary. The thimble tubes are opened at

the ten-path transfer device for the insertion of the neutron

monitor, and the development of tube leakages could result

in a nonisolable coolant leak.

Tube thinnings were detected by the eddy-current inspection method. When tube thinning is detected, the current

remedy is to retract the thinned segments to move them

away from the vibrating region. Thicker-walled tubes also

have been used. Tubes that were seriously degraded will be

removed from service by closing the isolation valves.

Many utilities have instituted inspection programs to monitor thimble-tube thinnings. At the present time there Is no

permanent solution to the thimble-tube thinning problems.

Flux thimble and guide-tube thinning problems are discussed In JE Information Notice No. 87-44.6

5.2-3.2 B&W SSHTs

Excessive wear was detected in SSHTs in B&W units in

the mid-1970.1 The tubes were mounted on the outside

surface of the thermal shield and exposed to pump-generated pressure pulsations. The failure was considered as

generic in nature and was attributed to FIV. The tubes were

modified so they would be less sensitive to the pressure

excitations; this was accomplished by changing the stiffness of the tube and/or adding additional structural supports. No problems have been reported since the corrective

measures were implemented.

Survey

5.3 Failure Information Survey Results

More specific information will be provided on major internal component failures. In addition to the major agingrelated degradation mechanisms, the total number of

reactors reporting the failure and their unit age will also be

identified. The unit age of a reactor is measured in

commercial operation years, which is the difference

between the commercial starting date and the event date of

the first LER on the reported failure. Unit age is a

parameter that can be used to represent the age of the

reactor when the first failure occurred. Corrective actions

taken will also be included in the survey results. LERs

from 1980 to 1990 provide the bulk of the failure

information. Useful information was also obtained from

EPRI reports on nuclear unit operating experiences.

5.3.1 WE Internals

53.1.1 Thermal Shield Support Bolts

Major aging-related degradation mechanisms: bolt failures caused by fatigue. Pump-generated pressure pulsations

and oscillatory hydrodynamic forces are the primary

stressors.

Number of units reporting failure: 6, mostly early reactors with the old thermal shield design.

Unit age at firt failure: not known for four units. Of the

remaining two units, one reported thermal shield support

bolt failures after 12 years, and the other unit after 21 years

of commercial operation.

Corrective actions: The pin-keyway support system was

replaced by a flexure shield support system. New reactor

design replaces thermal shields with neutron shield pads

attached directly to the core barreL

53.1.2 Baffle Plate Water-Jetting

Major aging-related degradation mechanisms: Highcycle fatigue. Baffle plate gaps enlarged by FIV. Fuel rods

damaged by vibrations caused by high-intensity turbulent

flows through gaps.

Number of units reporting failure: 6

Unit age at rt failure: 4, 5, 6, 13, and 20 years.

Corrective actions: Several remedies have been tried with

varying degrees of success, Including peening the baffle

plate joints, replacing fuel rods in the impingement areas

with solid stainless steel rods, and using partial grids for

43 NUREGICR-6048
Survey

additional supports to the fuel rods. One unit has converted

the downward bypass flow into an upward bypass flow in

the core baffle-core barrel region. New reactors use the

upward bypass flow scheme.

53.1.3 Control Rod Guide Tube Support Pins

Major aging-related degradation mechanisms: IGSCC

caused by crevice conditions, improper heat treatment, and

overtorquing of the split pin nuts.

Number of units reporting failures: 5

Unit age at first failure: 4. 7, 7, 8, and 14 years.

Corrective actions: Heat treatment at a higher temperature. New split pins are larger in size. The nuts are peened

and installed with a reduced torque.

5.3.1.4 In-Core Thimble and Guide Tube Thinning

Major aging-related degradation mechanisms: mechanical wear caused by contacts with flux thimbles and adjacent guide tubes. Motions induced by FIV of the unshielded portion of the thimble between the fuel assembly

and lower core plate.

Number of units reporting failure: 5

Unit age at first failure: 4, 4, 8, 11, and 17 years

Corrective actions: Tube sections with excessive mechanical wears are retracted and moved away from the vibrating

regions, and thicker-walled tubes are used. There is no

permanent solution at the present time. Utilities have set up

inspection programs to monitor thimble-tube thinnings.

5.3.1.5 Core Baffle Bolts in KWU-Built Units

Major aging-related degradation mechanisms: IGSCC

caused by welding-induced sensitization of nickel-alloy

bolts, coupled with crevice condition and high bolt tensile

stresses.

