NUREG/CR-5754

ORNL/TM-11876

Boiling-Water Reactor

Internals AgingDegradation Study

Phase 1

Prepared by

Prepared by

K. H. Luk

Oak Ridge National Laboratory

Prepared for

U.S. Nuclear Regulatory Commission
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NUREG/CR-5754

ORNIJTM-11876

RV

Boiling-Water Reactor

Internals Aging

Degradation Study

Phase 1

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-ACO5-840R21400
Abstract

This report documents the results of a study on the effects

of aging degradation on 25 selected Boiling-Water Reactor

(BWR) internal components. The operating environment

inside a BWR pressure vessel produces stressors that could

lead to the development of aging-related degradation

mechanisms. A data base containing aging-related failure

information for the selected internal components is

established using data from Licensee Event Reports.

Results of the failure information survey identified two

major aging-related degradation mechanisms for reactor

internals: stress corrosion cracking (SCC) and fatigue.

SCC includes intergranular SCC and irradiation-assisted

SCC (IASCC).

Strategies for controlling and managing aging degradations

are based on understanding the relationship between

stressors and the associated aging-related degradation

mechanisms. The implementation of a plant Hydrogen

Water Chemistry (HWC) program is considered to be a

promising method for controlling SCC, which is the more

prevalent problem for BWRs. Flow-induced vibration

(FMy) is the major cause of fatigue problems in BWR

internals. FIV problems are resolved either by eliminating

the excitation sources or by detuning the structure from

input excitations. Questions remain concerning the

effectiveness of HWC in mitigating SCC (including

IASCC) in internals and in the assessment of high-cycle

fatigue in a corrosive environment.

Vibration monitoring, based on neutron noise

measurements and trending studies, is an inspection

method that can provide early failure detection capability

and can improve the effectiveness of current plant inservice inspection programs. However, the large water gap

and the lack of existing ex-core neutron flux monitors may

hinder the use of neutron noise vibration measurements in

BWRs.

iii iii ~~~~~~NUREG/CR-5
Contents

Page

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

List of Flgures ........ ;.vi.

List of Tables..ix

Acknowledgments..xi A ckn men. wled .... .............................................................. ......................... x

Summary .Xi..

1 Introduction .1

References1...............................................................................................................................................................1

2 BWR Internal Components .3

2.1 Functions of BWR Internals .3

2.2 Component.Descrip..on.4

References..17

3 P..ma.yStressors 19

3.1 Applied.Loadings.19

32 Environmental Stressors .19

3.3 Manufacturing Stressors .20

References..20

4 Aging-Related Degradation Mechanisms .21

4.1 Stress Corrosion Cracking .21

4.2 Fatigue.22

4.3 Embrittlement.24

4.3.1 Radiation Embrittlement .24

4.3.2 Thermal Embrittlement .25

4A. Erosion.25

4.5 Creep and Stress Relaxation .25

References............................................................................................................................................................... 26

5 Survey of Aging-Related Failures .29

5.1 ISI Program for Reactor Internals .29

5.2 Reported Failure Information Survey .29

5.3 Internal Component Failure Information Survey Results .30

5.3.1 Reported SCC Failures .30

5.3.2 Reported Fatigue Failures .31

References.33

.v NUREG/CR-5754
List of Tables

Table

4.1 BWR internals susceptible to SCC

4.2 BWR internals susceptible to fatigue .

4.3 BWR internals susceptible to embrittlement .

4.4 BWR internals susceptible to erosion .

4.5 BWR internals susceptible to radiation-induced creep and stress relaxation .

4.6 BWR internals and potential aging-related degradation mechanisms.

5.1 BWR internals with reported SCC .

5.2 BWR internals with reported fatigue failures .

Page

23

24

25

25

26

27

32

33

ix, ix ~~~~~~NUREG/CR-
Acknowledgments

The guidance and direction provided by D. M. Eissenberg,

D. A. Casada, W. S. Farmer, and R. D. Cheverton in the

conduct of this study are greatly appreciated. The author

also wishes to thank A. E. Cross, L. J. Luttrell, and W. P.

Poore III for their help in obtaining the component failure

data for BWR internals. he assistance of C. C. Southmayd

and E. W. Carver in preparing this report is gratefully

acknowledged.

xi xi ~~~~~~NUJREG/CR
Summary

Reactor internals are located inside the pressure vessel, and

the operating environment is favorable to the development

of time-dependent or aging-related degradation mechanisms. The main objective of this study is to assess the

effects of aging degradations on Boiling-Water Reactor

(BWR) internal components.

Twenty-five BWR internals have been selected for the

aging study; they serve five basic functions: (1) balance-ofplant steam quality control, (2) core support, (3) reactor

coolant flow control and core heat transfer enhancement,

(4) housings for in-core instruments and specimen samples,

and (5) core power shaping and reactivity control. Fuel

assemblies, control rods, and in-core monitors are not

included in the scope of the present study.

Stressors are conditions that can initiate and sustain the

growth of aging degradations. Reactor internals are

subjected to stressors associated with applied loadings

(thermal and mechanical), the reactor cooling water, the

exposure to fast (E > I MeV) neutron fluxes, and

manufacturing processes. Flow-generated oscillatory

hydrodynamic forces and temperature fluctuations are

major applied load stressors. The reactor cooling water can

create an environment that is conducive to the development

of stress corrosion cracking (SCC). Cast austenitic stainless

steels ca l become embrittled by prolonged exposure to

high temperatures. Long-term exposure to fast neutron

fluxes can lead to changes in the mechanical properties of a

material, which can become susceptible to brittle fracture

and SCC. Chromium depletion in grain boundaries of

austentic stainless steels and residual stresses in weld heat

affected zones (HAZs) are primary stressors associated

with manufacturing processes.

Potential aging-related degradation mechanisms for BWR

internals include SCC, fatigue, erosion, embrittlement,

creep, and stress relaxation. These aging mechanisms

develop at different rates and may not be equally important

in the design life of a reactor. Results of a component

failure information survey identified SCC and fatigues as

the two major aging-related degradation mechanisms for

BWR internals.

Major reported failures of BWR internals include cracks in

jet pump supports caused by SCC and fatigue, SCC in core

spray spargers, fatigue cracks in feedwater spargers, and

SCC (in some cases irradiation assisted) in in-core monitor

dry tubes. Some of these failures have led to lengthy repair

times, but there is no indication that these failures

compromised the safety of reactor operations.

Aging degradations, if left unmitigated, will eventually

cause a failure in an affected component. It is essential to

control or eliminate stressors associated with the major

aging-related degradation mechanisms. The development

of SCC requires three conditions: (1) a susceptible

material, (2) a corrosive environment, and (3) the presence

of tensile stresses. The elimination of any one of the three

conditions will reduce the likelihood of the development of

SCC. The effects of the welding-related chromium

depletion process, which can make stainless steel

components susceptible to SCC, can be reduced by the use

of low-carbon-content stainless steels. The tensile stress

level in a structural component can be lowered by the use

of larger components or by reducing the magnitude of the

applied loads. Heat sink welding methods can control the

level of residual stresses in a weld HAZ. The implementation of a plant Hydrogen Water Chemistry (HWC) program

is intended to reduce the corrosiveness of the reactor

cooling water by lowering its dissolved oxygen content.

HWC is effective in mitigating SCC in the reactor recirculating water piping system, but there are insufficient data to

assess its effectiveness in reactor internals. A major

disadvantage of the HWC program is the increase in

radiation level in the power-generating areas. Irradiationassisted SCC (IASCC) is not well understood, and many

active research projects are trying to gain more insight and

understanding into this phenomenon.

Flow-induced vibration is the major cause of fatigue

problems in reactor internals. Low-cycle fatigue is caused

by large-amplitude vibrations and can be prevented by

detuning the system's natural frequency from the dominant

flow-generated excitation frequency. Large-amplitude

vibrations are usually detected and corrected during reactor

preoperational flow testings. Reactor internals are

susceptible to high-cycle fatigue caused by smallamplitude vibrations, which are much more difficult to

control and eliminate. The synergistic effects of high-cycle

fatigue and a corrosive medium are not well quantified at

the present time. High-cycle fatigue is an active agingrelated degradation mechanism for reactor internals.

In the presence of active aging-related degradation

mechanisms, it is essential to maintain a vigorous

inspection program to ensure the structural integrity of

reactor internals. The plant in-service inspection (ISI)

program calls for the visual inspections of accessible areas

of internals during refueling outages. The limitations of the

visual inspection method are well known, and various

alternate inspection methods, such as ultrasonic and eddycurrent inspections, have been tried on an experimental

basis. Reactors licensed after 1978 are equipped with loose

part monitoring systems (LPMSs), which can indicate the

xiii xiii ~~~~~~NUREGIC
Summary

presence of loose parts in the reactor primary system.

Appropriate actions taken by plant operators can limit

further damage to other reactor components and systems.

LPMSs do not possess the capability of indicating the

current status of flaws that may exist in a component.

Neutron noise vibration measurements and trending studies

can provide the information that can be used to evaluate

the current status of flaws that may exist in selected core

internal components. However, these practices have not

been fully exploited by the domestic utility industry.

Visual inspection, supplemented by ultrasonic and eddycurrent inspection, remains the major method for

inspecting reactor internals. The implementation of a

monitoring system with early failure detection capability

will further enhance the safety and efficiency of reactor

operations.

NUREG/CR-5754xi xiv
1 Introduction

The effects of aging or time-dependent degradation on

systems, structures, and components of a commercial

nuclear power plant are of concern to the U.S. Nuclear

Regulatory Commission (NRC). The Office of Nuclear

Regulatory Research (RES) has initiated and sponsored a

Nuclear Plant Aging Research (NPAR) Program.1 The

main objective of the program is to study aging-related

degradation mechanisms and their effects on major reactor

systems, structures, and components. The NPAR Program

also includes evaluation of the effectiveness of inspection,

monitoring, and maintenance methods for managing aging

effects. The basic NPAR approach is to identify major

reactor systems, structures, and components that are susceptible to the effects of aging degradation. Detailed aging

studies are then conducted for these systems, structures,

and components. One of the Oak Ridge National Laboratory's (ORNL's) assignments is to evaluate the effects of

aging degradation on reactor intemals.

Reactor internals are components located inside the reactor

pressure vessel (thus, the term "internal'). The components

perform several different functions. Many internals are

parts of the core support structure, while other components

direct the coolant flow through the vessel and control the

heat transfer in the core. Housings for in-core monitors and

specimen samples are also considered as reactor internals.

Internals for Boiling-Water Reactors (BWRs) and

Pressurized-Water Reactors (PWRs) are different in design

and construction, and they will be treated in separate studies. This Phase 1 report will concentrate on the effects of

aging degradation on BWR internals.

Thirty-seven BWRs are licensed for commercial operations

in the United States.2 Using the commercial operation

starting date as the reference for counting reactor ages, 7

reactors (or -19% of the total) are over 20 years old; 16

reactors (or 43%) are between 10 and 20 years old; and 14

reactors (or 38%) are <10 years old. There is a total of

-522 reactor-years of BWR operations; the operating histories of these reactors provide useful information for

studying aging effects in selected reactor components.

components. The second step is to identify stressors that

are present in the operating environment of the selected

components. The third step is to establish the relationship

between the stressors and potential aging-related degradation mechanisms. The final step is the identification of

major aging-related degradation mechanisms based on a

review of the operating history of the reactors and reported

component failure information. The understanding of the

relationship between major stressors and associated agingrelated degradation mechanisms also provides the basis for

formulating strategies for controlling and managing aging

effects.

Twenty-five BWR vessel internals selected for the present

study are identified in Chap. 2, which will also provide a

brief description of the function, design, and construction

of each component Major stressors for these selected reactor internals are discussed in Chap. 3, and potential agingrelated degradation mechanisms associated with the stressors are identified in Chap. 4. Chapter 5 summarizes the

reported aging-related BWR internal failure information.

The NPAR Program also addresses issues involving the

inspection and maintenance methods used to control and

manage aging effects. The effectiveness of current inservice inspection (ISI) programs and strategies for managing aging degradations are discussed in Chap. 6. Important

results of this Phase 1 aging assessment of BWR internals

are summarized in Chap. 7.

References

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

Program Plan," USNRC Report NUREG-1 144,

Rev. 1, September 1987.*

2. M. D. Muhlheim and E. G. Silver, "Operating U. S.

Power Reactors:' Nucl. Saf. 32(3) (July-September

l9 91).t

Available for purchase from National Technical Information Service,

Springfield, VA 22161.

tAvailable in public technical libraries.

Aging assessment is a multiple-step process. The first step

is the identification and description of the selected internal

1 1 ~~~~~~~NUREG/CR
2 BWR Internal Components

This study considers the effects of aging degradations on

reactor internal components in BWRs designed by the

General Electric Company (GE). It includes reactor design

versions from BWR-2 to BWR-6. The only significant difference between these reactors is that jet pumps are used

beginning with BWR-3 and in all subsequent BWR

designs. Following are the 25 BWR internals selected for

this study:

1. steam dryer

2. steam separator

3. steam separator support ring

4. shroud head

5. shroud head bolts

6. top guide

7. core shroud

8. core plate

9. orificed fuel supports (OFSs)

10. shroud support including access hole covers

11. core spray internal piping

12. core spray sparger

13. feedwater sparger

14. jet pump assembly

15. in-core neutron flux monitor housings

16. in-core neutron flux monitor guide tubes

17. in-core neutron flux monitor dry tubes

18. control rod drive (CRD) housing

19. jet pump sensing line

20. neutron source holder

21. control blades

22. vessel head cooling spray nozzle

23. control rod guide tubes

24. surveillance sample holders

25. differential pressure and liquid control line.

Most of the 25 selected components are commonly identified as "internals" in the Final Safety Analysis Report

(FSAR) for a BWR plant; the three exceptions are the incore neutron flux monitor housings, the CRD housing, and

in-core neutron flux monitor dry tubes. The CRD and incore neutron flux monitor housings are welded to the reactor vessel and are usually considered parts of the pressure

vessel. Dry tubes can be considered as a part of the in-core

neutron flux monitor. These components are included as*

internals because they perform a basic internal function

(i.e., providing housings to in-core instruments).

The fuel assemblies, control rods, and in-core monitors are

reactor components also located inside the reactor vessel.

They perform unique functions that are different from

functions of reactor internals. Fuel assemblies, control

rods, and in-core monitors will not be included in the scope

of the present study.

A simplified sketch of the arrangement of major BWR

internal components with jet pumps is shown in

Fig. 2.1(a), and the arrangement of internals in a BWR

without jet pumps is shown in Fig. 2.1(b).

2.1 Functions of BWR Internals

The 25 selected BWR internals perform a variety of functions. They can be grouped into five general areas:

(1) balance-of-plant (BOP) steam quality control, (2) core

support structures, (3) reactor coolant flow control and core

heat transfer enhancement, (4) housings for in-core instruments and specimen samples, and (5) core power shaping

and reactivity control.

Two of the selected intemals, the steam dryer and steam

separators, are used for BOP steam quality control. Eight

of the 25 components provide structural supports to the

core and other internals: steam dryer support ring, shroud

head, shroud head bolts, top guide, core shroud, core plate,

OFSs, and access hole cover. The next largest group of

internals are housings for in-core instruments and specimen

samples: in-core neutron flux monitor housings, in-core

neutron flux monitor guide tubes, in-core neutron flux

monitor dry tubes, CRD housings, neutron source holder,

control rod guide tubes, jet pump sensing line, and surveil-

-lance sample holders. Six components are used for coolant

flow control and heat transfer enhancement: core spray

internal piping, core spray sparger, feedwater sparger, vessel head cooling spray nozzle, differential pressure and liquid control line, and the jet pump assembly. Control blades

and jet pumps are used for core power shaping and reactivity control.

