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SUMMARIES OF RESEARCH REPORTS SUBMITTED

IN CONNECTION WITH THE NUCLEAR AGING

RESEARCH (NPAR) PROGRAM

Submitted by:

Louis N. Rib

Nuclear Power Safety Consultant for:

ENGINEERING AND ECONOMICS RESEARCH, INC.

1801 Alexander Bell Drive

Reston, Virginia 22091

Contract No. NRC-04-84-120, Task No. 021

Submitted to:

U.S. NUCLEAR REGULATORY COMMISSION

5650 Nicholson Lane

Rockville, Maryland 20852

September 1986

- . _
TABLE OF CONTENTS

Page

ABSTRACT i

1.0 INTRODUCTION 1-1

2.0 CATEGORY I SUMMARIES - BACKGROUND EXPERIENCE AND DATA BASE 2-1

2.1 Aged Plant Survey 2-1

2.2 LER Data-Aging Trends 2-8

2.3 Aging/Risk Considerations 2-17

3.0 CATEGORY 2 SUMMARIES - ANALYSIS METHODOLOGY AND METHODS TESTING 3-1

3.1 Methodology for Aging Analysis 3-1

3.2 Signature Analysis Technique for Electric Motors (MOVATS) 3-14

4.0 CATEGORY 3 SUMMARIES - RESULTS OF PHASE I RESEARCH FOR COMPONENTS 4-1

4.1 Electric Motors 4-1

4.2 Battery Chargers/Inverters 4-10

4.3 Electrical Cables 4-23

4.4 Pressure Transmitters 4-35

4.5 Diesel Generators 4-46

4.6 Mocor-Operated Valves (MOVs) 4-54

4.7 Check Valves % 4-60

4.8 Auxiliary Feedwater Pumps 4-68

4.9 Snubbers 4-74

i
ABSTRACT

As a result of implementing the Nuclear Plant Aging Research (NPAR)

program plan, the results of Phase I efforts for selected electrical and

mechanical components have been published since 1984. To reduce the difficulty of maintaining cognizance of this wealth of information, summaries

of each of 14 NUREG/CR reports have been developed and collated into this

report. The purpose is to make the result of these studies more available

for rapid survey, direct attention to specific reports of interest and for

the utilization of research results in regulatory process.

The summaries of these reports are grouped into three categories. The

first category involves the early scoping and background studies and includes:

Survey of Aged Power Plant Facilities, Operating Experience Reviews of LERs

to Identify Aging Trends, workshops to obtain experts opinions and Aging/Risk

Considerations. The second category contains two reports. One on developing

a Methodology for Aging Analysis and a second one on evaluation and the use

of a signature analysis technique (MOVATS). The third category contains

(Phase I) results of aging research on 9 components including: Electric

Motors, Battery Chargers/Inverters, Electrical Cable, Pressure Transmitters,

Diesel Generators, Motor Operated Valves, Check Valves, Auxiliary Feedwater

Pumps and Snubbers. Each report summary has four sections: Background, Summary, Results/Findings, and Utilization of Research Results in'the Regulatory

Process.

This report is considered a living document. That is research results

and additional summaries of reports may be added. In addition, selected

future reports also may be summarized and incorporated periodically.

ii

_ s
I. INTRODUCTION

The results of the Phase I efforts of the Nuclear Plant Aging Research

(NPAR) program have been published since 1984. This report of summaries of

these published reports was developed to provide an efficient method to

communicate program results. This report is considered a living document.

Additional summaries of selected future reports are planned to be incorporated

into this report.

The overall United States Nuclear Regulatory Commission (USNRC) objective

of aging research, as given in the USURC Long Range Plan and the NPAR program

plan (NUREG-1144), is the identification of significant component/environment

aging mechanisms with respect to potential risk to public safety. This

research applies principally to the time-related degradation of electrical

and mechanical components and systems during service; and the potential

impacts of degradation upon public safety. Specifically, the aim of this

research is to develop methodologies to identify such potential impacts on

safety, including the prevention or correction procedures, well in advance

of their actual occurrence.

To meet this objective the Nuclear Plant Aging Research Program Plan

(NUREG-1144) has been implemented under the sponsorship of the USNRC, Office

of Nuclear Regulatory Research. The Nuclear Plant Aging Research Program

Strategy is illustrated in Figure 1, including the Phase I and Phase II

segments of the program. The program goals are to: (a) identify electrical

and mechanical component and systems level aging effects likely to impair

plant safety, (b) identify methods of inspection and surveillance of components and systems that will be effective in detecting significant aging

effects prior to loss of safety function so that proper maintenance and

1-1
I--I

1%)

FIGURE 1 Nuclear Plant Aging Research Program Strategy
timely repair or replacement can be implemented, and (c) identify and recommend acceptable maintenance practices that can be undertaken to mitigate the

effects of aging and to diminish the rate and extent of degradation caused

by aging. The initial effort in the Nuclear Plant Aging Research program

was directed at: reviewing operating experience and experts opinion; establishing a data base containing the known and necessary information for

aging assessments of nuclear power plant components and structures; and

identifying and prioritizing aging issues and future research needs. The

first category of summaries report on the results of this activity. They

include the Survey of Aged Plant Facilities, Operating Experience Review of

LERs to Identify Aging Trends, workshops to obtain experts' opinion and Aging/

Risk considerations.

The second category of summaries, report on analytical and monitoring

techniques that have been developed or are being tested as a part of the

Nuclear Plant Aging Research program. One summary covers a status report on

the development, by SEA, Inc. of a methodology for the evaluation of complex

aging effects. The approach consists of the combined application of an

interactive modeling technique, the N-Square diagram, and system engineering

to develop a methodology for evaluation of complex aging systems interactions.

The other summary covers the report that describes the results of a limited

field test program to evaluate the Motor-Operated Valve Analysis and Test

§ystem (MOVATS) developed and commercialized by MOVATS Inc.

The third category includes summaries of the studies of nine key nuclear

plant components. They include: electric motors, battery chargers/inverters,

electrical cables, pressure transmitters, diesel generators, motor-operated

valves, auxiliary feedwater pumps and snubbers. These reports characterize

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all the time dependent and environmental effects acting on these components.

Also, methods of age degradation detection and monitoring are identified and

evaluated for overall contribution to plant safety through mitigation of aging

effects.

For uniformity of presentation, each of the report summaries contain

four sections: background, a summary, results/findings and utilization of

research results in the regulatory process.

The status of specific activities in the Nuclear Plant Aging Research

program, as of March 1986, is presented in Table I.

1-4
TABLE 2

NUCLEAR PLANT AGING RESEARCH (NPAR)

Technical Progress Reports of Interest to NRR, IE, AEOD, REGIONS

/, COMPONENT/

SYSTEM/TOPIC

CONTRACTOR REPORT

DRAFT

PROJECT

STATUS

REMARKS

NUREG

I Motor Operated

Valves

2 Use of Signature Analysis

Technique-MOVATS

ORNL

ORNL

/CR-4234

/CR-4380

Phase I study

is complete.

Project

complete.

No additional

work is contemplated.

Some progress has been made in

Phase II activity. It will have

recommendations for S&M.

Results included in IE Bulletin

85-03.

Ie

3 Check Valves

4 Aux. Feed Water

Pumps

ORNL

ORNL

/CR-4302 Phase I study

is complete.

/CR-4597 DRAFT Phase I study

is complete.

5 Electric Motors

6. Battery-chargers/

Invercers

BNL /CR-4156 Phase I study

is complete.

Phase II, the effort will be confined to motors inside containment.

The next phase will include testing

of aged unit from Shippingport.

BNL /CR-4564 DRAFT Phase I study

, complete.

7. Snubbers PNL /CR-4279 Phase I study

is complete.
TABLE 2 (Continued)

COMPONENT/

SYSTEM/TOPIC

8 Diesel Generators

CONTRACTOR REPORT PROJECT REMARKS

NUREG DRAFT STATUS

PNL /CR-4590 DRAFT Part I of.

Phase I report

complete.

Have identified of sub-systems and

components susceptible to aging.

Has some recommendation for S&M.

9 Room Coolers PNL /CR- DRAFT Project complete. No

additional

work has been

comtemplated.

Motors, fan belts, cooling coil

have propensity for aging.

I

a% 10 Transmitters FRC

ORNL /CR-4257-2

DRAFT INEL will do

detailed

engineering

studies, i.e.,

Phase I and

Phase II.

These two projects were to review

condition monitoring techniques

only. Detailed aging assessments

and recommendations for S&M will be

addressed in Phase I - Phase II

studies.

11 Cables FRC /CR-4257-1

12 Aging/Risk

Consideration

13 Survey to

Identify Aging

Trends from LERs

14 Results of Aging

Workshop

INEL

PNL

ORNL

SNL

/CR-4144

/CR-3543

/CR-3818

This specific

task is complete.

Project complete.

Project complete.

INEL has new activity to identify

where in plant systems aging is a

safety concern.

Identified components and systems

which have propensity for aging.

Identified susceptible components

from experts' opinion.

.
TABLE 2 (Concluded)

#/ COMPONENT/

SYSTEM/TOPIC

CONTRACTOR REPORT PROJECT

DRAFT STATUS

REMARKS 

NUREG

15 Kick-off

Aging Workshop

16 Survey of Aged

Facilities

SNL

INEL

/CR-0036

/CR-3819

Project complete.

Project complete.

Comprehensive meeting with industry.

Corrosion, erosion, vibration and

contamination have resulted in aging

of fluid mechanical systems.

ORNL =

S&M =

BNL -

I.- PNL = I FRC =

INEL =

Oak Ridge National Laboratory

Surveillance and Maintenance

Brookhaven National Laboratory

Pacific Northwest Laboratories

Franklin Research Center

Idaho National Engineering Laboratory
2.1 Aged Plant Survey NUREG/CR-3819 June 1985 INEL/FIN No. A6389

Survey of Aged Power Plant Facilities

1.0 Background

As part of the overall Nuclear Plant Aging Research Program, the Idaho

National Engineering Laboratory (INEL) was asked to perform a survey of aged

light water reactor (LWR) power plant facilities. They were to determine

what, if any, loss of function can be attributed to aging and to evaluate

the potential for any identified aging process to be significant for LWRs.

The four major subtasks to be undertaken in conducting this survey were: (a)

to identify the facilities to be surveyed; (b) to identify the sources of

information to be used; (c) to design and implement an automated data system;

and (d) to actually conduct the surveys and to identify the components and

environments where aging was a factor in loss of function.

Since it would be impractical to survey all commercial reactors in the

United States to determine the direction future aging research should follow,

the number of plants surveyed was limited by use of plant selection criteria.

The facilities initially chosen included 32 operating commercial nuclear

plants at 23 sites and four test reactors at two INEL sites.

Licensee Event Report (LER) data either have previously been used to

identify aging trends or to estimate gross failure rates for specific plant

equipment. Hence, it did not seem to be particularly fruitful to expand

that effort as part of this study. After reviewing several other documents

and data bases in detail it was decided to use two primary sources of data.

The first of these was the Nuclear Power Experience (NPE) published by the

2-1
S.M. Stoller Corporation, Boulder, Colorado, and the second was all the age

related USNRC Inspection/Enforcement IE) documents published to date.

The published NPE data is a compilation of about 20,000 separate pieces

of information from periodicals, technical papers, technical reports, LERs

and correspondence between plant owner and USNRC pertaining only to plant

operating problems. By necessity the data is condensed somewhat before it

is published, but since the purpose of the NPE is to more fully and more

objectively explain problems and their suspected causes, we judged that the

information obtainable from this source would meet our criteria exceptionally

well.

A review of all the existing IE Bulletins, Notices, and Circulars ( 575

reports and supporting data) for age related problems was conducted. This

source was used because, if an IE document is written on a subject, it is

considered by the USNRC as (a) a potentially wide-spread problems or, (b) of

high enough significance that action should be taken to assure the problem

cannot arise.

2.0 Summary of Approach

The results of this report recommend methods to help formulate a comprehensive research program that will systematically identify aging and service

wear effects which are likely to affect plant safety. The survey centered

on safety related plant systems with regard to component failures from operating histories.

The age related failure information gathered from the plant histories

was analyzed for reoccurring failure patterns. Emphasis was on identifica2-2
tion of specific equipment with high failure rates and of failure mechanism

relationships. The methodology used was to survey eight older commercial

power plants by first analyzing plant operating experiences, as put forth in

the published literature, and then to corroborate these results by actual

field inquiry. This approach did not, nor was it intended to, produce results that could be considered beyond disput in all regards. The intent was

merely to go to the detail necessary to point out, with a basis in fact,

some currently unexplored aging issues and to recommend the direction in

which future aging research should proceed.

As part of this task, it became obvious that a considerable amount of

data associated with age related component failures would have to be filed

and managed. A review of existing automated data files was made and none

were found that appeared adequate for this specialized work. Consequently,

a new computerized data file was initiated for this task.

In a paper by E. J. Brown of the USNRC, Office of Analysis and Evaluation of Operational Data (AEOD), he noted that one of the findings of a previous AEOD study (AEOD/C203) was that after any failure, "plant staff efforts

are directed toward return (of the plant) to operational status rather than

finding the root cause" of the failure. The implication is that the real

causes of failure are not typically being determined and corrected. Mr. Brown

encouraged the industry and the regulatory agencies to use "evidence from

operating plants to identify aging mechanisms" as a realistic approach to

accommodating the aging problem. This directed our research toward identifying

the causes of component failure. However, since valves, pumps, and pipes

were the components displaying the highest failure rates (see figure below)

we limited this portion of the search to these items.

2-3
REPORTED FAILURES BY COMPONENT FOR 4 BWRs AND 4 PWRs

V '' '~~ Note:

co 25 Components with <3%(<14)-

u @ _ K| failures not shown

VL 5 - 13%

to <4 q .- s4 '. 30

3.0 Results/Finding

reported pln ~ ~ eqipen ~ ~ ~ ~ ~~A 53042infudmehncl ytm. hs e

cam rel orothen earansiswas vbasn expcd, rign areement wihther

chemical or solid contamination).

A data sort was made to determine which plant systems showed the most

significant failure rates. From this information three plant systems were

chosen for a more detailed investigation. The systems chosen were: (1) the

Residual Heat Removal (RHR) Systems for BWRs; (2) the Safety Injection System

(SIS) for PWRs; (3) the Cooling Water System (CWS) for PWRs.

2-4
Figures

C

C 

0

_

in

0

0.0_

iu 0)

U t

gn V0

u-co

Reported failures by mechanisms for RHR

systems in 13 BWRs.

C

0

in

.0w

c

C

c c

.2 0

0 en C

w y 5 -

-a Ms

Reported failures by mechanisms for SI

systems in 17 PWRs.

C

O

(n

ccC

_. -

n W

a.-

Uts) 

DE

Reported failures by mechanism for CU'

systems from 17 PWRs.

2-5
Additional findings include:

1. High incidence of component failure in a plant system may or may not

indicate a weakness in a component but rather a change in the system,

its maintenance, or its mode of operation.

2. There appears to be strong correlation between cause of failure for

components and the functional system in which they operate, [e.g.,

failures due to vibration and foreign materials, in Residual Heat

Removal (RHR) Systems, 90%; in Safety Injection Systems (SIS), 61%;

and in the Cooling Water Systems (CWS), 51%. The CWS has an additional

30% due to corrosion and erosion].

3. Data from interviews with personnel from commercial plants suggests

that with certain failure mechanisms (e.g., water hammer, overnormal

vibration, and chemistry control) the heatup and cooldown cycles and

cold shutdown modes of operation are associated with high failure

rates.

4. Test reactor facilities do not experience the same magnitude of

failure due to foreign materials as commercial plants. This is

probably because efforts are made to keep them clean.

Conclusions drawn from the study are:

1. Since commercial power plant system environments are directly responsible for most age related component failures, examination of

individual components to determine failure mechanisms should be

supplemented with aging/systems interaction studies. System design,

maintenance, and operational problems are so predominant that it is

probable that failures due to the aging of component materials could

not be identified with any certainty. Only after the effects of the

major failure mechanisms are mitigated could material analysis,

coupled with the understandings of stressors and environment, yield

definitive results.

2. System cleanliness with regard to foreign materials and chemistry

control should have strict limits placed on it and should be monitored as part of the normal maintenance procedures.

3. Judging from the number of vibration failures evidenced in the survey, flow, and equipment induced vibration is a problem in plant

operation. Prevention of vibration and thermal cycle effects could

be enhanced by anti-rotation features being added to all fasteners

on safety or safety-related components in the plant.

4. Any changes contemplated for the system, or component design, operation, or maintenance must take into account possible adverse effects

on every component in the system and related systems. System and

component interactions are much more prevalent and subtle than most

realize.

2-6
5. Condition monitoring has obvious advantages and should be considered

as part of a comprehensive surveillance program. Because many conditions that govern component performance in todays' plants are system

effects, component condition monitoring alone is not adequate. To

be of maximum benefit each component and system should have its

degradation rate characterized.

4.0 Utilization of Research Results in the Regulatory Process

The results of this and subsequently planned studies will be useful to

the regulatory process and can be factored into this process through support

of implementation of:

a. OCFR.50.49(e)(5) - Aging

b. IOCFR.50.36(e) - Surveillance Requirements

c. 1OCFR.50, Appendix B - Quality Assurance Criteria

2-7
2.2 LER Data-Aging Trends NUREG/CR-3543 January 1984 ORNL/FIN No. B0822

Survey of Operating Experience from LERs to Identify Aging Trends -

Status Report.

1.0 Background

The objective of this study was to review the currently available sources of light-water reactor operating experience information contained in the

Licensee Event Reports (LERs) to identify and evaluate age-related events

and trends. The review focused on time-related degradation mechanisms that

affect mechanical, structural, and electrical systems and/or components

which could result in compromising a safety function. The study was conducted to evaluate the suitability of LERs as a source of data for evaluation

of aging trends and is in response to recommendations made at the NRC-sponsored Workshop on Aging, conducted in August 1982.