Number of units reporting failure: 5

Unit age at first failure: Not known.

Corrective actions: Nickel-alloy bolts were replaced by

austenitic steel bolts. Replacement bolts are secured by a

mechanical means instead of tack welds.

Table 5.1 Summary of WE internals

failure infomation

Failure Aging-related Component description degradation

mechanisms

Thermal shield Cracked bolts Fatigue

support bolts

Core baffle plate Enlarged gaps be- Fatigue

tween baffle

plates

Flux thimbles and Excessive thin- Mechanical wear

guide tubes ning

Control rod guide Cracked pins Crevice-assisted

tube support pins SCC

Core baffle bolts Cracked bolts Crevice-assisted

SCC

5.3.2 B&W Internals

5.3.2.1 Thermal Shield Support Bolts

Major aging-related degradation mechanisms: bolting

failure caused by crevice-assisted IGSCC.

Number of plants reporting failure: 7

Unit age at first failure: 6, 7, and 5 units at 8 years

Corrective actions: using replacement bolts that feature

new boll design with reduced stresses in the shank region,

peening the bolts to reduce surface tensile stresses, reducing the torque used in the installation of the bolts, and using Inconel X-750 instead of A-286 stainless steel as the

material of construction.

5.3.2.2 Core Barrel-to-Core Support Shield Bolts

Major aging-related degradation mechanisms: creviceassisted IGSCC.

Number of plants reporting failure: 2

Unit age at first failure: 7 and 8 years

Corrective actions: Replacement bolts are made with the

same material, grade A-286 stainless steel. New bolts were

machined, while old bolts were made by a hot-headed operation. The torque applied to the new bolts is also significantly reduced.

5.3.2.3 SSHT

Table 5.1 is a summary of the reported component failure

information on WE reactor internals.

Major aging-related degradation mechanisms: mechanical wear caused by FIV.

NUREG/CR-6048 44
Number of plants reporting failure: 6

Unit age at first failure: 1 2 2, 2 2 and 3 years

Corrective actions: New SSHTs are structurally detuned

to flow-induced excitations; this is accomplished by

changing the stiffness of the tube and adding new supports.

53.2.4 SHHT Bolts

Major aging-related degradation mechanisms: bolting

failure caused by crevice-assisted IGSCC.

Number of plants reporting failure: I

Unit age at first failure: 6 years

Corrective actions: SSHT bolts were replaced with stud

and nut fasteners made of Inconel X-750.

Table 5.2 is a summary of the reported component failure

information on B&W reactor internals.

5.3.3 CE Internals

S.3.3.1 Thermal Shield Support Bolts

Major aging-related degradation mechanisms: highcycle fatigue bolt failures caused by FIV.

Number of plants reporting failure: 2

Unit age at Srst failure: 7 and 8 years

Corrective actions: Thermal shields and support lugs

were removed from reactors.

Table 5.2 Summary of B&W internals

failure Infomation

Survey

53.3.2 COe Support Barrel

Major aging-related degradation mechanisms: fatigue

caused by FIV. Failure of thermal shield support pins led

to increased loadings on the core barrel support lugs, and a

through-the-wall crack was detected at two lug locations.

Number of plants reporting failure: 1

Unit age at first failure: 8 years

Corrective actions: The through-the-wall crack was arrested by drilling a bole at each end of the crack. The core

support barrel was put back into service.

5.3.33 Core Internal Hold-Down Ring

Major aging-related degradation mechanisms: mechanical wear caused by FIV. FIV was attributed to insufficient

bold-down spring force.

Number of plants reporting failure: 1

Unit age at first failure: 2 years

Corrective actions: New hold-down ring made of type

403 stainless steel was installed with an increased internal

hold-down spring force. No new failures were reported.

Table 5.3 is a summary of the reported component failure

information on CE reactor internals.

Table 5.3 Summary of CE internals

failure infomation

Failure Aging-related Component description degradation

mechanisms

Thermal shield Cracked bolts Fatigue

support bolts

Core support bar- Through-the-wall Fatigue

rel crack

Hold-down ring Excessive wear Mechanical wear

Failure Aging-related Component description degradation

mechanisms

Thermal shield Cracked bolts Crevice-assisted

support bolts SCC

Core barrel to Cracked bolts Crevice-assisted

core support SOC

shield bolts

SSHT Excessive Mechanical wear

thinning

SSHT bolts Cracked bolts Crevice-assisted

SCC

References

1. American Society of Mechanical Engineers, ASME

Boiler and Pressure Vessel Code, Section XI, Rules

for Inservice Inspection of Nuclear Power Plant

Components.