The majority of the BWR internals are made of type 304

stainless steel. They are designed in accordance with the

rules and regulations of Sect. III of the American Society of

Mechanical Engineers Boiler and Pressure Vessel (ASME

B&PV) Code.1 Specifically, internal components that are

parts of the core support system and those that are used for

reactor coolant flow control and core heat transfer enhancement are treated as safety-class items and are designed to

meet the requirements as stipulated in Appendix 1 of

Sect. III of the ASME B&PV Code. CRD housings, control

rod guide tubes, and in-core monitor housings that are a

part of the reactor primary pressure boundary are also

designed to meet all Sect. III requirements. Allowable

stress values from Sect. III are used as design guides for

internal components that are not safety-class items. Safetyclass internal components are fabricated in accordance with

requirements of Sect. III, Subsection NG; other internals

are fabricated to meet requirements of Sect. IX of the

3 NUREG/CR-5754
BWR

ORNL-DWG 91-2959A ETD

Figure 2.1(a) BWR internals with jet pumps

ASME B&PV Code.2 For reactors that were designed and

built before the establishment of Sect. III, analyses are

performed to ensure that calculated stress values for such

reactor components meet the intent of the Code under

specified design conditions.

The following is a brief description of the function, design,

and construction of the 25 selected BWR reactor internal

components.

2.2 Component Description

Information on the design of the 25 selected BWR internals

is obtained from various plant FSARs and supplemented

by the Electric Power Research Institute (EPRI) report on

the Monticello pilot plant life extension study.3

Reactor design is an evolving process, and many design

features are plant specific. Internal components included in

this report are those for a "typical" reactor designed by the

vendor. The report will attempt to identify major changes

NUREGICR-57544 4
BWR

OSPNt-OMi 9~3-34 ETU

VDUEL HADSLMDL

-VESSEIL 300 n

STEAM PRESSURE

A D~~~~~~~~~fD~MA UREMNTTA

5 ~ ~~~~~~~~~~CNTRO G/L5NW

5s NUREG/CR-5754
BWR

in internal components of the different reactor design versions, but no attempt will be made to identify all design

changes.

Component No. 1: Steam Dryer

Function: The steam dryer removes the excess moisture

in the steam exiting from steam separators. It is not needed

for steady-state operations in the newer BWRs.-

Description: The steam dryer assembly consists of the

drying vanes and the top and bottom structural support

members. They form a single structural unit that is

mounted above the steam separators. The entire unit is

supported by brackets extending inward from the pressure

vessel wall. The moisture removed by the dryer is carried

by a system of troughs and drains to the downcomer,

which is the annular space between the core shroud and the

pressure vessel wall. A sketch of a steam dryer unit is

shown in Fig. 2.2.

Design Code: The steam dryer is designed to meet the

intent of Sect. III of the ASME B&PV Code.

Material of Construction: The steam dryer is made of

type 304 stainless steel.

Component No. 2: Steam Separators

Function: The steam separators separate water droplets

from the steam in the steam-water mixture generated in the

core.

Description: The steam separator assembly consists of

129 standpipes welded to openings in the shroud head, and

an axial-flow steam separator is attached to the top of each

standpipe. Fixed vanes are located inside the separator. As

the steam-water mixture rises through the standpipes, a

vortex is generated by the spinning action imparted to the

mixture by the vanes. The vortex flow separates the water

from the steam in each of the three stages of the separator.

The steam passes through the top of the separators and into

the dryer assembly. The water flows down the standpipes

and into the annular space between the core shroud and the

pressure vessel wall. A sketch of the steam separator

assembly is shown in Fig. 2.3.

Design Code: The steam separator is designed to meet the

intent of Sect. III of the ASME B&PV Code.

Material of Construction: Standpipes and steam separators are made of type 304 stainless steel.

ORNL-DWG 91-2960A ETD

STEAM

FLOW

COLLECTING '

TROUGHDRAIN TUBES

STEAM

FLOW

Figure 2.2 Steam dryer assembly

NUREG/CR-5754 6
BWR

ORNL-DWG 91-2961 ETD head bolts. Steam separator standpipes are welded to openings in the shroud head. A sketch of the shroud head is

shown in Fig. 2.4.

Design Code: The shroud head is designed to meet the

intent of Sect. III of the ASME B&PV Code.

Material of Construction: The shroud head is made of

type 304 stainless steel.

ORNL-DWG 91-2962R ETD

RETURNING - WATER

CORE DISCHARGE

PLENUM

Figure 2.3 Steam separator assembly

Component No. 3: Steam Dryer Support Ring

Function: The steam dryer support ring supports the

steam dryer assembly. The ring is attached to four support

brackets that are welded to the pressure vessel wall.

Description: The circular support ring is'made of coldformed steel sections. The steel section is rectangular and

the dimensions are 3 by 6 in. Some BWR plants have'

steam dryer support rings made of non-cold-worked steel

sections. -'

Design Code: The support ring is designed to meet the

intent of Sect. III of the ASME B&PV Code.

Material of Construction: The steam dryer support ring

is made of type 304 stainless steel.

Component No. 4: Shroud Head

Function: The shroud head covers the discharge plenum

region'located at the top of the core. The plenum serves as

a mixing chamber for the steam-water mixture before it

enters the steam separators. The shroud head also provides

structural support to the steam separators.

Description: The shroud head is a dome-shaped steel

structure attached to the core shroud top flange by shroud

Figure 2.4 Shroud head

Component No. 5: Shroud Head Bolts

Function: Shroud-head bolts fasten the shroud head to the

core shroud top flange. Typically, 36 bolts are used; however, the exact number of bolts and bolt sizes can vary

from plant to plant.

Description: A typical shroud-head bolt is 1.75 in. in

diameter and 14 ft long. A nut is screwed onto one end of

the bolt, and a tee head is welded to the other. A sleeve

covers the rest of the bolt, and the base of the sleeve' is

joined to a collar that is welded to the shaft near the tee

head. A part of the collar is cut out to provide space for the

alignment pin window. A sketch of a shroud-head bolt is

shown in Fig. 2.5.

Design Code: The shroud head bolt is designed to meet

the requirements of Sect.-Ill of the ASME B&PV Code.

Material of Construction: The bolt, the nut, and the tee

head are made of Inconel 600 alloy. The sleeve and the

collar are made of type 304 stainless steel.

;7 7 ~~~~~~~NUREG
BWR

ORNL-DWG 91-2963

BOLT

NUT

.SLEEVE

- SLEEVE BASE

ORNL-DWG 91-2964 ETD

Figure 2.6 Top guide

WINDOW

Design Code: The top guide is designed to meet the

requirements of Sect. III of the ASME B&PV Code.

Figure 2.5 Shroud head bolt

Component No. 6: Top Guide

Function: The top guide provides lateral support and

maintains proper spacing of the upper ends of the fuel

assemblies.

Description: Top guides, with the exception of those in

BWR-6s, are made by welding intersecting steel beams to

a circular rim. At the intersecting joints, the upper beams

have slots cut on the lower part, and the lower beams have

slots cut on the upper part. The upper and lower beams are

interlocked to form the grid. A crevice-free design, in

which the rim and intersecting beams are machined from a

single piece of steel, is used in BWR-6s. Each square

opening formed by the intersecting beams provides lateral

support and maintains proper spacing for four fuel assemblies. Holes are drilled into the bottoms of the beams, and

they provide support to in-core neutron flux monitor dry

tubes and the start-up neutron sources. The top guide is

aligned by positioning pins that fit into slots in the top of

the core shroud. A simplified sketch of a top guide is

shown in Fig. 2.6.

NUREG/CR-5754

Material of Construction: The top guide is made of type

304 stainless steel.

Component No. 7: Core Shroud

Function: The core shroud provides lateral restraint to the

core. The top guide and the core plate are also supported

by the core shroud. The core shroud serves as the partition

separating the upward coolant flow from the downward

recirculating flow in the pressure vessel.

Description: The core shroud is a segmented cylindrical

steel structure. Each segment has a different shroud diameter. The steam separator/shroud head assembly is bolted to

the flange of the top segment, which has the largest diameter. Core spray spargers are attached to the inside surface

of the top shroud segment. The top guide is the lower -.

boundary of the top segment. The middle shroud segment

contains the core. It is the longest segment with an inter--

mediate diameter. The top guide and the core plate define

the top and bottom boundaries of the middle segment. The

bottom segment of the core shroud contains the upper portion of the reactor lower plenum. The bottom shroud segment has the smallest diameter of the three. The bottom of

the core shroud is welded to the shroud support.

8
BWR

For older reactors that are not equipped with jet pumps, the

bottom shroud segment is in the form of a truncated cone.

The top end of the cone segment is welded to the lower end

of the middle shroud segment, and the bottom of the cone

is welded to the reactor vessel wall.

Design Code: The core shroud is designed to meet the

requirements of Sect. III of the ASME B&PV Code.

Material of Construction: The core shroud is made of

type 304 or 304L stainless steel plates. The shell sections

are solution heat-treated and then joined together by longitudinal and circumferential welds.

Component No. 8: Core Plate

Function: Perforations in the core plate provide lateral

support and guidance to control rod guide tubes, peripheral

fuel support pieces, in-core neutron flux monitor housings,

and start-up neutron sources.

Description: The core plate assembly consists of a perforated top plate stiffened by a circular rim and grid support

beams. With the exception of BWR-6s, the top plate is also

supported by tie rods. Control rod guide tubes, peripheral

fuel support pieces, and the in-core neutron flux monitor

housings are inserted into the perforations of the top plate.

Grid support beams are attached to the core shroud, and

they are fillet-welded at regular intervals to the top plate.

Tie rods, when they are used, provide lateral support to the

grid support beams. One end of the tie rods is fillet-welded

to grid support beams through holes in the top plate, and

the other end is attached to the rim at the periphery of the

top plate. A simplified sketch of a core plate is shown in

Fig. 2.7.

Design Code: The core plate is designed to meet the

requirements of Sect. III of the ASME B&PV Code.

Material of Construction: The core plate assembly is

made of type 304 stainless steel.

Component No. 9: OFS Pieces

Function: OFS pieces support the weight of the fuel

assemblies, as well as distributing cooling water to them.

Description: There are two types of OFS pieces. The

standard or four-lobed OFS is a cylindrical structure with

four internal compartments and a central opening for the

positioning of a control blade. Each of the four internal

compartments also has a lobe-shaped opening on the top

surface. The standard OFS piece rests on top of a control

rod guide tube. The bottom of a fuel assembly is inserted

ORNL-DWG 91-2965 ETD

A A

RODS

GRID BEAMS

SECTION A- A

Figure 2.7 Core plate

into one of the lobe-shaped openings, and the OFS piece

provides lateral support and alignment to the bottom of the

fuel assemblies. The weight of the fuel assembly is transferred to the control rod guide tubes through the OFS

piece. The coolant flow into the fuel assemblies is regulated by an orfice located on the side of the lower portion

of the standard OFS piece. The peripheral OFS piece is a

one-opening cylindrical structure that provides lateral support to only one fuel assembly. The peripheral OFS is

welded to perforations in the core plate. The orifice of the

peripheral OFS, which regulates coolant flow into the fuel

assemblies, is located directly below the top opening. A

simplified sketch of a standard OFS piece is shown in

Fig. 2.8.

Design Code: The OFS is designed to meet the requirements of Sect. III of the ASME B&PV Code.

Material of Construction: The center or standard OFS

piece is cast from Grade CF-3 or CF-8 steel. The peripheral OFS is made of type 304 or 304L stainless steel.

9 9 ~~~~~~~NUREG/
BWR

ORNL-DWG 91-2966A ETD ORNL-DWG 91-2967 ETD

ACCESS HOLE COVER

A

L

I

SECTION A- A

Figure 2.8 Standard OFS piece

Component No. 10: Shroud Support

Including Access Hole Covers

Function: The shroud support, also known as the shroud

baffle plate, carries the weight of the shroud, shroud head,

steam separators, peripheral fuel assemblies, startup neutron sources, core plate, top guide, and jet pump diffusers.

The shroud support also provides lateral support to the fuel

assemblies. Jet pump diffusers penetrate the plate to inject

cooling water to the inlet plenum below the core.

Description: The shroud support is an annular plate. The

outer edge of the plate is welded to the inside surface of the

reactor pressure vessel, and the inner edge is welded to the

bottom segment of the core shroud. The inner edge of the

annular plate is also supported by columns welded to the

vessel bottom head. Jet pump diffusers are attached to

openings in the shroud support plate. Two access holes,

1800 apart, provide access to the jet pumps during construction. The access hole covers plugged the two access

openings. The access hole cover is a circular steel plate. It

is put into the access opening and rests on a ledge near the

bottom of the hole. A full penetration weld attaches the

cover to the shroud support. A sketch of the access hole

cover is shown in Fig. 2.9.

NUREG/CR-5754

Figure 2.9 Access hole cover

Design Code: The access hole cover is designed to meet

the requirements of Sect. III of the ASME B&PV Code.

Material of Construction: The access hole cover is made

of alloy 600.

Component No. 11: Core Spray Line Internal

Piping

Function: The core spray line internal piping, commonly

referred to as the core spray line, is a component of the

core spray system that supplies cooling water to the reactor

fuel assemblies during a loss-of-coolant accident (LOCA).

The objective is to supply and distribute sufficient coolant

to the fuel assemblies so that the maximum fuel cladding

temperature of 1204TC (22001F) is not exceeded during a

LOCA.

Description: The core spray line connects the external

core spray piping to the core spray spargers. It is made of a

5-in. Schedule 40 steel pipe, and the content in the pipe is

in a stagnant condition during normal plant operations.

Two core spray lines enter the reactor vessel through two

core spray nozzles. The two nozzles are located 1800 apart

in the vessel wall. Upon entry into the reactor vessel, a'

core spray line divides into two halves that are routed to

opposite sides of the vessel. Along the way, the pipe lines

are supported by clamps attached to the vessel wall. The

pipes then go down the downcomer and are butt-welded to

one end of the tee box pipe sections. The tee box pipe sections enter the top core shroud segment below the top

flange. They then pass through the rim of the top guide and

4

I

10
BWR

are connected to the center of a semicircular core spray

sparger. The tee box pipe section is a component of the

core spray sparger. The route of the other core spray line is

identical, except it is connected to a core spray sparger at a

different elevation. When the core spray system is activated in a LOCA, the core spray spargers inject cooling

water into the core.

Design Code: The core spray line internal piping is

designed to meet the requirements of Sect. HII of the

ASME B&PV Code.

Material of Construction: The core spray line internal

piping is made of type 304 stainless steel.

Component No. 12: Core Spray Sparger

Function: The core spray sparger is a part of the Core

Spray System. It injects cooling water into the core in the

event of a LOCA. The objective is to supply and distribute

sufficient coolant to the fuel assemblies, so the maximum

fuel cladding temperature of 12041C (22001F) is not

exceeded during a LOCA.