The scope of this study was the review of LER(s) and, prior to 1976,

LER predecessor abstracts for failures resulting from identified age-related

degradation mechanisms. An LER is generated by a licensee upon a deviation

from the plant Technical Specifications and generally only includes failures

that affect safety-related components or systems.

Although abstracts of documents dating back to 1969 were reviewed, the

majority of the data was obtained from the more detailed reports available

from 1976 to 1982. The abstracts were obtained from the Nuclear Operations

Analysis Center (NOAC) [formerly the Nuclear Safety Information Center (NSIC)]

file of over 35,000 LERs and LER predecessors. The selected LER abstracts

were reviewed, and data were collected on plant, system, component, subpart,

failure mechanism, severity of failure, and method of failure detection.

2-8
The data were entered into a computer file, which allowed sorts of various

parameters to identify predominant age-related failure mechanisms, affected

systems, failed component/subpart, or other data sorts of interest.

The review process involved elimination of non-aging effects, consolidation of information from multiple abstracts concerning a single event, and

discarding events involving other than commercial nuclear power plants. For

each event judged to be an age-related failure, the reviewer prepared an

input record for entry into a data file established in the Oak Ridge National

Laboratory (ORNL) computer for the aging study project.

2.0 Summary of Data Evaluation

In this study, specific time-related degradation mechanisms are identified a possible causes of a reportable occurrence. Data collected on domestic commercial nuclear power plants covering 1969 to 1982 yielded over 5,800

events attributable to age-related failures. Of these events, 2,795 were

attributable to instrument drift, which are addressed separately in the

report. The remaining events (3,098) were reviewed, and data were collected

for each event, identifying the specific system, component, and subpart; the

information included age-related failure mechanism, severity of failure, and

method of detection of the failure. About two-thirds of the failures were

judged to be degraded, with one-third listed as catastrophic failures. No

events were found to be incipient failures, because and LER is prepared only

on degraded or catastrophic failure conditions that place plant operation

outside the Technical Specifications. The study found that information

desired for evaluation of aging effects (equipment, age, service life, and

environment) was seldom available from LERs. This reflects the intent of

2-9
the LER system as a regulatory instrument, rather than an engineering data

collection system.

Distribution of Aging Study LERs by Year

Total number Percentage examined

Year of LERsa in aging study

1982 4784 12

1981 4632 14

1980 3835 12

1979 3543 13

1978 3567 12

1977 3414 12

1976 2740 13

1975 2518 9

1974 2007 9

1973 1327 ' 8

1972 763 7

1971 447 9

1970 265 11

1969 238 12

1968 218 11

1967 198 12

1966 170 8

aNumber of NOAC accessions for LERs

or LER predecessors including updates and

revised reports, as of March 1983.

2-10
3.0 Results/Findings

The Tables below show the distribution of failure mechanisms for the

3,098 events selected as age-related failures.

Failure Mechanisms for Age-Related Events

Failure Mechanism Number of

events

Wear 522

Corrosion 414

Contamination, internal 382

Contamination, external 331

Fatigue 324

Crack* 259

End of Life* 226

Contamination contact 205

Vibration 165

Stress Corrosion 110

Erosion 102

Other Miscellaneous mechanisms 58

*While not actual failure mechanisms, these

classifications are discussed also.

The 3,098 events described in the proceeding paragraph were each associated with one of 68 system classifications. The most frequently reported

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systems are listed in the Table below with those ten systems representing

53.4% of the events.

Age-Related LERs by System

Number of

System age-related designationa System events Percent

involving

system

ST Emergency core cooling ystc- 227 7.3

and controls

E Emergency generator system and 222 7.2

controls

SD Contaimwnt solation system nd 215 6.9

controls

PC Chemical/volume control and 212 6.8

liquid poison systems nd

controls

VA Station service water systems 144 4.6

and controls

Ca Coolant recirculation systems 13S 4.5

and controls

Si Containment heat removal systems 134 4.3

and controls

HI ain steam supply system and 130 4.2

controls (other than WR

steam supply)

CT Residual heat removal systems 129 4.2

and controls

SA Reactor containment systems 104 3.4

SUBTOTAL 1655 53.4

Balance of systems (56) 1443 46.6

TOTAL 3098 100.0

aFro Itruction for reparation of Data Entiry

Euent Report (LR) File. NUREC-0161, July 1977.

Sheets for Licens.

. 2-12
Age-Related LERs by Component

Rank Component Number Percent

1 Pipe 446 14.4

2 Valves, Other 277 8.9

3 Monitor 273 8.8

4 Valve, isolation 243 7.8

5 Pump 239 7.7

6 Diesel 151 4.9

7 Valve, Check 101 3.3

8 Steam generator 87 2.8

9 Heat exchanger 83 2.7

10 Snubbers 69 2.2

Age-Related LERs by Part

Rank Part Number Percent

1 Weld 324 10.5

2 Miscellaneous

subcomponent 266 8.6

3 Pipe and tubing 233 7.5

4 Valve seat 212 6.8

5 Contacts 192 6.2

6 Packing, seal 163 5.3

7 Wall (pipe) 137 4.4

8 Shaft 113 3.6

9 Housing 102 3.3

10 Bearing 69 2.2

Another study also utilized the LER data. The initial Accident Sequence

Precursor (ASP) program examined about 19,400 Licensee Event Reports (LERs)

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from the period 1969-1979. The initial screening of these LERs resulted in

selection of 529 events for detailed review. The detailed review identified

169 events considered to be precursors of potential severe core accident

sequences. These events were either initiating events for the sequences or

failures that could have affected the course of postulated off-normal events

or accidents. These are the events that the ASP program subsequently quantified.

Eleven events, from the common LER data base were selected in both the

ASP and Aging studies. These events are summarized in the table on the

following page. The reason only 11 events are common to the results of the

two studies is due to the strict selection criteria of the ASP effort.

Unless the event caused a transient or affected a safety function response

to a transient of interest, it was not included in the ASP events. The ASP

program also focused on events where multiple failures occurred, particularly

common-cause failures. Although aging processes affect all components to

some degree, there generally is not a coordinating mechanism to cause failure

to occur simultaneously in multiple components. An exception to this is

standby components where demand for operation itself provides the coordination.

The LER system is not an engineering data collection device; rather, it

is a regulatory event reporting scheme for the purpose of measuring licensee

compliance to their respective plant Technical Specifications. Consequently,

any analysis of events and failures from LERs must be tempered by the nonstatistical nature of the data.

This study utilized the LERs to determine failed components, the age-related failure mechanisms responsible, the severity of the ailure, and the

2-14
SUMMARY OF ASP/AGING STUDY COMMON EVENTS

Event

No.

1

2

3

4

5

6

7

8

9

10

11

Accession

NO.

152563

147400

120293

105540

97107

89205

60227

44751

128906

124222

93553

Event description

Steam generator tube break

Both main steam stop valves

fail

Plant service water strainers plugged

Low flow feedwater line

severed

Safety injection valve

failed to open

Reactor coolant pump shaft

failed

Steam isolation valves fail

to close '

Three of four safety system

level sensors fail

HPCIb fails due to governor

actuator drift

Six vain steam relief

valves fail to lift

Diesel generator fails due

to time set point drift

Failure mechanism

Wear from external

debris

Corrosion product

buildup

Contamination

buildup

Vibration

Corrosion product

buildup

Fatigue

Contamination of

pilot valves

RCSa crud buildup

Drift

Drift

Drift

aReactor coolant system.

igh-pressure coolant injection.

2-15
failure detection methods for reportable occurrences due to possible age-related failures. The data are by no means statistically accurate, but serve

to indicate possible areas where further study should be performed to better

characterize aging of components and systems.

These general conclusions can be drawn from the study.

1. Surveillance testing is an effective technique for detecting degradation.

2. Study data support recent ASME Code emphasis on pump and valve testing.

3. Confirmation of In-Plant Reliability Data System (IPRD) study emphasizes

reliability of pumps, valves, and diesel generators.

4. More research is needed to characterize instrument drift causes and

piping failures.

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2.3 AGING/RISK CONSIDERATIONS NUREG/CR-4144 APRIL 1985 PNL/FIN NO.B2865

Importance Ranking Based on Aging Considerations of Components Included

in Probabilistic Risk Assessments

1.0 Background

This study utilizes existing probabilistic risk assessments (PRAs) to

gain insights on the relationships between aging of nuclear power plant

components and public risk. A method is developed and applied to determine

the potential risk significance of aging effects. The study objective is to

identify components in nuclear power plants that adversely affect risk if

aging processes decrease component reliability or degrade performance characteristics. This objective does not include identifying specific aging

processes or describing aging effects (time dependence) on component failure

rates.

PRAs are a method to mathematically estimate the likelihood and the

consequences of potential accidents at nuclear power plants. In the process

of performing a PRA, the potential accident initiators (LOCAs, transients,

loss-of-offsite power, etc.) are identified and their likelihood quantified.

The safety systems and their support systems that must function to safely

shut down the reactors are then identified for each initiator. The safety

systems and their support systems are modeled using event tree and fault

tree methodology. The safety systems generally considered in a PRA are the

reactor protection system, main and auxiliary feedwater systems, high pressure

and low pressure injection systems, residual heat removal systems, containment

sprays, containment coolers, and accumulators. Support systems include

2-17
electric power, service water, and engineered safety feature actuation systems.

Operator actions are also included in the-models.

The event tree and fault tree model solutions determine the combinations

of component failures that lead to a core melt for each of the initiators.

The combination of an accident initiator and the system failures that result

in core melt is referred to as an accident sequence. The combinations of

individual component failures that cause "the required systems to fail" is

referred to as a cutset.

The probability of each individual component being unavailable is referred

to as its unavailability. The probability of the cutset is the product of

the unavailability of the individual events. The frequency of an accident

sequence can be approximated by the sum of all the cutsets that result in

failures of the same set of safety systems. The overall plant risk is simarily

approximated by the sum of the accident sequences, or equivalently, the sum

of all the accident cutsets. %

In addition, a probability of containment failure can be assigned to

each accident sequence. In some PRAs, the consequences of accident sequences

are evaluated in terms of man-rem, fatalities, or economic impact.

The scope of PRAs vary greatly. Some consider internal events only;

others include seismic events, floods and fires, etc. The depth of the

analyses of the systems and sequence consequences also varies considerably.

The scope of the PRA, as well as the level of detail considered, limit the

information that can be extracted from the analysis.

The aging program in general and the risk significance task in particular

can benefit from the products of other NRC and industry programs including

the Accident Sequence Evaluation Program (ASEP) and the data gathering programs

2-18
(LER's, NPRDs, and others). The ASEP program is designed to provide analysis

of the -dominant accident sequences for most LWR's in the United States. As

a part of this program the cutsets for the dominant sequences are identified

and risk importance measures calculated for a large number of components.

When the results are made available it will be possible to apply the methods

outlined in this report to a broad range of plants. This wll provide a

good basis for assigning priorities to component classes based on the risk

estimates at a large number of plants rather than the three (Oconee, Calvert

Cliffs and Grand Gulf) analyzed here. The approaches used in ASEP will

allow identification of the most risk significant components and systems

based on plant design and other operating characteristics. This information

will assist in making specific recommendations as to what type of inspection

and preventive maintenance programs will be most effective in controlling

risk at different plants based on plant design.

The output from this study cn be combined with ocher studies (data,

analytical or experimental) that identify the components that are most susceptible to aging mechanisms. The combination of identification of risk

significance and aging susceptibility will provide a good basis for effectively

focusing resources.

2.0 Summary of Approach

The approach taken in this study uses the results of existing probabilistic risk analyses (PRAs) to gain insights about the relationship between

risk and component aging. PRAs performed to date do not explicitly model

risk as a function of time, but calculate an average risk level. This report

defines a risk importance measure, that measures the sensitivity of risk to

2-19
changes in a component failure rate. This measure is the partial derivative

of the core melt frequency with respect to the failure rate of a specific

component. Those components that have the highest sensitivity have the

potential for causing the greatest change in risk if their failure rates

increase due to aging.

The information presented in a standard PRA does not include time dependent effects. In determining the risk level at a plant, PRA's generally

use a time averaged unavailability. Aging issues deal with the time dependent

nature of risk. This limits the nature of the information that can be extracted from a PRA without extensively modifying the PRA. This report suggests a

method for determining the potential risk significance of aging effects chat

is based on determining the sensitivity of risk to increases in failure

rate. This adaptation of PRA results enables us to identify the components

that have the most significant impact on risk if their failure rates increase

due to aging effects without describing the time-dependent behavior of the

failure rate. The information extracted from PRA's in this manner can be

quite useful in guiding research efforts if used in context.

The term component' can be considered to be individual pieces of hardware, e.g., a valve casing, a valve stem, wiring, etc. The component" can

also be considered as a functional unit such as a motor operated valve that

consists of a number of component parts. Components as defined in most PRAs

and in this report represent functional units. A motor operated valve for

instance is interpreted as consisting of the valve, the motor operator, the

circuit breaker, and the electrical cable and control circuitry specifically

associated with the valve. A brief description of the component boundaries

for each type of component is included in Table 1.

2-20
Frequently, components are subject to a number of different failure

modes. For instance, motor operated valves could fail to function by several

modes including: failure to open, failure to close, and gross leakage.

Table 1 also includes the most important failure modes for each component

type. These failure modes represent component functional failures and do

not indicate the root cause of the failure or the failure mechanism. From

an aging perspective, the time dependent processes that lead to a functional

failure are of the most concern.

3.0 Results/Findings

The results of risk aging sensitivity measure calculations are presented

for plants analyzed as part of the Reactor Safety Study Methodology Application

Program (RSSMAP). These studies represent limited-scope PRAs in that they

do not include external events and do not specifically include analysis of

piping and wiring. The plants icluded in this analysis are two PWR's,

(Oconee and Calvert Cliffs) and one BWR (Grand Gulf). Also included are

bounding calculations for three other components: a reactor vessel, steam

generator tubes, and snubbers.

Table 2 presents the aging sensitivity rankings for component groups at

PWR's. These results are obtained by adding the results of the component

groups at the two PWR's. Check valves of the auxiliary feedwater systems

and breakers/contactors and trip-relays/trip modules of the reactor protection

system have the highest potential risk impact as measured by the aging sensitivity measure.

2-21
TABLE 1: COMPONENT BOUNDARIES

Component Boundary Failure Modes of Concern

Pumps

(Electric)

Includes pump, motor, and the

control circuitry and electric

power components specificl]y

associated with the pump- 13

* failure to start on demand

* failure to run

* gross leakage

Pumps

(Turbine

Driven)

Includes pump, turbine, and

control circuitry specifically

associated with the pump.

* failure to start

* failure to run

* gross leakage

on demand

Motor Operated

Valves

Control Valves

(Air Operated)

Check Valve

Relief Valve

Circuit

Breaker/

Contactor (RPS)

Relay (RPS)

Trip Module/

Includes valve, motor operator

and the control circuitry, and

electric power components

specifically associated wtih

the valve.(L)

Includes the valve, the air

actuator, and the control circuitry specifically associated

with the valve. %

Includes the check valve only

Includes the relief valve only

The circuit breakers that

provide power to the control

rod drive mechanisms.

The relays that actuate the

trip breakers on signal from

trip module.

Includes the sensors, cables,

bistables, and relays that

measure plant parameters such

as reactor coolant pressure

and send a trip signal to trip

breakers.

* failure to open on demand

e failure to remain open

* Failure to go to the

"fail safe" position on

signal

* failure to provide control

capability

e failure to open

e stuck open

* failure to open

* failure to open

* failure to send trip

signal when plant parameters require

2-22
TABLE 1: (Concluded)

Component Boundary Failure Modes of Concern

Actuation Includes the sensors, cables, * failure to send ESAS

Channel/ bistables, and relays that signal when required

Subchannel measure plant parameters and

send out an Engineered Safety

Feature actuation signal.

Battery Includes the battery and the s failure to provide DC

battery charger. power to components

requiring DC power (given

loss of AC power)

Diesel Includes the diesel and its . failure to provide AC

Generator support sytems (lube oil power to components

cooling, fuel supply, etc.). requiring AC power (given

loss of off-site power)

Room Coolers Includes the fan and cooling * failure to cool pump room

coils that provide room cooling to pump rooms.

(1) The electrical components specifically associated with the pump or motor

operated valve would include the connector, cable, and circuit breaker that

power the motor, but does not include the electric power distribution

system that feeds the circuit breaker.

2-23
Tabl-e 2. Aging sensitivity of component groups in PWR's.

Rank Type System Aging Sensitivity

1 Check Valves

2 Circuit Breaker/Contractor

3 Trip Relay/Trip Module

4 Control Valves (air operated)

5 Motor Operated Valves

6 Pumps

7 Motor Operated Valve

8 Motor Operated Valve

9 Pumps

10 Actuation Channels

11 Check Valve

12 Motor Operated Valve

13 Turbo Generator/Diesel

Generator

14 Check Valve

15 Batteries

16 Pumps

17 Room Coolers

18 Pumps

19 Relief Valves

20 Check Valves

Auxiliary Feedwater

Reactor Protection

Reactor Protection

Auxiliary Feedwater

Auxiliary Feedwater

Auxiliary Feedwater

High Pressure ECC

Service Water

Service Water

Safeguards Actuation

Low Pressure ECC

Low Pressure ECC

Emergency Power

5.5

3.2

2.2

1.4

1.4

1.4

4.7

2.9

2.6

2.1

1.8

1.7

x 10-3

x 10-3

x 1o-3

x 10-3

x 10-3

x 10-3

x 10-4

x 10-4

x 10-4

x 10-4

x 10-4

x 10-4

1.6 x 10-4

High Pressure ECC

Emergency Power

High Pressure ECC

Service Water

Low Pressure ECC

Reactor Coolant Pressure

Boundary

Service Water

1.0 x

7.3 x

5.3 x

3.3 x

2.0 x

1.5 x

10-4

10-5

10-5

10-5

10-5

10-5

1.3 x 10-5

---

2-24
Table 3 presents the combined results for component types of the two PWR

plants. Check valves, circuit breakers/contactors, trip modules/actuation

channels, motor operated valves, pumps, and air operated control valves have

the highest values of the aging sensitivity measure.