2. "Characterization of the Performance of Major LWR

Components:* EPRI NP-5001, a frnal report prepared

45 NUREG/CR-6048
Survey

by the S. M. Stoller Corp., for the Electric Power

Research Institute, Palo Alto, Calif., 1987.

3. "Fuel Rod Degradation Resulting from Baffle WaterJet Impingements," IE Information Notice No. 82-27,

USNRC Office of Inspection and Enforcement,

Washington, D.C., August 1982.

5. Nucleonics Week 28(10),1-3 (March 5, 1987), a

McGraw-Hill publication.t

6. "Thimble Tube Thinning In Westinghouse Reactors:

IE information Notice No. 87-44, USNRC Office of

Inspection and Enforcement, Washington, D.C.,

September, 1987.

4. "Control Rod Drive (CRD) Guide Tube Support Pin

Failures at Westinghouse Plants," IE Information

Notice No. 82-29, USNRC Office of Inspection and

Enforcement, Washington, D.C., 1982.

Available friom the National Standards Institute, 1430 Broadway, New

York, NY 10018, copyrighted.

tAvailable in public tdhnical liraries.

NUREG/CR-6048 46
6 ISI and Monitoring Programs

The ISI program stipulates visual Inspections for reactor Internals during refueling outages. Visual inspections, while

relatively simple to perform, have obvious limitations.

They are used mainly to detect surface flaws in accessible

areas of a conponent. Visual examinations are not effective in detecting subsurface or partial through-the-wall

cracks and cannot detect failures in Inaccessible areas.

Alternate methods have been tried, on an experimental basis, to improve the effectiveness of detecting failures in reactor internals. UT inspections have been used with some

success in detecting cracks in bolts and in areas that are not

accessible to visual inspections. However, UT Inspections

also have limitations and access problems. In addition, the

interpretations of UT examination results are much more

complicated and difficult than those for visual inspections.

The eddy-current method is another inspection method that

has potential applications in examining reactor internals. It

has been used to inspect long tubular structures for excessive wall thinnings. The establishment of an effective inspection program for reactor components is an evolving

process. When a new inspection method has been used

successfully, the responsible organization will issue bulletins and letters to inform plant operators of the latest developments and their potential applications to reactor component inspection programs.

A major concern in reactor operations is the presence of

loose parts in the reactor primary coolant system. Reactor

operation is disrupted when loose parts are lodged in critical locations such as inlets to pumps and heat exchangers.

Loose parts were generated in some of the reported agingrelated failures. Although none of these incidents had endangered the safety of reactor operation, the possibility

cannot be ruled out In future occurrences. In addition to

safety considerations, these failures have resulted in economic losses caused by extensive outages for repair work.

For these reasons, NRC and plant operators are interested

in the development of inspection and monitoring methods

that can be used to detect failures in reactor components.

Research and development work to detect loose parts1 has

led to the issuance of NRC Regulatory Guide 1.133,2

which outlines the operating requirements for loose part

monitoring systems (LPMS) in U.S. commercial nuclear

power plants. LPMSs are required for reactors licensed

since 1978. Many plants licensed before 1978 have installed LPMS on a voluntary basis.

There is also a considerable amount of research and development work In the areas of preventive maintenance methods for reactor components.-An effective preventive-maintenance program can enhance the safety and efficiency of

plant operations. Vibration monitoring and trending studies

of key reactor components can provide the basic information needed for the decision-making process in an effective

preventive-maintenance program. These studies are used

extensively in ISI programs for French and German reactors.3 They have been used on an experimental basis in

domestic reactors, but vibration monitoring and trending

studies have currently not been formally incorporated Into

U.S. plant ISI programs. Some individual plants are required to make vibration measurements periodically as a

condition of their operating Technical Specifications.

6.1 LPMS

The LPMS Is an in-service detection system designed to

indicate the presence of loose metallic parts in the reactor

primary system. The system can provide diagnostic information to the plant operator concerning abnormal conditions in the reactor so that appropriate actions can be taken

to ensure safety in plant operations. Early warnings can

also minimize the risks to other reactor components and

systems.

The collision of a loose part, carried along by the reactor

coolant flow, with a stationary reactor component will

generate sound waves, primarily bending waves, which

will propagate to other structural components. The effec-

,tiveness of an LPMS will depend on the system's capability

to detect, capture, and interpret these structure-borne sound

waves.