Description: The core spray spargers consist of two circular headers at different elevations and two tee box pipe sections for each header. The upper header has bottommounted nozzles, and the lower header has upper-mounted

nozzles. Each circular header is composed of two 1800

segments made of 3-1/2-in. Schedule 40 steel piping. Each

of the 180° header segment is supplied by a tee box pipe

section. The tee box pipe section, made of a 5-in. Schedule

40 steel pipe, connects the core spray spargers to the core

spray line internal piping. One end of the tee box pipe section extends through the shroud wall and is butt-welded to

the core spray line. The other openings of the tee box pipe

sections are connected to the 1800 header segments. The

tee box pipe section is attached to the shroud by the seal

rings with the attachment welded to the 5-in. pipe and the

exterior surface. of the shroud. The content in a core spray

sparger is in a stagnant condition during normal operations.

A simplified sketch of a core spray sparger is shown in

Fig. 2.10.

Design Code: The core spray sparger is designed to meet

the requirements of Sect. 11 of the ASME B&PV Code.

Material of Construction: The core spray sparger is

made of type 304 stainless steel.

Component No. 13: Feedwater Sparger

Function: During normal operation, the feedwater sparger

distributes cooler feedwater to the saturated reactor recirculating water before it comes into contact with the vessel

ORNL-DWG 91-2968 ETD

-LOWER SPARGER

TEE BOX

VEW A- A

Figure 2.10 Core spray sparger

wall. The mixing of the two flows produces a flow of subcooled water to the jet pump and recirculation pump inlets

and prevents the occurrence of pump cavitation problems.

The homogeneous and uniform temperature mixture will

also help to prevent the development of an asymmetrical

power distribution in the reactor core. In the event of a

LOCA, the feedwater sparger becomes a component of the

High-Pressure Coolant Injection'(HPCI) System. The

HPCI System injects cooling water through the feedwater

sparger into the core to maintain an adequate reactor water

level in a LOCA.

Description: The feedwater sparger is a segmented circular header made of 5-in. Schedule 40 curved pipe sections.

The number of segments is either four or six and is plant

specific. Each segment receives water from a feedwater

nozzle. The water inlet is located in the middle of each

sparger segment, and the segment is shaped to fit the contour of the reactor vessel wall. A thermal sleeve is welded

to the water inlet; the other end of the thermal sleeve is

connected to the safe end of the feedwater nozzle by a slipfit joint. Each sparger is supported by the thermal sleeve

and a bracket mounted to each end of the segment. The end

brackets are bolted to'vessel wall brackets. The weight of

11 11 ~~~~~~~NUREG/
BWR

the segment is supported by the vessel wall brackets, which

also locate the header segments away from the vessel wall.

Radial differential thermal expansions of the header segments are accommodated by the thermal sleeves; tangential

slots are cut into the end brackets to account for tangential

differential thermal expansions. The sparger segments are

mounted at one elevation in the vessel. Feedwater is

ejected through multiple small elbow and nozzle assemblies in an inward and downward direction. A simplified

sketch of a feedwater sparger segment is shown in

Fig. 2.11.

Design Code: The feedwater sparger is designed to meet

the intent of Sect. III of the ASME B&PV Code.

Material of Construction: The feedwater sparger is made

of types 316L and 316NG stainless steel.

ORNL-DWG 91-2969A ETD

throat or mixing section, and a diffuser. The inlet section

consists of a transition piece casting, a holddown beam, an

inlet elbow, and a driving nozzle. The inlet section, the

suction inlet, and the throat section form a removable unit.

The diffuser is welded to an opening in the ring segment of

the reactor vessel shroud support structure.

The riser pipe is welded to the thermal sleeve of the recirculating water inlet nozzle; the riser is also supported by

two brace arms welded to beams extending from pads on

the vessel wall. The connection between the throat and the

diffuser is a slip-fit joint. The throat section is also

restrained laterally by a bracket attached to the riser pipe.

A metal-to-metal, spherical-to-conical seal joint is used

between the riser pipe and the transition piece casting. A

tight contact at the seal joint is maintained by clamps that

fit under ears in the riser pipe, and a holddown beam is

used to exert a downward force on a pad on top of the

transition piece casting. A simplified sketch of ajet pump

assembly is shown in Fig. 2.12.

The riser pipe receives high-pressure water from the recirculating water inlet nozzles and delivers the flow up to a

transition piece casting. The flow from the riser pipe is

turned downward into the driving nozzle by the inlet

elbow. The high-speed nozzle exit flow entrains additional

ORNL-DWG 91-2970 EMD

Figure 2.11 Feedwater sparger

Component No. 14: Jet Pump Assembly

Function: Jet pumps provide coolant flow for forced convection heat transfer in the reactor core, and they also provide power distribution shaping and reactivity control.

They are used beginning with BWR-3 designs.

Description: Jet pump assemblies are located in two

semicircular groups in the downcomer annular region

between the core shroud and the reactor vessel wall. Ajet

pump assembly is composed of the riser pipe (shared with

an adjacent jet pump), an inlet section, a suction inlet, the

NUREG/CR-5754

-RPV

RECIRCULATION

INLET NOZZLE

(1 TO EACH JET

PUMP RISER)

Figure 2.12 Jet pump assembly

12
BWR

ORNt-DWO 93-3404 ED cooling water through the suction inlet. The two flows mix

in the throat or mixing section before discharge through the

diffuser. The hydrodynamic forces acting on the jet pumps

are reacted by holddown beams placed across the top of the

transition piece casting.

Design Code: The jet pump is designed to meet the

requirements of Sect. III of the ASME B&PV Code.

Material of Construction: Most components of the jet

pump are made of type 304 stainless steel. Parts that were

cast, such as the transition piece casting and inlet elbows,

are made of either Grade CF-3 or CF-8 steel. A portion of

the diffuser is made of alloy 600. The holddown beam and

two parts at the inlet of the mixing section are made of

alloy X-750.

Component No. 15: In-Core Neutron Flux

Monitor Housings

Function: In-core neutron flux monitor housings provide

a path for the insertion of neutron flux monitors into the

reactor pressure vessel. Source Range Monitors (SRMs),

Intermediate Range Monitors (IRMs), Local Power Range

Monitors (LPRMs), and calibration monitors enter the

reactor vessel through monitor housings.

Description:' The in-core neutron flux monitor housing is

a pipe segment with a seal-ring flange at the bottom. The

pipe segment is inserted into the reactor vessel through

penetrations in the vessel bottom head and welded to the

inside surface of the bottom head. An'in-core neutron flux

monitor guide tube is welded to the top of each housing.

Either an SRMAIRM drive unit, an LPRM, or a calibration

monitor is bolted to the seal-ring flange at the bottom. The

housing unit is considered part of the reactor primary pressure boundary. The configuration of a typical in-core neutron flux monitor is shown in Fig. 2.13. A simplified

sketch of an in-core neutron flux monitor housing is shown

in Fig. 2.14.

Design Code: The in-core neutron flux monitor housing is

designed to meet the requirements of Sect. III of the

ASME B&PV Code.

Material of Construction: The in-core neutron flux monitor housing is made of type 304 stainless steel.

Component No. 16: In-Core Neutron Flux

Monitor Guide Tubes

Function: The in-core neutron flux monitor guide tube

provides a path for the insertion and positioning of neutron

Figure 2.13 In-core neutron flux monitor

flux monitors into the core. IRM, SRM, LPRM, and calibration monitors are inserted into the core through monitor

guide tubes.

Description: The in-core neutron flux monitor guide tube

is a pipe welded to the top of an in-core neutron flux monitor housing. The pipe segment extends from the top of the

neutron flux monitor housing to the top of the core plate. A

latticework of clamps, tie bars, and spacers provide structural support to the guide tube. To prevent loosening during reactor operations, bolts and clamps used in building

the support latticework are welded after assembly. The incore neutron flux monitor guide tube is considered part of

the reactor primary pressure boundary. A simplified sketch

of the in-core neutron flux monitor guide tube is shown in

Fig. 2.14.

Design Code: The in-core neutron flux monitor guide tube

is designed to meet the requirements of Sect. Ill of the

ASME B&PV Code.

Material of Construction: In-core neutron flux monitor

guide tubes are made of type 304 stainless steel.

13 NUREG/CR-5754
BWR

ORNL-OWG 93-3405 EM ORNL-DWG 91-2971 ETM

guide tube

, -

, housing

Reactor vessel ,. _

-GUIDE

TUBE

"k"Fe Figure 2.15 In-core neutron flux monitor dry tube

Figure 2.14 In-core neutron flux monitor housing and

guide tube

Component No. 17: In-Core Neutron Flux

Monitor Dry Tubes

Function: In-core neutron flux monitor dry tubes house

neutron monitors inside the core. Monitors include the

IRM, SRM, LPRM and calibration monitors.

Description: The dry tube is a thin-walled tubing. The

lower end of the tubing is welded to a thick-walled tube

that contains the instrument cavity. The thick-walled tubing is a part of the reactor primary pressure boundary. A

guide plug is welded to the top of the dry tube. An adapter

unit is inserted into the dry tube through the guide plug and

is pressed against a spring located inside the dry tube. The

top of the adapter is inserted into slots in the top guide. A

simplified sketch of a typical dry tube is shown in

Fig. 2.15.

Cbmponent No. 18: Control Rod Drive

(CRD) Housing

Function: The CRD housing provides access into the

reactor pressure vessel for the CRD. It supports the CRD

and the control blade. Ibe CRD housing also provides the

necessary restraint needed to counteract the forces caused

by the activation of the CRD system.

Description: The CRD housing is a tubular structure. The

main portion consists of two tubes welded together. A

flange is welded to the lower end, and a cap is welded to

the upper end. The CRD is bolted to the lower flange. The

CRD housing penetrates the pressure vessel bottom head

and is J-welded to the top of the bottom head-stub tube. In

the newer BWR plants, the CRD housing unit is welded

directly onto the bottom head. There is a thermal sleeve in

the inside surface along the entire length of the CRD housing. The CRD housing is a part of the reactor primary pressure boundary. A simplified sketch of a CRD housing is

shown in Fig. 2.16.

Design Code: The in-core neutron flux monitor dry tube is

designed to meet the requirements of Sect. III of the

ASME B&PV Code.

Material of Construction: The dry tube is made of

annealed type 304 stainless steel.

Design Code: The CRD housing is designed to meet the

requirements of Sect. III of the ASME B&PV Code.

Material of Construction: The CRD housing is made of

type 304 stainless steel.

NUREG/CR-5754 141
BWR

ORNL-DWG 91-2972A ETD Component No. 20: Neutron Source Holder

Function: The neutron source holder contains the

antimony-beryllium start-up sources that are needed to

provide additional neutrons for attaining criticality in the

first operating cycle. Start-up sources are not needed after

the first fuel cycle.

Description: The neutron source holder is a cylindrical

capsule containing the antimony gamma sources and the

beryllium sleeve. The holders are fit into a slot in the top

guide and inserted into a hole in the lower core plate. The

reactor vendor has recommended that neutron source holders be removed from the reactor after the first fuel cycle.

Design Code: The neutron source holder is designed to

meet the intent of Sect. III of the ASME B&PV Code.

Material of Construction: The neutron source holder is

made of type 304 stainless steel.

Component No. 21: Control Blade

Function: Control blades are used for power distribution

shaping and reactivity control. Power distribution shaping

is achieved by the manipulation of the pattern of neutronabsorbing rods in the control blades. The neutronabsorbing rods can also be positioned to neutralize the

effects of steam voids in the top of the core for reactivity

control.

Description: The major component of a control blade is

the neutron-absorbing rod assembly. Neutron-absorbing

rods are made of stainless steel tubings filled with boroncarbide powder. The tubes are seal-welded with end plugs

at the two ends. The rods are held in a cross-shaped pattern

by a steel sheath extending the full length of the rod. The

sheath is spot-welded to the top of the tubes. An upper

handle aligns the tubes and provides the necessary structural rigidity at the top of the rod assembly. The lower

handle and the velocity limiter serve the same purpose at

the bottom of the assembly. Control blades are inserted

into the core from the bottom of the pressure vessel. A

simplified sketch of a control blade is shown in Fig. 2.17.

Design Code: The control blade is designed to meet the

requirements of Sect. III of the ASME B&PV Code.

Material or Construction: Control blades are made of

type 304 stainless steel.

Figure 2.16 CRD housing

Component No. 19: Jet Pump Sensing Line

Function: The jet pump sensing line provides housing for

instruments that are used to monitor the jet pump flow.

Description: The jet pump sensing line is made of a

1/4-in. Schedule 40 pipe (larger pipe for BWR-6s). The

sensing line is welded to the jet pump diffuser at two

square support brackets. The line is routed down the diffuser, around the jet pump, and exits the reactor pressure

vessel through one of the two jet pump instrumentation

nozzles. There is one sensing line for each jet pump in the

reactor.

Design Code: The jet pump sensing line is designed to

meet the intent of SecL III of the ASME B&PV Code.

Material of Construction: The jet pump sensing line is

made of type 304 stainless steel.

15 15 ~~~~~~~NUREG
BWR

ORNL-DWG 91-2973 ETD Material of Construction: The vessel head cooling spray

UPPER nozzle is made of type 304 stainless steel.

HANDLE

Component No. 23: Control Rod Guide

Tubes.

Function: A control rod guide tube provides lateral guido a ance to a control rod and vertical support for a four-lobed

lo OFS piece and the four fuel assemblies surrounding a cono \o NEUTRON trol rod. I ~NEUTRON

ABSRODS Description: The control rod guide tube is a cylindrical

ROdS 1 tube extending from the top of the CRD housing up

o ~~~~~~~~through holes in the core plate. The bottom of the guide

oa tube is supported by the CRD housing. A thermal sleeve is A o inserted into the CRD housing from below and then rotated

SHEATH to lock the control rod guide tube in place. The top cf the

SHEATH 9- tPguide tube goes through a hole in the core plate, and a

four-lobed OFS piece is inserted into the top opening of the

guide tube. A simplified sketch of a control rod guide tube

is shown in Fig. 2.18.

COUPLING Design Code: The control rod guide tube is designed to

RELEASE meet the requirements of Sect. IlI of the ASME B&PV

HANDLE Code.

Material or Construction: The control rod guide tube is

VELOCITY made of type 304 stainless steel. LIMITER

COUPLING

SOCKET ORNL.DWO 93-3406 ETM

Figure 2.17 Control blade

AngnmentK

Component No. 22: Vessel Head Cooling md

Spray Nozzle

Function: The vessel head cooling spray nozzle maintains

saturated conditions in the reactor head region by condensing steam being generated by the hot reactor vessel walls

and other internals. The spray also has the effect of reducing thermal stratification in the reactor vessel coolant so

that the water level in the vessel can rise if needed. A

higher water level can provide effective cooling to more of

the reactor components inside the reactor vessel. . dr

Description: A part of the reactor coolant flow returning Contma od ddve

to the reactor vessel can be diverted to a spray nozzle in the God St""

reactor head. The spray nozzle is mounted to a short length

of pipe and a flange. The spray nozzle flange is bolted to a

mating flange on the reactor vessel head nozzle. Figure 2.18 Control rod guide tube

Design Code: The vessel head cooling spray nozzle is

designed to meet the requirements of Sect. III of the

ASME B&PV Code.

NUREG/CR-5754 16
Component No. 24: Surveillance Sample

Holders

Function: The surveillance sample holders contain impact

and tensile specimen capsules. The test specimens are used

to measure irradiation effects on material properties of the

reactor vessel wall materials.

Description: The surveillance sample holders are welded

baskets hanging from brackets welded to the inside of the

reactor vessel at the middle level of the core. The circumferential positions of the holders are chosen to expose the

test specimens to the same maximum neutron fluxes as

those experienced by the vessel wall.