Table 4 shows the combined results for component groups at the Grand

Gulf BWR. Motor operated valves of the low pressure emergency core cooling

system and service water system and actuators of the engineered safety actuation system have the highest potential risk impacts as measured by the aging

sensitivity measure.

Estimates are listed in Table 5 for the aging sensitivity measure for

three additional component types: the reactor vessel, steam generator tubes,

and snubbers using existing PRA's and related studies. The calculations are

bounding calculations intended to compare the importance of these components

to other components at the plant. Table 5 presents the results of these

calculations.

Reactor Vessel

The reactor vessel has the highest potential impact on risk of any

component in the plant. PRA's generally make the conservative assumption

that a failed reactor vessel results in an uncoolable configuration that

leads to core meltdown. The aging impact as measured by the aging sensitivity

measure is high compared to the other components in the plant.

Steam Generator Tube

A rupture in a steam generator, as an initiating event, results in a

small LOCA and consequently loss of heat removal capability of one steam

2-25
Table 3. Aging sensitivity of component types in PWR's.

Rank Type Aging Sensitivity

1 Check Valves 5.8 x 10-3

2 Circuit Breaker/Contactor 3.2 x 10-3

3 Trip Module, Relay/Actuation Channel 2.4 x 10-3

4 Motor Operated Valves 2.3 x 10-3

5 Pumps 1.7 x 10-3

6 Control Valves (air operated) 1.4 x 10-3

7 Turbo Generator/Diesel Generator 1.6 x 10-4

8 Batteries 7.3 x 10-5

9 Room Coolers 3.3 x 10-5

10 Relief Valves 1.5 x 10-5

TABLE 10. Aging Sensitivity of Component Groups at Grand Gulf

Rank Type System Aging Sensitivity

1 Motor Operated Valves Low Pressure ECC 2.3 x 10-4

2 Motor Operated Valves Service Water 1.3 x 10-4

3 Actuators Safeguards Actuation 9.9 x 10-5

4 Pump Service Water 5.9 x 10-5

5 Check Valves Service Water 5.9 x 10-5

6 Motor Operated Valves High Pressure ECC 5.4 x 10-5

7 Check Valves High Pressure ECC 5.4 x 10-5

8 Check Valves Low Pressure ECC 2.8 x 10-5

9 Batteries Emergency Power 2.4 x 10-5

10 Pump Low Pressure ECC 2.4 x 10-5

11 Pump/Tubine Pump High Pressure ECC 1.3 x 10-

12 Diesel Generator Emergency Power 9.5 x 10-6

13 Relief Valves Reactor Coolant Pressure 2.6 x 10-6

Control

2-26
generator. In this situation, core cooling requirements generally are the

operation of the auxiliary feedwater system and at least one high pressure

injection pump. Table 6 gives an estimate of the tube aging impact based on

the cooling requirement for four plants. Consistent with the aging sensitivity

measure definition, these estimates are based on simply adding the conditional

failure probabilities of the auxiliary feedwater system and the high pressure

injection system. The average value from these four plants is included in

Table 5. The potential risk impact of steam generator tubes as measured by

the aging sensitivity measure is higher than that of the standby components.

Snubbers

In order to determine the aging impact of snubbers the results of the

Seismic Safety Margins Research Program were reviewed. The case of snubber

failure is specific in that it has been done for the Zion plant.

The risk associated with snubber failures is characterized by an increased

likelihood of a LOCA induced by an earthquake. The earthquake also degrades

the safety system that cools the core in the event of a LOCA. In this situation, it is assumed that snubber failure will result in a large or medium

LOCA for any earthquake with a magnitude larger than design basis. The

dominant core melt sequences for an earthquake induced LOCA, contain the

failure of the Safety Injection System (SIS) to cool the core.

The aging sensitivity measure for snubbers as calculated here is moderately high. This calculation is an approximation and subject to high uncertainty.

Further, the information used is for only one plant that is not located in a

high seismic activity zone. The potential risk significance of snubbers

will be very site-dependent in general.

2-27
TABLE 5. Aging Sensitivity Measures for Selected Components

Component Aging Sensitivity

Reactor Vessel 1

Steam Generator Tube 3 x 10-3

Snubber 1.8 x 10-5

TABLE 6. Aging Sensitivity Measure Calculations for Steam Generator Tubes

Plant Name Cooling Requirements Aging Sensitivity

ANO 1/2 EFWS 6.5 x 10-4 +

1/3 HPIS 4.0 x 10-4=

1.1 x 10-3

Oconee 1/2 AFWS 2.4 x 10-4 +

1/3 HPIS 1.4 x 10-3=

1.6 x 10-3

Calvert Cliffs 1/2 AFWS 3.0 x 10-3 +

1/3 HPIS. 1.7 x 10-3 =

4.7 x 10-3

Sequoyah 1/3 AFWS 4.3 x 10-5 +

1/3 HPIS 3.5 x 10-3=

3.5 x 10-3

2-28
Limitations and Assumptions

The most important limitations of this study are the limited number of

plants analyzed and limiting the scope of components studied to those analyzed

in the PRA's. The analysis is limited to the effects of complete failure

(loss of function); the effects of degradation are not specifically addressed.

Also common-cause failures attributed to aging are not specifically addressed.

This report considers only some of the components that are potentially

important to risk. Components whose primary purpose is to mitigate the

consequences of severe accidents such as containment spray nozzles, piping

and pumps were not considered. The importance to risk of components that

mitigate accident consequences is not easy to determine in light of the

large uncertainties associated with the phenomenology and fission product

behavior of severe accidents. Structural components such as the containment

and containment lining were also not considered. Piping and wiring are not

explicitly considered in these analyses and components such as the reactor

vessel, steam generator tubes and snubbers are treated only superficially

for example purposes.

4.0 Utilization of Research Results in the Regulatory Process

The results of this study can be used to provide a relative sense of

the unavailability of selected components. Future efforts on the time dependent portion of aging for these component types are needed to more completely describe the risk impact due to component aging.

2-29
3.1 A METHOD FOR AGING ANALYSIS NUREG/CR- Jan. 1986 SEA CONSULTANTS,INC.

A methodology for Evaluation of Complex Aging Effects on Systems' and

Components' Performance and Interactions.

1.0 Background

A primary objective of the Nuclear Plant Aging Research (NPAR) Program

is to analyze the nature and extent of aging on safe reactor operation.

Simulation of the aging process is a method by which understanding can be

obtained. Correct simulation is predicated upon accurately modeling the

simultaneously occurring inherent and operational aging effects (stressors)

in combination with typically complex physical interactions that result from

the component and system design configuration. Similar to finite element

methodologies, aging simulation must begin with reduction of the problem to

the lowest state or element where aging effects can be reliably demonstrated.

The lowest element for components important to safety, for our purpose, is

the lowest unit with aging significance (LUWAS). A LUWAS may be defined as

the lowest constituent part that is susceptable to aging effects that must

perform within specification for safe nuclear operation. As with finite

element analysis, the initial step is development of the mathematical representation or transfer function of the input-output ration of the LUWAS. The

transfer function represents the LUWAS behavior at the beginning of life or

period prior to installation and startup. The transfer -function is empirical

(ratio of input to output) and no knowledge of the LUWAS internal structure

is required. At the component level the combination of all LUWAS transfer

functions provide a relationship describing the dynamics of the component.

3-1
Aging is a time domain phenomenon that alters the state of the LUWAS and

can be represented as a set of state variables (temperature, radiation, voltage, humidity, position, etc.). Utilizing time-domain methods, the transfer

function is written as a differential equation in terms of the state variables. Empirical relationships, government and industry supported test programs and plant experience are the bases to establish the differential changes

in LUWAS response and the resulting component behavior. State variables

representing the time domain aging effects provide basis for prediction of

the future state and output of the component.

The NPAR strategy recognizes that the evaluation of the affects of

aging on safety reactor operation must also include system interactions.

Evaluation of spatially and/or functionally coupled system interactions

require a methodology that is structured and interactive such that system

interdependencies are reliably displayed. Aging effects, analogous to system

interactions, are the result of internal and external interactions within a

component. Evaluation of aging effects also require a structured and interactive approach. The N-Square diagram is an interactive modeling technique

that is both structured and interactive.

An evaluation of complex aging effects on components and systems performance and systems interaction is developed under the auspices of the NRC's

Office of Nuclear Regulatory Research, Nuclear Plant Aging Research NPAR)

Program. To assist the Nuclear Plant Aging Research purpose, a comprehensive

approach has been developed that consists of the combined application of an

interactive modeling technique, the N-Square diagram, and systems engineering

to develop a methodology that facilitates understanding and evaluation of

3-2
complex systems interactions, such as those that result from the effects of

aging.

2.0 Summary

This report presents a comprehensive approach for evaluating complex

aging effects on components and systems performance and systems interaction.

This approach was developed under the auspices of the NRC Nuclear Plant

Aging Research (NPAR) Program and consists of the combined application of an

interactive modeling technique (the N-Square diagram) and systems engineering.

The resulting methodology facilitates analysis of complex systems interaction,

with specific illustrative applications to components (differential pressure

transducer, compressor and an invertor) and to systems (High Head Safety

Injection system and the low pressure coolant injection (LPCI) mode of the

Residual Heat Removal system). Emphasis is placed on evaluating the effect

of component aging on functional and spatial systems interactions.

The systems interaction methodology consists of development of several

levels of N-Square diagrams to obtain a dynamic plant model at the plant

function level. Consecutive levels of N-Square models are developed to

reduce the system to the component level and then to the Lowest Unit With

Aging Significance (LUWAS). From here the effects of postulated and empirically generated age stressors can be assessed, as potentially critical

systems and component interactions are identified.

The results of a comprehensive model of the components and systems

interactions required to complete a safety or safety-related plant function

could be used to identify inspection, surveillance, and monitoring methods

or corrective actions needed to assure plant performance.

3-3
3.0 Results

3.1 The NSQ Program

The approach taken in Phase I was to design and develop an N-Square

analytical tool using the Turbo Pascal language, version 3.01A. Turbo Pascal

is a standard pascal implementation that supports programming access to the

PC-DOS/MS-DOS operating system, (the hardware interface to the IBM PC) and

the math co-processing chip. Also, the Turbo Pascal compiler leads to code

which executes extremely fast (about 3 times faster than the Lotus 1-2-3

example investigated).

The program has been called NSQ, and consists of a single executable

file (NSQ.COM), less than 36K bytes in size, and two screen image files.

When invoked, NSQ presents a screen permitting up to 4 LUUAS components to

be identified, and a 4 x 4 - 16 cell array of transfer functions to be entered. All models were developed and tested within this prototype limitation.

The user can also open up and work within 4 "windows" that overlay the

N-Square matrix. These windows are invoked through use of 4 of the 10 function keys. Each window pertains to a group of parameters that relate to the

N-Square calculation.

Once the model has been entered, it can be saved on a disk, again by

means of a function key. Similarly the model can be retrieved from disk for

manipulation, much like a word processor. Use of a function key begins the

execution of the model. Once started the program switches into color graphics

mode (199 x 359 pixels) and displays the magnitude of up to 3 color-coded

LUWAS as a function of time. An N3 model (simple R-L-C circuit diagram)

with 100 time steps runs in about 30 seconds, and an N4 model (the differential pressure transmitter) with 50 time steps executes in 110 seconds

3-4
(1 minute 50 seconds). These benchmarks are for an 8088 CPU without 8087

co-processing support.

A number of improvements to the program are under consideration to

extend the software beyond its prototypical testing stage.

3.2 Illustrative Component Analysis

The results of the Phase I program are presented through application of

the methodology, presented on three (3) safety related components common to

a majority of commercial reactor safety systems: 1) a differential pressure

transmitter, a primary component in LOCA and ESFAS instrumentation and control

systems; 2) an air compressor used in instrument air systems; and 3) a 125

Vdc to 120 Vac invertor, typically applied as a 120 Vac emergency power

source. These examples are provided for demonstration only. The vendor

documentation utilized in the preparation of these examples is uncontrolled

and no representation of product acceptability or plant impact should be

pursued. The N-Square diagram and the summary table of Independent Stress

Effects are illustrative of the evaluation of the differential pressure

transmitter.

