f

NUREG/CR-5379

PNU-6560

Vol. 1

Nuclear Plant

Service Water System

Aging Degradation Assessment

Phase I

Prepared by D. B. Jarrell, A. B. Johnson, Jr., P. W. Zimmerman, M. L. Gore

Pacific Northwest Laboratory

Operated by

Battelle Memorial Institute

Prepared for

U.S. Nuclear Regulatory

Commission
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i

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NUREG/CR-5379

PN~jr-6560

Vol. 1

RM., R9

Nuclear Plant

Service Water System

Aging Degradation Assessment

Phase I

Manuscript Completed: May 1989

Date Published: June 1989

Prepared by D. B. Jarrell, A. B. Johnson, Jr., P. W. Zimmerman, M. L. Gore

Pacific Northwest Laboratory Richland, WA 99352

Prepared for

Division of Engineering

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

NRC FIN B2911
ABSTRACT

The initial phase of an aging assessment of nuclear power plant service

water systems (SWSs) was performed by the Pacific Northwest Laboratory to

support the Nuclear Regulatory Commission Nuclear Plant Aging Research

(NPAR) program. The SWS was selected for study because of its essential role

in the mitigation of and recovery from accident scenarios involving the

potential for core-melt. The objectives of the SWS task under the NPAR

program are to identify and characterize the principal aging degradation

mechanisms relevant to this system and assess their impact on operational

readiness, and to provide a methodology for the mitigation of aging on the

service water aspect of nuclear plant safety. The first two of these

objectives have been met and ar¢ covered in this Phase I report.

A review of available literature and data-base information indicated

that motor operated valve torque switches (an electromechanical device) were

the prime suspect in component service water system failures. More extensive

and detailed data obtained from cooperating utility maintenance records and

personnel accounts contradicted this conclusion indicating that biologic and

inorganic accumulation and corrosive attack of service water on component

surfaces were, in fact, the primary degradation mechanisms.

A review of the development of time dependent risk assessment (aging)

models shows that, as yet, this methodology has not been developed to a

degree where implementation is reliable. Improvements in the accuracy of

failure data documentation and time dependent risk analysis methodology

should yield significant gains in relating aging phenomena to probabilistic

risk assessment.

. , .~~~~~~~i

ir j
ACKNOWLEDGMENT

The authors wish to thank the cooperating utility that shared its experience and insight on service water system operation and maintenance with

the investigating team. We would like to express our appreciation to the

Electric Power Research Institution Service Water Working Group for their

review and comments on this report. The authors would also like to

acknowledge the editorial assistance of John W. Nageley.

V
SUMMARY

The service water system (SWS) represents the final heat transfer loop

between decay heat generated in the nuclear core and the safe dispersal of

that heat energy into the environment. It is the objective of this assessment performed by the Pacific Northwest Laboratory to demonstrate that aging

phenomena in SWSs can be identified and quantified such that aging degradation of system components can be detected and mitigated before the reduction

of system availability to below an acceptable threshold.

The following are the SWS task goals which were directly derived from

the Nuclear Plant Aging Research (NPAR) Program plan:

1. Identify the principal aging-degradation mechanisms, then focus on indepth study to identify and characterize the phenomena involved.

2. Examine the current surveillance specifications and make recommendations

on their accuracy to provide accurate reliability information.

3. Provide a means to evaluate the effectiveness of maintenance on

mitigating aging degradation phenomena.

4. Produce an inspection plan which optimizes the effectiveness of

inspections based on system risk reduction.

5. Utilize the information generated by this task to resolve related

generic issues and provide-guidance for aging and life extension

regulatory criteria.

The following was the approach used during Phase I:

* Perform a literature search of government and private sector reports

which related to service water, aging-related degradation, and potential

methodologies for analysis.

* Assemble a data-base that contains a listing of all commercial power

plants in the U.S.--their SWS configurations, characteristics, and water

source.

* Obtain and examine the available service water data from large generic

data-bases, i.e., NPRDS, LER, NPE, inspection reports, and other relevant plant reference data. Analyze the SWS of a specific power plant

for aging-related degradation phenomena based on the available data

obtained from this data-base.

* Perform a fault tree analysis of a typical plant SWS to examine failure

propagation and understand specific input requirements of probabilistic

risk analyses.

vii
* Develop an in-depth questionnaire protocol for examining the information

resources at a plant which are not available in the standard data-bases.

Subsequently, visit a central station power plant and solicit the

required information.

* Analyze the information obtained from the in-depth plant interrogation

and draw contrasts and conclusions with the data-base.

* Utilize the plant information to perform an interim assessment of SWS

degradation mechanisms and focus future investigations.

The following is a summary of Phase-I conclusions relative to the stated

goals:

1. Aging-related degradation of nuclear plant open (direct system interface

to raw water without chemical control) SWSs is prevalent and constitutes

a valid safety concern. Based on actual s~ecific plant data, the primary degradation mechanism found in open SWSs is corrosion, compounded

by biologic and inorganic accumulation. This conclusions directly contradicts the results of a failure analysis based on information

obtained from the Nuclear Plant Reliability Data System (NPRDS) database.

2. Based on multiple plant samplings, the current level of surveillance and

post-maintenance testing performed on the SWS is not sufficient to

accurately trend or detect system degradation due to aging phenomena.

3. While post-maintenance surveillance does give some measure of the

effectiveness of system modification/repair efforts, sufficient operational condition monitoring, and design basis post-maintenance testing,

information is not available to characterize SWS maintenance

effectiveness.

4. To improve the accuracy of current data to a point which would allow a

high degree of confidence in aging degradation analysis, a root cause

logic scheme needs to be developed that can be used to define the depth

of knowledge and documentation required to accurately characterize an

aging-related component failure event.

5. Clear resolution of the relevant SWS aging-related safety issues will

require the specification of additional failure data documentation and

regulatory requirements to ensure adequate system safety margin under

aged or extended-life conditions.

viii
CONTENTS

ABSTRACT . . . . . . . . . . . . . . .

ACKNOWLEDGMENT . . . . . . . . . . . .

SUMMARY . . . . . . . .. . . . . .

1.0 INTRODUCTION . . . . . . . . . .

1.1 NPAR PROGRAM GOALS . . . . .

1.2 REGULATORY PERSPECTIVE . . .

2.0 SYSTEM DEFINITION AND DESCRIPTION

2.1 OPEN SYSTEMS . . . . . . . .

2.2 OPEN RECIRCULATING SYSTEMS

2.3 CLOSED SYSTEMS . . .

2.4 COMPONENT DESCRIPTION . . .

3.0 PLANT OPERATIONS EXPERIENCE

i11

. . . o .. . . . . . . . . .

I . . . . . . . . 6

. . . . . . 0 . .

* . . . .

* . . * .

* . . ,. -..

. . . . . . . . . 1. . 0 . . .

3.1 THE IMPORTANCE OF INFORMATION ACCURACY

3.2 GENERIC DATA BASES . . . . . . . .

3.2.1 Generic Data Bases . . . . . . .

3.2.2 IEEE Standard 500 . . . . . . . .

3.3 COMPONENT HISTORY DATA . . . . . .

3.4 OTHER SOURCES OF INFORMATION . . . . . .

4.0 INTERIM AGING DEGRADATION ASSESSMENT . . . .

4.1 USES OF PLANT INFORMATION . . . . . . .

4.1.1 In-Plant Component Failure Follow

4.1.2 SWS Plant Aging Analysis . . . .

4.1.3 Probabilistic Risk Assessments

4.2 EVALUATION OF PLANT DATA . . . . . .

4.2.1 Failure Classification . . .

. . . . . . . .

v

vii

1.1

1.1

1.2

2.1

2.3

2.3

2.5

2.5

.3.1

3.1

3.1

3.2

3.4

3.3

3.4

4.1

4.1

4.1

4.1

4.2

4.5

4.5

... . . . . . ..

. . . . . . ..

. . . . . . .

. . . . . . . .

. . . . . . . .

ix
4.2.2 Aging Assessment Based on Plant Data . . . . . . . 4.9

4.2.3 PRA Input Illustration . . . . . . . . . . . . . . 4.10

5.0 CONCLUSIONS AND RECOMMENDATIONS ...... . . . . . . . . . . 5.1

5.1 CHANGES RECOMMENDED FOR COMPUTATION OF PRA INPUTS . . . . 5.1

5.2 RECOMMENDED PLANT RECORD CHANGES FOR AGING ASSESSMENT . 5.2

5.3 CONCLUSIONS FROM PLANT DATA . . . . . . . . . . . . . . . 5.2

REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . R.1

APPENDIX A - ROOT CAUSE SCHEME . . . . . . . . . . . . . . . . . . . A.1

APPENDIX B - SERVICE WATER SYSTEM MATERIALS AND

ASSOCIATED CORROSION MECHANISMS . . . . . . . . . . . . B.1

x
FIGURES

1.1

2.1

2.2

2.3

2.4

4.1

4.2

4.3

A.1

NPAR Program Strategy. . . . . . . . . . . . . .

Functional Services Water System Boundary Definition

Open Service Water System . . . . . . . . . . . . . .

Closed Service Water System . . . . . . . . . . . . .

Service-Water Intake Structure . . . . . . . . . . .

A Time-Line Illustration of a Single Failure . .

Sedimentary Corrosion on 18-in. Gate Valve . . . . .

Through Wall Pitting in 3-in. Pipe . . . . . . . . .

Basic Failure Analysis Scheme . . . . . . . . . . . .

. . . .

1.3

2.2

2.4

2.6

2.10

4.4

4.12

4.13

A.5

. . . . .

xi
TABLES

2.1 Listing of All Active U.S. Nuclear Commercial

Reactors, Their Electrical Power Rating, Source

of Water Used for a Heat Sink, and Specific

SWS Configuration ..... . . . . . . . . . . . . . . . . . . 2.7

3.1 Information Contained in Generic Data Base and

Plant Maintenance Records, by Event . . . . . ... . . . . . . . 3.4

4.1 Current Status of Component Failure Analysis

Knowledge Requirements . . . . . . . . . . . . . . . . . . . . 4.4

4.2 Summary of Plant Maintenance Data from an

Aging Perspective ....... . . .. . .. . . .. . . .. . 4.6

4.3 Twenty-One Month Summary of Maintenance Records . . . . . . . . 4.8

4.4 Logic Tree Results of First Analysis . . . . . . . . . . . . . 4.11

4.5 Logic Tree Results of Second Analysis . . . . . . . . . . . . . 4.11

4.6 Failure Mechanisms Identified in Second

Analysis of Plant Data . . . . . . . . . . . . . . . . . . . . 4.14

4.7 Failures per Item per Million Hours . . . . . . . . . . . . . . 4.15

5.1 Summary of Suggested Changes in Plant

Recordkeeping System ...... . . . .. . . . . . . . .. . 5.3

B.1 Observed Damage Mechanisms . . . . . . . . . . . . . . . . . . B.2

B.2 Potential Corrosion Factors ..... . . . . . . . . . . . . . B.3

xii
1.0 INTRODUCTION

The Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory

Research (RES), Division of Engineering (DE), is implementing the Nuclear

Plant Aging Research (NPAR) Program plan (1) to resolve technical safety

issues related to the aging of commercial nuclear power plants. Aging in the

context of the NPAR plan is defined as follows:

The cumulative changes with the passage of time that may occur within a

component or structure due to one or more of the following factors:

- natural processes during operation - external stressors caused by storage or operation - service wear caused by operational cycling - excessive testing - improper installation, application, operation or maintenance.

The specific topic of inyltigation in this task, performed by the

Pacific Northwest Laboratory, is the safety-related portions of the

nuclear plant service water system (SWS). During a loss of coolant accident

(LOCA), or similar core threatening postulated accident scenario, this system

is relied on to transfer heat from vital plant equipment, such as the residual heat removal (RHR) heat exchangers and the emergency diesel generators

(EDGs), to the ultimate heat sink.

Emphasis in this investigation has been placed on identification and

characterization of the mechanisms of material and component degradation

during service and the evaluation of methods of inspection, surveillance,

condition monitoring, and maintenance as a means of mitigating these effects.

1.1 NPAR PROGRAM GOALS

The specific goals of the NPAR program are

1. to identify and characterize the primary aging mechanism(s) which could

cause safety-related component degradation

2. to identify methods of inspection, surveillance, and monitoring which

will ensure timely detection of significant aging effects before loss of

safety function

3. to evaluate the effectiveness of storage, maintenance and replacement

practices in mitigating the rate and extent of aging degradation.

(a) Operated for the U.S. Department of Energy by Battelle Memorial

Institute.

1.1
The SWS Aging Assessment utilizes the NPAR phased approach to system

research shown in Figure 1.1. The shading in this diagram suggests the

degree to which the task initiative have been investigated. This report

describes the results of Phase I for the SWS task and includes information

on

* available subject literature

* available generic data-bases

e utility machinery history

* utility system expert inputs

* commercial expertise.

This information is analyzed and evaluated to identify the principal SWS

aging mechanisms that will direct the focus of the remainder of the task

goals.

1.2 REGULATORY PERSPECTIVE

The impetus for this study originates from the significant number of

documented SWS degradation-related events that have seriously impaired the

ability of this system to carry out its intended safety functions (2,3,4).

As previously stated, the safety function is vital to the successful termination of many potential core-melt scenarios. The cited references illustrate

the potential for a partial or complete loss of SWS operability from a common

failure mode. Additionally, fouling by either mud, silt, corrosion products,

*or aquatic bivalves has led to plant shutdowns and reduced power operation

for repairs of modification as well as degraded modes of operation (5,6).