Design Code: The surveillance sample holders are designed to meet the intent of Sect. III of the ASME B&PV

Code.

Material of Construction: The surveillance sample

holders are made of type 304 stainless steel.

Component No. 25: Differential Pressure and

Liquid Control Line

Function: The differential pressure and liquid control line

serves two functions: to sense the differential pressure

across the core plate and to provide a means for injecting

liquid control solution into the coolant stream.

Description: The differential pressure and liquid control

line enters the reactor vessel as two concentric pipes at an

elevation below the core shroud. The two pipes separate in

the lower plenum. The inner pipe ends near the lower

shroud with a perforated section below the core plate. The

perforated pipe section senses the pressure below the core

plate during normal operations and also serves as the outlet

for the injection of liquid control solution, as needed.

Injection of the cooler liquid control solution through the

inner pipe will reduce the effects of thermal shock on the

BWR

vessel nozzle. The outer pipe ends above the core plate and

is used to sense the pressure in the region outside the fuel

assemblies.

In some BWRs, the core spray line is used as the conduit

for injecting liquid control solution. In these reactors, the

differential pressure sensing lines enter the reactor vessel

through two bottom head penetrations. One line ends near

the lower shroud with a perforated pipe section for sensing

pressure below the core plate. The other line ends just

above the core plate and senses the pressure in the region

outside the fuel assemblies.

Design Code: The differential pressure and liquid control

line is designed to meet requirements of Sect. Inl of the

ASME B&PV Code.

Material or Construction: The differential pressure and

liquid control line is made of type 304 stainless steel.

References

1. American Society of Mechanical Engineers, "ASME

Boiler and Pressure Vessel Code, Section m, Nuclear

Power Plant Components, Div. 1," 1992.*

2. American Society of Mechanical Engineers, "ASME

Boiler and Pressure Vessel Code, Section IX, Welding

and Brazing:' 1992.*

3. Northern States Power Company and Multiple

Dynamics Corporation, "BWR Pilot Plant Life Extension Study at the Monticello Plant Interim Phase 2,"

Interim Report, EPRI NP-5836M, October 1988.

Available from American National Standards Institute, 1430 Broadway,

New York, NY 10018, Copyrighted.

17 17 ~~~~~~NTUREG/C
3 Primary Stressors

Reactor internals are submerged in a hot and moving liquid

coolant during normal plant operations. The nominal system pressure inside the reactor vessel is 7.3 MPa

(1055 psia). The coolant saturation temperature corresponding to the system pressure is 288*C (550'F). The

average core inlet flow velocity is about 2.1 m/s (7 ft/s).

Internal components located in the vicinity of the core are

also exposed to fast neutron (E > I MeV) fluxes and

gamma radiation. The operating environment inside a reactor pressure vessel generates many stressors that are considered as necessary to the development of aging-related or

time-dependent degradation mechanisms.

The term "stressor" is used to represent conditions that

may contribute to the initiation and the sustaining of the

subsequent growth of an aging-related degradation mechanism. Applied loading is probably the most common

stressor for structural components. In addition to applied

loadings, environmental effects and manufacturing processes can also impose stressors on reactor internals.

Environmental effects and manufacturing processes may

cause changes in the mechanical and physical properties of

the materials of construction. Changes in material properties can have a significant impact on the aging process.

Abnormal operating conditions can impose much more

severe conditions on reactor components. However, they

will not be investigated in the present study where the

emphasis is on the assessment of aging effects under normal plant operations. Normal plant operations constitute

the great majority of a reactor's operating history.

3.1 Applied Loadings

Applied loadings are mechanical and thermal loads associated with normal and transient (startup or shutdown) plant

operations. External mechanical loads include static differential pressure loadings, preloads in bolts, and hydrodynamic forces produced by coolant flows in the pressure

vessel. An important internally applied load is one caused

by welding-induced residual stresses.

Flow-gencrated hydrodynamic forces can be static (steady

state) or oscillatory in nature. Steady-state hydrodynamic

forces are usually referred to as lift and drag forces; they

are counterbalanced by the weight of the component and

structural supports. Oscillatory hydrodynamic forces are

generated by flow separations. Pump-generated pressure

pulsations can also act as periodic external excitations to

reactor internals. In the BWR environment, oscillatory

hydrodynamic forces are the major concern because they

can cause a component to vibrate. FIV is a major cause of

fatigue problems in reactor internals.

Preloads on bolts are applied loads acting on reactor internals. Tensile stresses produced by the preloads may

become a contributing factor to the development of cracks

in bolts. Most of the BWR internals are not part of the primary pressure boundary and are not subjected to large

static differential pressure loads. Exceptions are in-core

instrumentation guide tubes and housings; differential

pressure loads are incorporated into design loadings for

these components.

Thermal loads are caused by the existence of temperature

gradients in a component, by thermal expansions of different materials, and by restricted thermal expansions. Reactor internals are designed to accommodate differential

thermal expansions; as a result, constraint-induced, steadystate thermal loads are kept at a low level. Some internal

components are exposed to the mixing of fluid flows at different temperatures. The mixing actions produce rapid

temperature changes in the components. These thermal

cyclings can lead to the development of alternating stresses

and fatigue cracks.1

The applied loadings of major concern to BWR internals

are flow-generated cyclic loads. They produce alternating

stresses that can lead to fatigue failures.

3.2 Environmental Stressors

Major environmental stressors imposed on reactor internals

are the result of their contact with a hot and potentially corrosive coolant. The corrosiveness of the coolant is controlled primarily by the presence of dissolved oxygen in

the reactor cooling water. Dissolved oxygen is a product of

radiolytic reactions in the core. The level of dissolved oxygen in the BWR cooling water, 100 to 300 ppb, is sufficient to produce the electrochemical driving force needed

to promote SCC in sensitized austenitic stainless steel

components. The electrochemical potential of an austenitic

stainless steel is increased by the buildup of the dissolved

oxygen content in the cooling water, and the stainless steel

becomes more susceptible to corrosion. Type 304 stainless

steel, the most common material of construction for internals, is susceptible to SCC in the sensitized condition. The

presence of other impurities such as chlorides may accelerate the corrosion process.

The relatively high operating temperature can produce

physical changes in some reactor internal components.

19 19 ~~~~~~NUREG/CR
I

Primary

Parts made from cast austenitic stainless steel (CASS) can

become embrittled after prolonged operations in the BWR

operating temperature.2 Embrittled materials are more susceptible to fracture failures.

The exposure to fast neutron (E > 1 MeV) fluxes is another

important environmental stressor for internal components.3

The effects are most noticeable for components located in

close proximity to the core. Long-term exposure to fast

neutron fluxes can change the mechanical and physical

properties of the materials of construction for internals.

Irradiation effects can increase the yield and ultimate

strengths and decrease the ductility and fracture toughness,

as well as the critical flaw size, of a material. Neutron

irradiation effects can also lower the temperature at which

creep can become a significant deformation mechanism.

These changes may adversely affect the structural integrity

of reactor internals.

3.3 Manufacturing Stressors

Methods for fabricating reactor internals may impose stressors on the components. Manufacturing stressors can also

promote the development of aging-related degradation

mechanisms. Welding and cold-working are two common

manufacturing methods used in making internals, and they

can introduce stressors to the finished parts.

The conventional welding process sensitizes austenitic

stainless steels (such as type 304) and makes them susceptible to corrosion. Residual stresses in weld heat-affected

zones (HAZs) may contribute to the development of SCC.

Chromium and nickel are the two major alloying elements

used in the making of austenitic stainless steels. Carbon is

also used as a minor alloying element. Austenitic stainless

steels, such as type 304, have generally good corrosion

resistance. However, sensitization can make a welded

austenitic stainless steel component vulnerable to SCC.

When a structural component made of an austenitic stainless steel is cooled down slowly through the temperature

range from -820 to 4800C (1500 to 9001F), as in a welding

process, chromium carbides precipitate at grain boundaries.

The precipitation of chromium carbides will lead to a

decrease in the chromium content in regions adjacent to

grain boundaries. When the chromium content in type 304

stainless steel drops below 12%, the steel becomes susceptible to corrosion attacks at the grain boundaries.4 The

chromium depletion process is known as the sensitization

process, and weld HAZs are common locations for sensitized austenitic stainless steel.

In addition to the material sensitization process, residual

stresses are also generated in weld HAZs during cooldown.

Residual stresses may contribute to the development of

SCC.

Cold-working introduces new manufacturing stressors.

Plastic strains are accumulated in the workpiece during

cold-working, and excessive plastic strain accumulation

can lead to the development of cracks. Contacts with a hot

and corrosive coolant may accelerate the crack initiation

process. Evidence also suggests that cold-working can

induce martensite formation in the workpiece. Martensite

formation may help the material sensitization process and

make a cold-worked component more susceptible to corrosion attacks.5

Applied loadings, environmental effects, and manufacturing processes create stressors that, by themselves or in conjunction with each other, create conditions favorable to the

development of aging-related degradation mechanisms in

BWR internals.

References

1. E. Kiss and T. L. Gerber, "Status of Boiling Water

Research Structural Integrity Program at the General

Electric Company," Nucl. Eng. & Des. 59,27-45

(1980).*

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

Cast Stainless Steel: Mechanisms and Resulting

Properties," Trans. Fifteenth lWater Reactor Safety

Research Information Meeting, Gaithersburg, Md.,

USNRC Conference Proceedings, NUREG/CP-0090,

October 1987.t

3. R. E. Robins, J. J. Holmes, and J. E. Irvin, "Post Irradiation Tensile Properties of Annealed and ColdWorked AISI-304 Stainless Steel," Trans. Amer. NucL.

Soc. (November 1967).*

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

Corros. Sci. Tech. 3 (1973).*

5. "BWR Pilot Plant Life Extension Study at the Monticello Plant Interim Phase 2," Interim Report, EPRI

NP-5836M, Northern States Power Company and

Multiple Dynamics Corporation, October 1988.

Available in public technical libraries.

tAvailable for purchase from National Technical Information Service,

Springfield, VA 22161.

NUREG/CR-5754 20
4 Aging-Related Degradation Mechanisms

The presence of stressors is essential to the development of

aging-related degradation mechanisms. Major potential

aging-related degradation mechanisms for BWR internals

are identified as SCC, fatigue, embrittlement, erosion,

creep, and stress relaxation. SCC includes intergranular

SCC (IGSCC) and IASCC. Fatigue is caused by vibrations

and thermal cycling. Embrittlement can develop as a result

of prolonged exposure to fast neutron fluxes and thermal

aging. Contacts with a high-speed liquid flow or flows with

droplet impingements are common causes of erosion problems. Creep and stress relaxation are caused by prolonged

exposure to high temperatures and fast neutron fluxes.

They can lead to changes in material properties and deformation mechanisms. Stressors associated with these agingrelated degradation mechanisms are present inside a BWR

pressure vessel.

4.1 Stress Corrosion Cracking

Corrosion is the weakening of a structural component as a

result of material deterioration caused by electrochemical

reactions with the surrounding medium. The effects can be

global or highly localized. Global effects are referred to as

general corrosion. The localized effects usually involve

some form of crack development.

The surface of a metallic structure can become oxidized

when it is submerged in a corrosive fluid medium. The removal of corrosion products by the fluid medium will result in a thinning of the wall. This general corrosion process is global and occurs over the entire contact surface in

a somewhat uniform manner. Austenitic stainless steels,

such as type 304, have good resistance to general corrosion

and wastage.

SCC is a common form of highly localized corrosion phenomena. SCC can occur in ductile materials with little or

no plastic strain accumulation associated with the process.

The development of SCC in a structural component requires the simultaneous presence of three conditions: a

conducive environment, a susceptible material, and tensile

stresses above a threshold level. SCC is not likely to develop when any one of these three conditions is absent

from the operating environment. Thus, the elimination of

one condition is the basis for formulating strategies to control SCC. Depending on the alloy compositions and the nature of stressors presented in the system, cracks can develop along grain boundaries; such SCC is known as

IGSCC. When cracks propagate along certain crystallographic slip planes within the grains, the failures are

known as transgranular stress corrosion cracking

(TGSCC). More detailed discussions on the fundamentals

of SCC can be found in texts by Logan1 and Romanov. 2

The hot and oxygenated reactor cooling water creates a

corrosive environment in the BWR pressure vessel. The

dissolved oxygen in the reactor cooling water increases the

electrochemical potential of type 304 stainless steels and

makes them vulnerable to corrosion attacks. The presence

of impurities, such as chlorides and sulfates, in the coolant

may accelerate the crack development process.

Most of the SCC failures in BWR internals are found in

weld HAZ. In addition to the oxygenated reactor cooling

water, the welding process can provide the other two conditions that are needed for the development of SCC.

When a weld is cooled down through the temperature

range from 820 to 4800C (1500 to 9000F), type 304 stainless steel undergoes a sensitization process characterized

by a chromium depletion at grain boundaries. 3 The sensitization process makes austenitic stainless steels susceptible to corrosion attacks. The presence of residual stresses

in weld HAZs supplies the third requirement for SCC. The

level of residual stresses may be reduced by proper heat

treatment Proper heat treatment for internals is not always

possible because of their large size. As a result, residual

stresses remain in weld HAZs. Residual stresses are selfequilibrating, and tensile stresses are always present in

them. The threshold tensile stress level that is needed for

the development of SCC varies, but it is generally accepted

that the level is near the yield stress of the material.

Residual stresses in weld HAZs can approach the yield

stress value, and they can contribute to the development of

SCC. Tensile stresses generated by preloads in bolts can

also attain the threshold value needed for the development

of SCC.

The increase in electrochemical potential of the material,

the chromium depletion sensitization process, and the presence of tensile stresses provide the three necessary conditions needed for the development of SCCs. SCC is probably the most common aging degradation mechanism in

BWR internals. The majority of SCC failures observed in

BWR internals are intergranular. Other conditions may

also have some influence on the development of SCC. The

more significant ones are crevice condition and prolonged

exposure to fast neutron fluxes.

A crevice is a general term describing a small and narrow

region containing a stagnant fluid. High concentrations of

21 21 ~~~~~~~NUREG
Aging

impurities (such as chloride and sulfate) may be trapped in

the stagnant fluid, creating a local environment that is

favorable to the development of SCC. The impurities produce a highly acidic solution that is anionic in nature and

can destroy the protective thin oxide film and increase the

likelihood of corrosion attacks. The presence of crevices in

an internal component can be a design feature, or they can

actually be small cracks initiated by other degradation

mechanisms. Narrow gaps between contact surfaces, such

as those found in bolted joints, are design features that may

create crevices in an internal component.

SCC has been observed in nonsensitized stainless steel internal components. In some cases, the stress levels in these

components are relatively low and below the material yield

stress. A common factor in these nonsensitized, low-stress

internal components with SCC failures is a prolonged exposure to fast neutron (E > I MeV) fluxes. Reactor internals that are exposed to fast neutron fluxes in a corrosive

environment are susceptible to IASCC.4 Most IASCCs are

intergranular.

Basic understanding of IASCC is not complete, and no

unified theory can account for all radiation effects on SCC.

Evidence suggests that a threshold neutron fluence level

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

estimate for the threshold value is about 5 x 1020

(neutrons/cm2) (Ref. 5). BWR internals, such as control

blades, OFS pieces, the core shroud, the core plate, the top

guide, and start-up neutron sources can experience lifetime

neutron fluence levels exceeding the threshold values.

These components are susceptible to IASCC.