3.3 System N-Square Diagrams

Figure 1 and Table 1 are N-Square diagrams representing the functional

and spatial system interactions for reactor core water level maintenance for

a LaSalle vintage BWR (LPCI) and a Watts Bar vintage PWR (HPSI). It is easily

3-5
I.,~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~

~~~~~~~~C w N A 1 1 6

52 a dzCELLtI II CAPACITApCE 0' c 6,2 6,3 ~~~~~~~~~~~~~~~CLL 2

. < RE~O LATED 

D FF FVpp (C 2-cI )OIFFERETIAL 

Process Reference IC3

(a)VCE) B)Q4 03/(RCE)03 = (1C)Q3Pressure Pressure (I B)Q4 =(V CE)Q3 /(R CE)Q33C4TR

~UI~ =~(vcE) ( 002 3/(Rc)QJ IADWIRES A A 9.A.rJ 9 ( 1 ~~C Q 03/ CE CE Q 3 R 18 CAAIO tO IEVEATE 

QH ) (IB)03 = (IIFFZ XAG)/ 4 .7K -

B)Q3 G 01 FF SEN~~~~~~~~~~~~~~~fI~lSIG
INDEPENDENT STRESS EFFECTS

DEVICE: Pressure Transmitter I.Layne-6/85

LUWAS I STRESS I EFFECTS OF LUWAS | EFFECTS ON PERFORMANCE | RELATIONSHIP

1- ISOLATING

DIAPHRAGM

2 - PROCESS

FLANGE

3 -SILICON

OIL

4 -SENSING

DIAPHRAGM

5 -CAPACITOR

PLATES

6 - RIGID

INSULATION

7 -- CELL

HOUSING

8 - ELECTRONIC

HOUSING 

COVER

9 - LEAD WIRES

Neutron

Radiation

None

Gamma Radiation

Temperature

Neutron

Radiation

None

Enbrittles stainless

diaphragm.

1. Gassing

2. Viscosity increases.

3. Viscosity decreases.

Embrittlement

Decreases diaphragm deflection per unit pressure which

reduces signal for given pressure.

la. Increases copressibility which reduces coupling

between sensing and isolating diaphragms.

lb. Dielectric constant decreases, causing capacitance

to decrease which increases signal current.

Ic. Decrease in capacitance causes oscillator frequency

to increase and Vpp to decrease.

2. None (sonic velosity s unchanged).

3. None (sonic velosity is unchanged).

Decreases deflection per unit pressure.

None - Capacitance between capacitor plates and

housing changes, but circuit is not grounded to

housing.

Capacitance between leads changes.

I a K x Deflection K* Sig rT

la. Response time K x f(sonic

velosity)

lb. K x Iref

lc. fpp C j+r

C1 - C2

C1 . C2

Gamma Radiation Dielectric constant changes.

2. None

3. None

I a K x Deflection 

SigTjn

None

C K x f(gama radiation) I

Page I of 3

None

None

Gamma Radiation Properties of insulation

changes.
INDEPENDENT STRESS EFFECTS (Continued)

I.Layne-6/85

LUWAS STRESS EFFECTS OF LAS | EFFECTS ON PERFORMANCE T RELATIONSHIP

In - 0 RINGS

11 - Oscillator

Control

o IC-1

o Resistors

12 - Oscillator

o Switching

Transistor Q1

o Resistors

o Transforrer

o Capacitors

13 -Voltage

Regulator

o ener

o 1C2

Ganma Radiation

Temperature

Gama Radiation

@ 106 Rad

Neutron

Radiation

Insignificant

Gamma or Neutron

Radiation

Insignificant

Insignificant

Embrittlement and change in

dimensions.

Hlgh temp will have a similar

but lesser effect as gamma

radiation.

Changes properties of semiconductor.

Changes electronic characteristics.

Increases freq of switching.

Moisture could enter instrument and cause degraded

function or failure.

Changes freq/voltage 41 which changes Iref. but Idiff

does not change since XCI and XC2 change dentically.

Changes freq/voltage or causes failure.

Literature search or tests are

required to determine whether 0

rings in compression will lose

their sealing ability at the

radiation and temperature levels

encountered. Enter as a table.

I

Run SPICE'

table.

Run SPICE'

table.

and enter results as a

and enter results as a

Changes Iref but not Idiff which changes

back will mitigate.

Isig. FeedGamma or Neutron Changes capacitance within

Radiation tolerance.

Changes freq which changes Iref which changes Isig.

Feedback will mitigate.

Changes biasing and setpoints which could cause malfunction or failure.

Changes voltaqe approx. 1% which could cause malfunction.

Run SPICE" and enter as a table.

Run "SPICE" and enter as a table.

Run "SPICE" and enter as a table.

Run "SPICE" and enter as a table.

Radiation Changes regulated voltage.

Gamma Radiation Changes

(106 Rad) ties.

semiconductor properPage 2 of 3
INDEPENDENT STRESS EFFECTS (Concluded)

I.Layne-6/85

LIMAS { STRESS . EFFECTS OF LAS EFFECTS ON PERFORMANCE RELATIONSHIP

I .

Voltage

Regulator

fCont'd)

o Resistors

o Capacitors

14 -Zero Span

Adj

o Resistors

o Pot

15 - Current

Control

o C3,Q2,Q3

o Resistors

o Capacitors

16 - Current

Limiter

o Q4

o Resistors

o Capacitors

17 - Linear Adi

o Diodes

o Resistors

o Capacitors

None

Insignificant

None

None

Gamma Radiation

9 106 Rad

None

Insignificant

Gamma Radiation

None

Insignificant

Changes semiconductor proper- Improper signal current generated which could cause

ties. I loss f calibration or failure.

Changes semiconductor properties.

Destroy Junction.

Change capacitance.

Allows overcurrent condition or improperly limits

signal current.

Failure of linearity circuit.

The first order non-linearities of capacitance as a

function of pressure would not be correctly compensated.

Run SPICE" and enter as a table.

Run "SPICE" and enter as a table.

Run "SPICE' and enter as a table.

Run "SPICE" and enter as a table.

Gamma

None

Gamma

Radiation

Radiation

Page 3 of 3
0

I I

I 

I. ii I

S I t I, I I

I. I I. S S

I I I.
-- - --.- - -- -- a - - .-

I

I I

I~~~~~~~~~~

i~~~~~~~~~~~~~~~~~~

l In

II!

I

iI

I

I i

I

I q

: I4

II

I

I

i

I

I

III
determined by visual observation that system components such as circuit

breakers, LOCA initiation instruments and system interlock components have

multiple interactions in the performance of reactor core water level maintenance. The detailed analysis is discussed in the report.

4.0 Utilization of Research Results in the Regulatory Process

4.1 Application for Inspection, Surveillance and Monitoring Methods (ISMM)

The preoperational and startup phase of a commercial nuclear power plant

provide an initial insight into unplanned plant unique system and component

interactions. During this phase of plant construction, components and systems

are tested for conformance to test specifications which are developed to

demonstrate proper performance. Nonconformances discovered during acceptance

testing typically result from adverse interaction(s) that are readily detected.

The system interaction model for effect of aging can identify potentially

critical interactions within components or systems that directly affect

plant function. The results of a comprehensive model of the system and

component interactions required to complete a plant function, (e.g., water

level, depressurization, scram, etc.), could identify ISMM actions or corrective actions needed to assure plant performance.

4.2 Applications for PRA Enhancements

The nuclear industry and Nuclear Regulatory Commission rely on probabilistic risk assessment (PRA) for assurance of acceptable risk to public

health and safety during plant operation. PRA seeks to quantify the public

risk due to plant operation. Risk quantification is typically achieved

3-12
through use of component failure rates obtained from plant specific data,

Mil. Std 217 and IEEE 500. The nature and extent of the aging process on

component failure rates has been acknowledged as an important element in

risk quantification. Statistical testing to determine the nature and extent

of the aging process on the component failure rate and performance is necessary for acceptable PRA quantification and removal of uncertainties associated

with environmental qualification. Recognizing the complexity and cost of

regulated statistical testing, the N-Square interactive model at the LUWAS

level is a cost effective methodology to simulate and quantify the effect of

aging on component behavior.

3-13
3.2 MOV Analysis & Test System (MOVATS) NUREG/CR-4380 Jan. 1986 ORNL/FIN

No.B0828

Evaluation of the Motor-Operated Valve Analysis and Test System (MOVATS)

to Detect Degradation, Incorrect Adjustments, and Other Abnormalities

in Motor-Operated Valves.

1.0 BACKGROUND

Current surveillance requirements for determining the operability of

motor-operated valves (MOVs) installed in nuclear plant safety systems are

described in plant Technical Specifications. These consist of periodic

functional tests in accordance with the in-service inspection provision of

Section XI Division 1 of the ASME Boiler and Pressure Vessel Code. Numerous

events have been documented in which MOVs have failed to operate as required

either during the above surveillance tests or as a result of actuation of

safety systems during plant operation. This has raised concerns about the

general state of readiness of MOVs to perform their safety function under

all anticipated accident conditions.

Oak Ridge National Laboratory (ORNL), as a part of its support of the

NRC/RES Nuclear Plant Aging Research (NPAR) program, has been investigating

MOV operating experience in nuclear plants. The purpose is to identify the

factors which contribute to MOV failure, and to evaluate and recommend suitable monitoring methods. These methods may be used to detect and diagnose

degradations and other abnormalities in MOVs prior to loss of operability.

One such monitoring method that has been identified is the Motor-2perated

Valve nalysis and Test ystem (MOVATS) developed and commercialized by

MOVATS Inc. of Marietta, Georgia.

3-14
2.0 Summary

This report describes the results of a limited field test program carried out as part of the ORNL/Nuclear Plant Aging Research program to learn

what MOVATS can provide about safety related MOV operational readiness above

and beyond the currently used ARME Section XI methods. The limited test

program was carried out at the request of NRC/NRR to support their ongoing

review of plant in-service testing process and also to assist in resolution

of generic issue II.E.6.1, "In-Situ Testing of Valves."

The test program consisted of obtaining signatures, using the MOVATS

method, from 36 MOVs located in four commercial nuclear power plants. The

signatures are analyzed to determine the types of degradations present and

evaluated using the MOVATS method. Based on the results obtained, they determine what it states about MOV operational readiness. The valve signatures

and their analyses were provided by MOVATS Inc. under subcontract to ORNL.

The MOVATS method utilizes portable data sensing, acquisition and storage equipment which is connected to an MOV during a plant shutdown. The

sensors measure motor current, switch actuations and the axial displacement

of the worm in the motor operator, all during actuation of the valve. The

data is analyzed in-situ and also later at the MOVATS office using additional

data reduction equipment. Detection and diagnosis of abnormalities is made

based on comparison with signatures obtained from normal valves and from the

experience of the MOVATS technicians.

Valve abnormalities can be classified into two types, each of which can

affect operational readiness. The first type consists of time dependent

degradation of MOV parts resulting from the effects over time of environmental stressors and service wear. The second type consists of incorrect

3-15
adjustments and other similar abnormalities which either can lead to acceleration of the time dependent degradation or to failure of the valve to operate

-under some anticipated operating condition.

3.0 Results/Findings

ABNORMALITIES DETECTABLE BY MOVATS CLASSIFIED BY TYPE

Time-dependent

degradations

Incorrect adjustments &

other abnormalities

Bent stem

Gear wear

Motor pinion binding

Stem wear

Grease hardening

Motor degradation

Excessive inertia

Inadequate stem

lubrication

Improper seating

Valve backseating

Incorrect torque-switch

calibration*

Unbalanced torque switch*

Excessive spring-pack gap*

Excessive packing

tightness*

Improperly set bypass

switch*

Loose stem-nut locknut*

*Abnormalities that can cause valve failure under some anticipated operating

conditions.

3-16
SUMMARY OF SIGNIFICANT MOV ABNORMALITIES IDENTIFIED BY MOVATS

Abnormality Percent*

Improperly set bypass switch

Incorrect torque-switch calibration

Unbalanced torque switch

Excessive spring-pack gap

Excessive packing tightness

Excessive inertia

Loose stem-nut locknut

Valve backseating

Steam wear

Grease hardening X

Gear wear

Motor degradation

Miscellaneous abnormalities

75

50

33

17

8

8

8

8

8

8

6

3

35

*The percentages shown in this table are based on a

valves available for use in this test program. As

should be considered as indicators of frequency of

listed abnormalities rather than absolute measures

to all MOVs.

limited sampling of

a result, the values

occurrence of the

applicable generally

3-17
Based on the field tests carried out as part of this program and on an analysis of the capabilities and limitations of the MOVATS methods, the following

conclusions are drawn:

1. The MOVATS method can provide valuable diagnostic information regarding the operational readiness of MOVs well beyond that obtainable from

ASME Section XI type surveillance tests.

2. The field tests carried out as part of this task demonstrated that

incorrectly adjusted bypass switches and torque switches are common.

Both of these abnormalities, but particularly the incorrectly adjusted bypass switch, can directly affect MOV operational readiness under

some anticipated operating conditions.

3. The MOVATS method can provide useful diagnostic information regarding

time dependent degradation (aging) of MOV parts. However, in order

to utilize that information in determining operational readiness, the

method must be applied periodically and trending data obtained from

which to extrapolate operability into the future.

4. The MOVATS method has limitations with regard both to its ability to

detect all abnormalities and its convenience. Neither limitation

affects its utility in providing diagnostic information regarding MOV

operational readiness well beyond that obtained using the ASME Section

XI tests.

4.0 Utilization of Research Results in the Regulatory Process

This report, NUREG/CR-4380, in conjunction with NUREG/CR-4234 are contributors towards the resolution of generic issue II.E.6.1. "In-Situ Testing

of Valves."

3-18
4.1 ELECTRIC MOTORS NUREG/CR-4156 JUNE 1985 BNL/FIN NO. A3270

Operating Experience and Aging-Seismic Assessment of Electric Motors

1.0 Background

Nuclear power generating stations utilize electric motors as vital system components both for the performance of normal operations as well as for

the accomplishment of safety-related functions. It is therefore, imperative

that motors remain capable of driving required loads.

Motor design and materials of construction are reviewed to identify

age-sensitive components and to characterize the dielectric, rotational, and

mechanical hazards on motor performance and operational readiness. Operational and accidental stressors are determined, and their effect on promoting aging degradation is assessed. The functional indicators which can be

monitored to assess motor component deterioration due to aging or other

accidental stressors are identified. Failure modes, mechanisms, and causes

have been reviewed from operating experiences and existing data banks.

The study has also included consideration for the seismic correlation of

age-degraded motor components.

Only a few categories of electric motors are of direct safety significance in nuclear power plants: (1) three phase induction motors; (2) direct

current (dc) motors; and (3) three-phase synchronous motors. The squirrel

cage induction motor is the work-horse" of the nuclear industry, comprising

nearly 90% of the total population. Synchronous and dc motors constitute an

additional 9% with the balance comprised of specialty applications. The

percentage of motor failures of the total population in each category ranges

from 2.4% for synchronous motors to 6.3% for dc motors. The dc motor failure

4-1
rate is higher than normal and attributable to commutator related.problems.

One of the data bases indicated that motors with ratings of 1-99.9 horsepower (hp) represent nearly 47% of the total motor population. Fractional

hp (< 1.0 hp) motors and motors with ratings of 100-999 hp represent another

41% of the total population. Large motors (> 1000 hp) essentially make up

the balance. It can be deduced from the data analysis that failure rate

typically increases with horsepower rating, even though large motors are

often equipped with sophisticated surveillance, monitoring and protection

systems.

Three-phase induction motors are versatile and reliable, and speed can

be selected to suit the load. Dc motors are reliable and have accurate

speed control as well as efficient performance over the entire speed range,

but commonly require more maintenance. Where constant speed is an absolute

necessity, the synchronous motor is available.

With regard to motor applications, valves and pumps constitute nearly