A typical LPMS consists of a series of sensors (piezoelectric accelerometers) mounted on the outside of the

reactor pressure vessel and steam generators to detect

collision-generated structure-borne bending waves. A ring

of three sensors is mounted around the top and the bottom

of a PWR pressure vessel. Two sensors, separated by at

least 3 ft typically, are recommended at the primary inlet

tube-sheet for steam generators. Inputs from these sensors

are amplified and then fed to a monitor that records and

analyzes the signals. When iterprted properly, loose

parts collision signals can provide information on the mass

and energy of the moving object as well the Impact location. The input signals contain information that can be used

to evaluate the dominant frequency and amplitude of the

structure-borne sound waves detected at the sensor locations. Differences in arrival time at the various locations

can also be determined. Using the dominant frequency and

the amplitude of the structure-borne sound waves and a

predetermined calibration curve, the mass and energy of

the loose part are estimated. The location of the Impact

point can be determined by using differences in arrival

47 NUREGCR-W8
IsI

time at the various sensors and a triangulation process.

Many uncertainties are associated with the signal processing procedures. In general, the uncertainty level is low

when the loose part is small, and the sound wave propagation path is simple and straightforward. The uncertainty increases with the size of the loose part and the complexity

of the sound wave propagation path. More detailed information on LPMSs can be found in reports by Kryierl and

Mayo. 4

The performance of LPMS is mixed. This is not unexpected because of the complexity and difficulties in interpreting the structure-borne sound waves generated in a

collision process. EPRI has conducted a research project

with the goal of improving the performance of LPMSs.

Key results and recommendations of the project can be

found in an article by Weiss and Mayo.5

Note that a LPMS is not considered as a safety-related system. LPMSs can provide signals indicating the presence of

loose parts in the reactor primary system. These signals,

when properly processed and analyzed, can lead to actions

that could minimize damages to other reactor components

and systems. Minimizing the damage in a loose-part incident would reduce the outage time for repair work. When

viewed in this manner, the incentive of improving the performance of LPMSs is economic.

6.2 Vibration Monitoring and

Trending Studies

The safety of reactor operations and plant availability can

be improved with a monitoring system that can detect system degradations at an early stage so that failures or malfunctions can be effectively prevented. Vibration monitoring and trending studies, which are common practices in

preventive maintenance programs for rotating machineries,

have potential applications in reactor inspection and maintenance programs.

The hostile environment inside a reactor vessel would preclude the use of sensors attached directly to internal components for long-term vibration measurements. Instead, vibrations of selected reactor internal components can be inferred from neutron noises measured by ex-core detectors.

Neutron noises are fluctuations in the neutron flux around

a mean value. The neutron flux is moderated by the water

layer between the core and the pressure vessel. Vibrations

of some reactor internals (e.g., the core barrel) can change

the thickness distribution of the water layer surrounding

the core. Changes in the water layer thickness lead to variations in the moderating effects and can be correlated to

neutron noises as measured by ex-core detectors. Ex-core

neutron noise measurement is considered an effective

NUREG/CR-6048

method for studying vibrations in reactor internals such as

core barrels and thermal shields. For more information on

the theory and application of neutron noise analysis for reactor diagnosis, consult Ref. 6.

Results of an ex-core neutron noise analysis at a given

power level and a known sensor location are usually presented in the form of a noise spectrum, which is a plot of

the normalized power spectral density (NPSD) curve over

a specified frequency range. The ability to correctly associate characteristic features or spikes on the noise spectrum

with natural frequencies of reactor components is essential

to the success of the method. The structural frequency

identification process can be accomplished by correlating

in-core and ex-core vibration measurements during preoperational vibration testing or by analytical modeling.

Temporary in-core sensors are often used during reactor

preoperational testing and can provide information on the

structural natural frequencies of major reactor internals.

Neutron noise analysis is of limited value for reactor components whose natural frequencies do not form identifiable

features or spikes on the noise spectrum.

When the component natural frequencies are clearly identified on a noise spectrum, trending studies can be performed with noise spectrums at different time intervals.

Typically, three to five measurements are made in one fuel

cycle. Deviations of an identifiable point in an actual noise

spectrum from that of a reference spectrum can be interpreted as indications of degradations in the component.

The process will require the establishment of a detailed and

accurate reference noise spectrum for a specific reactor.

The reference noise spectrum may be obtained from results

of preoperational vibration testings. The ability to correlate

deviations in the noise spectrum with the severity of a system degradation is the key to the success of trending studies. The establishment of a knowledge-based expert system

on normal and abnormal behaviors of reactor components

can aid in the decision-making process.