It has been suggested that irradiated materials are weakened by gas bubbles formed by transmutation processes.6

Trace quantities of boron, which can react with thermal

neutrons to form lithium and helium,6 are found in

austenitic stainless steels. Hydrogen is also produced by a

transmutation process involving fast neutrons and elements

such as N, Ni, and Cr (Ref. 4). Bubbles are formed when

these gases precipitate from the solid. The bubbles have a

tendency to move to dislocations and grain boundaries, 6

and they create discontinuities that can weaken the material. The solid may disintegrate in the absence of high

tensile stresses when the bubbles are fused together. While

the gas bubble formation mechanism can account for some

of the observed characteristics of IASCC, questions remain

on bubble sizes, their weakening effects on the solid, and

the driving forces behind the bubble movement

The presence of impurities such as phosphorous and

silicone increases the susceptibility of unirradiated

stainless steels to IGS CC. It has been suggested that highNUREG/CR-5754

purity stainless steels are more resistant to IASCC than

commercial-grade stainless steels.4 However, conclusive

evidence to support this assertion is lacking.

The basic understanding of IASCC is not complete, and

there is still much active research work in the field.

General agreement is that the exposure to fast neutron fluence above a threshold value, together with a corrosive

environment, can provide conditions that are favorable to

the development of IASCC. The threshold tensile stress

level needed for the development of IASCC may be

significantly lower than the material's yield stress.

In summary, most welded BWR internals are susceptible to

SCC. Crevice conditions may aid the SCC process.

Prolonged exposure to fast neutron fluxes can lead to

IASCC in nonsensitized steel components with low stress

levels. Of the 25 selected BWR internals, 18 operate in

conditions that would make them susceptible to SSC; they

are summarized in Table 4.1. The last six components

listed in Table 4.1 are also vulnerable to IASCC.

4.2 Fatigue

Reactor internals are subjected to time-dependent or

dynamic loads. A structure will vibrate as a response to

dynamic loads. Vibrations can lead to crack developments,

and structural failures caused by vibrations are known as

fatigue failures. Cyclic loadings and flow-induced vibrations are the root causes of fatigue problems in BWR internals. The loadings can be either mechanical or thermal in

nature. Vortex shedding and other unsteady flow effects

can generate oscillatory hydrodynamic forces, while cyclic

thermal loads are produced by contact with the turbulent

mixing of flow streams at different temperatures. Fatigue

failures can be divided into two types: high- and low-cycle

fatigues. When stress is in the linear-elastic range and the

amplitude of the vibration is small, the resulting failure is

classified as high-cycle fatigue. Little or no plastic strain

accumulation is associated with high-cycle fatigue failures.

High stress values, large amplitude vibrations, and significant plastic strain accumulation are important characteristics associated with low-cycle fatigues. In addition to the

stress and strain criteria, the number of cycles of vibrations

before a failure occurs is used to identify low- and highcycle fatigue. When a crack is initiated between 103 and

104 cycles, the failure is classified as a low-cycle fatigue

failure.

The fluid flows inside the reactor pressure vessel can impose mechanical and thermal loads to internals submerged

in the flow streams. A common mechanical load is the drag

force, which is a steady-state hydrodynamic force acting on

a submerged structure. In transient operations, the propagation of pressure waves through the system can generate

;

22

2-
Aging

Table 4.1 BWR Internals susceptible to SCC

Component Stressors

Steam dryer supportring Corrosive coolant, cold working, weld HAZ

Shroud head bolts - Corrosive coolant, bolt preloads, crevice conditions

Access hole cover Corrosive coolant, crevice conditions, weld HAZ

Steam separator Corrosive coolant, weld HAZ

Core spray sparger - Corrosive coolant, weld HAZ, cold-working

Shroud head Corrosive coolant, weld HAZ

CRD housing Corrosive coolant, weld HAZ

Feedwater sparger Corrosive coolant, weld HAZ

Core spary line internal piping Corrosive coolant, weld HAZ, crevice conditions

Jet pump Corrosive coolant, preloads, weld HAZ

In-core neutron flux monitor housing Corrosive coolant, weld HAZ

In-core neutron flux monitor guide tubes Corrosive coolant, weld HAZ

In-core neutron flux monitor dry tubes Corrosive coolant, weld HAZ, fast neutron fluxes

Neutron source holder Corrosive coolant, fast neutron fluxes, crevice conditions

Core shroud Corrosive coolant, fast neutron fluxes, weld HAZ

Core plate Corrosive coolant, weld HAZ, fast neutron fluxes,

crevice conditions

Control blades Corrosive coolant, fast neutron fluxes, weld IIAZ,

- crevice conditions

Top guide Corrosive coolant, fast neutron fluxes, crevice conditions,

OFS pieces Corrosive coolant, fast neutron fluxes

oscillatory hydrodynamic forces. Core internals are designed to withstand such steady-state and transient loads.

The primary cause of high-cycle fatigue failures in BWR

internals is attributed to a phenomenon known as FIV.

When a fluid medium flows over a blunt body, flow separations lead to the shedding of vortices from the body.

Vortex shedding generates a cyclic load acting on the

structure that would cause it to vibrate. In most cases, the

amplitudes of the vibrations are small. A structure that is

subjected to small-amplitude vibrations is susceptible to

high-cycle fatigue failures. In rare instances, destructive

large-amplitude resonant vibrations could develop in the

structure when its fundamental natural frequency coincides

with an input excitation frequency. However, reactor internals are designed to avoid operating under a resonant.

vibration condition. Key reactor internal components are

monitored during reactor preoperation testings to ensure

the absence of flow-induced resonant vibrations.

The operation of pumps in the reactor primary coolant system introduces flow-related excitations into the reactor

coolant flow stream. They are pressure pulsations with

dominant frequencies closely related to the pump rotating

speed, blade passing frequency, and their harmonics. The

effects of these pressure pulsations are strongest at the entrance regions around vessel inlet nozzles. Coincidence of

a dominant pressure pulsation frequency with a structural

fundamental frequency can also lead to resonant vibration

problems. In any event, these pump-generated pressure

pulsations are excitation sources for small-amplitude vibrations for reactor internals.

There are regions inside the reactor pressure vessel where

reactor coolant flow streams at different temperatures are

mixed in turbulent conditions. Reactor internal components

located in these regions are exposed to a fluid medium with

rapidly changing temperatures. The changing temperatures

can impose thermal loads on components submerged in the

mixing flow streams. The effects of the varying stresses

caused by thermal cycling loads are similar to those caused

by vibrations. The components are susceptible to highcycle fatigue failures.7 Usually little or no plastic strain

accumulation occurs in a structure undergoing rapid

thermal cyclings.

23 NUREG/CR-5754
Aging

Low-cycle fatigue failures are mostly associated with excessive structural deformations under cyclic applied loads.

The components are usually undersized for the loads that

they are required to carry, and plastic strains are accumulated as a result of large structural deformations. Failure,

generally in the form of crack development, will occur if

the plastic strain accumulations exceed a critical value.

Strengthening or "beefing up" the component is a common

and effective solution to low-cycle fatigue problems.

BWR internals that are subjected to cyclic loadings are

susceptible to fatigue failures. Major cyclic loadings include flow-generated oscillatory hydrodynamic forces and

thermal cyclings caused by turbulent mixing of coolant

flow streams at different temperatures. Of the 21 reactor internals, 6 are susceptible to fatigue failures and are summarized in Table 4.2.

Table 4.2 BWR internals susceptible to fatigue

Component Stressors

Jet pump sensing line Flow-generated oscillatory,

hydrodynamic forces

Steam dryer Large applied cyclic pressure

loads

Feedwater sparger Flow-generated oscillatory

hydrodynamic forces, leakage

flow-induced rapid thermal

cyclings

Core spray sparger Flow-generated oscillatory

hydrodynamic forces

Jet pump Flow-generated oscillatory

hydrodynamic forces

In-core neutron flux Flow-generated oscillatory

monitor housings hydrodynamic forces

In-core neutron flux Flow-generated oscillatory

monitor guide tubes hydrodynamic forces

In-core neutron flux Flow-generated oscillatory

monitor dry tubes hydrodynamic forces

4.3.1 Radiation Embrittlement

In addition to fast neutron fluence level, effects of radiation

embrittlement are also determined by the irradiation temperature and material compositions. Austenitic stainless

steels, such as type 304, are considered susceptible to radiation embrittlement. The assessment of radiation embrittlement effects is based primarily on the interpretations of

experimental results. The experimental data base is obtained from testings using surveillance specimens from

commercial power reactors and other test reactor experiments. The reduction in uniform elongation is used as an

indication of the decrease of ductility in a material after a

prolonged exposure to fast neutron fluxes. Changes in fracture toughness are deduced from results of Charpy V-notch

impact testings and other types of fracture-toughness testings.

Experiments8 were conducted at a temperature range from

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

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

cm2. At -3000C (5700F), the unirradiated uniform elongation for type 304 stainless steel is -38%. At the same test

temperature, the uniform elongation was reduced to -22%

at a neutron fluence of -1.5 x 1020 neutrons/cm2. At -1.5

x 1021 neutrons/cm2 the uniform elongation at 300'C was

further reduced to -0.5%. The decrease in uniform

elongation seemed to level off at higher neutron fluence

levels. At a temperature of -300'C (570'F), testing results 8

using type 304 stainless steel test specimens indicated that

effects of radiation embrittlement begin to manifest at a

neutron fluence level of -5 x 1020 neutrons/cm2. The

expected fast neutron fluence for reactor internals is

determined by the distance of the components from the

core. In 40 years of power operations, the estimated maximum neutron fluence is -1 x 1022 neutrons/cm 2 and is

found at the central region of the top guide.9 The estimated 40-year neutron fluence levels for other components

located close to the core,9 such as the core plate, core

shroud, and fuel support pieces, range from -3 x 1020 to

5 x 1021 neutrons/cm2. These internal components are

susceptible to effects of radiation embrittlement. The

expected 40-year fast neutron fluence levels for other

internals are below 1.5 x 1020 neutrons/cm 2, and their

materials of construction should retain sufficient ductility

and fracture toughness to ensure that brittle fracture is not

likely to occur.

4.3 Embrittlement

Embrittlement is the loss of ductility and fracture toughness of a material, and it is usually accompanied by increases in the material yield and ultimate strengths.

Prolonged exposures to fast neutron fluxes and high temperatures (thermal aging) can cause embrittlement in components made of stainless steels. The process can transform

a ductile material to one that is susceptible to brittle fracture failures.

NUREG/CR-5754 24
4.3.2 Thermal Embrittlement

Thermal embrittlement is caused by a prolonged exposure

to high temperatures. It is a time-dependent degradation

mechanism that affects mostly reactor internals made of

CASS. 10 CASS is a two-phase alloy composed of austenite

and ferrite. The determining factor is the ferrite content in

the alloy; when it is <20%, CASS is not sensitive to thermal embrittlement. When the ferrite content is >20%, and

when the operating temperature is in the range from 427 to

4821C (800 to 9000F), a chromium-rich precipitation in the

ferrite phase would lead to a decrease in the fracture

toughness of the material. Also indications are that the ferrite phase could become brittle when a CASS component is

aged at a temperature of -3161C (600oF). Thermal aging is

a potential aging-related degradation mechanism because

the ferrite content in CASS BWR internal components can

go up to 25%, and the required temperature ranges are

within the service conditions inside the pressure vessel.

BWR internal components that are susceptible to radiation

and thermal embrittlements are summarized in Table 4.3.

Table 4.3 BWR Internals susceptible to

embrittlement

Aging

two potential problem areas associated with erosion in

BWR internals are bubble formation in a liquid flow

stream and liquid droplet impingements in two-phase

flows.

Bubbles in a liquid flow stream can become attached to

solid surfaces and can cause erosion by cavitation. In addition, the formation and collapse of bubbles can create disturbances in the flow that can disrupt the protective boundary layer film adjacent to solid surfaces and expose fresh

metal surfaces to the erosion process. High-speed flow regions, such as those in the jet pump throat, nozzle, and

diffuser, are vulnerable to bubble-induced erosion.

Erosion also can occur in a two-phase flow regime, such as

the steam-water flows in steam separators. The impact of

liquid droplets on a metal surface is analogous to the

impact of solid particles and can cause abrasion problems.

Of the 25 selected BWR internals, 2 are susceptible to

erosion and are summarized in Table 4.4.

Table 4.4 BWR internals susceptible to erosion

Component Stressors

Top guide Fast neutron fluxes

Core plate Fast neutron fluxes

Core shroud Fast neutron fluxes

Control blades Fast neutron fluxes

In-core neutron flux Fast neutron fluxes

monitor dry tubes

Cast components of the Thermal aging and ferrite

jet pump assembly precipitation in CASS

OFS pieces Fast neutron fluxes, thermal aging, and ferrite

precipitation in CASS

components

Cast components of the Thermal aging and ferrite

steam separator as- precipitation in CASS

sembly

Component Stressors

Throat, nozzle, and dif- Cavitation-induced

fuser sections in jet, bubbles

pump

Steam separator Liquid droplet

impingements

4.5 Creep and Stress Relaxation

Creep is the progressive deformation of a structure under a

constant stress. Stress relaxation is the reduction of internal

stresses in a component under constraints. Creep can lead

to fracture failures in a structural component. Stress relaxation can loosen bolts in bolted joints, which, in turn, could

result in a loss of structural integrity and leakage problems.

Creep and stress relaxation are generally associated with

structures that operate in a high-temperature environment.

Exposures to fast neutron fluxes may also affect the development of creep and stress relaxation: As a general rule,

when the operating temperature is less than half of the

melting point temperature of the material, effects of creep

and stress relaxation are not significant. The melting point

temperature for a typical austenitic stainless steel is

-1430'C (2600'F). The normal operating temperature for

BWR internals is about 2880C (550'F), less than half of

the stainless steel melting point temperature. Thermally

NUREGICR-5754

4.4 Erosion

Erosion is a potential aging-related degradation mechanism

when an internal component comes into contact with liquid

flows. The abrasive effects are well-known when solid particles are present in a flow stream. Reactor cooling water is

of high purity; solid particles, if they are present in the

flow stream, will be very small in quantity. Solid particle

abrasion is not a serious problem for BWR internals. The

25
Aging

induced creep and stress relaxation are not considered as

major aging-related degradation mechanismns for reactor

internals.

Prolonged exposure to fast neutron fluxes lower the temperature at which creep and stress relaxation can become

significant deformation mechanisms. Results of in-pile experiments conducted at 2880C (5500F) showed that effects

of stress relaxation can be observed in bolts made of type

304 stainless steelsl at a neutron fluence of 6 x 1019 neutrons/cm2. At about the same temperature, significant

stress relaxation has been observed in type 304 stainless

steel test specimens at neutron fluence above 5 x 1020 neutrons/cm2. BWR internals located in the vicinity of the

core have 40-year neutron fluence levels in the range from

6 x 1019 to 5 x 1020 neutrons/cm2. Of the 25 selected

components, 4 are susceptible to creep and stress relaxation and are summarized in Table 4.5.

Table 4.5 BWR internals susceptible to radiationInduced creep and stress relaxation

Component Stressors

Top guide Fast neutron fluxes and high

temperature (2880C)

Core shroud Fast neutron fluxes and high

temperature (2880C)

OFS pieces Fast neutron fluxes and high

temperature (2880C)

Control blades Fast neutron fluxes and high

temperature (2880C)

In-core neutron flux Fast neutron fluxes and high

monitor temperature (2880C)

dry tubes

2. V. V. Romanov, "Stress Corrosion Cracking of

Metals," The Israel Program for Scientific Translation

and the National Science Foundation, Washington,

D. C., 1961.

3. R. L. Cowan and C. S. Tedmon, "Intergranular

Corrosion of Iron-Nickel-Chromium Alloys," Adv. in

Corr. Sci. & Tech 3 (1973).