95% of the total motor population and are predominantly driven by squirrel

cage induction motors.

The insulating system of a typical electric motor consists of various

materials in association with conductors and supporting structural parts.

Insulating systems are NEMA designated as A, B F and H, in ascending order

of maximum operating temperature for a given life. Class B insulation systems

are consistently in the highest failure category while Class F and exhibit

significantly lower failures. This is partly because a large population of

motors in a typical nuclear plant have Class B insulation.

4-2
2.0 Summary of Applications

REACTOR-FLUID SYSTEM-MOTOR FAILURES (LER Data: 1974-1983)

*Residual Heat Removal (RHR) 41

Service Water (SW) 38

*High Pressure Coolant Injection (PCI)l 35

Gas Radiation Waste Management 20

Containment Heat Removal (CHR) 19

Emergency Generating System (EGS) 18

Chemical Volume and Control System (CVCS)2 16

Heating Vent. Air Conditioning (HVAC) 12

Containment Isolation System (CIS) 11

Containment Gas Control (CGC) 11

*Reactor Core Isolation Coolant (RCIC)1 10

*Core Spray (CS)1 9

Reactor Coolant Recirculationl 9

Fire Protection 8

Radioactive Monitoring 8

Control Room Habitability 8

Others <6

* Emergency Core Cooling System (ECCS)

1 BWR Application only

2 PWR Application only

(unmarked) Both BWR and PWR application

4-3
3.0 Results/Findings

During motorized system operation, key parameters can be utilized to

generally assess motor integrity. Various performance or functional indicators can serve to characterize the behavior of any electric motor. When

normal values for these parameters are observed to adversely change, the

incipient stage of degradation, potentially leading to ultimate failure, is

occurring. Therefore, characteristic parameter performance can be linked to

failure modes, mechanisms, and causes that are representative for all types

of motors.

A comprehensive survey table is available in NREG/CR-4156. The table

length (pages 4-23 through 4-30 precludes its presentation in the summary.

However, it is recommended as a reference for more detail on failure modes,

failure causes, failure mechanisms, aging, aging-seismic correlation, categorization of failure and relative probability of occurrence for electric

motor components. %

The following three figures present a summary of failure modes, mechanisms and performance indicators.

4-4
Environment * Degradation of Mechanical'

• Temperature and Electrical Strength

• Roldity with Time.

* Radiation e Shorted windings'

* Chemical

* Contaminants

Vibration * Turn-To-Turn Electrical

Loos* Laminations, Short

Wedges' * WInding-To-Ground Short

Blockings, Balance

Welghts'

t~l

STATORMOTOR

INSULATIN3\

( SYSTEM 

Dominant

Failure

Mechanism

Excess Current

* Frequent Starts

* verload

* Overheating, Burnout

Windings

* Excess Losses

High Voltage Gradient * Corona or oniatlon of

Insulation

* Electrical Short and

Breakdown

Functional Indicators:

* Temperature (indIngs)

* Vibration signature

* Current signature

* Voltage gradient, corona

* Insulation Resistunce

* Power factor/loss factor

* polarization Index

improper Ventilation * Overheating or Windings

* Design * Turn-To-Turn Short

* Foreign Particles

Water/Stem * Shorted Electrical Circuits

* Corroslon of Laminations

* Poor Electrical Connections'

* 41 |~~~~~~~~~~~~~~~~~~' Performance - |-. Fallurn Functlonal ( Motor Hazards Effects Indicators

Subsystem)

' Aging Susceptibility per Table 4-15

Functional Indicators for Dielectric ntegrity
Excessive or Under Lubrication * Overheating

o Damage Ol Sling Ring * Cracking, Scoring, Brineiling,

* Sal or Gasket Leakage or Splitting of bearings

Water/Steam * Corrosion

BEARINGS

LU8RICATION

(ROTOR ASSEMBLY)

Vibration

* lsaligment

* Noism

* Cracking

Functional Indicators:

-, 9

Bearing Insulation Damage * Bearing Current Leading to

Overheating

9 I.

Improper Coolinq

* Seal or Gasket Leakage -

* Overheating

* Lubrication

* Lubricity

. Viscosity

* Vibration signature

* Temperature (Bearlng,

oll)

* Eddy current losses

* AlIgroent

* Clearances

* Speed (torque)

0'-

Mechanical Problems * Jamming/Freezing of Bails

* OI/Forelgn Particle Deposits * Stator Lamination Damage

* Rotor-Stator Interferance * Tripping of Accessories

* Excessive Axial Mvement * Shifting of Rotor C.G.

* Too Mach Clearance Between

Bearing and Shaft

* Frozen Bishing

Dominant

FaIlure

Mechanisa

Performance - Fellure |& Funetlonol ( Hbtor Harards Eff fcts Indicators

Subsystuu/

* Aging Susceptibiltiy per Table 4-15

Functional Indicators for Rotational Intagrity
Vibration * Loose or Daeiged Leads'

* Bolts Weld Failure

* Weiding/Srazlng * Loss of Balance Weight

Dowel pins * Bending or Distortion

of Shaft

* Sheer of dowel pins

Environmental * Inproper Electrical

Contealnatlon/Corroslon of Contacts

Leads

MISCELLANEOUS

FAILtURES

( ALL 

\CC14PONENTS

Functional Indicators:

* Vlbrat1on signature

* Environmental conditions

* Corrosion

* Cracks

* Loeks (tseols gaskets)

* Wear

Damaged Seals Gaskets Oil or Water Leaks'

4

Wrong Adjustments

and Settings

* Overloading, Oerheating

-j

Faulty Heaters * Too Rich Mbisture

MlsapplIcatlon/llsoperation * Overloading, verhoatIng

* Valve Seating

* Phase Error

* Frequent Starting

Commutator Brushes (wear) * Poor Connections'

Dominant

Failure

Mechanism

-

l (otor

I)byse

- Performance | -.- Fallure | - - Functional

Hazards Effects Indicators

* Aging Susceptiblitly per Table 4-15

Functional IndIcators for Mechanical Integrity
lor

To -

r 

VALVE MOTORS

E3 PUMP MOTORS

E MISC. MOTORS

30

0 40

0

20

10

FI-_

go

21

14t

40

12 M 1 10

32

3 0 O . , _ * * , do \ , -\ . . . -^ ^ , , _ . .

STATOl ROTOR BEARING ACCESSORIES MINTtIANCE UNKNOWN

HIN FACTOR

Motor Component Failure Distribution (LER Data: 1974-1983)

Analysis of the LER data provided the following:

o Motor failures that occur inside of the defined motor boundary are

significantly greater than those that occur outside the boundary.

o For BWR systems, pump and valve motors are equally prone to failure

during normal plant operation, whereas for PWR systems, pump failures

are more likely to occur.

An IPRDS data review revealed the following:

o Pump motor failures are often control related.

o The stresses caused by continuous operation in a motor have less

deleterious effect on aging and service wear than intermittent or

infrequent operation.

o Vibration and moisture in-leakage are the prime causes of motor

failure.

o Most reported pump and valve motor failures are catastrophic (Note:

This indicated that incipient failures are not being identified.)

4-8
A comparison of NRC license SALP (Systematic Assessment of Licensee

Performance) ratings with LER motor failures indicates that licensees receiving below average maintenance ratings also experienced a comparatively higher

number of motor failures. This condition demonstrates that improved preventative maintenance could prolong motor life and reduce the overall number of

failures.

A review of additional data sources utilized served to reinforce foregoing conclusions:

o Motor failures typically occur, and are detected, while the machine

is in the operational mode.

o Most degraded motor conditions are either in advanced stages or have

resulted in catastrophic failures.

o Stator grounding and bearing related problems are the primary causes

of motor failures.

4.0 Utilization of Research Results in the Regulatory Process

Focus on updating national codes and consensus standards including:

a. IEEE-323, Qualifying Class E Equipment for NPGS.

b. IEEE-344, Seismic Qualification of Class E Equipment for NPGS.

c. IEEE-334, Type Test of Continuous Duty Class E Motors for NPGS.

d. Reg. Guide 1.89, Environmental Qualification of Certain Electric

Equipment Important for Nuclear Power Plants.

e. ASME-OM8, Performance Testing of Electric MOV Assemblies used in

Nuclear Power Plants.

4-9
4.2 BATTERY CHARGERS AND INVERTERS NUREG/CR-4564 Dec. 1985 BNL/FIN NO. A3270

Operating Experience and Aging-Seismic Assessment of Battery Chargers

and Inverters

1.0 Background

This report provides an aging assessment of battery chargers and inverters, and was conducted under the auspices of the NRC Nuclear Aging Research (NPAR) Program.

In characterizing the aging and service wear effects of battery chargers

and inverters, this study considers the predominant designs utilized by the

nuclear industry including the size and arrangement of this equipment in

operating stations. This equipment is typically located in mild environments

and is not subject to containment level environmental parameters. Because

of their importance for safe shutdown of the plant, this equipment is required

to be environmentally and seismically qualified. A discussion of the effects

of component performance under operational and environmental conditions is

provided. Several failure data bases are reviewed including Licensee Event

Reports (LER), In-Plant Reliability Data Systems (IPRDS), Nuclear Plant

Reliability Data Systems (NPRDS), and Nuclear Power Experience (NPE). They

are reviewed to determine the failure modes, causes and mechanisms experienced

in recent years by the nuclear industry and to prioritize the most significant

modes of failures. The study also Includes a discussion of manufacturer

recommendations for maintaining reliable equipment. A review of industry

and government standards relating to testing and maintaining of this equipment

is included as well. Standards from the Institute of Electrical and Electronics Engineers (IEEE) and the National Electrical Manufacturers Association

(NEMA) specifically address battery chargers and inverters.

4-10
Nuclear power plants utilize battery chargers and inverters to supply

power to safety-related equipment, instrumentation, and controls. A battery

charger converts alternating current (ac) to direct current (dc) to provide

power to dc-driven equipment and components as well as to keep the standby

batteries in a fully charged condition. Some plants are designed with a

standby charger in addition to the required number of units (typically 2-4

per plant). On the other hand, inverters are used to supply ac-power to

safety related equipment and equipment important to plant operation after

converting the dc-power source to an ac output. A typical plant design

requires at least two such units to distribute power to various control

equipment vital to power and safe shutdown operations. Plant systems such

as the Reactor Protection System (RPS), Emergency Core Cooling System (ECCS),

Reactor Core Isolation Cooling (RCIC) System, and the ac/dc distribution

system utilize these devices to provide the power to satisfy their safety

functions.

Both battery chargers and inverters are considered together in this

study because of their similarities in design, construction, and materials.

The subcomponents, particularly the electronic elements, such as diodes,

relays, capacitors, integrated circuits, etc. are the same in both equipment.

They also serve related safety functions in the plant and experience the

same environment as well as similar operational stresses.

The combination of required and desired options provided by the manufacturer results in each nuclear inverter and battery charger being somewhat

unique, although certainly the major circuitry involved with each are similar

for the specific models and manufacturers.

4-11
Four types of inverters and three battery charger types are presently

used in nuclear applications. This equipment is discussed in detail with

the advantages and disadvantages of each type noted. The size or the equipment varies with the electrical bus configuration used at the plant. Exceptions to this occur when individual utilities add non-safety but operationally

important loads to the inverters or chargers which dictate that a larger

unit be employed. The data received indicated a range of station battery

charger sizes from ratings of one hundred amps to six hundred amps, while

station inverter sizes varied from ratings of five kilowatts to two hundred

kilowatts. For inverters, data were also collected for specific application

inverters such as those used with the HPCI, RCIC, and Auxiliary Feedwater

systems. These smaller inverters are generally rated less than one kilowatt.

Some lower voltage battery chargers (24 and 48 volts) were also included in

the data if a significant effect was clearly indicated, such as a loss of

nuclear instrumentation. Manufactured by the same suppliers of the larger

station chargers which usually operate at 125 volts, these smaller units

were found to operate on the same principles and contain the identical components as the larger units.

Most of the battery chargers and inverts utilized in nuclear power

plants were manufactured by six primary vendors. Although there exists

several different designs, the physical appearance and construction of the

units including the mounting of subcomponents inside the cabinet are similar.

The size of inverters are designated by their voltage-ampere rating. The

weight of an inverter varies from as low as 30 lbs. for a 0.5 KVA rated unit

to 4,000 lbs. for a 200 KVA unit. Similarly, battery chargers have weights

4-12
ranging from 60 lb. to 5,000 lbs. The physical size of the equipment also

varies significantly.

2.0 Summary

To accomplish the identification of aging and service wear effects and

appropriate inspection/surveillance/monitoring techniques, it was necessary

to examine potential failure modes, mechanisms and causes. This was achieved

by reviewing Battery Charger and Inverter design and materials of construction, by establishing the stressors that are both operational and accident

related, and by reviewing existing failure related data. Aging-seismic

correlation was addressed during this phase of the program. An interim

review of current standards, manufacturer's recommendations, and condition

monitoring techniques was performed in order to aid in the determination of

future work.

Despite the use of redundant equipment and buses, failures, especially

in the case of inverters, have resulted in reactor trips, inadvertent safety

system injections, and emergency core cooling system unavailability. The

consequences of battery charger failure is not as immediate, since a fully

charged battery can supply dc power when a battery charger is unavailable.

Depletion of the battery due to charger unavailability can, however, result

in the loss of safety systems under post accident conditions. Inverter

failure, on the other hand, directly interrupts the 120 volt ac power supply

to vital controls, logic and annunciators.

The three types of battery charger designs are the Silicon Controlled

Rectifier (SCR) solid *state type, the controlled ferroresonant, and the

magnetic amplifier (mag amp). While all three types are used at nuclear

4-13
facilities, the SCR or hyristor solid state charger is the most widely

used, making up nearly 75% of the population, and, in fact, is the only

charger type that is qualified to IEEE-323 and IEEE-650.

Four basic inverter designs are currently in use: the ferroresonant

transformer, the pulse-width modulated, the quasi-square wave, and the step

wave. The former two types are most often used, with the latter two types

making up less than 20% of the inverter population.

Since battery charger and inverter are closely coupled in most nuclear

plant application they are subjected to very similar system level stresses.

That is, the charger output normally is connected to the inverter input, and

the ac supply to the battery charger also provides alternate power to the

vital bus either directly or through a rectifier in the inverter module.

This is illustrated in Figure 1 which also depicts the alignment of offsite

and onsite emergency ac sources.

3.0 Results/Findings

The charger and inverter subcomponents that are most susceptible to

aging are capacitors, transformers and inductors, silicon controlled rectifiers (including diodes), and fuses. High voltage, current or temperature

will affect all of these components, while transformers and inductors are

also susceptible to excessive moisture. Fuses may fail due to thermal fatigue.

As one would expect, plant configurations help determine battery charger

and inverter reliability. Those plants that have a standby charger or a

second full capacity charger are the most reliable. For inverters, those

4-14
.Offsite

AC Source

480 VAC BUS - SAFETY RELATEDL Emergency Diesel

Generator

I) r- _ _-r - ---- 1

I I

.1

L

)Input-Circuit

Breaker I

I.

'GER | 

I

) Output Circuit Rr IkPfl

Battery

Charger

Boundary

480 VAC BUS

I) BUS

480/120 VAC

Transformer

r--- I

I Breaker

'R I

IInverter

Boundary

I