Neutron noise analysis can be used on many reactor components and systems including reactor internals. One of the

successful applications of the method is the vibration assessments of core barrels in PWRs.7 The core barrel is

specially suited for neutron noise analysis because of its

size and the effects of its vibrations on the distributions of

the water layer thickness between the core and the vessel

wall. The vibrational characteristics of the core barrel are

also well understood. Inspections of the core barrel during

refueling outages confirmed the general correctness of the

neutron noise analysis results. There is practically no

published information on the results of trending studies on

reactor internals in U.S. reactors. One of the reasons could

be because results of vibration monitoring and trending

48
studies may contain information that is considered as proprietary in nature by reactor vendors and plant operators.

Proprietary information is not published in the open literature.

Vibration monitoring and trending studies have been incorporated into ISI programs for French and German reactors. They have been used on an experimental basis in

many domestic plants. The ASME Operation and

Maintenance (OM) Committee is currently studying the

possibility of requiring in-service vibration monitoring for

PWR internals.

References

1. R. C Kryter et al.,'Loose Parts Monitoring: Present

Status of the Technology, Its Implementation in U.S.

Reactors and Some Recommendations for Achieving

Improved Performance," Progress in Nuclear Energy

1(2-4), (1977).*

2. "Loose Parts Detection Program for the Primary

System of Light-Water Cooled Reactors," USNRC

IsI

Regulatory Guide 1.133, Washington, D.C., May

1981.

3. D. Wach, "Vibration, Neutron Noise and Acoustic

Monitoring in German LWRs," Nucl, Eng. and Des.

129 (1991).*

4. C. W. Mayo, "Loose Part Signal Theory," Progress in

Nuclear Energy 15 (1985).*

5. J. M. Weiss and C. W. Mayo, "Recommendations for

Effective Loose Part Monitoring," Nucl. Eng. and Des.

129 (1991).*

6. J. A. Thie, "Power Reactor Noise," American Nuclear

Society, 1981.

7. D. N. Fry, J. March-Leuba, and F. J. Sweeney, Union

Carbide Corp. Nucl. Div, Oak Ridge Nail. Lab., "Use

of Neutron Noise for Diagnosis of In-Vessel

Anomalies in Light-Water Reactors," USNRC Report

NUREG/CR-3303 (ORNLITM-8774), January 1984.t

Available in public technical libraries.

tAvflabl. fr purchase fom National Technical Infcmation Service,

Springfield VA 22161.

49 NUREG/CR-6048
7 Discussion and Conclusions

Reactor Internals operate Inside the pressure vessel and are

subjected to many environmental as well as physical stressors. Primary stressors for PWR internals are related to the

reactor primary coolant flow and exposures to fast neutron

fluxes (E > I MeV). Stressors can initiate and sustain the

growth of aging-related degradation mechanisms.

Aging-related degradation mechanisms develop at different

rates, and conditions inside the pressure vessel may favor

the development of selected dominant mechanisms in internal components. Fatigue, SCC, and mechanical wear are

the major aging-related degradation mechanisms for PWR

internals. They are identified based on a review of reactor

operating histories and reported component-failure information.

Flow-induced cyclic hydrodynamic loads are important

stressors for PWR internals. These loadings include pumpgenerated pressure pulsations, vortex-shedding oscillatory

forces and high-intensity turbulent flows. These cyclic

loads can excite internal components into vibrations. When

they are not properly mitigated, Fly can lead to fatigue and

mechanical failures.

Because of Its low dissolved oxygen content, the water

chemistry of a PWR is not favorable to the development of

SCC in the bulk of the reactor cooling system. However,

narrow gaps and other small regions in some internal components can trap a stagnant fluid medium with locally high

concentrations of corrodents. These crevices create a corrosive environment and can cause SCC when tensile

stresses are also present in the system. Preloads in bolts

can generate tensile stresses needed for the development of

SCC. Crevice-assisted SCC is found in bolted joints and

other tightly fitted connections in reactor intemals.

Long-term exposure to fast neutron fluxes is a stressor that

can lead to embrittlement and radiation-assisted SCC. A

basic understanding of these potential aging degradation

mechanisms has not yet been achieved. There are many

active research works whose main goal is to increase the

understanding of neutron Irradiation effects on the aging of

reactor components.

When they are not mitigated, aging effects will eventually

lead to a failure in the affected components. Understanding

the stressors associated with these aging degradation

mechanisms can lead to the establishment of effective

programs to control and manage these aging effects.