Service water system regulatory concerns are prioritized in reference (7) in

the form of Generic Safety Issues. Suggested solutions to service-waterrelated Generic Safety Issues, and the regulatory format for their implementation, are the final phase objective of this study.

1.2
IIIIIII

Conslealn

for

Life

Extensions

Phase 11

Conprehensive Aging

Assessment and

Recommendations for

1,S aid M (A) IIIII

Phase I

a

- Post Serice

Eamninations

* Lab Taes ad

Analyses;

* Efort Continued

In Phase 11

Phase 11

38905049.1 M

FIGURE 1.1. NPAR Program Strategy
2.0 SYSTEM DEFINITION AND DESCRIPTION

The function of the service water system is to transfer the heat loads

from various sources in the plant to the ultimate heat sink. The three

safety-related heat sources served by this system are identified as

* core decay heat

* decay heat removal components

* emergency power sources.

Because of the wide variation in the nature of each plant's ultimate

heat sink and the application of a multiplicity of system design approaches,

the system is defined from a functional standpoint as follows: all components, their associated instrumentation, controls, electrical power, cooling

and seal water, and lubrication, and other auxiliary equipment which comprises the final heat transfer loop between the heat sources and the ultimate

heat sink.

A overall perspective of this functional definition is given by examining the elementary diagram of a SWS shown in Figure 2.1. The dashed boundary

shows the range of components considered by this study: -

* intake structure including canals or other diversion structures from the

ultimate heat sink to the pump debris removal mechanism

* the pump galley and structures with all associated water-level control

devices (weirs, gates, valving, etc.) and instrumentation

* the service-water pump, shafting and motive source including controls,

cabling and electrical distribution system

* the piping distribution network, from the pumps to the heat exchangers,

including all valving, manifolds, instrumentation, and logic networks

* the service (secondary or cooling) water side of a actual heat exchange

devices (primary or cooled fluid-side aging is treated by other NPAR

system studies)

* all discharge piping, valves, and manifolds from the heat exchangers to

the outlet or discharge structure

* the discharge structure, gates and associated effluent channeling

devices.

Only those components which are designated important to reactor safety

(nuclear safety class 3, see ref 8) and designated Seismic Category I are

examined in this investigation. Seismic Category I requires that plant

structures, systems, and components be designed to withstand a design basis

earthquake.

Safety-related service-water cooling loops in commercial nuclear

reactors are also designed to meet the single failure criterion. That is,

2.1
-I INDICATION AND

CONTROL SIGNALS

ELECTRIC . POWER

FIGURE 2. 1. Functional Services Water System Boundary Definition
redundant components are provided such that the failure of any single active

component in an SWS will not prohibit the adequate removal of heat from any

of the safety-related loads.

A somewhat more detailed discussion of practical SWS layouts follows,

illustrating three basic configurations, usually selected based on the nature

of the available ultimate heat sink. These descriptions are given to provide

a broad knowledge of the design and function of SWS components and do not

represent any specific BWR or PWR design.

2.1 OPEN SYSTEMS

A diagram of a typical open SWS is shown in Figure 2.2. This type of

arrangement is often referred to as a "straight through" system and is generally characterized by the availability of a large volume of water as the

ultimate heat sink. The obvious advantage of this configuration is its

relative simplicity and resultant lower initial cost to the utility. The

absence of a large capacity intermediate heat exchanger and a requirement for

a secondary set of component cooling water pumps, as found in the closed

cycle configuration, make this layout attractive for both its economy of

installation and theoretical component maintenance cost. The major offset

here is the potential for problems associated with the exposure of a large

number of components to a potentially aggressive raw water environment.

Many of the component cooling subsystems are throttled to maintain

required component temperature limits, resulting in low-flow velocities or,

in many cases, in intermittently used components, resulting in completely

stagnant loops. This allows solids deposition and various forms of corrosion

to accelerate in these areas (these considerations are addressed in

Appendix B).

2.2 OPEN RECIRCULATING SYSTEMS

A variation of the open system is the recirculating type which maintains

a self-contained ultimate heat sink. This is frequently achieved through the

use of a spray cooling pond or a dedicated cooling tower. The advantage of

this arrangement is two-fold: 1) through settling the make-up water filtration, the water purity (turbidity) is vastly improved leading to significantly reduced siltation in low-low velocity areas, and 2) chemical control

of the circulated water is achievable allowing a reduction in corrosion and

biofouling without the limitations imposed by environmental discharge

restrictions.

2.3
River

Source

(UHS) Emergency

Feed Water

FIGURE 2.2. Open Service Water System
2.3 CLOSED SYSTEMS

The basic difference between an open and a closed system is that, in a

closed system, plant personnel can control the coolant chemistry which comes

with contact with the system load heat exchangers, whereas in an open system

they cannot. This type of system is frequently used when adverse environmental conditions (saline or other corrosive) are prevalent.

The closed system (shown in Figure 2.3) is obviously more expensive

initially and contains a larger number of components, but it is not as susceptible to premature aging through extensive corrosion attack within the

secondary loop.

A compilation of all active commercial reactors, their electrical power

rating, source of water used for an ultimate heat sink, and specific SWS

configuration is given in Table 2.1. This listing is primarily sorted by SWS

cycle type, with plant names listed in alphabetical order for each of the

types. Again, the distinction between open and closed systems is the practicality of maintaining control of cooling water chemistry, thus an CLOSED

POND system could be controlled via additive chemicals, whereas a OPEN POND

is considered to be too large for effective control. The information listed

was extracted from Final Safety Analysis Reports (FSARs) and may be updated

by plant modifications.

2.4 COMPONENT DESCRIPTION

While arrangements and materials utilized in construction differ widely

among plants, the design of each functional area of all SWSs is basically

the same. The following discussion is generic and does not represent any

specific plant.

The intake structure serves to admit water to the SWS, provides for the

necessary degree of debris removal, and houses the service-water pumps and

their associated switchgear. For the open system shown in Figure 2.2, this

basically consists of identical bays which house bar gates, stop logs,

traveling screens, screen wash pumps, chlorine injection equipment, and the

circulating and service-water pumps. An elevation drawing showing a typical

arrangement of these components is shown in Figure 2.4. A separate

ventilation system for the service-water pump area is designated class I to

ensure that the equipment remains operable during a design basis accident and

is capable of functioning after a design basis earthquake. The service-water

pumps and their associated bays are also designated class I seismic

equipment.

The parallel traveling screens function to remove small debris that

penetrate the larger (-3 in. opening) the bar gate strainers. The screens

are cleaned by an automatic high-pressure spray system which senses

differential pressure across the screens.

2.5
Pur

Mol

Ocean

Source

(UHS)

Alps H;HI

1toLrI Lli Aid

QT C', T (!Yj .

Underground

_ ______

-F----- .

.4*- - _ --- r-- . r

I ¶3-

I I I

I I

I I

I I

Underground I 13

Iz -

II

4-

4-

4-~

FIGURE 2.3. Closed Service Water System
TABLE 2.1. Listing of All Active U.S. Nuclear Commercial

Electrical Power Rating, Source of Water Used

and Specific SWS Configuration

Reactors, Their

for a Heat Sink,

Reactor/

Rating

Ultimate

Heat Sink

Service

Water

Plant Name Parent Utilitv CYcl e

Clvrt Clfs-1,2

Cook-I

Cook-2

Fermi-2

Fort St Vran-1

Hope Creek

Limerick-1,2

Maine Yankee

Pilgrim-1

Rancho Seco-I

River Bend-i

Salem-1

Salem-2

San Onofre-1

San Onofre-2

San Onofre-3

Shoreham

St Lucie-1,2

Waterford-3

Diablo Canyn-I

Diablo Canyn-2

Palo VR-1,2,3

WNP-2

Byron-1,2

Vogtle-1(2)

S TX Proj-1,2

Big Rock Point

Catawba-1,2

Clinton-1

Davis-Besse-1

Dresden-2,3

Fitzpatrick

Ginna

Kewaunee

Lasalle-1,2

McGuire-1,2

Nine Mile-1

Nine Mile-2

North Anna-1,2

Oconee-1,2,3

Palisades

Perry-1,2

(CE-845)

(W-1030)

(W-1090)

(GE-1093)

(DA-330)

(GE-1067)

(GE-1055)

(CE-790)

(GE-655)

(B&W-916)

(GE-940)

(W-1090)

(W-1115)

(W-436)

(CE- 1070)

(CE-1080)

(GE-819)

(CE-810)

(CE-1165)

(W-1084)

(W-1106)

(CE-1270)

(GE-1100)

(W-1120)

(W-1125)

(W-1250)

(GE-71)

(W-1145)

(GE-955)

(B&W-880)

(GE-794)

(GE-821)

(W-470)

(W-535)

(GE-1078)

(W-1180)

(GE-610)

(GE-1080)

(W-890)

(G&W-860)

(CE-798)

(GE-1205)

Baltimore Gas & Elec

Indiana & Mich Elec

Indiana & Mich Elec

Detroit Edison Co

Colorado Public Serv

Public Service Elect

Philadelphia Elec Co

Maine Yankee Atomic

Boston Edison Co

Sacramento Muni Util

Gulf States Util

Public Serv Elec NJ

Public Serv Elec NJ

Southern CA Edison

Southern CA Edison

Southern CA Edison

Long Island Lighting

Florida Pwr & Light

Louisiana Pwr and Li

Pacific Gas & Elec

Pacific Gas & Elec

Arizona Public Serv

WA Pub Pwr Supply

Commonwealth Edison

Georgia Power Co

S TX Proj Nuc Gen

Consumers Power Co

Duke Power Co

Illinois Power Co

Toledo Edison Co

Commonwealth Edison

Pwr Auth State of NY

Rochester Gas & Elec

Wisconsin Pub Servic

Commonwealth Edison

Duke Power Co

Niagara Mohawk Power

Niagara Mohawk Power

Virginia Elec & Pwr

Duke Power Co

Consumers Power Co

Cleveland Elec Illum

Chesapeake Bay

Lake Michigan

Lake Michigan

Hyperbol/Erie

FD TWR/(Mech)

Hyperbol/Delawar

Hyperbol/Schukil1

Back R/(Atl 0)

Cape Cod Bay

Hyperbol/Canal

FD TWR/Miss R

Delaware River

Delaware River

Pacific Ocean

Pacific Ocean

Pacific Ocean

L I Sound

Atlantic Ocean

Mississippi River

Pacific Ocean

Pacific Ocean

Basin Filter-Well

FD TWR/Columbia

Hyperbol/Rock

FD TWR/Savna R

Reser/Colorado

Lake Michigan

FD TWR/(Mech)

Lake Clinton

Tower/L Erie

Cooling Lake

Lake Ontario

Lake Ontario

Lake Michigan

Reservoir

Reservoir

Lake Ontario

Hyperbol/Lake M/U

Reservoir (Anna)

Reservoir

FD TWR/Michigan

Hyperbol/L. Erie

Closed

Closed

Closed

Closed

Closed

Closed

Closed

Closed

Closed

Closed

Closed

Closed

Closed

Closed

Closed

Closed

Closed

Closed

Closed

Closed

Closed

Closed Pond

'Closed Pond

Closed River

Closed Well

Open/Closed

Open Lake

Open Lake

Open Lake

Open Lake

Open Lake

Open Lake

Open Lake

Open Lake

Open Lake

Open Lake

Open Lake

Open Lake

Open Lake

Open Lake

Open Lake

Open Lake

2.7
TABLE 2.1. (contd)

Plant Name

Reactor/

Rating Parent Utility

Ultimate

Heat Sink

Service

Water

Cycle

Point Beach-1,2

Robinson-2

Summer

Wolf Creek

Zion-1,2

Brunswick-1,2

Crystl River-3

Millstone-I

Millstone-2

Millstone-3

Oyster Creek

Seabrook-1

Turkey Pnt-3,4

Callaway-1

Sharon Harris-1

Yankee-Rowe

ANO-1

ANO-2

Arnold

Beaver Valy-1,2

Belfont-1(2)

Braidwood-1,2

Brwns F-1,2,3

Coman Peak-1,2

Cooper

Ct Yanke

Farley-1,2

Ft Calhoun-1

Grnd Glf-1(2)

Hatch-1,2

Indian Point-2

Indian Point-3

La Crosse

Monticello

Peach Bot-2,3

Prairie I.-1,2

Quad City-1,2

Sequoyah-1,2

Surry-1,2

Susquehana-1,2

TMI-1 ,2

(W-497)

(W-665)

(W-900)

(W-1150)

(W-1040)

(GE-790)

(B&W-825)

(GE-660)

(CE-830)

(W-1150)

(GE-620)

(W-1150)

(W-728)

(GE-1140)

(W-900)

(W-175)

(B&W-836)

(CE-858)

(GE-538)

(W-833)

(B&W-1213)

(W-1120)

(GE-1065)

(W-11i1)

(GE-778)

(W-582)

(W-860)

(CE-478)

(GE-1250)

(GE-770)

(W-873)

(W-965)

(GE-48)

(GE-536)

(GE-1065)

(W-507)

(GE-789)

(W-1148)

(W-775)

(GE-1050)

(B&W-792)

Wisconsin Elec Pwr

Carolina Power & Lig

S Carolina Electric

Kansas Gas and Elec

Commonwealth Edison

Carolina Pwr & Light

Florida Power Corp

Northeast Utilities

Northeast Utilities

Northeast Utilities

Jersey Central Pwr

Public Serv Co of NH

Florida Pwr & Light

Union Electric Co

Carolina Power & Lig

Yankee Atomic Electr

Arkansas Pwr & Light

Arkansas Pwr & Light

Iowa Elec Light

Duq4esne Light Co

Tennessee Valley Ath

Commonwealth Edison

Tennessee Valley Ath

Texas Util Gen Co

Nebraska Public Pwr

Ct Yankee Atomic Pwr

Alabama Power Co

Omaha Public Pwr Dis

Mississippi Pwr & Li

Georgia Power Co

Con Ed Co of NY

Pwr Auth State of NY

Dairyland Pwr Coop

Northern States Pwr

Philadelphia Elec Co

Northern States Pwr

Commonwealth Edison

Tennessee Valley Ath

Virginia Elec & Pwr

Pennsylvania Pwr & L

Metropolitan Edison

Lake Michigan

Reser/Robinson

Reser/Monticello

Cooling Lake

Lake Michigan

Atl 0 Outfall

Gulf of Mexico

L I Sound

L I Sound

L I Sound

Barnegat Bay

Atlantic Ocean

Canal/Byscane Bay

Hyperb/Missouri R.