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

Stress Corrosion Cracking as a Factor in Nuclear

Power Plant Aging," J. Mat. Eng. 9,4 (1988).*

5. Northern States Power Co. and Multiple Dynamics

Corp., "BWR Pilot Plant Life Extension Study at the

Monticello Plant: Interim Phase 2," EPRI NP-5836M,

October 1988.

6. J. Gittus, Irradiation Effects in Crystalline Solids,

Applied Science Publisher, London, 1978.

7. E. Kiss and T. L. Gerber, "Status of Boiling Water

Reactor Structural Integrity Program at the General

Electric Company," Nucl. Eng. Des. 59, 27-45

(1980).*

8. R. E. Robins, J. J. Holmes, and J. E. Irvin, "Post

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

Soc. (November 1967).

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

Assessment of Major Light Water Reactors

Components-Overview," USNRC Report

NUREG/CR4731, November 1989 .t

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

of Cast Stainless Steel: Mechanisms and Resulting

Properties," Trans. Fifteenth Water Reactor Safety

Research Information Meeting, Gaithersburg, Md.,

USNRC Proceeding NUREG/CP-0090, October

1987.t

11. Structural Integrity Associates, Inc., "Component Life

Estimation: LWR Structural Materials Degradation

Mechanisms," Interim Report, EPRI NP-5461,

September 1987.

In summary, 19 of the 25 selected BWR internals are susceptible to SCC, including IASCC. Specifically, 6 out of

the 19 are susceptible to IASCC. Of the 25 components, 7

are vulnerable to fatigue failures. Embrittlement is a potential aging-related degradation mechanism for eight components, and erosion is a potential aging-related degradation

mechanism for two internal components. Five components

are susceptible to the effects of creep and stress relaxation.

The 25 selected components and their potential agingrelated degradation mechanisms are shown in Table 4.6.

References

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

Wiley & Sons, New York, 1966.

Available in public technical libraries.

tAvailable for purchase from National Technical Information Service,

Springfield, VA 23161.

NUREG/CR-5754 26
Aging

Table 4.6 BWR internals and potential aging-related degradation mechanisms

Component SCC Creep Fatigue Embrittlement Erosion

Steam dryer .. _____ *

Steam separator * *

Shroud head *

Shroud head bolts *

Steam separator support ring

Top guide * *

Access hole cover *

Core shroud * *

OFS piece *

Core plate

Core spray line internal

piping

Core spray sparger

Feedwater sparger* *

Jet pump * * * *

In-core neutron flux monitor

housings

In-core neutron flux monitor * *

guide tubes .

In-core neutron flux monitor * * *

dry tubes

CRD housing

Neutron source holder _

Jet pump sensing line *

Control blade

27 27 ~~~~~~~NUREG/C
5 Survey of Aging-Related Failures

Stressors and potential aging-related degradation

mechanisms have been identified for the 25 selected BWR

internals. Aging-related degradation mechanisms do not

develop at the same rate in the environment inside the

reactor vessel. Therefore, component failure information is

needed to properly assess the relative importance of the potential aging-related degradation mechanisms. The identification of major aging-related degradation mechanisms can

also provide information that can be used to formulate

strategies for managing aging effects.

As components of an operating nuclear power reactor, reactor internals are regularly inspected as part of the plant

ISI program. The ISI programs for domestic nuclear power

plants are established under the rules and regulations of

Sect. XI of the ASME B&PV Code.1 The Code is also the

basis for ISI programs in most countries that utilize U.S.-

developed reactor technologies. 2

The program outlines specific procedures and requirements

for inspecting reactor components, structures, and systems

to ensure safe plant operation. The programs may include

orders, rules, criteria, and guidelines established by regulatory agencies as well as industry groups.

5.1 ISI Program for Reactor Internals

The plant ISI program calls for the visual inspection of reactor internals. A complete inspection cycle is 10 years,

and selected components are inspected at refueling outages.

Section XI of the ASME B&PV Codel specifies three

classes of visual inspections: VT-1, VT-2, and VT-3. A

VT-I 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 boundary components, and, as a result, 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 verifications 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. More detailed information of

the inspection methods and procedures can be found in the

appropriate subsections of Section XI of the ASME B&PV

Code.1

During a refueling outage, some internal components are

removed from the pressure vessel and stored in a pool.

VT-i and VT-3 visual examinations are performed on accessible areas of reactor internals when they are in the storage pool. The underwater inspection is performed with remote television cameras under proper lighting. Internals

that remain in the vessel are also visually examined by remote television cameras.

When a crack or flaw is detected, detailed information

concerning the crack, such as its location and suspected

causes, are recorded. The information is also sent to NRC

as a Licensee Event Report (LER). Information on corrective actions taken are also included in the LER. For these

reasons, LERs are considered reliable sources for reactor

component failure information.

5.2 Reported Failure Information

Survey

The survey of the aging-related component failure information for BWR internals is compiled from LERs and nonproprietary EPRI reports on nuclear power reactor operating experiences. Most of these failures were reported in a

10-year period from 1980 to 1989. It is difficult to extract

reactor internal failure information from LERs, and it is

possible that a few failure cases may have been missed.

Therefore, the survey results are not exhaustive in nature,

but they should provide a representative picture of the

aging-related failure situation for BWR internals.

Because of access limitation problems, four BWR internals

are not inspected: the OFS piece, steam separators, shroud

head, and core plate. The other 21 components were inspected in accordance with the plant ISI program. Agingrelated failures were detected in 19 of the 21 components.

The survey results indicated that the core shroud and the

top guide are the two regularly inspected components that

had no reported failures during the survey period.

However, the core shroud and the top guide are subjected

to many significant stressors and are considered vulnerable

to aging-related degradations.

29 29 ~~~~~~~NUREG/C
Survey

5.3 Internal Component Failure

Information Survey Results

A review of the survey results indicates that most reported

aging-related failures in the 10-year survey period can be

attributed to two aging-related degradation mechanisms:

SCC (including IASCC) and fatigue. Radiation-induced

embrittlemcnt may have contributed to the failures of components that were exposed to high-energy neutron fluxes.

5.3.1 Reported SCC Failures

Of the 25 selected BWR internals, 19 are susceptible to

SCC. SCC was detected in 11 of the 19 components. Most

of the cracks were located at weld HAZs. The following

sections provide a brief description of the more significant

reported failures.

53.1.1 Jet Pump Holddown Beanis

SCC was detected in jet pump holddown beams in six

BWRs in the early 1980s. The beam is made of a nickel alloy (Inconel X-750). The cracks were located across the

ligament of the beam section at the mean diameter of the

bolt thread area as illustrated in Fig. 5.1. The cracks were

intergranular. Low heat-treatment temperature (8850C),

which led to the sensitization of the beam material, and

high pre!oads on the beam bolt were major contributing

factors to the development of SCC in jet pump holddown

beams. Replacement holddown beams and beams in later

models of BWRs are solution heat-treated at a higher temperature (1090'C), and preloads in the beam bolts were

also reduced. These measures were effective in preventing

IGSCC problems in new holddown beams.-

A failure of the holddown beam could result in the disassembly of the jet pump and may also lead to a reduction of

ORNL-DWG 92-2913 ETM

the operation safety margin during postulated accident situations. NRC has issued IE Bulletin No. 80-07 (Ref. 3) on

IGSCC problems in jet pump holddown beams; also, the

bulletin established requirements for the inspection of jet

pump holddown beams during refueling outages.

5.3.1.2 Core Spray Spargers

Core spray sparger crackings were detected in six BWRs.

The cracks were initiated and propagated by IGSCC. Cold

working and sensitization during the fabrication of the

spargers and stresses incurred during installation were

considered as major factors in the development of the observed cracks.

Cracks in the core spray spargers could alter the flow distribution to the core and might lead to the generation of

loose parts in the pressure vessel. NRC IE Bulletin No. 80-

13 (Ref. 4) addresses issues concerning the core spray

sparger cracking problems and outlines requirements for

the inspection of core spray spargers during refueling outages.

5.3.1.3 Access Hole Covers in Shroud Support

Ultrasonic (UT) inspections of one BWR showed indications of partial through-the-wall cracks in the welds attaching the access hold covers to the shroud support plate.

These cracks were not detected by visual inspections.

Welding-induced residual stresses and crevice conditions

on the weld are major contributing factors to the initiation

and propagation of cracks in the access hole cover to

shroud support welds.

A failure of the access hole cover to shroud support weld

could lead to a separation of the cover plate from the

shroud support. The potential exists that a severed access

hole cover may be swept into the recirculation pump suction line, causing damage to the pump. The unplugging of

the access hole covers will also have the undesirable effects of creating an alternate flowpath that would allow

some of the recirculating system flow to bypass the core

during normal and accident operating conditions. NRC has

issued Information Notice No. 88-03 (Ref. 5) on the importance of inspecting the access hole cover to shroud support

welds, but the Information Notice did not specify any specific inspection requirement.

53.1.4 In-Core Neutron Flux Monitor Dry Tubes

Indications of cracks were detected by visual inspections of

IRM and SRM dry tubes in several BWRs. The cracks

TRUNNION

Figure 5.1 Crack locations of JET pump holddown

beams

NUREG/CR-5754 30
were located in the thin tube segment surrounding the -

compression spring. Due to the high neutron fluence level

at the locations of these components, it is generally accepted that IASCC and radiation-induced embrittlement

may have contributed to the development of the observed

cracks. The accumulation of cruds and oxide formation in

crevices between the guide plug and the thin tube segment

may have produced the tensile stresses needed for the development of SCC.

The thin tube is a nonpressure-retaining boundary of the

dry tube. There was no indication of cracks in the adapter,

shaft, guide plug, and thick instrument cavity (a primary

pressure retaining boundary). Cracked dry tubes are replaced during refueling outages.

53.1.5 Stub Tube to CRD Housing Welds

During routine inspections of the pressure vessel of two

older BWRs, leakage was detected in the gap between the

vessel wall and the CRD housing. Reactor cooling water is

leaked to the outside of the pressure vessel through a slipfit clearance between the inside surface of the stub tube

and the outside surface of the CRD housing. The leak is

considered as unisolable.

Leakage was attributed to the development of through-thewall cracks in the stub tube in the vicinity of the J-weld

that joins the CRD housing to the top of the stub tube. The

cracks were intergranular. A typical crack is illustrated in

Fig. 5.2.

ORNL-DWG 92-2914 ETD

Survey

The leakage was stopped by using mechanical seals sized

to slip over the outer surface of the stub tube and by rollswaging the CRD housing to close the gap with the vessel

wall.

Stub tubes are parts of the reactor pressure vessel, and their

failures are included in the report because they were caused

by the attachment of the CRD housing. This is considered

as a generic problem for older BWRs with stub tubes. Stub

tubes are not used in newer BWRs where the CRD housing

is welded directly to the bottom bead of the reactor vessel.

A search of the reported failure data base did not locate any

failure of the CRD housing to vessel bottom head welds.

Other SCC failures were also reported and are summarized

in Table 5.1. The type of reactors involved and the number

of failure cases were not available when the information

was taken from EPRI reports.

Specifically, IASCC failures were reported in three BWR

internals: control blades, IRM and SRM dry tubes, and

neutron source holders. Radiation-induced embrittlement

may also have contributed to the failures of the dry tubes.

5.3.2 Reported Fatigue Failures

Of the 25 selected BWR internals 7 are susceptible to fatigue, and failures were detected in 6 of the 7 components.

Most of the reported cases are classified as high-cycle

fatigue caused by flow-induced vibrations. Low-cycle fatigue was the cause of failure of cracks detected in thin

steam dryer hoods, and excessive structural deformations

under cyclic pressure loads were the responsible failure

mechanism. The following is a brief description of the

more important fatigue failure cases.

53.2.1 Feedwater Sparger

In 1972, large circumferential cracks were detected in a

feedwater sparger segment in a BWR-3 unit. Subsequent

inspections revealed similar cracks in other BWR-3 and -4

units. The cracks were located near the sparger water inlet

in the vicinity of the feedwater nozzle. Failures were attributed to flow-induced vibrations and rapid thermal cycling.

The thermal sleeve connects the feedwater sparger to the

feedwater nozzle. One end of the thermal sleeve is welded

to the sparger water inlet, and the other end is joined to the

safe end of the feedwater nozzle by a slip fit joint. A simplified sketch of the feedwater sparger to feedwater nozzle

CRD

HOUSING

VESSEL

WALL

N/ J-WELD

STUB TUBE

11 -. .

Figure 5.2 Crack In CRD stub tube

31 31 ~~~~~~~NUR
Survey

Table 5.1 BWR internals with reported SCC

Component Cases reported Failure location

Core spray sparger 6 (BWR-3, BWR-4) Circumferential cracks in sparger

(distribution unknown) ring and at header arm to 'T' box

weld HAZs

CRD housing 2 (1 BWR-2, 1 BWR-3) Stub lube at HAZ of CRD housing

to stub tube weld

Shroud head bolts 2 (2 BWR-4) Near root of collar to bolt weld

Access hole cover 1 (1 BWR-4) Access hole cover plate to shroud

support welds

Core spray line internal piping 2 (2 BWR-4) Weld HAZs

Steam dryer support ring Several Cold-work locations

Feedwater sparger Several Weld HAZs in ejection nozzles

Jet pump assembly holddown beam 6 (5 BWR-3, 1 BWR-4) Bearn ligament sections at beam bolt

diameter

Jet pump assembly riser pipe 2 (1 BWR-3, I BWR-4) Riser pipe to safe end welds

Control blades Several Sheath to tie rod and handle welds,

bail portion of handle

IRM dry tube 4 (2 BWR-2, 1 BWR-3, I BWR-4) Top of thin tube segment surrounding compression spring

SRM dry tube 3 (2 BWR-2, I BWR-4) Top of thin tube segment surrounding compression spring

Neutron source holder Several Holder near top of beryllium

chamber

connection is shown in Fig. 5.3. In the original design, the

slip fit joint is loose, and feedwater leaked into the narrow

region between the outside surface of the thermal sleeve

and the inside surface of the feedwater nozzle. Instability

of the leakage flow caused the feedwater sparger to undergo vibrations. In addition, the cooler leakage feedwater

is mixed with the hot recirculating downcomer reactor water in the region surrounding the sparger water inlet. Flowinduced vibrations and rapid thermal cyclings led to the

development of fatigue cracks in the feedwater sparger and

around the feedwater nozzle comer.

ORNL-DWG 92-2915 ETD

REACTOR

The feedwater sparger fatigue cracking problem was resolved by the implementation of a new sparger with a tight

slip-fit joint, effectively reducing the leakage flow down to

a very low and manageable level. A new feedwater nozzle

was also used to prevent crackings in the nozzle with the

new feedwater sparger design.

The BWR vendor has conducted extensive studies on the

feedwater sparger cracking problem, and the results are

summarized in an article by Kiss and Gerber.6

5.32.2 Jet Pump

Jet pumps in most early BWR-3 units were affected by

flow-induced vibration problems, which led to the development of fatigue cracks in the pump support system. The

design deficiency was recognized early, and the fatigue

cracking problem was resolved by using a new and

stronger holddown beam. The improved jet pump support

was implemented to many units before they become operational.