7.1 FIV

PWR internal components have failed as a result of FIV.

Fatigue and mechanical wear are the two important failure

mechanisms. FIV problems are generally resolved by

eliminating the excitation sources or by changing the

vibrational characteristics of the affected structural

components. The practical choice for most reactor internals

is the second option where the component is detuned from

the external excitations.

Some of the FIV problems were detected very early in reactor operations and can be regarded as a part of the reactor

"debugging" process. Excessive mechanical wear observed

in CE reactor internal hold-down rings is an example of a

problem of this nature. The problem was resolved by installing a new hold-down ring and increasing the holddown force during installation. The SSHT in B&W units is

another successful example of detuning the structure from

input excitations.

Other FIV problems are more difficult to resolve. Thermal

shields in many early CE and WE units have encountered

FIV problems. Fatigue cracks In support bolts were the

most common reported failure. Detuning the shields by increasing the stiffness of the support systems did not prove

to be effective in minimizing the effects of flow-induced

excitations. Thermal shields had been removed from some

CE units. Neutron-shield pads, instead of thermal shields,

are now used In the new WE reactors. However, the removal of the thermal shield without an effective replacement may increase the neutron fluence at the wall of the

reactor pressure vessel and adversely affect its long-term

structural integrity. It should be noted that analysis results

indicated that removal of the thermal shields from the CE

reactors will have no undesirable effect in the remaining

design life of the affected reactors.

A more recent FIV-related problem is the excess mechanical wear observed in flux thimbles and thimble guide tubes

in WE units. The problem has been traced to FIV in the

unshielded portion of the thimble between the fuel assembly and the lower core plate. The current practice is to retract the thinned thimble sections away from the vibrating

region. No permanent solution has been proposed, and the

flux thimble and guide tube thinning problem is still under

active investigation.

51 51 ~~~~~NUREG/CR-
Discussion

The baffle plate water-jetting problem observed in several

WE units is an example in which the coolant flow field

was modified to eliminate flow-generated excitations. The

driving force behind the jetting problem was eliminated by

changing a downward bypass flow to an upward bypass

flow in the core baffle-core barrel region. New WE reactors use the upward bypass flow scheme.

Most of the serious FIV problems in PWR internals have

been resolved by well-established engineering practices.

Flow-induced mechanical wear in flux thimble and guide

tubes is an unresolved problem. Many plants have initiated

inspection programs to keep track of the thimble and guide

tube thinning problems.

Due to the relatively high flow speed and high-intensity

turbulence level in the reactor coolant flow, it is not possible to eliminate all FlVs in PWR intemals. Some internal

components are subjected to small-amplitude vibrations

during reactor operations. They may be susceptible to

high-cycle fatigue failures. The effects of a corrosive environment on high-cycle fatigue failures are not well understood. Research results in this area have not reached the

state that they can be incorporated into design codes such

as the ASME B&PV Code. High-cycle fatigue may become

a more important factor in determining the useful working

life for reactor intemals.

7.2 SCC

Most SCCs in PWR internals were detected in bolted joints

and tightly fitted connections. Crevice conditions at these

locations create a highly localized corrosive environment

that is conducive to the development of SCC. The use of

molybdenum disulfide (MoS2) as a thread lubricant can

contribute to the corrosiveness of the crevice condition.

The susceptibility of the material and the corrosiveness of

the cooling fluid are not sufficient to cause SCC. Preloads

in bolts and the tensile stresses that they generate are also

needed for the development of SCC.

Failures have been detected in thermal shield support bolts

in B&W units and in control rod guide tube split pins of

WE reactors. The B&W thermal shield support bolts were

made of A-286 stainless steel, and the split pins are made

of Inconel X-750 alloy. The responsible aging degradation

mechanism is SCC, specifically IGSCC. Improper heat

treatment of the split pins may have also contributed to the

corrosion problems in guide tube split pins.

A two-prong approach has been taken to resolve SCC

problems in PWR internals. The first approach is to reduce

NUREGICR-6048

the susceptibility of the material, and the second is to decrease the tensile stress level in the component. The material of construction for support bolts in the B&W thermal

shield was changed from A-286 stainless steel to Inconel

X-750 alloy. A higher temperature is now used for the beat

treatment of WE guide tube split pins. In both cases the

size of the bolt shank region was enlarged, and preloads

were reduced during installation. These design and fabrication changes should lower the tensile stresses below the

threshold level in the bolts. Together with the use of less

susceptible materials, they should provide adequate protection against IGSCC problems in support bolts.