Hyperbol/Reser

Deerfield River

Reservoir, Ark R

Hyprbol/River M/U

FD TWR/(Mech)

Hyprbol/Ohio River

Hyperbol/Tenn Riv

Reser/Kankakee R

Combcycle/Tenn Rv

Reser/River

Missouri River

Connecticut River

FD TWR/(Mech)

Missouri River

Hyperbol/Miss R

FD TWR/(Mech)

Hudson River

Hudson River

Mississippi River

FD TWR/Miss R

FD TWR/Susq R

FD TWR/(Mech)

Spray Canal

Comb Cycle/Tenn

James River

Hyperbol/Susquhan

Hyperbol/Susqu

Open

Open

Open

Open

Open

Open

Open

Open

Open

Open

Open

Open

Open

Open

Open

Open

Open

Open

Open

Open

Open

Open

Open

Open

Open

Open

Open

Open

Open

Open

Open

Open

Open

Open

Open

Open

Open

Open

Open

Lake

Lake

Lake

Lake

Lake

Ocean

Ocean

Ocean

Ocean

Ocean

Ocean

Ocean

Ocean

Pond

Pond

Pond

River

River

River

River

River

River

River

River

River

River

River

River

River

River

River

River

River

River

River

River

River

River

River

Open River

Open River

2.8
TABLE 2.1. (contd)

Service

Reactor/ Ultimate Water

Plant Name Rating Parent Utility Heat Sink Cycle

Trojan (W-1130) Portland General Ele Hyperbol/Columbia Open River

Vermont Yankee (GE-514) Vt Yankee Nuc Pwr Co FD TWR/Conn R Open River

Watts Bar-1,2 (W-1170) Tennessee Valley Ath Hyperbol/Tenn Riv Open River

Hyperbol = Hyperbolic Cooling Tower.

FD TWR = Forced Draft Cooling Tower.

R = River.

Reser = Reservoir

0 = Ocean.

Three service-water pumps are typically utilized to provide the necessary head for flow requirements of the various heat exchangers in the system.

Normal operation of the system is with two of the three SWS pumps in operation supplying all steady-state cooling requirements and the third pump in a

standby condition (one out of two and three out of four is also common).

Pump motor power ranges from approximately 200- to 600-hp and flow from 4,000

to 10,000 gal/min per pump depending on configuration and plant size. Pumps.

are powered from plant vital AC busses (with diesel backup) to ensure a continuous electrical supply. Pump motor winding temperature, current, discharge pressure, and bearing temperature instrumentation is generally available at local and remote readout panels.

Common practice is to pass the water through a discharge strainer following the pump to ensure that any remaining particles are small enough so

they are not capable of plugging the smallest heat exchanger tube in the

system. These strainers are monitored for excess differential pressure which

could indicate a plugged condition. The swing check valves located just

downstream of the strainers are designed to prevent reverse flow through the

idle pump and strainer.

The crosstie valves located on the discharge header permit full operation of both service-water loops with any two SWS pumps running. These are

motor operated valves powered from the plant vital bus. It is a common

practice to provide for emergency makeup to the reactor vessel or steam

generators from the SWS because it is engineered as a high-reliability watersource. Redundant headers then provide the necessary flow distribution network to the individual cooling loads. The discharge header then collects

the system flows and directs them back through the cooling water flume. Some

arrangements allow makeup to the emergency cooling water reservoir from the

discharge header as illustrated in Figure 2.4.

2.9

J
Top View

TRAVELING SCREENS

RACKS

11

WATER FLOW

TO MAIN

TRASH RAKE

TRAVELING

HOIST CAR

NORMAL WATER

RACK

CIRC WATER PUMPS SERVICE WATER PUMPS

TUNNEL

Side View

FIGURE 2.4. Service-Water Intake Structure

2.10
The emergency cooling water source (river, cooling pond, ocean, etc.)

may take various forms depending on local conditions, but it must meet the

requirements of Reg. Guide 1.27 (9) in capacity, availability, and accessibility under seismic conditions.

2.11
3.0 PLANT OPERATIONS EXPERIENCE

The first step in conducting phase one of the NPAR plan on SWSs was to

investigate the extent and accuracy of the generic data bases already in

existence for the topic at hand. The generic data analysis results did not

appear reasonable in light of the investigators' previous plant experience.

Subsequently, a broader search for information led the researcher to specific

plant data bases and final to the system expert.

This section examines the importance of accuracy in aging data, the

utility of the various data sources, the impact of detailed machinery history

data, and other sources of relevant information and expertise.

3.1 THE IMPORTANCE OF INFORMATION ACCURACY

Considering the breadth of the potential aging stressors in SWSs, it is

vital that the available research resources be focused somewhat narrowly on

only those key areas where a high return can logically be expected. To

accomplish this, researchers have examined the data available to provide some

guidance as to what is and is not, of vital importance to a specific examination task. It is important for the investigator to understand the intended

purpose, features, and limitations of the data base being considered.

Due to the operational background of the investigators in this task, a

logical expectation based on experience in the plant as to the principal

aging degradation mechanisms for the SWS served as a check on the initial

data base conclusions. The resulting search for more accurate and definitive

data resulted in a complete reorientation of the task focus.

It is extremely important that this and other NPAR investigations (10)

avoid being mislead by incomplete or inaccurate data. Since the end product

of the NPAR Program revolves around aging and license renewal regulatory

issues, there appears to be a real need for technical specifications, surveillances, and other necessary licensee requirements aimed at resolving the

primary aging issues.

3.2 GENERIC DATA BASES

Component unavailabilities computed from generic data bases may be used

for input to a PRA configured for a specific plant. Resulting risk levels

could then be examined to see how changes in certain component unavailability

rates affect overall plant risk. This would identify key components that

have the potential of significantly increasing plant risk, if their

unavailabilities increase due to aging.

3.1
If the level of plant risk calculated using unavailability values from

generic data bases is acceptable, and if it can be assumed that the actual

component unavailabilities of a plant's components are at or below the corresponding unavailabilities computed from the generic data base, then the

actual plant risk is acceptable.

This approach allows for the impact of high unavailability of a specific

component to be assessed. If a specific component at the plant has an

unusually high unavailability or failure rate, overall plant risk may be

calculated using unavailabilities from the generic data bases and the one

specific unavailability from the plant. It can then be determined whether

this increased unavailability raises the plant's risk significantly.

Use of data from generic data bases in PRA calculations results in a

high level of uncertainty in calculated plant risk. There are several

reasons for this:

1. None of the data bases contain a comprehensive listing of all failures

that have occurred.

2. Some are composed of random failure histories, while others only list

failures which have violated certain specifications.

3. The failures which are listed often have incomplete information.

3.2.1 Generic Data Bases

Common generic data bases investigated during the SWS assessment include

the following:

* NPE - Nuclear Plant Experiences (11) - A quarterly updated indexed

system containing over 50,000 BWR and PWR events.

* LER - Licensee Event Report system (12) - Basically a compilation of

nuclear plant Technical Specification violations; an NRC data system.

* NPRDS - Nuclear Plant Reliability Data System (13) - An industry-wide

voluntary system for monitoring performance of selected systems and

components, currently under the direction of INPO.

Comparisons of these data bases have been conducted elsewhere (14,15, 16).

Several key conclusions can be made from these references:

1. LERs do not contain sufficient information to support aging data

analysis.

2. While corrective maintenance records contain virtually all of the component failure occurrences, generic data bases contain only 10 to 30% of

that information.

3.2
3. Little or no root cause documentation is found in generic data bases.

This results in no clear definition of aging events.

3.2.2 IEEE Standard 500

Mechanical unavailabilities given in IEEE Standard 500 (17) may be used

in the same way as unavailabilities computed from generic data bases. The

unavailabilities in this standard are computed using the Delphi method, a

technique for collecting and summarizing reliability information. The method

uses information and data from classical statistical data bases whenever

possible, and supplements this with estimates provided by plant experts. The

experts are allowed to use any information they deem useful in arriving at

their estimates. Data from different sources is weighted so that more

detailed data are then sent back to the experts, giving the experts a chance

to modify their original estimates, or reaffirm them. These results are

again collected and synthesized. This process was used incorporating data

summaries of Licensee Event Reports, NPRDS data, IPRDS data, input from over

200 experts, and several other sources of unavailability data.

Because approximately 80% of the data listed in the IEEE Standard 500 is

taken from available data bases, including the generic data bases discussed

previously component unavailabilities obtained from the IEEE Standard 500

have relatively high uncertainties. Combination of the different data bases

should weaken biases of specific data bases, although it cannot get rid of

them completely. One advantage of this system is that the source of each

unavailability is given. For instance, if an unavailability is based completely on expert opinion, this fact is stated. If the unavailability is

based on information from three generic data bases, these data bases are

referenced.

3.3 COMPONENT HISTORY DATA

The use of a computerized component history data base (plant specific

corrective maintenance record file) has the greatest potential to provide the

most complete picture of age-related documentation available. While the

plants of primary interest, i.e., older plants, may not have records of sufficient detail or quality to allow time-related degradation to be derived,

the information points out degradation mechanisms with considerably more

reliability than does other more accessible data.

The following paragraphs illustrate and contrast the differences

observed between corrective maintenance records and the most commonly used

aging-study data base.

The number of event entries during the 21-month period of available

plant data (18) are shown in Table 3.1. This time interval was selected

because of the availability of computerized plant information. Prior information resides exclusively on microfilm and is extremely difficult to

extract. While the studied system had no reportable LERs during the time

3.3
TABLE 3.1. Information Contained in Generic Data Base and

Plant Maintenance Records, by Event

Number of Entries During 21-Month Period

Event Categorv Plant Generic Data Base

Functional Failure 1 113 19

Surveillance or 2 79 23

Inspection, Problem Found

Surveillance or 3 64 0

Inspection, No Problem

Found

Work Done in Support of Not Included 98 0

One of Above in Analysis

Work Done; Reason Not 4 63 0

Documented

Cancelled Not Included 32 0

in Analysis

Record Unclear 5 5 0

TOTAL 454 42

interval, the generic data base contained approximately 25% of the problems

discovered through surveillances and only 10% of the total functional failure

events reported in the maintenance history.

The circumstances surrounding the submittal of events were found to have

a great impact on the number and completeness of the reported events.

Generic data-base report frequency and, consequently, the apparent failure

frequency increased dramatically in 1985. Taking the increased failure rate

as an indicator of significant aging degradation would have been misleading;

in reviewing the corrective maintenance logs, the failure rate was seen to be

essentially constant. The actual reason for the increase was the plant's

switch to computerized recordkeeping which included the desired generic data

base formatted fields; this allowed entries to be submitted to the central

processing facility from the computer terminal in only a few minutes.

3.4 OTHER SOURCES OF INFORMATION

Several other sources of component degradation and failure information

were investigated in an attempt to secure additional information for

analysis:

3.4
* Associated Research Investigations - Several research programs (19,20)

which are directly applicable to service-water-related phenomena are

available. These programs are both NRC and industry supported and will

be used extensively in the corrosion investigation sub-task of the

phase II assessment.

* Incident Investigation Reports - These reports evaluate nonroutine plant

events or incidents that are considered to have a potential impact on

reactor safety, either through component failure or a more implicit

scenario (managerial breakdown). This evaluation may take the form of a

fairly detailed root cause analysis and can therefore shed more light on

occurrences than generic data. These reports are not, however, designed

to provide information specific to aging research and cover only a

small fraction of the plant failure inventory. For these reasons, their

use of this task is quite limited.

* Subject Matter Experts - Experts in the subject of service water systems

are considered to be those individuals who, through intimate association

with the design, operation, or modification of system functions, have

become uniquely qualified to make technical or intuitive judgments

regarding the causes, results, and mechanisms of component degradation.

Two categories of this type of expertise were explored for insights into

further understanding of system phenomena.

- Utility Operations and Maintenance Personnel - The level of understanding that comes from learning system function and design, and

then from actually living with the results of day-to-day operation

of that system, cannot be overemphasized. A great deal can be

gained from the collective knowledge of the operators, maintenance

personnel, and engineers whose efforts to maintain the system in a

high state of operational performance gives them a substantial

backlog of practical experience.

In an effort to extract maximum benefit from out interviews with

plant personnel, a questionnaire protocol was developed to record

the full extent of previously undocumented plant knowledge. These

sessions, lasting up to six hours, included key plant personnel--

design engineering, operating supervision, equipment operators,

chemists, maintenance foremen, and maintenance personnel. The

information obtained during these sessions gave considerably deeper

insights into failure occurrences than other sources in that they

took advantage of the inherent root cause analysis performed by

plant personnel. The dramatically differing results of attempting

an even moderately rigorous root cause investigation with and

without this information is illustrated in.Section 3.