Due to its operation characteristics, fatigue caused by flowinduced vibrations is always a concern with jet pumps in

BWRs.

SLIP FIT

JOINT I

Figure 5.3 Feedwater sparger and feedwater nozzle

connection

NUREG/CR-5754 32
Survey

53.2.3 LPRM Dry Tube References

Indications of leakage from three Local Power Range

Monitor (LPRM) dry tubes were detected during a cold

shutdown of an overseas BWR-6 unit. The reactor was operating in the low-pressure coolant injection (LPCI) mode.

Damages to the LPRM dry tubes were confirmed in subsequent inspections of reactor internals, and 46 pieces of a

broken dry tube were retrieved. The damaged dry tube also

represented a degradation of a portion of the reactor

coolant pressure boundary. Failures were attributed to fatigue caused by flow-induced vibrations generated by the

LPCI flow.

The problem was unique to BWR-6 units and was resolved

by the installation of a flow deflector at the LPCI discharge

openings inside the shroud. The flow deflector shields the

LPRM dry tubes from excitations generated by the LPCI

flow.

The flow deflector was installed in domestic BWR-6 units.

The reported failure data base showed no LPRM dry-tube

failure in domestic BWR-6 units.

Fatigue failures in other BWR internals were reported and

are summarized in Table 5.2. Although some reactor internals are susceptible to erosion, creep, and stress relaxation

failures, an extensive search of the LER data base did not

locate any reported failure in BWR internals that can be

attributed to these particular aging-related degradation

mechanisms.

1. American Society of Mechanical Engineers, ASME

Boiler and Pressure Vessel Code, Section XI, "Rules

for Inservice Inspection of Nuclear Power Plant

Component."*

2. Science Application International Corp., "Nuclear

Plant In-Service Inspection Requirements and

Practices in Different Countries: A Comparative

Review," Interim Report, EPRI NP-5919, July 1988.

3. "BWR Jet Pump Assembly Failure," IE Bulletin No.

80-07, USNRC Office of Inspection and Enforcement,

Washington, D.C., April 1980.t

4. "Cracking in Core Spray Sparger," IE Bulletin

No. 80-13, NRC Office of Inspection and

Enforcement, Washington, D.C., May 198 0.t

5. "Cracks in Shroud Support Access Hole Cover

Welds," Information Notice No. 88-03, NRC Office of

Nuclear Reactor Regulation, Washington, D.C.,

February 1988.t

6. E. Kiss and T. L. Gerber, "Status of Boiling Water

Reactor Structural Integrity Programs at the General

Electric Company," Nucl. Eng. and Des. 59, 27-45

(1980).t

Avaulable from American National Standards Institute, 1430 Broadway,

New York, NY 10018, Copyrighted.

tAvailabe in public technical libraries.

Table 5.2 BWR internals with reported fatigue failures

Component Cases reported Failure location

Steam dryer 8 (BWR-2, BWR4, BWR-5) Cracks in thin dryer hoods and lift-

(distribution unknown) ing rod straps

Steam dryer support ring 1 (BWR4) Cracks in support bracket

Jet pump sensing line 5 (1 BWR-3,4 BWR-4) Attachment welds to diffusers and in

sensing lines

Jet pump restrainer gate 2 (2 BWR-3) Tack welds in outboard clamp bolt

keeper

In-core neutron flux monitor 1 (BWR-6) Cracks in dry tube at upper cooling

(LPRM) dry tube holes

Feedwater sparger Several Circumferential cracks in sparger

near reactor pressure vessel feedwater nozzle

33 33 ~~~~~~~NUREG/CR-5754
6 ISI and Aging Management Program

Visual inspection is a relatively simple inspection method

used mainly to detect surface flaws and cracks. However,

the method is limited to easily accessible regions of the

component being examined and is not capable of detecting

subsurface or partial through-the-wall cracks. Due to access limitations, interior regions of steam separators, OFS.

pieces, the shroud head, and the core plate are not inspected.

Alternate methods have been tried, on an experimental basis, to detect partial through-the-wall cracks, as well as

flaws in inaccessible regions in BWR internals. A method

that has been used with some success is ultrasonic (IT) inspection. A UT inspection has located partial through-thewall cracks in the access hole cover that were not detected

by visual examinations; however, UT inspection has its

own access limitation problems, and the interpretations of

UT inspection results are more complicated and difficult

than are those for visual inspections. Many active research

works are investigating alternate inspection methods; the

objective is to improve the overall effectiveness of the

plant ISI programs. Establishment of a plant ISI program is

an evolving process. When new information or new inspection methods become available, the responsible organization will issue bulletins and letters to inform plant operators of the latest developments and their implications in inspection procedures and maintenance requirements.

One of the major concerns in reactor operations is the presence of loose parts in the primary system. Because of the

long interval between inspections and the inability of visual inspections to detect subsurface or partial through-thewall cracks, the scenario that a reactor may operate with

flawed or cracked internal components cannot be ruled out.

The failure of reactor internals during operation may generate loose parts that can cause damage to components in

the reactor primary system. Such failures have occurred in

PWRs,1 but there is no reported incident of this nature in

BWRs.

Loose parts can endanger the core support and core cooling

functions of reactor internals. In addition to safety implications, loose parts in the reactor primary system can lead to

extensive outages for repair work. For safety reasons,'NRC

is interested in the development of monitoring methods

that can detect the presence of loose parts in the reactor

primary system. Research and development in detecting

loose parts has led to the establishment of NRC Regulatory

Guide 1.133,2 which outlines the operating requirements

for the establishment of Loose Part Monitoring Systems

(LPMSs) in domestic commercial nuclear power plants;

Reactors licensed since 1978 are equipped with LPMSs.

LPMSs have also been installed in many plants licensed

before 1978.

6.1 LPMS

The LPMS is an acoustic-based monitoring system designed to detect the presence of loose metallic parts in the

reactor primary system. It provides diagnostic information

to the plant operator so the operator can take appropriate

actions to ensure a safe plant operation. These corrective

actions can also minimize the risks to other reactor components and systems.

Sound waves will be generated when a loose part, carried

along by the reactor coolant flow, collides with a stationary

reactor component. The sound waves, basically bending

waves, will propagate to other parts of the reactor, including the pressure vessel. The effectiveness of the LPMS will

depend on the system's capability to detect these impact

waves and the operator's ability to interpret these structureborne signals.

The LPMS uses a series of sensors (piezoelectric accelerometers) mounted on the outside surface of the reactor

pressure vessel to detect collision-generated, structureborne sound waves. Four rings of three sensors each are

recommended for BWRs. One ring is located in the top

head region and another at the bottom head region of reactor vessels. The other two rings are mounted in the cylindrical portion of the vessel, usually in the vicinity of outlet

nozzles. Inputs from sensors are filtered and then fed to a

monitor that will record and analyze them. More detailed

information on LPMSs can be found in the reports by

Kryter3 and Mayo.4 -

The impact sound waves contain information that can be

used to estimate the mass and energy of the moving object,

as well as the location of the point of impact. A dominant

frequency and the amplitude of the structure-borne sound

waves at a sensor location can be extracted from the input

signals. Differences in arrival times at various sensor locations are also measured. The mass and energy of the moving part are estimated based on acomparision of the measured dominant frequency and amplitude of the input

sound waves with a known calibration curve. The impact

location is determined by using differences in arrival times

at known sensor locations and a triangulation process.

Many uncertainties are associated with the signal processing procedures. As a general rule, the uncertainty level is

low when the mass of the moving object is small and the

propagation path of the structure-borne sound waves is

35 35 ~~~~~~~NUREGIC
Is'

simple. The uncertainty level increases with the mass of

the loose part and the complexity of the sound wave propagation path.

The performance of LPMS is mixed. Because of the complexity and difficulty in processing and interpreting the

structure-borne sound waves generated in a collision process, mixed performance is not unexpected. EPRI has conducted a study with the goal of improving the performance

of LPMSs; recommendations from the study can be found

in the paper by Weiss and Mayo.5

While an LPMS can provide information that can be used

to detect the presence of loose parts in the reactor primary

system, it lacks the capability of indicating the current status of system degradations that may exist in reactor components, structures, and systems. The safety of reactor operations and plant availability can be enriched by a monitoring system that has the capability of providing information regarding the status of system degradations in a component. Based on the information, appropriate corrective

actions can be initiated before failures and malfunctions

occur. 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.

6.2 Vibration Monitoring and

Trending Studies

The hostile environment inside a reactor pressure vessel

would rule out the use of accelerometers attached to internals for long-term vibration measurements. Most reactor

internal vibrations are measured through indirect means. A

common method is to infer internal component vibrations

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 internal components located around the core can change

the thickness distribution of the water layer surrounding

the core. Changes in the water layer thickness can lead to

variations in the moderation effects of the water and can be

correlated to neutron noises as measured by ex-core detectors. Ex-core neutron noise measurement is considered an

effective method for detecting vibrations in internal components such as the core shroud, in-core neutron monitor

dry tubes, and fuel channel boxes.

The key to the neutron noise vibration measurement

method is the establishment of a reference noise spectrum

at a given reactor power level and at specified sensor locations. The noise spectrum is usually given in the form of a

plot of the normalized power spectral density (NPSD)

curve over a specified frequency range. The ability to

identify characteristic features or spikes in the noise spectrum with structural natural frequencies of internals is essential to the success of the method. Neutron noise vibration analysis is of limited value to internals whose natural

frequencies cannot be clearly identified on the noise spectrum. Fundamentals of the theory and practice of reactor

neutron noise vibration measurement can be found in the

text by Thie.6

A reference noise spectrum can be obtained from results of

reactor preoperational testings. Temporary in-core sensors

can provide information that can be used to identify characteristic natural frequencies of internal components in the

reference noise spectrum. The structural natural frequencies of reactor internals are also computed by analytical

models. Trending studies are performed by comparing

noise spectrums obtained during plant operations with the

reference spectrum. Three to five neutron noise measurements are made during a fuel cycle. Deviations of an identifiable point in an actual noise spectrum from that in the

reference spectrum can be interpreted as indications of

degradations in an internal component. The ability to correlate deviations in the noise spectrums with the severity of

system degradation is essential to the success of trending

studies.

It may be more difficult to establish an effective neutron

noise measurement system in BWRs than in PWRs. By

comparision, the water layer thickness between the core

and the pressure vessel wall is larger in BWRs than in

PWRs. The increase in water gap may reduce the sensitivity of the water layer thickness vs changes in moderation

effects and make it more difficult to detect neutron noise

signals. Also, there are no existing ex-core neutron detectors in BWRs' As a result, neutron noise vibration monitoring and trending studies have not gained acceptance with

BWR plant operators and reactor vendors.

Vibration monitoring and trending studies are used extensively in inspection and maintenance programs for German

and French reactors,7 but their potential applications to

plant ISI programs have not been fully explored in the

United States.

The ISI program is an inspection and maintenance program

that is designed to detect failures before they can become a

threat to the safety of plant operations. It offers little help

in controlling or reducing the effects of aging degradations.

Aging-related degradation mechanisms can be managed by

NUREG/CR-5754 36
controlling or eliminating stressors associated with the aging degradation mechanisms. Strategies used to manage the

two major aging-related degradation mechanisms will be

addressed separately.

6.3 Strategy for Managing SCC

Reported aging-related failure information survey results

show that SCC is probably the most common aging degradation mechanism for BWR internals. Austenitic stainless

steels, such as type 304, are used extensively in the construction of BWR internals. These steels provide good

general corrosion resistance capability; however, the use of

welding in the fabrication process and the reactor cooling

water may combine to create conditions that are conducive

to the development of SCC. As noted previously, three

conditions are needed for the development of SCC: a corrosive environment, a susceptible material, and the presence of tensile stresses. All three conditions are present in

BWR internals when weld HAZs come into contact with

the reactor cooling water.

The basic strategy for controlling SCC is the elimination of

any one of the three conditions. Material sensitization,

which is responsible for the susceptibility of austenitic

stainless steels, can be minimized by the use of stainless

steels with a low carbon content and heat sink welding

methods. However, for many existing operating reactors

these are not viable options. The two remaining options for

these reactors are the reduction of the corrosive agent in

the coolant and the lowering of the tensile stress level in a

component.

Tensile stresses in an internal component are caused by externally applied loads and by residual stresses in weld

HAZs. Proper heat treatments can reduce residual stresses

in weld HAZs, but they are not considered practical for

most reactor internals because of the size of the components involved. When the dominant tensile stresses are

generated by applied loads, the stress level in a component

can be lowered by the use of larger structural components.

Reducing the magnitude of the applied load is also effective when it is possible to do so. These remedies have been

used to reduce the tensile stress level in jet pump holddownbeams. The reduction of the tensile stress level is an option

for controlling SCC, but it has limited applications.

The remaining option for controlling SCC is to lower the

dissolved oxygen content in the reactor cooling water. The

dissolved oxygen content can be reduced by the injection

of hydrogen gas into the coolant flow stream. The hydrogen will react with oxygen to form water. The process beIsI

comes known as the Hydrogen Water Chemistry (HWC)8

program.

The dissolved oxygen content in the reactor cooling water

is a product of radiolytic reactions in the core and is

generic to the reactor operation. A direct relationship exists

between the dissolved oxygen content and the electrochemical potential (ECP) of austenitic stainless steels. When the

ECP is increased, the steel becomes more susceptible to

SCC. The removal of the dissolved oxygen and lowering

the ECP to a value below -0.23 V (standard hydrogen electrode) will reduce the risk of the development of SCC in

internals made of austenitic stainless steels.

Hydrogen injection is an effective method to remove the

dissolved oxygen from the coolant flow stream.

Measurements in high-purity water showed a significant

drop in the ECP for austenitic stainless steels when the dissolved oxygen content is reduced to below 40 ppb.9 The

susceptibility to SCC for type 304 stainless steels is greatly

reduced when the dissolved oxygen content is kept below

40 ppb. Laboratory testing resultsl 0 indicated that HWC

can inhibit the initiation of SCC, as well as retard the

growth rate of existing flaws.

The implementation of a plant HWC program is an effective method for controlling SCC in primary reactor coolant

piping systems. In principle, it should also be an effective

method for mitigating SCC in internal components. The

radiolytic reaction rate is higher around the core region,

where most internals are located, and a higher hydrogen

gas injection rate may be required to reduce the local dissolved oxygen content. However, field operating data with

BWR internals in an HWC environment are very limited.

As more plant HWC programs are implemented, a larger

operating data base will become available, and it can be

used to assess the effectiveness of HWC for reactor internals. The BWR vendor, GE, has issued a Service

Information Letter (SIL)1

I to BWR plant operators recommending the implementation of a HWC program in all

BWR plants.'

Note that the implementation of a plant HWC program will

increase the radiation fields in the power-generating areas

such as the turbine building. The HWC program may come

into conflict with other programs aiming at reducing radiation exposure levels to workers in nuclear power plants.

The presence of impurities such as chlorides and sulfates

may also contribute to the development of SCC. Limits are

imposed on the BWR water chemistry parameters to control the impurity contents in the reactor cooling water. The

conductivity of water, which is a reliable indicator of the

overall level of impurities in the water, is monitored during

37 NUREG/CR-5754
ISI

reactor operations. During normal BWR power operations,

the operating specifications stipulate that the chloride and

sulfate contents shall be maintained at 20 ppb or less, and

the water conductivity at <0.3 uS/cm.8 When the water

conductivity exceeds 0.3 gS/cm, the reactor water cleanup

system is activated to lower the water conductivity to below its allowable limit. Prolonged operation in an off-standard condition (in excess of 96 h) will require the implementation of an approved program of corrective measures

that will bring the level down below the allowable value.