7.3 Other Potential Aging-Related

Degradation Mechanisms

Aging-related degradation mechanisms develop at different

rates, and, in addition to the three identified major degradation mechanisms, other aging mechanisms may emerge in

future reactor operations.

Neutron irradiation effects may manifest themselves in the

forms of embrittlement and IASCC. The expected lifetime

neutron fluence level for some reactor internals can attain

the value in which neutron irradiation effects are important.

This is of special concern to components in the immediate

vicinity of the core, such as the core baffle, in-core instrumentation guide tubes, and possibly the core support

plates. The existence of crevice conditions, bolt preloads,

and exposure to fast neutron fluxes would make the core

baffle a prime candidate for embrittlement and IASCC.

The exposure to fast neutron fluxes may also lower the

temperature at which creep and stress relaxation can become important aging-related degradation mechanisms for

reactor internals. At the PWR, the normal operating temperature ranges between 288 and 3160C; the best estimate

for the threshold neutron fluence level for creep and stress

relaxation in stainless steels is about 6 x 1019 (neutrons/

cm2). The expected neutron fluence levels for reactor

internals located in the core region can exceed the threshold value and are susceptible to the effects of radiationinduced creep and stress relaxation.

Reactor internal components made of cast austenitic stainless steel (CASS) are susceptible to thermal aging and embrittlemlent. PWR CASS internal components have ferrite

contents in excess of 20% and operate at a temperature of

-3160 C (600TF). CASS components can becomes brittle

under these conditions and may be susceptible to thermal

embrittlement. Many CASS components have been replaced by machined parts.

52
I -

Discussion

In-service inspections are relied upon to detect cracks and

other signs of aging-related failures in PWR internals that

are susceptible to neutron Irradiation id thermal aging

effects. The adoption of a preventive maintenance program

with the capability of detecting system degradations will

enhance the safety as well as the efficiency of plant operations. Such a program can also provide useful information

to the decision-making process concerning the repair

and/or replacement of a degraded component.

Note that most reactor internals are not components of the

reactor primary containment system. Failures of reactor Internals may iesuli in conditions that can challenge the integrity of the reactor primary containment system, but such

failures cannot weaken the system itself. Key internal

components can be replaced if necessary. Therefore, a vigorous and comprehensive in-service inspection program

with the capability of detecting and monitoring system

degradation will ensure the safe working of reactor internals components.

53 NUREG/CR-6048
NUREG/CR-6048

ORNLITM-12371

Dist. Category RV

Internal Distribution

1. D. A. Casada 41. W. P. Poore m

2. D. D. Cannon 42. C. E. Pugh

3. R. D. Cheverton 43. 1. S. Rayside

4. D. F. Cox 44. T. L. Ryan

5. C. P. Frew 45. C. C. Southmayd

6. R. H. Greene 46. J. C. Walls

7. H. D. Haynes 47. ORNL Patent Office

8. J. E. Jones Jr. 48. Central Research Library

9. R. C. Kryter 49. Document Reference Section

10. J. D. Kueck 50-51. Laboratory Records Department

11-40. K. H. Luk 52. Laboratory Records (RC)

External Distribution

53. G. Sliter, Electric Power Research Institute, P.O. Box 10412, Palo Alto, CA 94303

54. J. W. Tills, Institute for Nuclear Power Operations, 1100 Circle 75 Parkway, Atlanta,

GA 30339-3064

55. R. J. Lofano, Brookhaven National Laboratory, Bldg. 130, Upton, NY 11973

56. R. P. Allen, Battelle-PNL, MS P8-10, P.O. Box 999, Richland, WA 99532

57. J. P. Vora, U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory

Research, Electrical and Mechanical Engineering Branch, 5650 Nicholson Lane,

Rockville, MD 20852

58. G. H. Weidenhamer, U.S. Nuclear Regulatory Commission, Office of Nuclear

Regulatory Research, Electrical and Mechanical Engineering Branch, 5650 Nicholson

Lane, Rockville, MD 20852

59. E. J. Brown, U.S. Nuclear Regulatory Commission, Office for Analysis and Evaluation

of Operational Data, Reactor Operation Analysis Branch, Maryland National Bank

Building, 7735 Old Georgetown Road, Bethesda, MD 20814

60. C. Michelson, Advisory Committee on Reactor Safeguards, 20 Argonne Plasma Suite

365, Oak Ridge, TN 37830

61. M. Vagins, U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Electrical and Mechanical Engineering Branch, Division of Engineering,