- Industry Consultants - The other source of special knowledge is

those consulting personnel who specialize in narrower aspects of

the aging problem. Due to the broad perspective of PNL, we were

able to consult with both government (21) and private sector

representatives. This consultation took the form of managerial

3.5
guidance, based on previously successful projects of a similar

nature and more subject oriented workshops aimed at broadening-the

knowledge base of the NPAR group at PNL. This type of interdisciplinary expertise, both from within and outside the laboratory,

will be used extensively during the second phase of this project.

3.6
4.0 INTERIM AGING DEGRADATION ASSESSMENT

This section of the report examines current documentation; analyses

content and use of component failure information; outlines the approach taken

to analyze and interpolate the data; and presents the interrupted aging

assessment of SWSs based on limited Phase-I data, with a summary of conclusions based on that assessment.

4.1 USES OF PLANT INFORMATION

Because of the expense of data acquisition and the perception that SWS

performance and material condition change at a relatively slow and constant

rate, the histories of components are not well documented compared to what is

considered to be needed for a "safety prominent" system. With the emergence

of service water as the "Achilles's heel" of accident scenario heat transfer

to the ultimate heat sink, added attention is being given to measurement of

critical system parameters. This sub-section examines the uses of servicewater failure information with an eye to better defining future data

requirements.

4.1.1 In-Plant Component Failure Follow

All commercial power plants, in one way or another, have a system for

analyzing plant component performance. From overall plant heat rate to the

establishment of component preventive maintenance frequency requirements,

component failure data plays a formative role in plant operation and maintenance. Nuclear plants generally take, this formulation a step further in

that the requirements of a regulated industry demand a safety as well as a

cost perspective. Component failure data is scrutinized not only for potential large capital drain and repeat offenders but is also subject to license

demands embodied in the plant technical specifications. All operating plants

meet these demands through carefully constructed operating procedures and

periodic surveillance procedures, which provide instruction for implementing

the Technical Specification requirements.

Increased component failure reporting and analysis has brought to light

the magnitude of the cost and safety implications brought on by aging degradation of service water components. Utility response to this problem has

largely been to mitigate these effects through corrective maintenance as

they become manifest through component failures. The cost of these repairs

has reached the point where inter-utility programs (22) have developed to

explore methods of early detection and mitigation of degradation in SWSs.

4.1.2 SWS Plant Aging Analysis

Two methods of assessing aging mechanisms on service water components

were examined. The first method analyzed specific failures using a root

cause approach and expert knowledge. This technique attempted to identify

the mechanisms responsible for each failure and then determine whether they

were age-related. The second method attempted to deduce the aging rates of

4.1
specific components from the failure rate information. The resulting timedependent failure rate could then be used in a computer model to determine

the effects of aging on PRA output values.

The first method examined seeks 1) to ascertain whether or not a specific failure is age-dependent through use of both the root cause logic tree

detailed in Appendix A, and 2) to verify this conclusions with the more

intuitive responses of plant system experts. Through use of this composite

data, root components, root parts, and failure mechanisms were identified for

all maintenance events identified as constituting a functional failure.

Specific failures and the associated failure mechanisms were then examined to

discover whether they cold be identified as age-related. Once the principal

age-related failure mechanisms are identified, they will then be examined in

greater detail in the second phase of the project. By focusing on the principal failure mechanism, the Phase-It study recommendations will yield the

maximum potential for increasing system reliability.

The second approach investigated the use of a computer software package

which examines the effects of aging on component failure rates. Although

root cause analysis is not specifically addressed, the Aging Data Analysis

for Reliability Evaluations (ADARE) (23) software package uses the data

obtained through root cause analysis of specific failures, as input. For

each failure of the component type being examined, the user must know the

failure time and whether the failure was age-dependent. (The failure time

for a specific failure is the length of time between the current failure and

most recent previous failure.) The program then computes a time-dependent

failure rate based on the assumption that the aging rate of the component is

linear.

4.1.3 Probabilistic Risk Assessments

Currently, the only method of quantitatively assessing the risk to the

public from the operation of nuclear power reactors is the Probabilistic Risk

Assessment (PRA). The quantitative aspect of the PRA provides a numerical

value for the probability that the safety systems will perform their intended

function for the environments and time period of interest. An acceptable

level of risk can then be established and used as a yardstick to compare the

operation of a given plant to the standard.

A probabilistic risk assessment is basically a statistical methodology

used to determine the probability of core melt per reactor year of operation.

The calculation of this overall core-melt probability relies on knowing the

component unavailability for each component in the given system. As shown

below, the overall component unavailability is the sum of mechanical

unavailability (fraction of time that the component was inoperable due to

failure) and test/maintenance unavailability (fraction of time the component

was inoperable due to testing and maintenance), which are calculated directly

from plant data.

4.2
QC = Qm + Qt/m (4.1)

where Qc - component unavailability

Qm = mechanical unavailability

Qt/m = test/maintenance unavailability

Mechanical unavailability is a measure of the unavailability of'a combonent caused by functional failures. It is theiresult 'o dividing the time

that a component is unavailable (inoperable), due to all failure events, by

the total operating list of the component. In equation form,

time inoperable due to all figures

Qm = total lifetime of component

This is a dimensionless fraction that is usually given in terms of days per

year of operation.

Test/maintenance unavailability accounts for the length of time a

component is unavailable for use due to testing, inspections, and/or maintenance work. An unavailability rate is obtained by dividing the total number of days the component was declared to be inoperable by the total number

of days that the component could have been operable if it had operated without any failures or testing. This unavailability rate, obtained from plant

data, is then stated in terms of days per year of operation to make it compatible with the previous term:

time inoperable due to test and maintenance 4.3

Qm records recording interval (4.3)

The total unavailability is then found by performing the addition indicated

in Equation (3.1).

A time-line illustration of a single failure, repair, test, and return.

to service sequence is shown in Figure 4.1.

Summarizing, the PRA inputs necessary to accurately evaluate risk are

the mechanical unavailability and the test/maintenance unavailability. These

two input values are calculated from failure intervals and repair and testing

unavailability periods which are calculated using plant data or the other

data sources outlined in Section 2.

Table 4.1 summarizes the knowledge level requirements for each of the

plant data uses discussed.

4.3
DISCOVERY

OF FAILURE

COMPLETE

REPAIR

BOL START

REPAIR

RETURN TO EOL

SERVICE FAILURE

SURIVEILLANCE

PERIOD OF OPERABILITY

FIGURE 4.1. A Time-Line Illustration of a Single Failure

TABLE 4.1.

Data Use

Determination

Failure Frequency

Unavailability

Root Cause

Determination

Trending Required

Component State

History

Primary Motivation

Current Status of Component Failure Analysis

Knowledge Requirements

Component Failure Probabilistic

In-Plant Risk Assessment

Yes

Usually

IntuitiveUndocumented

Usually Qualitative

No

Yes

Yes

No

No

No

Aging

Yes

Yes

Yes

Yes

Yes

Cost/Safety Safety Safety/Cost

4.4
4.2 EVALUATION OF PLANT DATA

The data used in this study for generating component unavailability

values and for determining the primary age-related failure mechanisms were

derived from the records of a single commercial PWR reactor and consisted of

324 plant maintenance records. These records document all SWS functional

failures, surveillance tests, and inspections which occurred over a 21-month

period. Relevant information was taken from these records and entered into a

database. Table 4.2 details what information was available, as well as how

complete it is, from an aging perspective.

4.2.1 Failure Classification

To begin the analysis process, each record was placed in one of the five

categories shown in Table 4.3. Category 1 designates records which indicate

that a functional failure occurred. Records indicating that a problem was

discovered during a surveillance test or an inspection are placed in category

2. Some of these "problems" are classed as a functional failure, while other

simply indicate a degraded state of the component. An attempt was made to

analyze the entries in the first two categories using a formalized root cause

logic (see Appendix A). The third category consists of records of surveillance tests for which no problems were found. Records which indicate that

work was performed, but do not indicate why the work was necessary, are

classed as category four. The final category is reserved for records which

are incomplete and cannot be placed in categories one through four. Records

of cancelled and support work were not included.

From this attempt at classification, it was found that the uncertainty

in failure rate figures based on plant data appear to occur for two principal

reasons:

1. In many cases, it was unclear from the maintenance records whether a

functional failure occurred during the 21-month data interval. Many

records state that maintenance work was carried out but do not-say why.

Other records state that work was performed after a surveillance test,

but it is not clear whether a functional failure occurred during the

test.

2. The plant records did not provide enough data points to calculate a failure rate for each distinct component. Calculations of failure rates must

be based on a time period in which a minimum of one, and better still, two

or three functional failures occur. Because some failures occur

relatively infrequently, many component types had no records of

functional failures during the 21 months examined, so a failure rate

could not be calculated for these component types.

In addition, three factors were found to be the main contributors to the

creation of uncertainty in the calculated unavailability rates:

4.5
TABLE 4.2. Summary of Plant Maintenance Data from an

Aging Perspective

Plant

Designation Information Contained Comnleteness of Information

Job Number

Component

Description

Component

Type

are

A unique number is assigned to

each maintenance record.

Describes where component is

located within the system.

Identifies type of component.

In some cases, maintenance

performed on multiple

components is listed on only

one maintenance record. In

these cases a separate database

record was created for each

component.

Frequently, this field only

indicates that the component

is located in the SWS, and no

further details are given.

No details are given. For

instance, all types of valves

simply identified as "valve."

Component

Number

Valve

TypeTa)

Individual component number.

Type of value. (Manual, air

operated, check, butterfly,

etc.)

Usually complete, although

only one number is listed when

multiple components were dealt

with. The other component

numbers can usually be found in

one of the narrative fields.

This information was obtained

from plant diagrams or

occasionally from narrative

fields.

Size Diameter of valves and pipes. Usually not given. Occasionally

found in narrative fields.

Subcomponent(a)Specific part within component

boundary.

Occasionally found in narrative

fields. Level of detail

varies. For example, some

valve maintenance records

listed "internals" as

subcomponents while others

identified a specific internal

part, such as the disk.

4.6
TABLE 4.2. (contd)

Plant

Designation Information Contained Comnleteness of Information

Functiona

Failure a]

Start Date

Stop Date

Category(b)

Specific functional failure,

if it can be determined that

a functional failure occurred.

Date that a work order was

obtained for given maintenance

event.

Date that maintenance event

ended.

Assigned based on the reason

Job order was placed and the

event outcome.

Must determine from narrative

fields. Some fields specify

that a failure occurred,

although no records indicate

that a failure did not occur.

Given most of the time.

(Approximately 95% of the time.)

Given on approximately twothirds of the records.

Assigned to all records, although

many records were difficult to

place in a category.

Coded

Fields

Fields consist of codes

identifying plant status,

effect of failure on plant,

method of discovery, etc.

Similar to NPRDS coded fields.

Listed frequently for those

records placed in categories 1

and 2. Almost never identified

for records in other categories.

Failure(c)

Description

Cause of

Failure c)

Work

Performed(c)

Used to describe maintenance

work and testing.

Used for identifying cause of

failure, problems, symptoms,

or other observations.

Work is described. Replacement

or adjustment of parts are

described

This field fails to identify

whether a failure occurred. It

is often used when no apparent

failure occurred.

Often it is not mentioned whether

a cause was identified. Other

times a cause is given when no

failure is identified.

Usually

for the

given.

given, although reasons

work are not always

(a) This information was taken from narrative fields and was not given in its

own specific field on the plant records.

(b) Categories are not assigned to maintenance events by the plant. Each

record was placed in a category to simplify analysis of the data.

(c) Narrative fields that are not used consistently, from which information

such as size, valve type, etc., were taken.

4.7
TABLE 4.3. Twenty-One Month Summary of Maintenance Records

Fraction of Total

Action Category Number Job Reports

Functional failure 1 93 0.29

Surveillance and 1ppection 2 99 0.31

(problem found)19

Surveillance and inspection 3 64 0.20

(no problem noted)

Work done (reason unknown) 4 63 0.18

Unclear(c) 5 5 0.02

(a) Maintenance records for support work or cancelled work are not

included. (no callout for this footnote)

(b) Twenty of the problems noted were identified as functional

failures.

(c) Not enough information to place in any of the previous categories.

1. Not all records list work order start and stop dates. The use of calculations based only on those records that list both start and stop dates

would clearly underestimate the actual unavailability rates. An attempt

at correction has been made by multiplying the calculated unavailability

rates by the inverse of the fraction of records which have both start

and stop dates listed. This method of correction is crude, and will be

replaced by a more accurate method in future phases of this study.

2. In some cases, components may have been unavailable before the listed

start date or after the stop date. It is unknown how long a component

was unable to function before discovery of a failure. It is also known,

for instance, that in some cases valve disks are removed from valves

with the intent of replacing the valve at the next outage. The stop

date on the maintenance record will be listed as the date that the disk

was removed, and a separate maintenance record will be started when the

valve is actually replaced. This method does not record that the valve

is actually unavailable during the time from when the disk is removed

until the valve is replaced.

3. Because specific failure mode information is not always available, an

additional uncertainty exists in that the failure mode category may be

incorrect.

4.8
In summary, the data provided by the plant was not sufficient to calculate an accurate failure rate or unavailability rate for each specific component. Improved recordkeeping and input from system experts could substantially decrease these uncertainties, producing data which would then be

suitable for use in aging or PRA input calculations.

4.2.2 Aging Assessment Based on Plant Data

It is essential, in the process of evaluating the effect of aging on the

plant, to be able to identify, with a high degree of confidence, the failure

mechanism responsible for each failure or problem and whether this mechanism

is age-related or not. Once this has been accomplished, the effects of aging

on risk may be evaluated. The techniques used to identify age-related failures and then assess the aging impact on risk are discussed in this section.