~~~~~~~~~~~~~I

)Output CircuitI

Breaker

l C... ~ sI

I

I

I .

(Alternate)

Figure -1

~~~~. I

_ _ _ _ _ _ _ _~~~~~~~~~~~~~~ 4 -0W A. LnLl

Battery Charger - Inverter Alignment

Vital Bus

(120 VAC)

4-15
plants with transfer switches, which allow a separate bypass ac feed appear

to be the most reliable. Those plants with rectifiers providing an alternate

dc feed to the inverters, have not been nearly as reliable.

The weak links which might be susceptible to seismic excitation are:

cabinet mountings to floor or wall, subcomponent mountings, wire and cable

connections, relays and circuit breakers, transformers, oil filled capacitors,

and fuse holders.

Battery charger and inverter failures exhibit the typical "bathtub"

curve when plotted against component age. That is, a high number of failures

occur in the first year of operation with a pronounced wear-out effect in

the fifth and sixth years of operation.

Increasing preventive maintenance scope and intervals, replacing troublesome equipment, and improving system designs are some of the actions taken

by utilities who have experienced inverter and battery charger failures

which have affected plant safety ad availability. Improvements in materials

and procedures also help to reduce the failure rate and could explain the

shape of the curve obtained when plotting inverter and charger failures

against plant age. Initial Aging

Operation Effects

Increased PM

Design Improv

Number of

Failures 1

Random Failure

Level

Plant Age-Years

A summary of the results of the data analysis, with regard to failure

modes and mechanisms and probability of occurrence are listed in Table 1.

and

,ements

I'

4-16
The conclusions drawn thus far from this study include: (1) improvements

in battery charger and inverter performance can be obtained through implementation of a comprehensive preventive maintenance program supported by

appropriate personnel training; (2) due to the extensive systems interactions

related to charger and inverter failures, it is recommended that plant procedures be in place to respond to these potential failures. Additionally,

periodic capacity testing should be conducted to insure that the capability

of this equipment to supply the required loads has not diminished due to the

aging of key components.

Future work will consist of testing naturally aged battery chargers and

inverters under normal and accident conditions to validate the performance

indicators. Additionally, final recommendations will be established for

inspection, surveillance, monitoring and maintenance programs.

4.0 Utilization of Research Results in the Regulatory Process

Present NRC interest in vital ac power, of which the inverter is an

integral part, is documented in Issue No. 48, LCO for Class E Vital Instrument Buses in Operating Rectors. Research on this issue has resulted in

recommendations such as the implementation of a 72 hour annual time limitation

for energizing a vital ac bus from an interruptible power source, i.e., an

alternate bus. Other potential recommendations include the standardization

of technical specifications of vital ac buses, and the definition of minimum

plant configuration requirements.

4-17
TABLE 1. Battery Charger/Inverter Failure Modes/Mechanisms/Causes

4.1.

1

00

AGING PROBABILITY

COMPONENT FAILURE MODES FAILURE CAUSES FAILURE 4ECHNISMS AGING SEI94IC OF OCCURRENCE

Falls to Build up of dirt, solldifica- Increase In friction, binding. Yes Yes Medium

Operate tIon of lubrication, bearing

CIRCUIT BREAKER wear.

(Consists of contacts

coll, mechanical Metal fatigue, mbrittlement & Trip coll force becomes less than Yes Yes Medium

linkages, case) cracking of Insulation, spring force.

Falls Open

Oxidation I pitting of contact Loss of continuity across Yes No Low

surfaces. contacts.

Falls Open Metal fatigue. Equipment load cyciing. Yes Yes Medium

FUSE

Maiting of link. Heat generated by surrounding No Ho Low

components.

Contacts Open Oxidation & pitting of contact Loss of continuity across Yes Yes Mdium

surfaces. contacts.

RELAY

Open Crcuit of Electromechanical action caus- Loss of continuity through coll Yes No Low

Coll Ing corrosion of fne wires. wires.

Loss of Capacl- Overheating by Internal Loss of electrolyte. Yes No High

ELECTROLYTIC tance stresses.

CAPACITORS

Open Circuit Vibration. Failure of leads. Yes Yes Low

Loss of Capacltance

Overheating forms gasses. Dielectric breakdown. Yes No High

OIL fILLED CPPACITORS

Open Clrcuit Vlbratlon. Fallure of leads. Yes Yes Low

Open Circuit Vibration. Failure of leads. Yes Yes Low
TABLE 1. Continued

-.

I0

AGING PRO3ABILITY

COMPONENT FAILURE MODES FAILURE CAUSES FAILURE MECHANISHS AGING SEISMIC OF OCCURRENCE

Short CIrcuIt- Temperature cycling/over Cracking of Insulation. Yes No Hedlui

(turn to turn heating.

KAGNETICS or to ground) Low tmperature. Cracking of moisture seals.

(Transformer

Inductor) Short circuit- High voltage stress. Insulating material deterioration No No Medium

(turn to turn

or to ground)

'hange In Vlbratlon/over temperature, Change In shunting. Fracture of Yes No Low

Inductance connecting wires.

SILICON CONTROLLED Short or Open Overheating. Overvoltaqg, overcurrent due to No No Medium

RECTIFIER Circuit transients.

RESISTOR Open Circuit Vibration. Lead falls. Yes Yes Low

Change In Value Internal or ambient temperature Decrease In resistance value as No No Low

changes, temperature Increases.

Change In Temperature cycling. Cracking of circuit lines. Yes Yes Medium

Output

PRINTED CIROUIT Corrosion. Open circuit at terminals or Yes No Low

eo0ARDS within pcb.

Vibration. Loose or open connection. No Yes Hiqh

SURGE SUPPRESSOR Short Circuit Semiconductor barrier break- Overvoltage, overcurrent. No No Low

down due to overheating.
TABLE 1. Concluded

_ _A ING PROBABILITY

COMPONENT FAILURE ODES FAILURE CAUSES FAILURE KACHANISMS AGING SEISMIC OF OCCURRENCE

MISCELLANEOUS

- Connectors Open or Short Installation stresses Fatigue of wire at terminals. Yes Yes Medium

CIrcuit

- Mtors No Response Buildup of dirt on movement. Increase In bearing friction Yes No Medlum (Stuck)

Overheating. Coll Insulation degrades causing Yes No Low

shortina.

- Stch Falls Open or Contact pitting/corroslon.

Closed

- Potentiometer Thermal degradation.

0
The Office for Analysis and Evaluation of Operational Data (AEOD) has

continued to follow inverter performance, and has found unsatisfactory improvement in failure rates and failure consequences despite substantial

regulatory and industry recommendations/requirements in this area. Transients, including reactor trips and safety injections, continue to occur

because of inverter failure or vital bus degradation. In addition, because

of the potential for a significant plant impact when an inverter fails, a

second NRC group studying issue A-17, "System Interaction," has also initiated

research into inverter performance on the systems level. This includes a

review of significant instrumentation which is provided by the vital bus and

the effect of its loss on operator performance during a severe transient

event.

Similarly, in the area of dc power of which the battery charger is an

integral part, the NRC has studied, and continues to research, the performance

of safety related dc power supplies and the impact that a loss of dc has on

plant safety.

In July 1977, the NRC issued a report (NUREG-0305) which addressed the

reliability of dc power supplies at operating nuclear power stations, and

discussed the likelihood and the consequences of a postulated failure of all

dc during normal operation of a plant. The NRC staff concluded that because

of the importance of ac and dc power systems, efforts to review the reliability of these systems should continue to be expended.

In April 1981, a second study of dc power was performed by the NRC

(NUREG-0666) as part of the work in Issue A-30, "Adequacy of Safety Related

DC Power Supplies." The issue here is primarily one of dc independence to

minimize the potential for common mode failure. This study was a proba4-21
bilistic safety assessment to determine the relative contribution of dc

power related accident sequences to the total core damage probability. A

significant finding of this PRA study was that a potentially large contribution could be reduced by requiring dc power divisional independence, and

improved test, maintenance, and surveillance of dc components, including

batteries and chargers.

4-22
4.3 ELECTRICAL CABLE-ISMM NUREG/CR-4257-1 Aug. 1985 FRANKLING RESEARCH CENTER

Inspection, Surveillance and Monitoring of Nuclear Power Plants - with

Applications to Electrical Cables

1.0 Background

This report discusses the types of cables that are used in nuclear

power plants, the causes and mechanisms of their failure, and the means of

detecting age-related deteriorations that may lead to failure. The general

concepts of equipment condition monitoring as applicable to the detection of

age-related deterioration of safety-related equipment are evaluated. The

goal of cable condition monitoring is to determine the degree of cable degradation and to predict the remaining useful life. In-situ nondestructive

testing and destructive laboratory testing are discussed. Interim recommendations are given for the implementation of a cable condition monitoring program.

Initially, electrical cable appear to be simple devices that should be

easy to monitor for age-related deterioration. The two primary components

are insulation and conductors. However, further evaluation indicates that

monitoring and evaluation of cable aging is a relatively complex subject.

Many types of insulations are presently in use, most of which are organic

nature. The types of deterioration vary for the different organic materials

and may be significantly different for the same material when it is subjected

to different stresses, combination of stresses, and different stress levels.

Besides being composed of different materials, different types of cables are

used for different functions. Instrumentation cable is significantly different from power cable both in construction and in the stresses to which it

is exposed.

4-23
The testing of cables in a nuclear plant is limited by many concerns.

The number of cable circuits is of a proportion that precludes testing of

all circuits. Disruption of circuits for testing can lead to reduction of

*safety if restoration is not properly coixtrolled. In some cases, tests that

are of interest may be precluded by practical limitations of test equipment

and suitability of the test to in-situ applications. Monitoring of aging

parameters of cables is not simple, however, it is not impossible nor impractical.

Four basic types of electrical cables are used in nuclear power generating stations:

o Low voltage power cables (600 V rating)

o Medium voltage power cables (usually voltages of 4 to 13 kV)

o Control cables (low voltage, below 600 V, at a few amperes. Used to

control devices such as valve operators.)

o Instrumentation cables (typically thermocouple wire, resistance

temperature detector [RTDJjcables, twisted pairs, and coaxial cables

for transmission of data or instrument power. Also used for currentcontrolled devices [e.g., 20 to 50 mA], signal transmission, etc.).

Table 1 provides further breakdown of cable types and applications. t

should be noted that 4-kV safety-related cable is not used inside primary

containment and that 13-kV power systems are generally not safety-related.

Some typical constructions of cable are illustrated in the report.

For a new BWR, the amount of cable within primary and secondary containment is approximately as follows:

Low voltage power 300,000 feet

Control 240,000 feet

Instrumentation (general) 1,300,000 feet

Instrumentation (neutron monitoring) 30,000 feet

4-24
Cable mounting configurations vary with the type and vintage of the

plant. Cable may be located in trays or conduits inside of containment. In

some newer BWR plants, most of the cable is located in conduits inside of

primary containment. However, small sections of this cable may be exposed

between the device housing and the end of the conduit section. Outside of

containment, most cable is located in cable trays.

The most prominent modes of age-related failures of cable are electrical

insulation degradation and termination failure. Cable insulation parameters

that can be evaluated are insulation resistance, polarization index, high

potential withstand capability, tensile strength, elongation, dielectric

strength, partial discharge level, and dissipation factor. Test methods and

implications of results are discussed as well as the limitations of performing

in-sit testing.

An acceptable correlation of cable inspection, surveillance, and monitoring results with the continued ability to withstand DBA conditions is not

presently available. Artificially aged cable has been used in accident

simulation tests. However, data regarding the level of deterioration in the

parameters of the aged cables prior to DBA simulation are not available to

allow determination of acceptable levels of deterioration in cables that are

in service.

Current utility practices pertaining to surveillance and testing of

cables are discussed. Tests to measure insulation resistance are being

conducted at some plants to identify trends for cable deterioration. Periodic

high-voltage testing of cables connected to large motors is performed by

some utilities' but is considered by other utilities to be a source of substantial degradation to the cables. The purpose of existing utility cable

4-25
testing programs is to determine whether significant deterioration has occurred and whether replacement or further investigation is required. These

tests are not condition monitoring tests in that remaining life is not estimated nor is the continuing ability to withstand DBA conditions evaluated.

Information obtained during this study was analyzed to develop interim

recommendations for condition monitoring, surveillance, and maintenance of

cables. It was found that sufficient data do not exist to formulate quantitative criteria, however, significant condition monitoring parameters were

identified, and various observations were made in this report that can help

formulate a basis for developing quantitative criteria. Discussion pertaining

to interim recommendations is provided; which concludes that further investigation is necessary to develop criteria for choosing between continued operation of cables and taking corrective measures.

2.0 Summary of Testing Status %

At present, an adequate correlation between the measurable aging parameters and the future functional capability during normal and DBA conditions

of many types of equipment does not exist. Therefore, replacement or repair

decisions are heavily based upon prior experience and engineering judgment

with regard to test results rather than firm predetermined criteria.

The parameters that are used to monitor the aging of equipment are

based on the aging mechanisms of the materials, parts, and interconnections

involved. Assuming that the critical parameters indicative of the degradation

of the equipment with aging can be identified and the correlation between a.

parameter and the functional capability is known, one must identify methods

that can be used to monitor the parameters of installed equipment. In many

4-26
cases, suitable techniques are not available that can be applied for continuous or periodic monitoring of the critical parameters. For example, in the

case of power cables, the ability to perform in-situ partial discharge testing

of long cables is limited. The available portable power supplies for the

test are limited to approximately 30 kVA, which may be insufficient for long

lengths of cable.

Condition monitoring of equipment is defined as continuous or periodic

observation and evaluation of critical parameters to assess the equipment's

ability to continue to perform its specified safety functions during a period

following the moment of observation. Observation may take the form of measurement or inspection. The specified functions of the equipment include

those required in the event of applicable accidents as well as normal service.

Condition monitoring is expected to reveal not only the functional

state at the moment of observation, but also the ability of the equipment to

remain capable of performing as specified for a period following the observation. Therefore, to be considered as condition monitoring techniques,

diagnostic techniques must provide a basis for assessing the expected functional capability of the equipment should a DBA occur subsequent to the

observation. Ideally, monitoring a single functional parameter would suffice

to indicate the functional capability and exact criteria would exist to

decide, based on the observed value of the parameter, what action to take at

the time of observation, i.e., to continue operating without change, to

perform maintenance or calibration, to repair the equipment, or to replace

the equipment.

Conditions monitoring requires a data-colleccion and analysis system to

determine the extent of degradation and the nearness of incipient failure.

4-27
The monitored parameter must be evaluated to decide what corrective actions,

if any, are to taken. Analysis of data collected over a long time can help

to determine the causes of degradation and the systematic preventive actions

(such as scheduled inspection, preventive maintenance, or replacement) that

will help avoid or reduce equipment degradation.

A brief summary of some of the important available tests pertaining to

condition monitoring of cable insulation is provided. Table 1 lists cable

type tests, their general application and specific application after cable

installation.

Cost-effective techniques for adequate monitoring and detection of

degradation and incipient failures are still in the developmental stage for

most electrical equipment. In many instances, research may be required

before practical techniques can be recommended. However, the existing state

of technology has not been fully exploited.

Effective use of condition monitoring in nuclear power plants can improve

plant safety. and availability. Condition monitoring technology has the

potential to allow relaxation of some of the stringent requirements for

preventive maintenance and replacement of safety equipment by providing

quantitative assessment and prediction of equipment degradation. Evaluation

of condition monitoring results will allow establishment of maintenance and

replacement intervals based on actual rates of equipment deterioration,

thereby allowing improved allocation of resources. Less frequent maintenance

will be performed on equipment found to deteriorate more slowly than expected,

while more frequent maintenance will be performed on those that degrade more

rapidly. The result will be on overall improvement in plant safety.

4-28
TABLE 1. Cable Tests and Applications

Specific Cable Application

Test Description General Application After Installation

IMSIILATtOW TESTS

Low Voltage Insulation

Resistance

Insulation Resistance (500 V)

Polarization Index

Troubleshooting, locating existing

short crcuits

Determination of general acceptability

of Insulation. Usually performed at

500 Vdc

Service test to determine general

state of insulation system. requently

used on larger motors. May b used on

cables. Uses 500 V dc text quipment

Locating failures (isolating to a particular

circuit) JNDt)

Requires shield or ground path to detect

deterioration of Insulation. If return path

not provided, dry air between

cable and ground will provide good

Insulation and may nvalidate results.

Can be used to determine general Insulation

level between conductors of paired cable.

Does not predict failure. May be

trended. ay require Isolation

of certain nterconnected equipment

from test circuits to prevent damage

INDE)

Compares charging and leakage current values

at 1- and 10-minute Intervals. Requires

return path such as shield for measurement

between conductor and ground. (NDE)

q),

Elongation Manufacturing test, provides

Indication of general acceptability

of Insulation

Can be used to determine If insulation

Is losing flexibility and Is becoming

embrittled (DI)

Tensile Strength anufacturing

indication of

of Insulation

test, provides

general acceptability

Less telling than elongation test

since cable conductor provides tensile

strength needed after insulation. May

Indicate excessive softening of insulation

(DE)

High Potential Withstand

Test

Manufacturing test to prove

Insulation capability

Usually limited to higher voltage power

circuits (4 kV and above). ay be performed with connected load in place (e.g..

transformer, motor). Limited to

applications where isolated from other

circuits and cables to prevent extensive

damage from flashover or Induced voltages.

Host often performed at O0t or less of

manufacturing test level. May be

destructive to Insulation and cause

failure if not properly performed (test

Is usually a DC test). If properly

performed, NDt
TABLE 1. Cable Tests and Applications (Cont.)

Specific Cable Application

Test Description General Application After Installation

INSULATION TtSTS

Dissipation Factor

Power Factor

Partial Discharge Test

Tests to determine electrical losses

within Insulation. More mportant

for higher voltage equipment.

This Is an ac test

Measures corona discharge In voids

of nsulation. Corona discharge

causes high speed ions to attack

surrounging Insulation and cause

further deterioration. A high

voltage ac test.

By using a pulsed signal, the location

of a fault failure) In a cable ay

be determined by Interpretation of the

reflected signal. Useful for finding

the location of a fault In a long cable

The test measures and compares resistive

leakage current to capacitive leakage

current. The measurements are more

important for high voltage systems since

losses are proportional to the square

of the voltage. Increasing dissipation

factors and power factors at higher

voltages Indicate Insulation deterioration

due to partial discharging

test is generally not performed on cable

In-situ. Would be reserved for suspected

problem equipment generally In 13 kVac and

above range. Requires shield or return

path

May be used to detect conductor

deterioration at remote termination or penetration. ay Indicate significant deterioration of portions of cable insulation.

May Indicate contamination of remote

terminals. May be used to detect deterioration of series connections or splices

4p.

0

Time -Domain Reflectometry

(Cable Radar)

CONDUCTOR/TERMINATION TEST

Continuity (Loop) Resistance Determination of circuit resistance

variations indicating conductor or

termination prohlems. For use on power

circuits

Useful for determining changes in circuit

resistance. ay be used on motor circuits

with windings In circuit. Significant

changes In resistance would Indicate the

need to evaluate terminations, windings,

and cable conductors for cause

NDE Nondestructive Evaluation

DE Destructive Evaluation
The types and extent of cable testing that can be performed in-situ at

a nuclear power plant using presently available techniques are limited by

practical concerns. However, they do not preclude surveillance and monitoring

of the aging of electrical cables through use of presently available techniques. The existence of the limitations requires thoughtful planning to

assure that test results are meaningful and that inadvertent damage to the

equipment does not occur. Care in selection of tests to be performed and

the sample of cables to be monitored will overcome these limitations and

help assure the usefulness of results.

3.0 Results/Findings

The results of the screening of representative failure modes and causes

are presented in Table 2.