5650 Nicholson Lane, Rockville, MD 20852

62. J. J. Bums, U.S. Nuclear Regulatory Commission, Division of Engineering Safety,

Office of Nuclear Regulatory Research, 5650 Nicholson Lane, Rockville, MD 20852

63. M. J. Jacobus, Sandia National Laboratories, P.O. Box 5800, Division 6447,

Albuquerque, NM 87185

64. H. L. Magleby, Idaho National Engineering Laboratory, MS 2406, P.O. Box 1625,

Idaho Falls, ID 83415

65. V. N. Shah, Idaho National Engineering Laboratory, P.O. Box 1625, Idaho Falls, ID

83415

66. Office of Assistant Manager for Energy Research and Development, Department of

Energy, Oak Ridge Field Office, Oak Ridge, TN 37831

67-68. Office of Scientific and Technical Information, P. 0. Box 62, Oak Ridge, TN 37831

55 NUREG/CR-6048
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

12-891 (AsIoned Irv NRIC. Add Val., Swap., Pev..

N2C1302 BIBLIOGRAPHIC DATA SHEET NUREG/CR-6048

See instructions on the ORNL/TH-12371

2. TITLE AND SUBTITLE

Pressurized-Water Reactor Internals Aging Degradation Study 3. DATE REPORT PUBLISHED

MONTH I YEAR

Phase 1 September 1993

4. FIN OR GRANT NUMBER

B0828

5. AUTHOR(S) 6. TYPE OF REPORT

K. H. Luk

7. PERIOD COVERED anchww Dates,

B. PERFORMING ORGANIZATION -NAME AND ADDRESS IN NRC. pfvide Division. Office erlion, U.S Nuckar Rguleto Cou nisvion and mWilig addred mifconwsct, provide owne end mailing adibusr

Oak Ridge National Laboratory

Oak Ridge, TN 37831-6285

9. SPONSORING AND ORGANIZATION ADDRESS -NAME Iii NRC. type maes eb we ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~if con rwsos~~~~~~~~~~~ provide NRC Divleion. Office es Region. U.S Nuclear Regulasosy Commission.~~~~~~~~~~

9 SPONSOR ING ORGANIZATION -NAME AND ADDRESS (if NRtyeaeo / etpeviwNCOvin Offk* Or Iionw U&S NaulrarltvuhotwCommtnnon,

and mailihn addre.)

Division of Engineering

Office of Nuclear Regulatory Research

US Nuclear Regulatory Commission

Washington, DC 20555

10. SUPPLEMENTARY NOTES

11. ABSTRACT 1200 was ar less)

Tkis report documents the results of a Phase 1 study on the effects of aging degradations on pressurized-water

reactor (PIR) internals. Primary stressors for internals ar generated by the primary coolant flow in the reactor

vessel, and they include unsteady hydrodynanic forces and pump-generated pressure pulsations. Other stressors

are applied loads, manufacturing processes, impurities in the coolant and exposures to fast neutron fluxes. A

survey of reported aging-related failure information indicates that fatigue, stress corrosion cracking (SCC) and

mechanical wear are the three major aging-related degradation mechanisms for PWR internals. Significant

reported failures include thermal shield flow-induced vibration problems, SCC in guide tube support pins and

core support structure bolts, fatigue-induced core baffle water-jet impingement problems and excess wear in flux

thimbles. Many of the reported problems have been resolved by accepted engineering practices. Uncertainties

remain in the assessment of long-tenn neutron irradiation effects and environmental factors in high-cycle fatigue

failures. Reactor internals are examined by visual inspections and the technique is access limited. Improved

inspection methods, especially one with an early failure detection capability, can enhance the safety and

efficiency of reactor operations.

12. KEY WORDS/DESCR PTORS Listwotdsorphramrharw.,isistefhev in locatng. the teaOn. I 13. AVAILABILITY STATEMENT

NPAR, reactor internals, stressors, aging-related degradation unlimited

mechanisms, component failure information, inservice inspections I4. SECURITY CLASSIFICATION

pressurized-water reactor (PWR), stress corrosion cracking (SCC), unclassifie

fatigue unclassified

Iaes Regotti

unclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 335 1249
Federal Recycling Program
NUREGICR-6048 PRESSURIZED-WATER REACTOR INTERNALS AGING DEGRADATION STUDY SEPTEMBER 1993

UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555-0001

SPECIAL FOURTH-CLASS RATE

POSTAGE AND FEES PAID

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PERMIT NO. G-67

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