Root cause analysis (RCA) is a method for gathering information concerning a component failure and analyzing it is such a way that the uncertainty in the interpretation of the exact nature of the failure mechanism is

minimized (see Appendix A). The failure mechanisms involved in a specific

component functional failure, if sufficient information is available, can be

established by performing a root cause analysis on the information provided

on plant maintenance records. To accomplish this goal, a logic flow diagram

(logic tree) was designed to define and facilitate the process. Using only

the information given in the plant records, a strict interpretation of root

cause analysis fails to identify most root causes or failure mechanisms with

even a low degree of confidence.

Using a questionnaire protocol to sample plant personnel knowledge has

shown that those involved directly with SWS maintenance are able to identify

root cause information which the maintenance data alone could not. Experience has shown that the plant personnel are quite accurate in their diagnoses. There are two reasons that satisfactory conclusions are frequently

reached by maintenance personnel in cases where strict application of root

cause analysis has failed. One is that maintenance personnel have access to

knowledge which is not always documented on plant records. At the actual

time of the event, the maintenance personnel perform a type of informal root

cause analysis of their own to determine the circumstances of the event and

what corrective action should be taken. They do not, however, document every

piece of information that they use in deciding whether a failure has

occurred, what caused a suspected failure, or what action should be taken.

Instead, they may only document the fact that a failure occurred and corrective action was taken, leaving out the facts that enabled them to make their

decisions. Therefore, when the records are examined later, a good part of

the information needed for the root cause analysis is missing.

A second advantage that experienced personnel have is their background

experience and knowledge or previous failures. Plant personnel often have

knowledge of previous maintenance work and/or failures of the component being

examined. This may enable them to immediately rule out certain possible

causes or to associate specific symptoms with certain causes.

4.9
Plant data classified as belonging to categories 1 or 2 was examined a

second time, incorporating previously undocumented information obtained

through conversations with in-plant experts. The logic tree results from

this analysis identified far more details than the results of the first

analysis, which did not incorporate undocumented information. A summary of

the results of the two analyses are shown in Tables 4.4 and 4.5.

The results of the second analysis of the plant data show that corrosion

and material accumulation and wear are the most frequently occurring failure

mechanisms. This outcome is consistent with the beliefs of plant experts,

who suspect that many of the unidentified valve failures are also due to

corrosion. The two photographs of typical open system service water shown in

Figures 4.2 and 4.3 bear witness to this conclusion. Figure 4.1 demonstrates

the degree of accumulation of sediment (approximately 1-in. thick) on an 18-

in. gate valve removed following several years of open system service. Figure 4.2 shows a through-wall pitting attack on a 3-in. pipe coupling, also

exposed in the same raw water environment. Table 4.6 gives a numerical

perspective of the failure mechanisms identified by the logic tree in the

second analysis of the plant data.

4.2.3 PRA Input Illustration

Failure rates (number of failures/unit time) were calculated from plant

data for all component types having records of failures during the examined

21 months. These time-specific failure rates were then time scaled to obtain

a mechanical unavailability for each component type which would allow a dir-,

ect comparison with IEEE Standard 500 data. Ninety-percent confidence intervals were calculated for these mechanical unavailabilities, again for

comparison to values in IEEE Standard 500, which are also presented using 90%

confidence intervals. The Methods used in calculating the IEEE values are

discussed in Section 3.2.

Table 4.7 provides a comparison of mechanical unavailabilities calculated from plant data and IEEE Standard 500 values. Only those components

for which comparable IEEE values were also available are shown. Table 4.7

shows that for generically comparable items (e.g., pump failures and all

valve failures) the plant specific data falls close to the best estimate and

within the 90% confidence range. As would be expected, a limited components

category (strainers) deviates more from the norm but still falls within the

90% values.

This example provides a single-case illustration for preferring IEEE

data, as is done in more PRA applications, over using the existing data base

that is commonly seen in most aging analyses.

4.10
Root Part(

Root Cause

Failure ME

Failure Me

Root Cause

Failure ME

TABLE 4.4. Logic Tree Results of First Analysis(a)

Not Percent

Entry Identified Identified Identified

(b) 8 140 0.05

1 147 0.01

echanism 78 70 0.53

'chanism Investigated 13 135 0.09

e Confidence(c) 0 148 0.00

!chanism Confidence(c) 0 148 0.00

(a)

(b)

(c)

Using only recorded plant information (categories 1 and 2) and

strict root cause analysis.

Number of root components identified was not recorded during this

analysis.

Identified as high confidence or not identified as high confidence.

TABLE 4.5. Logic Tree Results of Second Analysis(a)

Not Percent

Entry Identified Identified Identified

Root Component 171 21 0.89

Root Part 152 40 0.79

Root Cause 62 120 0.32

Failure Mechanism 111 81 0.58

Failure Mechanism Identified 67 125 0.35

Root Cause Confidence(b) 61 131 0.32

Failure Mechanism Confidence(b) 65 127 0.34

(a)

(b)

Using plant records (categories 1 and 2) and undocumented knowledge

gained from conversations with plant experts.

High confidence level (>95%) in occurrence of conclusion.

4.11
. 4 I A-I lo

I ' SgL

.4

A, .e: i ,

I k.

. -- I 

.

. 0

+a. X,, *~-

', V *. '

.

9 4

b . 0

'1⁄4*ir t 4.

I 41

I14;,j

4-

M.) . .-

! }!

5V1

'I

I8 - -x RI 

-4

- SCk

?

., t, r

Vt.

>aUr,

FIGURE 4.2. Sedimentary Corrosion on 18-in. Gate Valve
w~ I

A

*#i.

I E

t~~~~~~~FGR 

4 . I.

Through Wall Pitting in 3-in. Pipe
TABLE 4.6. Failure sMehanisms Identified in Second

Analysis a of Plant Data

Number of Number with Fraction with

Mechanism

Mechanism

Entries Functional Failure Given

Corrosion

Loss of Material

Properties

Environmentally Assisted

Crack Growth

Wear

Biological or Inorganic

Accumulation

Evaporation or

Degradation of Lubricant

Cause no previously

identified (not age

related)

Not appropriate,

insufficient evidence

49 18 0.26

1 0 0.01

1 0 0.01

18 13 0.09

32 25 0.17

2

5

81

1 0.01

2 0.03

54 0.442

(a) Entries listing a functional

2) examined using logic tree

failure or other problem (categories 1 and

and undocumented information.

4.14
TABLE 4.7.

Component

Motor driven pump

Pump motor

Strainer

Valve Operator

Valve (all types)

- Check

- Manual

- Butterfly

- Globe

- Gate

Failures per Item per Million Hours

(mechanical unavailability)

Plant(a)_ 1EE Standard 500(a)

Best

Low High Low Estimate High

156 418 52.9 228.0 880.0

0(b) -- 1.E-2 0.7 4.0E3

244 552 0.3 1.2 50.E3

11 30 1.0 62.0 180.0

6 10 0.03 1.4 3.2E3

16.3 83.2 -- 3.2 --

4.8 9.4 -- 0.2 --

0.4 14.0 0.3 1.2 345

2.4 12.3 0.2 3.5 174

4.1 8.3 0.2 1.9 46.1

(a) The low and high values are given for a 90% confidence interval

around the calculated mechanical unavailability. This means that

90% of the time the actual mechanical unavailability of the

component lies within this interval.

(b) No functional failures occurred during the time period examined.

4.15
5.0 CONCLUSIONS AND RECOMMENDATIONS

This section presents conclusions and recommended actions based on the

PNL Phase-I SWS aging degradation assessment.

5.1 CHANGES RECOMMENDED FOR COMPUTATION OF PRA INPUTS

Although recognizing the additional burden that will be placed on plant

personnel to document more detailed failure information, it is essential to

the resolution of safety-system aging and plant life extension issues to

obtain expanded data which will allow a meaningful analysis. A few changes

and additions in plant recordkeeping combined with implementation of root

cause analysis at the plant will greatly reduce inherent uncertainties,

making plant data adequate for calculating PRA input values and assessing

aging mechanisms.

Sufficient fields should be provided on maintenance records to allow the

calculation of both failure rates and unavailability rates for service water

system component types. Frequently, information in unavailability fields is

not complete, not present, or the information is randomly placed in the narrative fields, making it difficult to locate. Because the narrative fields

are used randomly, information intended for a particular narrative field may

be located elsewhere on the record or may not be recorded at all. Effectiveness of plant data in calculating PRA input values could be greatly

increased by four minor changes in current recordkeeping methods:

1. Record time actual work started - It was frequently difficult to determine from records the actual length of time that the component was

unavailable for service. The start date listed is the time that the

maintenance or test was requested, not the time that the actual work

started. This means that if the component is still functioning in a

degraded state, then the unavailability time would only cover the time

during which the component was being repaired, replaced, and/or tested.

2. Record the data of last test or repair - It is often unknown how long a

component has been unavailable before discovery of the problem. This

uncertainty may be corrected if it is known when the component was last

known to be functioning properly. Therefore, the date of the last test

or repair of each component should also be recorded on maintenance

records.

3. Add a failure category field - The failure description field is not used

effectively. It does not differentiate between cases in which a failure

occurred and cases in which the component has degraded but still works.

A category field, with functional failure categories similar to those

used in this study, would make the records much easier to work with. A

category field would make it possible to work only with those records in

which a failure occurred, or with only records of surveillance tests.

This field should contain a coded response to indicate whether a failure

has occurred, whether the component is degraded but still working, or

5.1
whether no problem was found. In addition, it would be helpful to

create a category indicating tests in which an actual failure occurred.

5.2 RECOMMENDED PLANT RECORD CHANGES FOR AGING ASSESSMENT

Information on current plant maintenance records is not sufficient for

assessing aging. The undocumented detailed information used by plant personnel in reaching their conclusions about each case is needed in order to

obtain useful results from a root cause analysis. These results may then be

analyzed to reach conclusions about component aging.

There is no field on current records for indicating which part of the

component failed or degraded. Sub-parts are sometimes listed in narrative

fields, but this information is often unspecific. For instance, when a valve

fails, the maintenance report frequently states that the valve internals have

corroded, but the specific internals which corroded are not named. The specific part within the component which failed needs to be known to perform a

root cause analysis. A field for this information should be added to plant

maintenance records.

Once a system is implemented which identifies failures and classes them

as aging-related or random, aging rates and time-dependent failure rates can

be used to ascertain the effects of aging on risk. Certain details about the

components' histories are necessary for calculating these rates. Fields

should be provided for entering this information if these rates are to be

calculated in the future. This information includes the date the current

component was installed, the date of the last failure or test, and what types

of maintenance the component has received since installation.

Implementation of the modifications recommended in this section, accompanied by an effort to fully complete maintenance records would greatly

improve the usefulness of plant data. This improved data would then be adequate for computing PRA inputs with acceptable levels of uncertainty, and

for assessing the effects of specific aging mechanisms on risk and plant

life. A summary of suggested changes in the plant recordkeeping system is

shown in Table 5.1.

5.3 CONCLUSIONS FROM PLANT DATA

The information available on plant maintenance records examined during

this task is insufficient by itself for a high level of confidence in an

accurate root cause analyses or aging assessment. Improvement of recordkeeping and implementation of a root cause analysis system, however, would

make both root cause analyses and aging assessment possible and accurate.

5.2
TABLE 5.1. Summary of Suggested Changes in Plant Recordkeeping System

Information Needed

Needed

For a

Currently

Available Suggested Changes

Component number A,U,F on records Use a separate record

for each component.

Component sub-part

Unavailability start

and stop dates

Failure Category

Mechanism

Component History

A no

U

U,F,A

F,A

A

no

no

no

no

The specific part of the

component which has

failed dr degraded

should be identified.

Dates indicating length

of time component was

unavailable should be

listed.

Records should be placed

in categories similar to

those used in this

report. This would

identify which records

to use in PRA

calculations.

The cause of failure

field should specify

what mechanism was

responsible for the

failure or problem.

Fields for indicating

when component was

installed, last

maintenance or test

date, and previous

failure histories

would aid in aging

assessment and root

cause analysis.

(a) The given information is needed for: A

U - Calculation of Unavailability rate,

Failure rate.

* Aging assessment,

and F - Calculation of

5.3
Work with the plant data has produced three main conclusions:

1. Close work with plant data has shown several areas in which plant

maintenance recordkeeping needs to be improved.

2. Attempted root cause analysis of the data suggests that this type of

analysis would be most effective if it were carried out at the plant by

those who actually participated in the maintenance work.

3. Finally, the results of root cause analysis done on the plant maintenance records and corroborated by industry experts and plant personnel

suggest that corrosion, material accumulation, and wear are the three

SWS failure mechanisms that warrant further investigation.

5.4
REFERENCES

1. B. M. Norris, and J. P. Vora. 1987. Nuclear Plant Aging Research

Program Plan. NUREG-1144, U.S. Nuclear Regulatory Commission, Division

of Engineering, Office of Nuclear Regulatory Research.

2. U.S. Nuclear Regulatory Commission. 1982. Report on Service Water

System Flow Blockages by Bivalve Mollusks at Arkansas Nuclear One and

Brunswick (Nuclear Power Stations), Office for Analysis and Evaluation

of Operational Data.

3. Office for Analysis and Evaluation of Operational Data. 1983.

Evaluation Report No. AEOD/E319, Boifouling at Salem Units 1 and 2.

4. U. S. Nuclear Regulatory Commission. 1982. A Study on Service Water

Fouling, Operating Reactors Assessment Branch.

5. D. Eisenhut. 1982. Proposed Recommendations for Imoroving the

Reliability of Open Cycle Service Water Systems. Memorandum from

S. Hanauer, March 19, 1982.

6. P. Hayes. 1983. Generic Safety Issue No. 51. Improved Reliability of

Open Service Water Systems. Memorandum from W. Minners, April 5, 1983.