When the water conductivity exceeds 1.0 pS/cm for a

period of 24 h, the reactor will be brought to a cold shutdown within 16 h. When the water conductivity exceeds

5 gS/cm, the reactor will be shut down in an orderly

manner as rapidly as other plant constraints will allow. The

reactor water cleanup system and operating procedures will

ensure that the cooling water is maintained at an acceptable

purity level so that the degradating effects of impurities

can be kept to a minimal level.

Other corrective actions can also be implemented to reduce

the risk of SCC. The presence of crevices in ar, internal

component is a feature that could aid the development of

SCC, as well as reduce the effectiveness of HWC. A

crevice-free design and manufacturing process, together

with the use of stainless steels with a low carbon content

and heat sink welding method, could produce internal

components that are less susceptible to SCC. This combined approach should be used in making internals for

newer reactors and for replacement parts.

IASCC is a major aging-related degradation mechanism

for reactor internals located around the core. The effects of

a plant HWC program on the development of IASCC are

not well understood. A reduction of the corrosive agent in

the cooling water may reduce the risk of IASCC, but there

are not sufficient operating data to quantify the effects.

6.4 Strategy for Managing Fatigue

BWR internals are susceptible to high- and low-cycle fatigue failures. High-cycle fatigue is the more prevalent

degradation mechanism-..

The development of cracks in thin steam dry hoods is one

of the few cases of reported aging-related failures in which

low-cycle fatigue was the suspected cause. The thin steam

dry hoods were subjected to large structural deformations

under cyclic pressure loads. As a result of these early problems, thicker steam dryer hoods are used in the newer reactors and as replacement parts. Failures have not been detected in these new and thicker steam dryer hoods.

A second potential source for low-cycle fatigue failures in

reactor internals is the development of flow-induced resonant vibrations. Reactor internal components are designed

to prevent the development of large-amplitude flowinduced resonant vibrations; this is accomplished by

separating the component structural fundamental natural

frequency from the dominant flow-induced excitation frequency. As an added measure, preoperation flow testings

are used to detect potential resonant and other types of vibration problems. The development of flow-induced resonant vibrations is a problem that has plagued the operations

of heat exchangers, but it is not a common problem for reactor internals. In one reported case of fatigue failure, the

development of flow-induced resonant vibrations may be

the suspected cause. This case involved a LPRM dry tube

in a BWR-6 reactor. The reported failure was detected during a shutdown operation with the reactor in the low-pressure coolant injection (LPCI) operating mode. Damage was

attributed to fatigue failure at the dry tube upper cooling

holes caused by FIV created by the LPCI flow. The problem was resolved by the installation of a flow deflector to

shield the LPRM dry tubes from flow-induced excitations.

No evidence suggests that the development of resonant vibrations is a common problem for BWR internals.

However, vessel internals are not immune to the effects of

flow-induced small-amplitude vibrations, which could lead

to high-cycle fatigue failures. The effects of small-amplitude vibrations are cumulative; depending on the stress

level in the component, they could increase the probability

of fatigue crack initiation in a component after a certain

number of cycles of operation. For most stainless steels

and at specified stress levels, the allowable number of cycles of operation can be evaluated by using the fatigue

crack initiation curves specified in the ASME B&PV

Code.12 The fatigue crack initiation curves, commonly referred to as the S-N curves, tend to approach asymptotic -

stress values as the number of cycles increases. The S-N

curves for some material level off beyond a certain number

of cycles of operation, and the corresponding asymptotic

stress values are known as endurance limits. Theoretically,

there is no limit to the number of cycles of operation when

the maximum stress level in a component is kept below the

endurance limit. Maintaining a low stress level in a reactor

internal component is the basic strategy for managing

high-cycle fatigue problems. When the stress level cannot

be kept below the endurance limit or when the S-N curves

do not level off, a counting procedure is used to keep track

of the number of cycles of operation; the information can

be used to estimate the time to fatigue crack initiation.

Note that environmental factors have not been taken into

account in the development of the fatigue crack initiation

curves. Flaws in a component could also affect the time to

fatigue crack initiation and alter the rate of crack propagation.

NUREG/CR-5754 38
A fracture-mechanics analysis can be used to evaluate the

time interval between crack initiation and unstable crack

propagation. The rate of fatigue crack propagation is sensitive to environmental conditions, and attempts are being

made to incorporate these factors into the analyses. At the

present time, the confidence level in these predictive methods is uncertain at best because environmental effects have

not been fully incorporated into the analysis.

IsM

6. J. A. Thie, "Power Reactor Noise:' American Nuclear

Society, La Grange Park, III, 1981.

7. D. Wach, "Vibration, Neutron Noise and Acoustic

Monitoring in German LWRs, "Nucd. Eng. and Des.

129 (1991).t

References

8. "BWR Hydrogen Water Chemistry Guidelines: 1987

Revision," EPRI NP-4947-SR, Electric Power

Research Institute, Palo Alto, Calif., December 1988.

1. "Characterization of the Performance of Major LWR

Components:' EPRI NP-5001, Electric Power

Research Institute, January 1987.

2. U.S. Nuclear Regulatory Commission Regulatory

Guide 1.133, "Loose Parts Detection Program for the

Primary System of Light-Water Cooled Reactors."*

3. 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," Prog. Nucl. Energy 1, 2-4

(1977).t

4. C. W. Mayo, "Loose Part Signal Theory," Prog. NucL.

Energy 15 (19 8 5 ).t

5. J. M. Weiss and C. W. Mayo, "Recommendations for

Effective Loose Parts Monitoring," Nucl. Eng. and

Des. 129 (199 1).t

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

Electrochemical Studies of the Stress Corrosion of

Type 304 Stainless Steel," Corrosion 35,288 (1979).t

10. B. M. Gordan et al., "Hydrogen Water Chemistry,"

Interim Report, General Electric Company, NEDE30261, September 1983.

11. General Electric Company, Service Information Letter

No. 408, July 9, 1984.

12. American Society of Mechanical Engineers, "ASME

Boiler and Pressure Vessel Code, Section VIII,

Pressure Vessels, Div. 1."*

Copies are available from U.S. Government Printing Office,

Washington. D.C. 20402. A1rN: Regulatory Guide Account. t

Available in public technical libraries.

*Available from American National Standards Institute, 1430 Broadway,

New York. NY 10018, Copyrighted.

39 NUREG/CR-5754
7 Discussions and Conclusions

BWR internals operate in an environment that is favorable

to the development of aging-related degradation mechanisms. The primary stressors are a high-temperature, corrosive, and moving liquid coolant; manufacturing processes; fast neutron fluxes; and cyclic loadings. Reported

aging-related failures can be attributed to two major degradation mechanisms: SCC (including lASCC) and fatigue.

The presence of crevices and cold-work locations in an internal component may contribute to and accelerate the development of SCC problems.

7.1 SCC

The three conditions needed for the development of SCC

are a corrosive environment, a susceptible material, and

tensile stresses. Dissolved oxygen is the primary corrosive

agent in the reactor cooling water, and material sensitization can make austenitic stainless steels susceptible to corrosion attacks. Preloads in bolts and residual in weld HAZs

are major sources of tensile stresses in an internal component.

The basis for formulating aging management strategies is

the control or elimination of any one of the conditions associated with the development of aging-related degradation

mechanisms. The use of stainless steels with a low carbon

content and heat sink welding methods can make internal

components less susceptible to the corrosive environment;

however, these are not viable options for existing reactors

that have not been built with these technologies. A popular

approach taken by the nuclear power industry to mitigate

SCC in BWRs is to reduce the dissolved oxygen content in

the reactor cooling water. This is accomplished by the implementation of a plant HWC program in which hydrogen

gas is injected into the coolant flow stream to remove the

dissolved oxygen. Testing results indicated that the plant

HWC is effective in suppressing SCC in reactor primary

coolant piping systems. More operating data will be

needed to assess the effects of HWC on reactor internals.

The use of crevice-free design and manufacturing methods

may aid in the control of SCC in reactor internals in new

reactors and in replacement parts.

The exposure to fast neutron fluxes is a major contributing

factor to the development of IASCC. IASCC has been detected in nonsensitized steel components with low stress

levels. By reducing the corrosiveness of the reactor cooling

water, the plant HWC program should be beneficial to the

control of IASCC. However, more operating data will be

needed to assess the effects of HWC in mitigating IASCC

in reactor internals.

The implementation of a plant HWC program will increase

the radiation fields around the power-generating areas.

This may limit the use of HWC because the increase in radiation level is in conflict with the goal of reducing radiation exposures to workers in nuclear power plants.

7.2 Fatigue

Reactor internals undergo vibrations as a response to dynamic loads. Dynamic loads can be mechanical or thermal

in nature. Vibrations will eventually lead to crack initiation

and crack growth. The time to crack initiation is determined by the levels of stress and structural deformations in

a component.

Most fatigue failures in BWR internals are caused by

small-amplitude FIVs. Unlike large-amplitude FIVs, it is

not possible to eliminate all small-amplitude vibrations,

and high-cycle fatigue is an active aging degradation

mechanism for reactor internals. Also uncertainties are associated with the synergistic effects of fatigue and a corrosive environment A vigorous inspection program is

needed to control fatigue cracking problems in reactor internals.

7.3 Inspection Programs

When an aging-related degradation mechanism is not controlled or mitigated, it would eventually lead to failures in

the affected components. Effective inspection and aging

management programs are essential to the prevention of

aging-related failures.

BWR internals are visually inspected in accordance with

the rules and regulations of Sect. XI of the ASME B&PV

Code. Visual inspections can detect surface flaws, but the

method has shortcomings. The major one is access limitation; namely, the inspection is only performed on parts that

can be visually observed. As a result, internal components

with a complex geometry or located in inaccessible regions

are not inspected on a regular basis. Another shortcoming

of visual inspections is that they cannot detect subsurface

or partial through-the wall cracks. Other inspection methods, such as UT and eddy-current inspections, have been

used to examine internals. Both the UT and eddy-current

inspection methods have their own limitations. For these

reasons, the scenario in which a reactor may operate with

flawed or cracked internal components cannot be ruled out.

Reactors licensed since 1978 are equipped with LPMS.

LPMS is an acoustic-based monitoring system and can detect the presence of loose parts in the reactor primary

41 NUREGICR-5754
Discussions

system. When a loose part is detected, corrective actions

can be taken to limit potential damages to other reactor

components and systems. LPMS does not possess the

capability of detecting incipient failures. Vibration

monitoring and trending studies can provide such

capability. Neutron noise measurement is a common

method for detecting vibrations in selected reactor

internals. Trending study is an assessment of the relative

vibration amplitude of an internal component as a function

of time. Increases over a predetermined limit are

interpreted as signs for the incipient failure of a

component. The large water gap between the core and the

reactor vessel wall may reduce the effectiveness of neutron

noise vibration measurements for BWR internals. The lack

of ex-core neutron flux monitors may also handicap the use

of neutron noise vibration measurement in BWRs.

An effective vibration monitoring and trending studies

program can enhance reactor safety, as well as plant

efficiency. These preventive maintenance practices have

been used sparingly but should warrant more consideration

and exploitation by the U. S. utility industry.

Reactor internals, while they perform safety-related functions such as core cooling and core support, are not components of the reactor primary containment system. They can

be replaced if necessary. As a result, reactor internals are

not considered to be high-ranking safety-class components.

Failures of internals could create conditions that may

challenge the integrity of the reactor primary containment

system, but they do not affect the effectiveness of the

primary containment systems. However, aging-related

failures may require extensive shutdown time for repair

works. The implementation of effective aging degradation

management and improved monitoring programs can improve the safety, as well as the efficiency, of long-term reactor operations.

NUREG/CR-5754 42
NUREG/CR-5754

ORNLIIM-11876

Dist. Category RV

Internal Distribution

1. D. A. Casada 41. W. P. Poore III

2. D. D. Cannon 42. C. E. Pugh

3. R. D. Cheverton 43. J. 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 C

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, ORO, Oak Ridge, TN

37831

67-68. Office of Scientific and Technical Information, P. O. Box 62, Oak Ridge, TN 37831

NUREG/CR-5754

43
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1 REPORT NUMBER 12 891 SA.signed by NRC. Add Vol.. Suva.. REV.. 3201,202, BIBLIOGRAPHIC DATA SHEET NUREG/CR-5 7tL'

(See instructions on the reverse)UREG/CR-5754 lORNL/TM-1 1876

2. TITLE AND SUBTITLE

Boiling-Water Reactor Internals Aging Degradation Study 3. DATE REPORTPUBLISHED

MONTH | EAR

Phase 1 September 1993

4. FIN OR GRANT NUMBER

B0828

5. AUTHORIS) 6. TYPE OF REPORT

K. H. Luk

7. PERIOD COVERED finclusiM Dates,

B. PERFORMING ORGANIZATION - NAME AND ADDRESS 1l NRC. providebjvision. Office or Region. U.S. Nuclear Regulatory Commission. and maihng address: ii contractor, provide

nats. nd mailing ag/ddrusr

Oak Ridge National Laboratory

Oak Ridge, TN 37831-6285

9. SPONSORING ORGANIZATION -NAME AND ADDRESS 1li NRC, type "Same as aboee-: if contractor. peovide NRC ivision. Office or Region. U5. Nuclar aRegulatory Commission,

and mailing address.)

Division of Engineering

Office of Nuclear Regulatory Research

US Nuclear Regulatory Commission

Washington, DC 20555-0001

10. SUPPLEMENTARY NOTES

11. ABSTRACT (200 words oriless

This report documents the results of an aging assessment study for boiling water reactor (BWR) internals.

Major stressors for BWR internals are related to unsteady hydrodynamic forces generated by the primary

coolant flow in the reactor vessel. Welding and cold-working, dissolved oxygen and impurities in the coolant,

applied loads and exposures to fast neutron fluxes are other important stressors. Based on results of a

component failure information survey, stress corrosion cracking (SCC) and fatigue are identified as the two

major aging-related degradation mechanisms for BWR intemals. Significant reported failures include SCC in jetpump holddown beams, in-core neutron flux monitor dry tubes and core spray spargers. Fatigue failures were

detected in feedwater spargers. The implementation of a plant Hydrogen Water Chemistry (HWC) program is

considered as a promising method for controlling SCC problems in BWR. More operating data are needed to

evaluate its effectiveness for internal components. Long-term fast neutron irradiation effects and high-cycle

fatigue in a corrosive environment are uncertainty factors in the aging assessment process. BWR internals are

examined by visual inspections and the method is access limited. The presence of a large water gap and an

absence of ex-core neutron flux monitors may handicap the use of advanced inspection methods, such as neutron

noise vibration measurements, for BWR.

12. KEY WORDSIDESCRIPTORS (List words or pA roses that wvil assist rsearchrn in locating the teport.) 13. AVAI LA81 LIT Y STATEMENT

unlimited

NPAR, reactor internals, stressors, aging-related degradation 14.SECURITYCLASSIFICATION

mechanisms, component failure information, inservice inspections (ThiisParl

boiling-water reactor (BWR), stress corrosion cracking (SCC), fatigue unclassified

(rhis Report)

unclassified

15. NUMBER OF PAGES

16. PRICE

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- -NUREG/CR-57,54' : ; -BOILING-WATER REACTOR INTERNALS AGING DEGRADATION STUDY, '- - -- :: : - SPEBR19 ,

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