Further in-depth investigations pertaining to data bases such as NPRDS,

TMI failure analysis reports, andtmaintenance records at power plants are

necessary to establish correlations of cable insulation aging to various

stresses.

Following are the recommendations for further study for developing

quantitative criteria for surveillance, maintenance, and replacement of

cables.

o Visit selected utilities to collect available data pertaining to

age-related degradation of cables concerning failures and recognized

flaws.

o Obtain a description of presently established cable testing programs

from selected utilities.

o Identify the relevant research activities on cables sponsored by

utilities (such as the recently initiated EPRI-sponsored research on

equipment aging at the University of Connecticut), or the government,

both in the U.S. and abroad, and extract the necessary information.

4-31
o Select a group of experts on cables, consisting of cable users,

manufacturers, and research laboratories, to provide input to formulate consensus quantitative criteria for condition monitoring,

surveillance, maintenance, and replacement of cables, and to develop

further practical tests to relate cable condition to remaining life.

4.0 Utilization of Research Results in the Regulatory Process

The principal application in the regulatory process is in support of

the implementation of 1OCFR50, Appendix B-Quality Assurance Criteria.

4-32
TABLE 2. Insulation Failure Causes and Mechanisms for Electrical

Cables in Nuclear Power Plants*

Cable Tyvpe Failure Causes Failure Mechanisms

Power

600 V,

4 kV, and

13 kV

Loss of flexibility

(cracks when flexed)

Chemical changes of polymeric

structure resulting from

exposure to heat, nuclear

radiation, humidity,

chemicals, etc.

Loss of imperviousness

(moisture readily reaches

conductor through pores

and cracks)

Loss of adequate dielectric properties, inability

to withstand voltage

stress

Same as above, plus possible

physical abuse of cable

(chafing, cutting, small

radius bending)

Excessive exposure to high

electric stress, high temperature, or large radiation

dose. Failure frequently

coupled with presence of

moisture or water. Loss of

dielectric properties

generally occurs after

deterioration of mechanical

properties. Loss of

dielectric properties would

be more rapid at points of

insulation stress conditions

such as mechanical abuse

locations or areas of

repeated small radius

bending. For 13 kV and above

cable treeing, water treeing,

and electrochemical treeing

may cause relatively rapid

breakdown of dielectric

capabilities.

* Failure of cable sheathing can also occur, but the sheathing is not relied

upon for operability of cables. The major failure modes for sheathing are

loss of flexibility and imperviousness; the causes are similar to those for

cable insulation.

3-33
TABLE 2. Cont.

Cable Type Failure Causes Failure Mechanisms

Control Loss of flexibility

(cracks when flexed)

Loss of imperviousness

(moisture readily reaches

conductor through pores

and cracks)

Same as for power cable

failure causesSame as for power cble

failure causes

Instrumentation

Loss of flexibility

(cracks when flexed)

Same as for power cable

failure causes

Loss of imperviousness

Adverse change in dielectric properties e.g.,

insulation resistance

and/or capacitance

between conductor(s) and

shield(s)] which may cause

attenuation of signal

Same as for power cable

failure causes

Exposure to heat,

nuclear radiation, humidity,

chemicals, etc., creating

chemical and/or physical

changes to the insulation.

Significant changes in

dielectric properties can be

expected during accident

conditions. These effects

may be additive with slower

changes due to aging.

3-34
4.4 PRESSURE TRANSMITTERS-ISMM NUREG/CR-4257-2 FRANKLIN RESEARCH CENTER

Inspection, Surveillance, and Monitoring of Electrical Equipment Inside

Containment of Nuclear Power Plants - Pressure Transmitters

1.0 Background

Most of the pressure-transmitters in use in nuclear plants are actually

differential pressure transmitters, in which a pressure is applied to both

sides of the sensing element and the difference between them is measured.

Pressure transmitters are commonly used to measure gage pressure, absolute pressure, differential pressure, liquid level, and flow rate. To

measure gage pressure, one side of the detector is exposed to atmospheric

pressure and the other side to process pressure. To measure absolute pressure, one side of the detector is evacuated and sealed, and the other side

is exposed to the process pressure.

To measure flow rate, a differential pressure transmitter may be connected across a flow-restricting orifice in a process line and the pressure

drop produced across the orifice measured. Within limits, the pressure drop

is proportional flow rate.

Differing techniques are required to measure liquid level with a pressure

transmitter. For nonpressurized vessels, the high-pressure port of the

pressure transmitter may be connected to a tap in the vessel located at or

below the minimum level to be detected and the low-pressure port of the

transmitter vented to atmosphere. The liquid head of the vessel then produces

a pressure indication proportional to the level of the liquid in the vessel.

For pressurized vessels, the high-pressure port of the transmitter may be

connected to a tap in the vessel located at or below the minimum level to be

4-35
detected as above. However, the low-pressure port is connected to the top

of the vessel by a sensing line filled with the liquid contained in the

vessel. The filled line acts as a pressure reference to allow the level of

the pressurized liquid to be determined from the differential pressure that

is indicated.

Pressure transmitters are used in nearly all safety systems having

liquid or gaseous process media. They are used to initiate and monitor

protective actions requiring reactor trips, containment isolation, and starting and stopping safety actions based on system conditions; to monitor and

control reactor pressure, coolant level, and main steam line flow rate.

Pressure transmitters prevent operation of low-pressure safety systems, such

as residual heat removal in a boiling-water reactor, until reactor pressure

is reduced to a safe operating level. Also, they are used to detect high

pressure in the containment building and to initiate containment isolation

and pressure control. Depending on the vintage of the nuclear power plant,

60 to 200 pressure transmitters are used per plant.

2.0 Summary

This report describes the types of pressure transmitters commonly used

in nuclear power plants. The applications of pressure transmitters for

detection of gage, absolute, and differential pressures; levels; and flow

rates are described.

Pressure transmitters are subject to stresses from their environment,

the process they are monitoring, and their electrical power supply. Under

normal plant conditions, the environmental stressors are temperature, humidity, and radiation. Nuclear application transmitters are sealed for DBA

4-36
environment steam conditions; therefore, normal environment humidity is

generally not a problem. However, elevated temperature and radiation may

affect the environmental seals and the electronic assemblies over a period

of time. The seals may harden, crack, or take a set, allowing steam inleakage under DBA environments or damp conditions during normal service.

The process to which the transmitter is connected may also affect the

temperature of the transmitter; however, many transmitter applications have

relatively long (50 to 200 ft) sensing lines that limit heat transfer from

the process. Stresses from process transient overpressure conditions, such

as water hammer, may cause significant zero shifts in transmitcer settings.

These stresses can cause permanent deformations of sensing elements or of

mechanical linkages within the transmitter mechanism. Generally, these zero

shifts can be removed by recalibration. However, recalibration is not possible for in-containment devices during periods of reactor operation.

Pressure surges may occur when the transmitter is valved in and out of

service if the valving operation is not done in the proper sequence. For

many applications, this condition would cause the sensing cell to be forced

out of the calibrated span and could cause a calibration shift. In general,

this shift would be correctable during the next calibration and would not

cause permanent damage. However, the shift would be present for the entire

period between calibrations. Calibration shifts may also occur if the transmitter is calibrated at one temperature but operates at a significantly

higher temperature.

Comparison of as-found and as-left calibration data is described as a

partial means of evaluating the level of deterioration of the transmitter.

To allow evaluation and trending of the calibration data, care must be taken

4-37
I~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~

to ensure that variations in method or procedure do not produce erroneous

data. The precision of the measurements must also be high.

The evaluation of calibration data alone will not ensure the capability

of operating under design basis accident (DBA) conditions. If steam or

moisture penetrates the transmitter housing, the transmitter electronics

will be affected, become inaccurate, and most probably fail. Therefore, the

integrity of the housing seal must also be evaluated periodically to be able

to predict continued capability to perform under DBA environments. Three

Mile Island-2 report GEND-INF-029 provides a good example of water incursion

damage to a transmitter. In this case, the water in-leakage is believed to

have been by way of the electrical conduit or fittings. The damage was

extensive; components were badly corroded, a transformer was badly burned,

and the structural integrity of one transistor was completely lost.

The boundary of the transmitter evaluated here is comprised of the

process medium connections (ports), not including the sensing lines or test

valve manifold; the support backed for the transmitter; and the electrical

signal terminals. The boundary also includes the housing seal system, including the wiring penetration seal.

Tables 1, 2, and 3 are a compilation of License Event Report (LERs) for

pressure transmitters by application (pressure, level, and flow rate sensing).

The utility personnel contacted and the ORNL survey indicated that between 60

and 200 transmitters are used in safety-related applications in each plant.

The LERs indicate that in a 3-year period 330 reportable failures occurred

throughout the nuclear power industry. During this same period,- 65 plants

were in operation. These numbers indicate that less than two reportable fail4-38
ures occurred per year per plant. The approximate failure rate indicated is

0.02 failures/year (2.4 x 10-6 failures/h) for each transmitter.

These values indicate that transmitters are relatively stable devices

under normal power plant conditions; However, the above information does

not include consideration of DBA environment conditions that may affect

failure rates significantly, nor does it include transmitter failure information that was nonreportable under the LER system.

Some of the originally installed transmitters in older plants are being

replaced because they do not meet present environmental qualification requirements. The replacement transmitters have been qualified to IEEE Standard

323-1974.

TABLE 1 PRESSURE TRANSMITTER FAILURE DATA FROM LERs FROM

SEPTEMBER 1982 TO APRIL 1984

(40 pertinent LERs)

Failure mechanism Number of Percent

events of total

Calibration drift 32 55

Transmitter failure 6 15

(exact cause not given)

Sensing cell failure 5 12.5

Amplifier failure 2 5

Frozen sensing line 2 5

Cause unknown 2 5

Gas in sensing line 1 2.5

4-39
TABLE 2 LEVEL TRANSMITTER FAILURE DATA FROM LERs FROM

JANUARY 1981 TO APRIL 1984

(128 pertinent LERs)

Failure mechanism Number of Percent

events of total

Calibration drift 68 53

Transmitter failure 23 18

(exact cause not given)

Electronics failure (most 17 13

frequently amplifier)

Sensing cell failure 6 5

Frozen sensing line or 3 2

transmitter

Sensing line plugging 2 1.5

Gas in sensing line 2 s 1.5

Force motor failure 2 1.5

Mechanical linkage problems 2 1.5

Corroded internals 1 <1

Unknown cause 1 <1

Loose connection 1 <1

TABLE 3 FLOW TRANSMITTER FAILURE DATA FROM LERs FROM

JANUARY 1981 TO DECEMBER 1983

(109 pertinent LERs)

Number of Percent Failure mechanism events of total

Calibration drift 47 43

Transmitter failure 23 21

(exact cause not given)

Electronics failure (most 12 11

frequently amplifier)

Sensing line plugged 8 7

Cause unknown 5 4.6

Sensing element failure 3 2.7

Mechanical linkage problems 3 2.7

Force motor and magnetic 4 3.6

transducer failure

Wet transmitter internals 2 1.8

Internal power supply failure 2 1.8

4-40
3.0 Results

Table 4 provides .a listing of failure modes, failure causes, and potential failure mechanisms and their effects for force balance, capacitance

cell, and strain gage type transmitters. While these lists indicate the

possibility of many failure mechanisms, the actual failure rate for many of

these mechanisms is very low.

The most common effects of the stressors on the transmitters are calibration shifts. The evaluation of failure data contained in Licensee Event

Reports indicates that total failure of pressure transmitters occurs relatively infrequently. Discussions with plant technical personnel confirmed that

transmitters are not considered a significant source of trouble.

Evaluation of in-service failures is recommended to allow further differentiation between sudden failures having no precursor and failures that

can be detected in the incipient state. Such evaluations would aid in the

further development of monitoring techniques. Because some of the transmitter

failures are of the sudden type, periodic operability checks are an important

means of detecting failures shortly after their occurrence so that a significant number of transmitter failures do not remain undetected.

A combination of operability monitoring and condition monitoring may be

used to improve the probability of successful weathering of DBA conditions.

4.0 Utilization of Research in the Regulatory Process

The results of this and subsequently planned studies will be useful in

the regulatory process and can be factored into this process through support

of the implementation of:

4-41
1. Technical Specifications

2. Surveillance Requirements - lOCFR50.36(e)

3. Updating appropriate regulatory guides, including these endorsing

national standards IEEE-317, 323, 383, and 535.

4-42
TABLE 1. Transmitter Failure Modes, Causes, Mechanisms, and Effectsa

Failure mode Subcomponent Failure cause Failure mechanism Effect on device

Fom-e IkZtance type tranamitter

Failure to operate Force bar

.ind linkage

Feedback coil

(force motor)

Wear of pvot points

Coil urnout

AmpitfIer

Housing sealq W

Shorting or opening of

electronic components

(e.g., capacitors,

transistors)

Inability of seal to

provide moisture and

pressure barrier

Diaphragm through

leakage

High vibration or

improper nstallatIon

Overheating of coil

insulation due to

coil nsulation failure

Thermal or voltage

overload due to power

source srges or

deterioration of components

Compressive set of

organic seals or

embrittlement and

cracking due to thermal

and radiation stresnes

Perforation of

diaphragm from

corrosion or flaw

Overpressurization

due to transient

Seal deterioration

from ompreARIve

or decomposition

Overpresstrization

due to transient

Failure to operate

as required

Decrcased accuracy

or complete failure

Loss f output

Diapragm

Depending on type

of failure, may fail

high or low, lose

accuracy, or fail

with steady output

Failure of electronls

due to shorting and

corrosion from Ingress

of environmental contaminants

Varaable nstrument

drift AS pressures

across diaphragm

equalize

Zero shift

Variable Instrument

drift as pressures

across daphragm

equalize

Zero shift

Diaphragm

sea l

Force bar

and linkage

Permanent deformation

of diaphragm

Inability to maintain

presqure barrier

Bending of components

In level system
TABLE 1. (Continued)

Failturp do Subcnmponent Failure cause Failure mechanism Effect on device

Capaitane type tinamitters

Falltre to operate Senising cell Leakage of cell fluid

through daphragm

Perforation due to flaw

Terminal cover

sent

Electronics

honsing seat

Eleetronics

Inability of seal to provide moisture ;nod pressure

boundary

Inability of seAl to providd moisture and pressurre

boundary

Circuit continuity lost

Z_I

T__

.P.

Compressive set of

organic seals or

embrittlement and

cracking due to thermal

and radiation stresses

Compressive set of

organic seats or

e-Ahrittlemont and

cracking due to thermal

and radiation stresses

Circuit hoard retaining

screvs Iproperly

tightened or loosened

due to vibration

Oxidation of contacts

due to conramlnation

during repair or after

housing seal filures

Thermal or voltage

overload due to power

source surges or

deterioration of

components

Perforation due to

corrosion or flaw

Loss of ec.racy

And drift; rupture

allows equalization

of forces on diaphragm

Loss of electronics

due to Ingress of

environmental

contaminants

Loss of accuracy

to complete flutre

possible due to Ingress

of environmental

contaminants

Loss of signal,

sporadic operation

Circuit continuity lost,

hridging of crcuits

Shorting or opening of

component (e.g., capacitor, transistor)

Loss of signal,

sporadic operation

Loss of output, may

fail high or low

Loss of accuracy

(contanination of

cell oil changes

capacitance of cell)

Zero shift

FaIlure to operate

as required

Sensing cell Leakage of process

fluid Into cell fill

fluid

Voltme of cavity changes

whcii seal moves

Dislodging of tithe

seal due to overpressurization
TABLE 1. (Concluded)

Fallire otdet Sihcomrnnent Failtore ce Fatlitre meclianism Effeet n device

Fal1ure to operate

as required

Selning cell Oil chemical anges nit breakdown due to

thermal or radiation

stress

Zero shift, redtuced

accuracy, or changes

In response time

Loss of accuraCV,

drift, zero shift

Electronics Change In component parnmeter (e.g., amplificatton,

resistance, capacitance)

Deterioration of component from thermal,

radiation, or voltage

stresses

Strnin :In.e tre tnaomittera

Failure tn operate Straln gage

oil-Aling 4eMI1 .

IS leAd wire seal

n

Lose of continuit In

bridge circuit

11;nhi ltv toi pr v1 .1.

notiure and prematire

barrier

Potentlometer

Electric

module

Corrosion of resistive

elements n potentiometer

Change in component

parameter (e.g., ampliflcation, resistance,

capacitance)

Permanent deformation of

tube

Bourdon tube leaks to

transmitter housing

Primary system

.transient nr flaw

Compressive met f

organic eals or

embrittlement or

cracking due to thermal

and radiation stresses

Wire-wound potentiometer

corrodes open due to

thermal stress and

corrosive lubricant7

(does not occur f proper

lubricant s used during

manufacture)

Component deterioration

or failure (transistor,

diode, capacitor,

resistor)

Overpressure transient

Fails over range

.oSA of fl output

Zero hift

Loss of output

Fl Iure o e.lctro.f)ni

*tie to contaminiation

Falure to operate

as required

8ourdon tube

Perforation due to

corrosion or flaw

Drift, contamination

of transmitter

internals, failure to

respond

"Failures external to the transmitter, such as power source and sensing line problems, are not ncluded.
4.5 DIESEL-GENERATORS NUREG/CR-4590 APRIL 1986 PNL/FIN No. B2865

Aging of Nuclear Station Diesel Generators: Evaluation of Operating and

Expert Experience

1.0 Background

Emergency diesel generators (EDG) at nuclear plants provide class 

emergency (1E) standby power to support the operation of emergency safety

and plant protection system loads. During loss of power to the emergency

buses from the main generator or offsite sources, the EDG system must provide

backup power to operate critical reactor safety equipment.

More than 200 diesel-generators of various models, by at least nine

manufacturers, are in service as EDG at nuclear plants in the United States.

The diesel models, and types currently in use at nuclear plants, their major

components, and their support systems are briefly identified in this report.

The number of engines in nuclear service evaluated in this report, are listed

below by manufacturer:

Number in Service

Manufacturer in Nuclear Industry

ALCO 18

Allis Chalmers 3

Caterpillar 5

Cooper Bessemer 11

Fairbanks Morse 49

Electro-Motive Division 84

Nordberg 8

Transamerica Delaval 31

Worthington 4

4-46
The diesel-generator is a complex, small power plant within a nuclear

facility. Like all mechanical and electrical equipment, diesel-generator

systems and components are subject to aging. Each system and component is

affected by its operational history and environment. Engine systems deteriorate, or age, both while operating and standing idle. Since diesel generators

in nuclear service are not operated for most of their installed lives, the

diesel generator and certain related systems should be considered unique

from the aging standpoint when compared to many other systems which are used

extensively.

2.0 Summary of Systems Associated with the Diesel Generator

~~Dsrbto Syte Suoe Suvilne Sn

Oisnbuson Smem Suplies systemn, in

___.__ ___ ____ _____ _ _ B_ Boundary I

Interfacing Subsystems of the Diesel Engine System

4-47
3.0 Results/Findings

The evaluation of nuclear plant operational experience and expert opinion concerning the aging of nuclear service emergency diesel-generators

indicates: that it is observable; follows recognizable patterns; shows changes

in the modes of aging degradation with time; is confined to few, relatively

major components; increases as a percentage of all failures with time; is

caused by normal operational stressors. As a result of the review of 1,984

failures associated with diesel-generator systems from 1965 through 1984,

1,336 were classified as abnormal failures and 629 as aging related failures.

OVERALL CAUSES OF DIESEL-GENERATOR FAILURE

Percent of

Failure Cause All Failures(a)

Poor manufacturing or construction

quality control

Adverse conditions: vibration, shock

Human error - maintenance

Adverse environment: dust humidity,

chemicals, etc,

Unknown

Maladjustment/misalignment

Other

22

17

13

13

11

8

16

(a) Includes aging and non-aging failures.

4-48
SYSTEMS & COMPONENTS CONTRIBUTING MOST TO ALL DIESEL-GENERATOR FAILURES

Percent of

Systems and Components All Failures

Instruments and Controls System 25

Governor 10

Sensors 3

Relays. 2

Startup components 2

Fuel System 11

Piping on engine 3

Injector pumps 2

Starting System 10

Controls 3

Starting air valve 2

Starting motors 2

Air compressor 1

Switchgear System 10

Breakers 3

Relays 5

Instruments and controls 1

Cooling System 9

Pumps 2

Heat exchangers 2

Piping 2

Lubrication System 7

Heat exchangers 2

Pumps 2

Lube Oil 1

Other Systems 28

100

4-49
CAUSES OF DIESEL-GENERATOR FAILURE DUE TO.AGING

CAUSES OF DIESEL-GENERATOR FAILURE DUE TO AGING

Percent of

Failure Cause Aging Failures

Adverse conditions: vibration, shock

Poor manufacturing or construction

quality control

Adverse environment: dust, humidity,

chemicals, etc.

Unknown

Human error - maintenance

Maladjustment/misalignment

Poor design: wrong application or component

Other

27

18

17

14

9

6

5

4

SUMMARY OF SIGNIFICANT FAILURE RATES() DUE TO AGING BY

MANUFACTURER, SYSTEM, & FAILURE CAUSE

Fal lws Causo

a. p desig: Wrs" appicati or < t b. Per awemfacteuelq or constrvetlo/ouality control

C. Adyrse coditionst vIbration. lock

d. Adverse onvroient: dst. Cftslcals. etc.