7. U.S. Nuclear Regulatory Commission. 1983. A Prioritization of Generic

Safety Issues, NUREG-0933, Office of Nuclear Reactor Regulation (with

updates through March 1987).

8. U.S. Nuclear Regulatory Commission. 1976. Oualitv Group

Classifications and Standards for Water-, Steam-, and Radioactive-WasteContaining Components of Nuclear Power Plants, Regulatory Guide 1.26,

Office of Standards Development.

9. U.S. Nuclear Regulatory Commission. 1976. Ultimate Heat Sink for

Nuclear Power Plants, Regulatory Guide 1.27, Office of Standards

Development.

10. S. Z. Bruske and L. C. Cadwallader. 1985. Trial Application of the

Candidate Root Cause Categorization Scheme and Preliminary Assessment of

Selected Data Bases for the Root Causes of Component Failures Program,

EGG-EA-6842.

11. EG&G, Inc., Idaho Falls, Idaho. Nuclear Power Experience. 1985-1988.

Published by S. M. Stroller Corp., 1919 14th St., Suite 500, Boulder, CO

80302.

12. U.S. Nuclear Regulatory Commission. 1986. Licensee Event Report

Compilation, NUREG/CR-2000, ORNL/NSIC-200, Office for Analysis and

Evaluation of Operational Data, December 1985 through September 1986.

R. 1
13. U.S. Nuclear Regulatory Commission. no date. Nuclear Plant Reliability

Data System. A Proprietary Institute of Nuclear Power Operations

database; portions made available through USNRC, Office of Nuclear

Regulatory Research.

14. L. C. Meyer, and P. T. Jacobs. 1987. An Evaluation of Information

Sources for Aging Research on Reactor Protection Systems, NUREG/CR-4804,

EGG-2487.

15. Murphy, G. A., R. B. Gallaher, M. L. Casada, and H. C. Hoy. 1984.

Survey of Operating Experience from LERs to Identify Aging Trends,

Status Report; NUREG/CR-3543 ORNL/NSIC-216.

16. Meale, B. M., 0. G. Satterwhite, J. L. Plum, and P. M. McGuire. 1986.

Identification of Aging Root Causes of Component Failures for Service

Water and 1E Electrical Power Distribution Systems, EGG.-REQ-7227, EG&G

Idaho, Inc., Idaho Falls, ID.

17. IEEE Standard 500. 1983. IEEE Guide to the Collection and Presentation

of Electrical. Electronic, and Sensing Component. and Mechanical

Equipment Reliability Data for Nuclear Power Generating Stations. The

Institute of Electrical and Electronics Engineers, Inc., New York.

18. D. B. Jarrell, J. Huenefeld, S. White, and P. Zimmerman. 1987. Service

Water System Site Information Visit, PNL-6502, Pacific Northwest

Laboratory, Richland, WA.

19. K. R. Hoopingarner, and J. W. Vause. 1987. Aging of Nuclear Station

Diesel Generators: Evaluation of Operating and Expert Experience.

NUREG/CR-4590, Vols. 1 and 2, PNL-5832, Pacific Northwest Laboratory, Richland, WA.

20. 0. A. Neitzel, K. I. Johnson, T. L. Page, J. S. Young and P. M. Daling.

1984. Bivalve Fouling of Nuclear Power Plant Service-Water Systems,

NUREG/CR-4070, PNL-5300, Pacific Northwest Laboratory, Richland, WA.

21. Bitten, E. R. 1985. N-Reactor Production Assurance Program-BOP

Evaluations: Task 1: Buried Piping, UNI-3133. United Nuclear, Inc.,

Richland, WA.

22. Electric Power Research Institute. 1987. Proceedings: 1987 Seminar on

Nuclear Plant Lavup and Service Water System Maintenance, Charlotte, NC.

23. W. E. Vesely. 1987. Risk Evaluations of Aging Phenomena: The Linear

Aging Reliability Model and its Extensions, NUREG/CR-4769 Scientific

Applications International Corporation, Columbus, OH.

R.2
APPENDIX A

ROOT CAUSE SCHEME
APPENDIX A

ROOT CAUSE SCHEME

A LOGIC TREE FOR INFORMATION ASSESSMENT AND AGING IDENTIFICATION

Before starting a root cause analysis for aging research using plantrelated data, several basic questions about the validity and reliability of

that data need to be answered. Without an a priori knowledge of the data,

the results of the research may be misinterpreted or even counterproductive

to the investigation effort. Root Cause Analysis (RCA) is not new to the

utility industry. The identification of the correct fundamental problem

associated with a component malfunction has always been a hallmark of the

astute engineer, foreman, operator, or mechanic. While intuitively shaped

through experience, very little formal, documentable structure has historically been generated to act as a guide to this central issue. As a result, a

wide spectrum of methods have been applied with varying degrees of documentation and consequently a wide band of uncertainty in the end product.

Root cause analysis is a subjective process because component environmental history and failure condition information is often incomplete and

multiple possibilities exist for the same symptoms. In order to reach a

valid conclusion, the analyst applies judgement and data interpretation to

rule out all but one possibility for the root cause of a component failure.

In some cases, the field engineer may feel obligated to make a root cause

identification when, in fact, it may be best to label the cause unknown. The

analyst may not be aware of the multitude of decisions that were made to

reach a conclusion. Thus the lack of information, communication, and unknown

and untraceable decisions, render root cause analysis highly uncertain,

potentially inconsistent, and potentially incorrect. An aging assessment or

license extension program will undoubtedly require more consistent, and indepth data, to instil that the program confidence that will produce adequate,

justifiable regulatory guidelines.

This appendix is intended to help define the RCA process, providing a

perspective necessary to formulate a uniform approach, depth of investigation, and documentation. The basic questions of what failed, why, when, and

how, must specifically relate component failure to the aging phenomena.

ANATOMY OF A FAILURE

Root cause categorization schemes (A-1 and A-2) have been performed

using various data for input (A-3, A-4) to fit component failures into

supposedly relevant domains. A step back must be taken to approach the

larger question of the root origin of the failure. To being this search, an

important definition must be accepted, that of the root cause analysis of a

component failure as

A.1
The process of determining the fundamental degradation mechanism

associated with a component failure, such that, if corrected, it will

prevent a recurrent failure of a similar nature.

From this definition, it follows that data on three elements must be

known to provide the essentials of root cause analysis: 1) the fundamental

component (ground element) that failed, 2) the degradation mechanism which

lead to failure, and 3) the operational history of the component.

The prospects for a component successfully accomplishing its design

mission hinges on three basic factors each of which has the ability to

produce a premature component failure: 1) design or application, 2) changes

in the component's environment to conditions beyond the design envelop, and

3) aging degradation. Each area is examined in the following in more detail

to provide a suitable definition in the context of an aging/failure related

study.

DESIGN INADEQUACY

Every component in a nuclear power plant is designed to operate within a

limited, specified (or implied) set of operating conditions. Specifying

placement of a 400-psi pressure gauge in a system that will be exposed to

2000 psig is obviously a design error. Design specifications apply to a wide

range of potential stressors that the component is expected to be exposed to

during its design life. Typically, these include specified envelopes for the

following:

* temperature

* pressure

* lifetime

-- cycles

-- wear conditions -- duration

* maintenance requirements

* deterioration (degradation) rate.

The failure of a component when it has clearly been operated within its

design envelope is by definition a design inadequacy.

ENVIRONMENTAL ABNORMALITY

The failure of a component through the violation of the design envelop

from improper maintenance or operation is termed an environmental abnormality. Even assuming that all predictable stressors are taken into account

A.2
in designing a component, the design envelop can be breached by operation or

maintenance of the component outside its design specifications. Typical

examples are a steam generator operated outside its pH limits, a pump that is

operated under runout conditions, or an incorrect fuse that immediately blows

when it is energized. Personnel errors in operation and maintenance of a

component generally fall in this category, because they frequently result in

material environments outside the design specification envelop.

AGING DEGRADATION

Aging degradation pertains to stressor that are either not anticipated

in the design process or are more severe than anticipated even though they

still do not exceed the design specifications of the device.

Aging degradation failure is defined as failure of a component at less

than design life resulting from manufacturing flaws or unanticipated environmental stressors. This catch-all category includes a large portion of the

age-related failures from what may be considered "implied" design specifications, i.e., assumptions on the part of the designer and/or the component

user concerning the quality of the manufactured product or the severity of

the component environment whichin practice, turn out to be ill-founded.

These include but are not limited to the following:

* manufacturing defects,-

* an operational parameter that results in premature (lifetime less than

design) failure of-component from such influences as

-- erosion

-- corrosion

-- biofouling -- organic attack -- vibration

-- embrittlement.

Examples here include 1) a heat-treatment-induced crack in a recirculation

pump shaft, 2) a service water pump impeller, which must be replaced every

eighteen months because river water contains a higher suspended solid concentration than anticipated, and 3) a emergency diesel generator which is

"tested to death" in accordance with technical specifications. These

failures are what are sometimes termed "aging failures" in the classical

sense.

APPROACH TO ROOT CAUSE

To begin formulating an overall approach to root cause analysis from in

aging perspective, the definition of aging as used in this study bears

repetition:

A.3
The cumulative changes with the passage of time that may occur within a

component or structure due to one or ore of the following factors:

- natural processes during operation

- external stressors caused by storage or operation

- external stressors caused by storage or operation

- wear caused by operational or test cycling - improper installation, application, operation or maintenance.

This definition of aging spans all three categories of root cause. The

problem is not merely one of aging in a narrow sense, i.e., old or worn out,

but encompasses all paths that lead to a component failure. A generalized

root cause scheme which can accommodate any abnormal event is, therefore,

required to meet the needs of a broad aging-research charter. To provide an

initial approach to the problem, a logic diagram (Figure A.1) was developed

to provide an explicit structure for a repeatable root cause analysis

methodology and to define what is meant by, and the information required for,

a viable root cause investigation. This figure is necessarily preliminary.

It is intended only as a framework on which to build a practical root cause

analysis system in future work.

The figure is designed to treat failure events after-the-fact, as if the

researcher were at the plant for the purpose of determining why component Z

has failed. It presupposes no knowledge of the event but has inherent to it

an intimate knowledge of the system and, particularly, of the component

function and environment. It is intended that one should start with collected failure knowledge at the top of page one and attempt to reach a conclusion box at the bottom. To state that a root cause analysis has been adequately performed for a given component failure, it must be shown that

process shown in Figure A.1 has been followed.

The extent that additional knowledge is pursued to define the root cause

of failure must be balanced by the monetary and risk significant of the component. This aspect will be treated in Phase II of the SWS study.

A.4
FIGURE A.1. Basic Failure Analysis Scheme

A.5
SUMMARY

A component's chance for realizing its design life requirements without

sustaining a failure depends on the following factors:

DESIGN

1. component engineered to meet all design parameters

2. component application within the design envelop

3. QC programs which guarantee defect-free manufacturing and

modification processes

ENVIRONMENT

1. proper equipment operation within design specifications

2. proper preventive maintenance of limited lifetime components

AGING

1. identification of all aging stressors and their magnitudes

2. maintenance programs that identify, monitor, and effectively

mitigate aging degradation.

While the above goals are obviously an idealization and unattainable, their

pursuit is, nevertheless the horizon of reliability engineering.

REFERENCES

A-1. Meale et al. "Identification of Aging Root Causes of Component Failures

for Service Water and 1E Electrical Power Distribution Systems." EGGREQ-7227, July 1986.

A-2. Murphy et al. "Survey of Operating Experience from LERs to Identify

Aging Trends." Status Report, NUREG/CR-3543, ORNL/NSIC-216, November

1983.

A-3. NUREG/CR-1363. "Data Summaries of Licensee Event Reports of Valves at

U.S. Commercial Nuclear Power Plants."

A-4. Kahl and Borkowski. "Component Aging Illustrated in Maintenance

Histories from IPRDS." Presented at 1984 American Nuclear Society

Annual Meeting, Conf-840614-84, DE84014294.

A.6
APPENDIX B

SERVICE WATER SYSTEM CORROSION MECHANISMS
APPENDIX B

SERVICE WATER SYSTEM CORROSION MECHANISMS

INTRODUCTION

Corrosion has emerged as a major aging-degradation factor in nuclear

plant service water systems (SWS). System materials are subject to relatively aggressive waters with a wide range of compositions, including the

following: sea, lake, river, and cooling pond. Normal operation and standby

conditions each have special corrosion considerations. The main report

identifies the two principal service water system types: open systems (predominant) (Figure 2.2) and closed systems (Figure 2.3). The open system and

primary side of the closed system are subject to similar environments, given

a common water source. The secondary side of the closed system Ais serviced

by treated water, so corrosion can be mitigated by proper choice and control

of the water chemistry. Service water systems are also exposed to a variety

of biological species that frequently enter into the corrosion phenomena.

This'section provides a brief overview of service water system corrosion

phenomena. This Phase-I NPAR SWS study is representative of an investigation

of a single plant (B.1). Literature was studied for all U.S. plants (see

Table 1.1 in Section 1.3.2), and expert opinion invoked, to gain insights on

generalizing the single-plant findings. The Phase-II study will address

corrosion and water treatment of other representative plant configurations in

more detail.

MATERIALS

The predominant fT -ucF In fa-t alV& fof krV! C 7WRt-ePysystems is carbon steel. Other materials commonly used are copper and copper-base alloys,

monel (a nickel-copper alloy), and stainless steels (300 and 400 series).

Copper-base alloys are widely used in heat exchanger applications because of

inherently good corrosion resistance combined with good mechanical properties, excellent thermal conductivity, and ease of soldering or brazing.

Stainless steels, particularly types 304 and 316, are used increasingly as

nickel-base replacement materials to improve corrosion resistance.

{ESt 82ismate7FromFSsrv~ons~~m~moetand~matel al to

rrone-p1int We are aware of one plant that is replacing carbon steel component's with' stainless steel to reduce the impact of corrosion. All pipe

and valves with diameters 3 in. and smaller, and valve internals for larger

valves, are being replaced with stainless steel. Heat exchanger leaks are

typically repaired by plugging leaking tubes. When too many tubes are plugged, the tube bundle is replaced. The copper-nickel tubes are sometimes

replaced with stainless steel. Valve stems sometimes are replaced with

17-4 PH steel.

B.1
TABt~E-.IvObserved Damage -Mechanismn

Component Material Observed Deqradation Damage Mechanism

Carbon Steel Pin-hole leak Tuberculation or

concentration cell

corrosion from deposits

Plugging Rust and biological

and/or inorganic

deposition

Check-valvej

swingFarRI

Carbon steel Swing-arm failure Rust which is removed

by chaffing as the arm

rotates

Gage valve-dip k

eatH emcii-hg?

sJo \JO .

Carbon steel

90-10 CuNi

Disk separated from

stem

Leak

Rust

Corrosion pitting

Leak

Plugging

Denickelification

Tuberculation

fragments

CORROSION MECHANISMS

There is potential for corrosion and other damage in SWSs that either is

not noticed or has not proceeded far enough to cause failures. Table B.2

shows potential corrosion mechanisms that are promoted by combinations of

material, environment, and stress. These are prime candidates for a systematic degra-dation evaluation planned for phase II. The damage mechanisms in

heat exchangers have been observed and identified by multifrequency eddy

current testing. The pipe damage mechanisms are identified from handbook

recommendations. Most subsystems contain carbon steel and stainless steel

pipe and fittings. Various stainless steel alloys are used: 304 SS for

pipe, and 410 and 17-4 PH SS for some valve stems. Heat exchanger tubes are

stainless steel, copper-nickel and copper. Copper-bearing alloys may be

susceptible to ammonia attack in raw waters where decaying organic matter or

fertilizers are sources of ammonia. Ammonia-induced stress corrosion cracking of Admiralty brass could be a consequence of exposure to ammoniated

waters. Stainless steels are susceptible to stress corrosion cracking in

oxygenated, chloride-bearing waters.

B.2
TABLE B.2. Potential Corrosion Factors

micro-biologically induced corrosion

oxygen concentration cell

grounding connections

dissimilar surface conditions

dissimilar metals

dissimilar soils

errosion corrosion

weldment HEZ areas

deposition

stagnant flow areas

chloride tunneling (stainless steel)

intergranular stress corrosion

hydrogen embrittlement

Some portions of the SWS are stagnant, inviting deposition and certain

types of corrosion. Other regions are subject to relatively high flow rates

resulting in erosive attack.

A review of aqueous corrosion mechanisms for copper-base and iron-base

alloys has been published (B.1), including uniform corrosion, galvanic

attack, stress corrosion cracking, intergranular attack, crevice corrosion,

pitting, dezincification, fretting corrosion, corrosion fatigue, erosion, and

corrosion. Corrosion product transport is also briefly reviewed. An EPRI

study of failure causes in condensers has relevance to corrosion in SWSs

(B.2). Biofouling has emerged as a major consideration in open water systems, reflected in Reference B.3 through B.9. Microbiologically-induced

corrosion (MIC) is treated in a later section of this appendix.

A review of aqueous corrosion mechanisms for copper-base and iron-base

alloys has been published (B.1), including uniform corrosion, galvanic

attack, stress corrosion cracking, intergranular attack, crevice corrosion,

pitting, dezincification, fretting corrosion, corrosion fatigue, erosion, and

corrosion. Corrosion product transport is also briefly reviewed. An EPRI

study of failure causes in condensers has relevance to corrosion in SWSs

(B.2). Biofouling has emerged as a major consideration in open water systems, reflected in References B.3 through B.9.

B.3
CORROSION STRESSORS IN SWSs

Of the wetted systems in NPPS, the SWS seems to have the most aggressive

combination of corrosive factors, even though the temperature range is relatively low (-0-500C or 32 to 120F). Effects of electrical, mechanical, and

thermal factors are reviewed briefly here. They will be addressed in detail

in the Phase II investigation, including potential effects of accident and

post-accident scenarios.

Electrical

To date this study has not addressed the importance of stray currents in

SWS corrosion, but the electrical effects need to be considered for the

buried and submerged structures. and piping (B.10). Galvanic factors from

dissimilar metal couples, differential aeration, etc., must also be considered. Cathodic protection systems can be effective in mitigating corrosion in soils and impure waters, if the systems are properly designed and

maintained (e.g., timely replacement of sacrificial anodes).

Mechanical

Service water system piping and components are subject to vibration.

While not generally excessive, in some cases the vibrations can contribute to

fatigue. In extreme cases, water hammer has caused obvious damage to specific components. Water flow is sufficiently high in other cases (12 to

15 fps) to contribute to erosion, exacerbated by chemical factors, and in

some cases, by biological species (e.g., mollusks) that cause local flow

perturbations. Cavitation damage, particularly on pump materials, is a

potential phenomenon.

Thermal

Water intake conditions are ambient, varying from hear freezing up to

-900F over the range of SWS locations. Outlet conditions also vary with

season and location, in the range of 90 to 1200F.

ENVIRONMENTAL FACTORS

The chemical, radiation, and atmospheric, or underground factors in the

corrosion of SWSs are briefly reviewed in this section. The expanded treatment in Phase II will incude analysis of possible variants in accident and

post-accident scenarios.

Chemical

Service water systems are subject to a wide range of relatively impure,

untreated waters (exception: the secondary side of closed systems). Dissolved oxygen, halides, carbon dioxide, ammonia, manganese dioxide, etc.,

provide aggressive combinations that contribute both to a range of corrosion

mechanisms and/or to deposition, leading to fouling and plugging. As

B.4
indicated in another section of this appendix, a range of biological species

(e.g., sulfur-reducing bacteria) contribute to the aggressive chemical environments in SWS.

Water flow varies from erosive to stagnant, contributing a range of

variants on the aggressive nature of the environment. For example, steel

corrosion increased by a factor of -7 as sea water flow increased from 0 to

20 ft/s (0 to 6 m/s) (B.11).

Radiation

In normal operation, corrosion in SWSs is not influenced by radiation.

Possible effects in accident or post-accident scenarios will be evaluated in

Phase II but currently appear to be nil.

Atmospheric Corrosion

Intake structures and exterior surfaces of components and piping are

subject to atmospheric corrosion. Moist areas and zones subject to wet/dry

cycling are particularly subject to atmospheric attack. Dead-leg, air-bound

locations are sometimes overlooked or difficult to inspect. A review of

atmospheric corrosion includes materials found in SWSs (B.1). In general,

the conditions on the wetted surfaces are expected to be more aggressive than

on the surfaces in contact with air, but air-side corrosion needs to be considered in Phase II.

Underground Corrosion

Sections of SWS piping and intake structures are in contact with soils.

As with wates, SWSs are subject to a wide range of soils, including site-tosite variations in pH, composition, moisture content, etc. Therefore, the

acceptable life of materials in contact with soils will vary from site to

site. The principal factors in soil corrosion are porosity (aeration),

electrical conductivity, dissolved species (including polarizers and

inhibitors), moisture, and acidity or alkalinity (B.11). Standard field

tests have been conducted that provide a basis for predicting corrosion

behavior for a range of materials in various soils (see Reference B.11,

p. 153).

A systematic assessment of corrosion on buried piping at the Hanford N

Reactor was conducted and reported on by Hurd (B.12). This report summarizes

surveillance and maintenance practices. The underground piping system

includes -56,000 ft-(17,070 m) of-carbon steel pipe with diameters from 3 to

108 inches (7.5 to 270 cm).

The range of wall-thinning mechanisms from both the inside (water-side)

and outside (soil-side) pipe surfaces were considered. The following observations illustrate selected considerations that were evaluated in the

N Reactor buried piping study (B.11):

B. 5
* Galvanic corrosion can develop from coupling different metals or from

adding sections of replacement pipe in contact with older sections of

the same material (new pipe tends to be anodic to the older passivated

pipe).

* Holidays in protective coatings resulted in local pitting.

* A survey of corrosion by a range of river waters indicates up to a

factor of five differences in steel corrosion rates at three U.S. river

sites. However, even larger factors are possible (e.g., rates up to

40 mpy).

* Partially filled horizontal pipes transporting air-saturated water tend

to pit at the apex above the water level and at the water-to-air

interface.

* Intermittent flow of air-saturated water produced a particularly

corrosive environment.

* Differential dissolved oxygen concentrations in soils (e.g., under paved

roads versus open soil) produced local corrosion phenomena.

* Corrosion may occur at grounding connections in buried steel pipe with a

galvanized central ground grid after the sacrificial zinc anode is

consumed.

* Dissimilar siX (e.g., clay versus sand) can set up local corrosion

cells.

CORROSION/FOULING PHENOMENA IN SWSs

Table B.2 illustrates the types of corrosion mechanisms that must be

addressed in a comprehensive SWS corrosion assessment. Each is described in

the following.

Uniform Corrosion

Wall thinning by uniform corrosion generally has been accounted for in

the original design, though allowances are not always consistent with 40-y

life. However, this mechanism generally does not result in serious problems

except in specific areas such as unprotected pipe and water boxes. Cases of

substantial uniform corrosion may result in secondary effects such as corrosion product transport and deposition, leading to plugging and localized

corrosion under the deposits. Examples of corrosion rates are cited here but

will be addressed for a broader range of materials and conditions in the

Phase II assessment.

At ambient conditions, steel corrosion in sea water varies from 0.001 to

0.008 inches per year (ipy) or 25 to 200 um per y (B.11). Steel corrosion,

as a maximum, would be expected to double for each 100C (200F) increase

B.6
(B.11). However, decreases in oxygen concentration and calcareous deposits

may mitigate the-corrosive attack as temperature increases.

Corrosion caused by a variety of river waters has been assessed.

Test data indicate that the general corrosion rate for carbon steel

exposed to untreated Columbia River water is approximately 3.6 mils/year.

These data were obtained from the results of a corrosion monitoring program

for carbon steel pipe at Hanford locations (B.10). Test data, reported for a

six-m period, showed that the initial corrosion rate was approximately

0.66 mils/m: the uniform steady-state rate after six months was approximately 0.30 mils/m. Theidata also indicated that pitting corrosion was

occurring.

Test data for three rivers located in the eastern United States were

reported for an eight-y period by Coburn (B.12). The three rivers were the

Allegheny, Monongahela, and Mississippi. The data showed that the general

corrosion rate of carbon steel pipe exposed to Columbia and Allegheny River

water was approximately equal. The general corrosion rate for exposure to

Monogahela River water was twice that for exposure to the Columbia River

water; the greater rate was attributed to ferric sulfate concentration from

mine drainage which resulted in a pH range of 3.5 to 4.0. The general corrosion rate for exposure to Mississippi River water was less than for exposure to Columbia River water; the lower rate was attributable to the test

method, not to the water mineral content. The data show that the initial

corrosion rate for each test decreases to a uniform rate; the value for the

Allegheny test was approximately 2.5 mils/y (-65 um/y).

A pipe wall wastage evaluation for a 20-yr-old, 6-in. Schedule-40 carbon

steel pipe showed that leakage may be expected in approximately 15 more years

(B.11). This estimate was based upon the Coburn data for the Alegheny River

water test which was conducted at room temperature. The applied steady-state

general corrosion rate was approximately 2.5 mils/y; the pitting corrosion

rate was approximately 8 mils/yr (200 um/y). This case illustrates that some

systems may not last for a full 40 y and, in fact, probably does not

represent the most aggressive cases.

Phase II should consider the impact of corrosion on such factors as

changes in internal pipe diameters, local chemical environments, and wall

morphology.

Pitting Corrosion

Tuberculation and pitting are common corrosion mechanisms for carbon

steel pipe exposed to natural water (B.10).

Pit and tubercule formation on carbon steel pipe at Hanford locations

has been assessed. A tubercule is a concentration of corrosion product

covering an anodic area where pit formation is progressing. Initially, the

tubercule causes an increase in corrosion penetration rate; later the

penetration rate decreases because the corrosion layer protects the pit.

B.7
However, the impenetrable corrosion layer may initiate an oxygen depletion

corrosion mechanism which promotes the anodic tendency at the pit. A damaged

tubercule is replaced by a layer of corrosion product and tubercule formation

continues.

A pitting rate of 8 mils/y (200 um/y) was estimated for carbon steel

pipe exposed to Columbia River water (B.10). In more aggressive waters,

higher pitting rates are to be expected (to be characterized in Phase II).

Corrosion products such as ferric oxide occupy approximately four times

the volume of the original metal (B.10). The corrosion products may cause

flow restrictions where they develop or they may spill and/or dissolve and

redeposit at other locations.

Scaling caused by depositable species such as carbonates is another

major consideration in raw water systems. Various indices (e.g., the

Langelier Saturation Index) provide a basis to predict the scaling behavior

of waters (B.15). However, as in the case of the Langelier index, the

application is relatively narrow, i.e., applying only to calcium carbonate.

INTERGRANULAR STRESS CORROSION CRACKING (IGSCC)

Stainless steels are being used to replace mild steel in smaller pipes

and components in SWSs to improve corrosion resistance. In some areas, e.g.,

uniform corrosion and erosion-corrosion the benefit is clear. However, even

at the relatively low temperatures characteristic of SWS, stainless steels

are prone to IGSCC if attention is not given to proper alloy selection,

metallurgical condition, and environmental control. For example, numerous

IGSCC failures of 304 stainless steel components have occurred in spent fuel

pools at less than 506C (122F) (B.13, B.14). Oxygenated, relatively impure

waters are a given in SWSs, suggesting a need for increased attention to

controllable factors that mitigate IGSCC in stainless steels. Choice of L

grades is an obvious positive to improve IGSCC resistance. Other important

considerations are selection of welding parameters and heat treatments to

minimize sensitization, and elimination of high stresses. Investigations of

IGSCC at low temperatures indicate effects of stress level and sensitization

of 304 SS IGSCC (B.13, B.14). Materials that release and re-deposit

relatively large amounts of corrosion products (e.g., carbon steel) may

contribute to IGSCC resulting from adsorption and concentration of aggressive

species in deposits on materials that are prone to IGSCC. It is worth noting

)( that 300-series stainless steels are prone o IGSCC while 400-series materials are largely immune.

Ammonia-induced cracking of copper-bearing alloys was mentioned earlier

and must be considered in systems where ammonia is a significant species.

B.8
EROSION CORROSION

High flow rates at some locations in SWS, sometimes exacerbated by local

flow perturbation, can lead to erosion corrosion., Carbon steels are particularly susceptible to erosive attack. Austenitic stainless steels are

essentially immune (B.11). Addition of even 1% chromium can reduce erosive

effects by an order of magnitude, but chromium contents of >2% are recommended. The most erosion-resistant copper alloys for sea water are the tine

bronzes (tin contents of 5-10% for cast alloys; 12% for heat exchanger

tubes) (B.11). For the brasses, resistance to erosion increases with zinc

content. Addition of iron improves the erosion resistance of the Cu-Ni

alloys.

BIOLOGICAL ATTACK

Service water systems are subject to a wide range of biological species,

including microorganisms (bacteria) and macroorganisms such as mussels, clams

and barnacles.

Surveillance and control of biological species have been rather

ineffective at some power plants (B.7, B.8). Although mollusk control may be

hampered by conflicting objectives (environmental versus biological controls), relatively effective new molluscicides are on the market. Measures

to mitigate biological attack sometimes have been included in plant design.

However, some cases of bivalve fouling have not been anticipated, due to

invasions to locations where the species previously was unknown. Use of

strainers and similar devices can aid in control of biofouling.

The reliability of open-cycle water systems can be improved by surveillance and control programs (B.7). Revision of plant technical specifications is an important aspect of effective biofouling control.

Strategies to control biofouling must take into account secondary

effects. For example, wash-off of bivalves killed in a chlorination campaign

may result in heat exchanger plugging (B.7). Secondary effects of biocides

injected to control MIC must be considered. For example, denickelification

of Cu-Ni alloys is a potential side effect of chlorination.

The occurrence of MIC in service water systems has been troublesome,

particularly in stagnant or deadleg locations (B.4). Bacteria colonies

attached to pipe walls develop nodules or tubercles, isolating the colony

from the environment, rendering biocides largely ineffective. If the

nodules are broken off without destroying the bacteria, a fresh influx of

nutrients further augments the colony and the attendant corrosion. It is

worth noting that current technology includes the use of wetting agents/

surfactants for use in treating service water systems subject to MIC problems. Specifically, the surface active agents are added to the water containing biocidal treatments, attempting to penetrate biota which are growing

in the Service Water System.

B.9
Materials considered to be corrosion-resistant are now recognized to be

prone to MIC. For example, a carbon steel SWS component is replaced with

Type 316 SS. Within six months the stainless steel had extensive MIC attack

(B.7). Some nickel-base alloys show evidence of resistance to MIC.

Bacterial corrosion is pervasive and can occur in local areas throughout

a system. It sometimes has features similar to other corrosion mechanisms,

so a systematic root cause analysis is often required to differentiate MIC

from other types of corrosion.

The importance of biological fouling to the safety of SWSs is addressed

in IE Bulletin No. 81-03 (B.8) and Generic Issue 51 (B.9).

FOULING

Fouling refers to all deposits on system surfaces that increase resistance to fluid flow and/or heat transfer. Sources of fouling include the

following:

* organic films of microorganisms and their products (microbial fouling)

* deposits of macroorganisms such as mussels (microbial fouling)

* inorganic deposites, including scales, silt, corrosion products and

detritus.

Scales result when solubility limits for a given species are exceeded.

Deposits r sult when coolant-borne particles drop onto surfaces due to

hydraulic factors.

The deposits result in reduced flow of cooling water, reduced heat

transfer, and increased corrosion. Sediment deposits promote concentration

cell corrosion and growth of sulfur-reducing bacteria (B.6). The bacteria

can cause severe pitting after one month of service. Piping systems designed

for 30 years have had their projected life reduced to five years due to

undersediment corrosion.

Chlorination is the predominant method to control biofoulants, but

federal discharge regulations limit the effectiveness of this approach

(B.6). Bromine is another biocide used in service water systems but is

similarly limited. Other methods include backflushing, organic coating, or

thermal shock (8.10). Chemical and mechanical cleaning methods are applied

periodically in some systems to remove the fouling materials. The relatively

large SWS size imposes a need to focus antifouling procedures on areas that

have the most significance to plant safety and efficiency.

B.10
I

MANAGING EFFECTS OF SWS CORROSION AND FOULING

Understanding Corrosion and Fouling Processes and Their Safety Impacts

The relatively aggressive conditions in SWSs result in several types of

corrosion and fouling phenomena, outlined in prior sections. In most cases,

the mechanisms are recognized. Some types of corrosion were anticipated in

plant designs; some, such as MIC, were not. Some root cause analyses of

component failures point clearly to corrosion; in other cases the cause was

less obvious, with corrosion as a suspected factor. Phase II of this study

will focus in more detail on corrosion and faulting as important elements in

the aging of SWSs. Potential effects on plant safety will be addressed

here.

Corrosion Detection and Monitoring Methods

Currently, corrosion is discovered by leaks, reduced flow surveillances,

or inspections. Coupons are used to monitor heat exchanger corrosion on SWS

heat exchangers. However, it seems that the coupon monitoring serves as an

indicator of current condition rather than as a tool to obtain corrosion

rates for predicting the life of the heat exchanger. Coupons are not used

generally elsewhere in the service water system. Pipe wall thickness is

periodically measured ultrasonically. Typically, the amount of wall thinning

is within an acceptable range, but corrosion product and silt buildup are

sources Of uncertainty in the measurements. Reduced flow area from material

buildup and pinhole leaks are not detected by either the coupon or the

ultrasonic wall thickness measurements. Wall thickness measurements would be

best performed after pipe cleaning.

Reference 8.12 summarizes methods used to assess and monitor corrosion

in carbon steel piping systems. Methods include deposit monitors, corrosion

coupons, bypass pipes, spool pieces, corrosion rate meters, and pitting

meters.

Trending of Corrosion and Fouling in SWSs

Trending of corrosion parameters and the magnitudes of corrosion on key

locations and components in SWS appears to be an area where substantial

improvements would yield valuable benefits, both to the assurance of system

integrity and to the economic operation of the system. Phase II will include

a summary of current trending practices, with recommendations for

improvements.

MAINTENANCE/REFURBISHMENT/REPLACEMENT

Numerous considerations are obvious in these areas:

* timely painting of exposed structures

* replacement of sacrificial anodes in cathodic protection systems

B.11
* proper attention to materials selection in components prone to failure

* coating application and repair

* chemical and/or mechanical cleaning

* water treatment.

REFERENCES

B.1. Johnson, Jr., A. B., D. R. Pratt, and G. E. Zima.

Materials Performance in Dry Cooling Applications.

1976. Survey of

BNWL-1958.

B.2. "Failure

Electric

Research

Cause Analysis - Condenser and Associated Systems." 1982.

Power Research Institute, Inc., Palo Alto, California,

Project 1689-2.

B.3. Hayner, G. O., D. H. Pope, and B. E. Crane. 1987. "Microbiologically

Influenced Corrosion in Condenser Water Boxes at Crystal River-3."

Paper presented at the Third International Symposium on Environmental

Degradation of Materials in Nuclear Power Systems -- Water Reactors,

August 30 to September 3, 1987, Traverse City, Michigan.

B.4. Stein, A. A. 1987. "Microbiologically Induced Corrosion." Paper

presented at the Third International Symposium on Environmental

Degradation of Materials in Nuclear Power Systems -- Water Reactors,

August 30 to September 3, 1987, Traverse City, Michigan.

B.5. "Biofouling Control Assessment - A Preliminary Data Base." 1982.

Electric Power Research Institute, Inc., Palo Alto, California,

Contract TPS 80-739.

B.6. Johnson, K. I., and D. A. Neitzel. 1987.

Open-Cycle Water Systems. PNL-5876, Vol.

Laboratory, Richland, Washington.

Improving the Reliability of

1, Pacific Northwest

B.7. Neitzel, D. A., K. I. Johnson, and P. M. Daling. 1986.

Reliability of Open-Cycle Water Systems. PNL-5876, Vol.

Northwest Laboratory, Richland, Washington.

Improving the

1, Pacific

8.8. U.S. NRC; I&E Bulletin No. 81-03, Flow Blockage of Cooling Water to

Safety System Components by CORBICULA SP. (ASIATIC CLAM) and MYTILUS

SP. (MUSSEL), July 1981.

B.9. "A Prioritization of Generic Safety Issues."

Issue 51 Proposed Requirements for Improving

Cycle Service Water Systems. NUREG/CR-0933,

Commission, Washington, D.C.

December 1984. Generic

the Reliability of Open

U.S. Nuclear Regulatory

B.10. Daling, P. M., and K. I. Johnson. 1985. Bivalve Fouling of Nuclear

B.12
Power Plant Service-Water Systems.

Laboratory, Richland, Washington.

PNL-5300, Vol. 2, Pacific-Northwest

B.11. Henager, C. H., P. M. Daling, and K. I. Johnson. 1985. Bivalve

Fouling of Nuclear Power Plant Service-Water Systems. PNL-5300,,

Vol. 3, Pacific Northwest Laboratory, Richland, Washington..

B.12. Hanford Engineering Development Laboratory. 1985. N Reactor

Production Assurance Program Balance of Plant Evaluation. UNI-3133,

Rev. 1, Vol. 1, Compiled by E. N. Hurd, UNC Nuclear Industries,

Richland, Washington.

B.13. The Electrochemical Society, Inc. 1986. The Corrosion Handbook.

Edited by H. H. Uhlig, Massachusetts Institute of Technology,

Cambridge, Massachusetts.

B.14. Coburn, S. K. 1965. Corrosion in Fresh Water. ASTM Metals Handbook,

Vol. 1, 9th Edition.

B.15. Jones, R. H., A. B. Johnson, Jr., and S. M. Bruemmer. 1982.

"Evaluation of Stress Corrosion Cracking of Sensitized 304SS in Low

Temperature Borated Water." In Proceedings of the 2nd International

Conference on Environmental Degradation of EnQineering Materials,

Blacksburg, Virginia, September 1981.

B.16. Bruemmer, S. M., and A. B. Johnson, Jr. 1983. Effect of Chloride

Thiosulfate. and Fouride Additions on the IGSCC Resistance of Type

304 Stainless Steel in Low Temperature Water. Presented at the

International Symposium on Environmental Degradation of Materials in

Nuclear Power Systems--Water Reactors, Myrtle Beach, South Carolina,

August 22-25, 1983.

B.17. Uhlig, H. H.

John Wiley &

1967. Corrosion and

Sons, Inc., New York,

Corrosion Control.

New York.

Published by

8.13

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Richland, WA 99352

NRC: FIN B29119

10. SPONSORING ORGANIZATION NAME AND MAILING ADDRESS Iat-WZO CoaF lIe TYPE OF REPORT

Division of Engineering

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission B PERIODCOVEREDtlulw dtnl

Washington, DC 20555

12 SUPPLEMENTARY NOTES

13. ABSTRACT 1200w'dsw I=)

Phase I of an aging assessment of nuclear power plant service water

systems, was performed by the Pacific Northwest Laboratory in support of the

U.S. Nuclear Regulatory Commission Nuclear Plant Aging Research program. The

service water system was selected for study because of its essential role in

the mitigation of and recovery from accident scenarios involving the potential

for core melt. The objectives of the Service Water System (SWS) task are to

identify and characterize the principal aging degradation mechanisms relevant

to this system, to assess their impact on operational readiness, and to provide

a methodology for the mitigation of aging on the service water aspect of nuclear

plant safety. The first two of these objectives are covered in this Phase I

report.

A review of available literature and data base information indicated

that motor operated valve torque switches (an electromechanical device) were

the prime suspect in component service water system failures. More extensive

and detailed data, however, obtained from cooperating utility maintenance

records and personnel accounts, contradicted this conclusion indicating that

organic accumulation and corrosive attack of wetted component surfaces were,

in fact, the primary degradation mechanisms.

14 DOCUMENT ANALYSIS -a. KEYWORDSIDESCRIPTORS 1E AVAILABILITY

Service Water System, Safety Components, Aging Degradation, 1.ATEMENT

Maintennace, Nuclear Plant Aging Research, Root Cause Analysis Unlimited

16. SECURITYCLASSIFICATION

b. iDENTIFIERSIOPEN ENDED TERMS Unclassified

Unclassified

17 NUMBER OF FAGES

IS. PRICE

*U.S.GCIVRNMNNT PRINTING DFOI0CE1g989.741..5o,OO111
UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555

SPECIAL FOURTH-CLASS RATE

POSTAGE & FEES PAID

LISNRC

PERMIT No. G467

OFFICIAL BUSINESS

PENALTY FOR PRIVATE USE, $300