0. IJstutelsal It, etc.

11 Human error - malate"nane

g* IJnkou/otfer

ALtD AC CAT M f F t tDI IRT

- - - - - - - 12 -

22 14 17 23 19 12 - 34 21

22 - 17 37 23 31 3 24 6

11 43 17 7 20 19 17 9 13 - - - - - - 15 - -

14 1< 17 9 a 9 - - -

is 29 33 10 17 12 it - 21

Syitens

Instrnts and coatrols

Cool ng Cyst.

fuel system

Lubr Icat I on systs

Generator

Swltchor

Engle structure

StartlIq system

Intake and 6ulOut cysts.

DrIve traIn

36 43 s0 36 27 21 34 24 29

Is 14 33 - 12 7 17 1 1 - 14 - - 19 7 19 19 1 21 - - 7 7 _ _ Id -_ -_ 7 - _

14 - - 7 - - -

_ - - 9 - 14 _ 8 -

_ - - 7 17 - - S 21 _ _ _ _ _ S _ I I _

_ _ _ _ _ _ _ S _

1) lamiled to qlng failures pr 100 aqlng failures 100.

4-50
An extensive Aging Failure Matrix is given in Table 4.1 of the original

report and extends from page 4.4 through 4.11. This table lists the system

and component, the component failure potential (high, medium, or low), the

type of failure (chemical or mechanical) plus some comments when appropriate.

In order to develop a potential list of measurable parameters they

purpose to conduct an active and thorough trend analysis program to:

o Identify potentially problematic components early in component life.

o Employ the latest methods for on-line system monitoring and diagnostics, including full-spectrum, fast Fourier transform (FFT) vibration monitoring.

o Digitize trend analysis readings for a computerized comparison of

results.

o Compare performance histories with similar installations, nuclear

and non-nuclear.

The figure below shows the percent of failures due to aging degradation

for any plant at a particular age. The resultant graph shows the changing

ratio of aging related failures to all failures. A plot of this ratio for

each year from construction and startup ( years) through 14 years of operation reveals that aging degradation is an increasingly large percentage of

total failures for nuclear service diesel generators. This figure also

shows the ranges and quality of data used for the plot.

100

Good Data Limited Data

x 80 

Age of Plant at Time of Failure (yr)

o . I~ue to Aging

LE~

10

4-51
4.0- Utilization of Research Results in the Regulatory Process

The results of this study and subsequent studies planned will be useful

to the regulatory process and can be factored into this process through:

o A NUREG report on diesel-generator aging

o Recommendations for diesel-generator maintenance

o Changes to RG 1.32, 1.108, 1.118

o Changes to applicability of ASME Section III vibration design standards to diesel-generators

o Changes to ASME Section XI applicability to diesel-generators

o A review of the safety related status of the diesel-generators governor systems

o A revision of the Class E electrical equipment qualifications standards as associated with diesel-generators

o Changes to DEMA standard practices

o Changes to related ASTM standards

o Other application to ongoing NRC efforts related to emergency dieselgenerators:

- Analysis of trends and patterns AEOD LEAD): Evaluate the applicability of methods of analyzing trends and patterns from the

NPRDS, (a) survival analysis, (b) multiway tables and log linear

modeling, (c) modeling of known reliability models to failure

data, (d) trend analysis, (e) detection of shifts in failure rates.

- Evaluation of Transamerica Delaval, Inc. (TDI) diesel engines

(NRR LEAD): Review the technical efforts of the TDI owners'

group centered on generic problems, and confirm the design and

quality of about 200 other key engine components.

- Unresolved safety issue A-44. Station Blackout -- (NRR LEAD):

The proposed resolution also includes a recommendation that the

reliability of each EDG should be maintained at or above specified

acceptable reliability levels. The objective of the EDG reliability program is to develop the details in support of improving

and maintaining EDG reliability consistent with the resolution of

USI A-44.

4-52
Diesel Generator Reliability. (RES LEAD): Evaluate means of

assuring diesel generator reliability. The study is analyzing

diesel failures and maintenance procedures and uses information

from Trojan Nuclear Plant.

Other Related Tasks -- (IE LEAD): Evaluate the results of NRC/

Industry meeting held in April 1985. Review industry responses

to: an NRR letter to utilities pertaining to possible changes in

R.G.1.108; evaluate ACRS concerns; and analyze Japanese experiences.

4-53
4.6 MOTOR-OPERATED VALVES (MOVS) NUREG/CR-4234 VOL.1 JUNE 1985 ORNL/FIN NO.

B0828

Aging and Service Wear of Electric Motor-Operated Valves used in Engineered Safety Feature Systems of Nuclear Power Plants

1.0 Background

This is the summary of the Phase I report on electric motor-operated

valves (MOVs) to be produced under the Nuclear Plant Aging Research Program

(NUREG-1144). The program includes the evaluation and identification of

methods for detecting, monitoring, and assessing the severity of time-dependent degradation (aging and service wear) of MOVs in nuclear plants. These

methods are to provide capabilities for establishing degradation trends

prior to failure and developing guidance for effective maintenance.

The topics of this report are failure modes and causes resulting from

aging and service wear, manufacturer-recommended maintenance and surveillance

practices, and measurable parameters (including functional indicators) for

use in assessing operational readiness, establishing degradation trends, and

detecting incipient failure. The results presented are based on information

derived from operating experience records, nuclear industry reports, manufacturer-supplied information, and input from a architect-engineer firms and

plant operators.

Failure modes are identified for both the valve and the motor-operator

assembly. For each failure mode, failure causes are listed by subcomponent

or subassembly, and parameters potentially useful for detecting degradation

that could lead to failure are identified.

4-54
2.0 Summary of MOV Application in Nuclear Power Plants

- ysten WAmbOr of T<V Bi* valve

sr Sov tV (im)a type

am (typical)

Low-pressure cote spray -12 C'b Ca°

1sgb-pressure eoolant lajectlon S-14 4-24 C, Cl

Lo-preasure colant injection 28-50 4-24 C, Cl

(include residual beat removal ad contalmnat spray)

Reactor recircnlation gyaten 5-10 2-28 C

eactor ore iolation cooling J-10 3-4 C, a

Contalent lsolaton 4-14 3-24 W' C

0r systems 50-150 2-60 C, C, N

PnU (typical)

£=liary fedwter 4-10 4- Cl

Cbmical d oluim control 8-10 3-6 C

Contalmset pray 5-12 5-14 C, C1, 

Delc generator coollng *-10 4-12 Cl

jigb-pressure coolant njection 10-24 2 112-4 C1

lasting, ventilation, nd air 5-10 5-48 3

conditionIng

Low-pressure coolant injectionl 12-30 *-10 C

residual bat rmoal

Safety injection 15-54 3-14

Refueling _ ter storage tank C

Contamet spray C Cotanet p C

a systms 50-150 2-60 oC, C, 

awai~ual pipe sire.

ba - gate valve.

OC - globe valve.

dv - btterf ly valve.

3.0 Results/Findings

The NPRDS, operated by INPO, contains component and failure data that

can be obtained on special request. In response to a special request, INPO

provided a data search that yielded 722 failure events for this study. The

events were then manually sorted as to failed component, which was usually

found in the narrative portion of the event description. Additional manual

sorting was performed for manufacturer, failure mode, cause, type, and method

of detection. Tables 3.1-3.6 summarize the results of this effort.

4-55
3.1 Failure vents Related to Failed Components

Failed component Number Percent

Limit witch/torque switch 275 38.1

Power supply/controls/wiring 133 18.4

Motor 125 17.3

Mechanical internals 90 12.5

Housing, bolts 27 3.7

Valve 17 2.4

Lubrication 13 1.8

Not determined 42 5.8

Total 722 100

3.2 Failure Events Related to

Manufacturers

3.3 Failure Mode Distribution

Failure modes Number Percent

Will not open 144 20

Will not close 122 17

Will not start/move 93 13

Out of adjustment 68 9

Out of limits 55 8

Will not operate per demand 45 6

Fracture/break 21 3

Spurious operation 18 2

False response 17 2

Crack 16 2

Miscellaneous (23 categories) 76 11

Other 47 7

Total 722 100

Manufacturer Number Percent

Limltorque 667 92

Rotork 18 3

Other 37 5

Total 722 100

3.4 Failure Types

Event (type) failure Numbera

Electrical 368

Mechanical 359

Electronic 34

Other instrument 18

Corrosion 10

Other chemical 3

Nuisance 2

Human error 2

aSome events had more than

one failure type indicated.

3.5 Failure Caused Distribution

Cause Number Percent

Burned/burned out 68 9

Set point drift 65 9

Abnormal wear 58 8

Short/grounded 45 6

Connection defective 43 6

Material defect 41 6

Open circuit 40 6

Instrument/switch miscalibration 39 5

Circuit defective 36 5

Miscellaneous (16 categories) 143 20

Other 124 17

None given 20 3

Total 722 100

3.6 Method of detection

Detection Number

Operational abnormality 257

Surveillance testing 247

Routine surveillance 88

-Incidental observation 57

Visual/audio alarm 57

Preventative maintenance 25

Special inspection 23

In-service inspection 20

Other 7

a~~~~~~~~~~~~~~~~~~~~

Some events had more than on

detection code listed.

4-56
3.7 Summry of NOV aging nformation available from L-based valve failure studies

Identified Frequencies P chanieol, Affected yste. er

failure of urveillance Re rks

mode or cause. failures odfeti WI

nherent No No rW and S Contatinent Iso- Data based on ECON keyword search for

aintenahce error conditions lation valve failure LRs

esign error Containment iso- Reactor coolant

lation ECCS

ailed to open or LER rate * No Contamnaent Iso- Containment Iso- Gross constant failure rates estimated

close as required 7 10- 3

/day lation lation for major valve types in selected

nternal leakage for MOVs~ Maitn tem Nonsafety safety systems - results should be

LPCI/RHI systems used with caution

Main steam

orque switch No Resize motor No No Recommended standards for:

failure operator NOV actuator specifications

etor damage Improve maintenance Sizing procedures

rong lube procedures aintenance and handling procedures

PAD of HOVs

versized operators No No Main steas Reactor coolant Recommended standards for:

orque switch ECCS Main steam nOV actuator specification.

failure or Siting procedures

mtsadjustment Maintenance and handling procedures

R&D of OVa

versized operators No Larger valve stems NSS system NSS systems Oversized actuators and misadjusted

mproper thermal Replace overload NSS relief IMP systems torque/limit switches are preoverload heaters valves Feedwater dominant cause of valve structural

dersized operators Resize motor SOP ystems

operators

Training of aintenance personnel

orque switches Yes, on a Evaluate thermal HPCI Auxiliary Some technical inaccuracies; disagreement

imit switches per-plant overload bypassing RCIC feedwater with NP-2'1 regarding oversized valve

otors basis Signature tracing operators: study suggests that changes

1978-1980 Improved torque in valve operating conditions lead to

switch procedures torque switch inaccuracies: propose

ON signature tracing

TW- Feedwater

MS - main steam

ECCS - Emergency core cooling system

NSS - Nuclear steam supply

RCIC - Reactor core solation cooling

6 Defined as natural failures such as those due to end of life, spurious, or unexplained.

It Is concluded that detection and onitoring ethods currently in

use for valves and motor operators are relevant, and their use should be

continued; the main issue is effective augmentation.

4-57
4.0 Utilization of Research Results in the Regulatory Process

4.1 General

Phase I Aging characterization and I, S MM of Components

a. Regulatory Guides

- 1.147, Inservice Inspection Code Case Acceptability-ASME Section

XI, Div. 1

- 1.148, Functional Specification for Active Valve Assemblies in

Systems Important to Safety in Nuclear Plants

b. Generic Issues

- Task II.E.6.1, In-Situ Testing of Valves

- Item B-58, Passive Mechanical Failures

- Item C-11, Assessment of Failure and Reliability of Pumps and

Valves

- Issue 54, Valve Operator-Related Events Occurring During 1978,

1979, and 1980

- Issue 105, Interfacing Systems LOCA at BWRs

- ASME Boilers and Pressure Vessel Code Section XI - Rules for

Inservice Inspection of Nuclear Power Plant Components

Phase II

ASME Operation and Maintenance Committee Standards

OM-2 Requirements for Performance Testing of

Nuclear Power Plant Closed Cooling Water System

OM-8 Requirements for Proportional and Periodic

Performance Testing of Motor-Operated Valve Assemblies

OM-10 Requirements for Inservice Testing of Valves in Light Water

Cooled Nuclear Power Plants

OM-15 Requirements for Performance Testing of Nuclear Power Plant

Emergency Core Cooling System

4-58
4.2- Specific

In nuclear power plants, testing requirements for identified MOVs are

contained in the plant Technical Specifications (TS), which specify the

overall In-Service Inspection (ISI) requirements for safety-related systems

and components. ISI and testing of ASME Class I, 2, and 3 components are

required to be in accordance with Sect. XI of the ASME Boiler and Pressure

Vessel Code. The purpose of TS requirements is to ensure operability of

components within specified limits required by readiness to perform the

desired safety function and not to specifically monitor degradation of performance.

Section XI testing requirements for MOVs involve only two parameters -

stroke time and seat leakage. With regard to stroke time, the code specifies

that the stroke time from test to test be compared. When the increase from

the previous test is sufficiently great (>25% or >50%, depending on stroke

time), the test frequency should be increased to once per month until corrective action is taken. Corrective action is required when the stroke time

exceeds its specified maximum value.

Leakage rate measurements also are compared with previous measurements,

and when the margin between measured and maximum permissible leakage is

reduced by 50%, the test frequency should be doubled. When projections

indicate that the leak rate will exceed the maximum permissible rate by >10%

prior to the next test, corrective action is required.

ASME Code Sec. XI contains surveillance intervals, frequencies, and

test requirements for Code Class 1, 2, and 3 systems.

4-59
4.7 CHECK VALVES NUREG/CR-4302 VOL. 1 DEC. 1985 ORNL/FIN NO. B0828

Aging and Service Wear of Check Valves Used in Engineering Safety-Feature

Systems of Nuclear Power Plants. Operating Experience and Failure

Identification.

1.0 Background

This is the summary of the Phase 1 report on check valves (CVs) to be

published under the Nuclear Plant Aging Research Program (NUREG-1144). The

program is concerned with the evaluation and identification of methods for

detecting, monitoring, and assessing the severity of time-dependent degradation (aging and service wear) of CVs in nuclear plants. These methods will

allow degradation trends to be established prior to failure and allow guidance

for effective maintenance to be developed.

The topics of interest for this report are failure modes and causes

resulting from aging and service wear, manufacturer-recommended maintenance

and surveillance practices, and measurable parameters (including functional

indicators) for use in assessing operational readiness, establishing degradation trends, and detecting incipient failure. The results presented are

based on information derived from operating experience records, nuclear

industry reports, manufacturer-supplied information, and input from plant

operators.

Failure modes are identified for CVs. For each failure mode, failure

causes are listed by subcomponent or subassembly, and parameters potentially

useful for detecting degradation, which could lead to failure, are tabulated.

4-60
2.0 Summary of CV Applications in Nuclear Power Plants

Valve Valve

Number Size Number Size

System of CVs (in.) System of CVs (in.)

BWR (typical) PWR (typical)

Low-pressure 10-18 2-28 Auxiliary feedwater 4-23 4-8

core spray

Containment spray 4-14 6-14

High-pressure 10-14 4-24

coolant injection HPCI 12-28 2-1/2-4

(HPCI)

LPCI/RHR 5-14 8-10

Low-pressure 10-21 4-24

coolant injection BOP systems 200-400 2-60

(LPCI) [includes

residual heat removal

(RHR) and containment

spray]

BOP systems 200-400 1/2-24

3.0 Results/Findings

3.1 Frequency of Failures

Two data bases contain failure frequencies: the NPRDS and the IPRDS

prepared by the Institute of Nuclear Power Operations (INPO). In their

"1981 Annual Report," for CVs up to 4 in., 3 failures/106 calendar hours

were reported during the time period of 7/1/74 to 12/31/81 for leak and

failure to stop. For 4- to 12-in. valves, 4 failures/106 calendar hours

were reported for the same period. Data for CVs of 12 inc. and larger were

insufficient to calculate failure rates.

4-61
FAILURE MODE DISTRIBUTION(1 )

Failure Mode Percent

Seat leakage 70

External leakage 16

Failed to close 8

Failed to open 2

Damaged internals 4

3.2 Methods of Detection

In an LER survey, surveillance testing found 32%, while 10 CFR 50 Appendix J leakage testing found 27%. Another 28% of the failures were detected during normal operation. Only 1% of the failures occurred during an

operational demand.

Detection( ) Percent

In-service and surveillance test 67

Incidental observation 4

Routine observation 14

Operational abnormally 11

Maintenance 2

Special inspection 2

(1) Based on 382 events involving NPRDS component valve, component engineering Code C (Check Valves)

4-62
3.3 Maintenance Actions

The IPRDS, which extracts repair information from plant maintenance

records, is the only data base that contains detailed information on maintenance actions performed on failed CVs. However, because of insufficient

entries for CVs, this data base could not be included.

Maintenance activity is sometimes stated briefly in the LERs. Based on

these reports, the valves were repaired 542 of the time and replaced 11%.

About 25% of the LERs did not indicate any maintenance activity.

Activity(1) Percent

Repair/replace 93

Modify/substitute 4

Temporary measure 3

3.4 Modifications Resulting from Failures

The operating experience data bases do not contain detailed descriptions

of postfailure modifications. Some IE publications have outlined a few CV

modifications, which are summaried below:

1. Improved soft-seated valve seals - Hard seat valves were modified to

a combination soft and hard seat configuration. Several types of

soft rings were tried before a molded (one-piece) seal provided a

satisfactory leaktightness.

2. Obturator attachment - The locking device that secures the obturator

to its hanger wore sufficiently to allow the obturator to fall free

of the hanger.

4-63
IDENTIFIED FAILURE CAUSE( )

Failure Cause Percent

Aging/cyclic fatigue 7

Normal/abnormal wear 50

Binding/mechanical damage 6

Lubrication problem 2

Previous repair/installation 2

Corrosion 4

Weld related 2

Dirty 14

Pa'rticulate contamination 1

Out of adjustment 3

Foreign/incorrect material 3

Unknown 1

Connection defect/loose part 3

Material defect 2

4-64
3.5

Summary of Valve Part Failure Assessments

Part Materials Significant Stressors & Measurable Parameters

Body, cap

(bonnet)

Fasteners

Stainless steel

Stainless steel

Mechanical: obturator guide

wear, galling body wear,

rupture

Chemical: corrosion, erosion

Mechanical: loosening,

breakage

Chemical: corrosion

Dimensions, appear

ance, roughness

cracking

Torque, appearance

Seat Stainless steel

hardened alloy

Resilient

material

Mechanical: wear

Chemical: erosion,

corrosion

Leakage rate, dimen

sions, appearance

cracking

Obturator

Obturator

hanger

Hanger pin

Hanger pin

bearing

Seals,

gaskets

Stainless steel

with hardened

alloy seating

Stainless steel

Stainless steel

Hardened alloy

Asbestos type.

Stainless steel

Resilient

material

Mechanical: wear

Chemical: erosion,

corrosion

Mechanical: wear, fracture

chemical; erosion, corrosion

Mechanical: wear, fracture

Chemical: erosion,

corrosion

Mechanical: wear fracture

Chemical: erosion,

corrosion

Mechanical: distortion,

compression

Thermal: hardening,

embrittlement (nonmetals)

Chemical: corrosion

Leakage rate, dimensions, appearance

cracking

Dimensions, appearance

roughness

Dimensions, appearance, roughness

Dimensions, appearance

External leakage,

appearance, noise

torque or force

applied for obturator movement,

packing gland

position

4-65
.........

4.0 UTILIZATION OF RESEARCH RESULTS IN THE REGULATORY PROCESS

4.1 General

4.1.1 Phase I Utilization

Regulatory Guides

1.147 Inservice Inspection Code Case Acceptability-ASME Section XI,

Div. 1

1.148 Functional Specification for Active Valve Assemblies in Systems

Important to Safety in Nuclear Power Plants

Generic Issues

Task II.E.6.1 In Situ Testing of Valves

Item B-58 Passive Mechanical Failures( )

Item C-11 Assessment of Failure and Reliability of Pumps and

Valves( 1 )

Applicable Standard Review Plan (SRP) Chapters:

3. Design of structures, components, equipment and systems

5. Reactor coolant system and connected systems

6. Engineered safety features

9. Auxiliary systems .

10. Steam and power conversion systems

4.1.2 Phase II Utilization

ASME Boiler and Pressure Vessel Code

Section XI - Rules for In-service Inspection of Nuclear Power Plant

Components

4-66
ASME Operation and Maintenance Committee Standards(2)

OM-10 Requirements for Inservice Testing of Valves

in Light Water Cooled Nuclear Power Plants

OM-2 Requirements for Performance Testing of Nuclear

Plant Closed Cooling Water Systems

OM-15 Requirements for Performance Testing of Nuclear Power

Plant Emergency Core Cooling Systems

(1) Closed out as of July 9, 1985, but will be addressed in NPAR program.

(2) Draft Maintenance Program Plan - RES resources (May 24, 1984).

4.2 Specific NRC Efforts Related to Check Valves

1. San Onfore Water Hammer Event (IE Lead): As a result of this event,

the staff has been directed to implement and coordinate actions

necessary to assure the reliability of safety related-check valves.

Including:

a. The adequacy of check valve design for application in the feedwater and other systems.

b. The adequacy of Inservice Testing (IST) Programs and procedures

to detect degraded and failed valves.

c. The adequacy of check valve maintenance programs to assure operability.

2. Generic Issue TMI II.E.6.1, In Situ Testing of Valves (NRR Lead):

Study and make recommendations to change current in situ testing

methods in order to provide assurance that valves (including check

valves) will perform their safety-related functions under design

basis as well as operating conditions.

3. Generic Issue 105, Interfacing System LOCA in LWRs (NRR Lead):

Perform a risk analysis and a cost benefit analysis of leak testing

reactor coolant boundary pressure isolation valves, including check

valves. The results of this effort will be used to, among other

things, support resolution of Generic Issue TMI II.E.6.1.

4. Generic Issue 93, Steam Binding of Auxiliary Feedwater Pumps (NRR

Lead): Determine an appropriate means to prevent steam leakage past

check valves which could result in steam binding of the auxiliary

feedwater pumps. This issue is being treated separately from the

more general statement of the problem (Generic Issue TMI II.E.6.1.)

because of its direct effect on plant safety.

4-67
4.8 AUXILIARY FEEDWATER PUMPS NUREG/CR-4597 MARCH 1986 ORNL/FIN

NO. B0828

Aging and'Service Wear of Auxiliary Feedwater Pumps for PWR Nuclear Power

Plants

1.0 Background

This report addresses time-related degradation of pressurized-water

reactor (PWR) power plant auxiliary feedwater pumps (AUXFPs). Since failures

of these components can reduce the amount of feedwater available for removing

heat when the usual feedwater supply is unavailable, such failures can result

in altered safety margins for PWR systems.

Auxiliary feedwater pumps (AUXFPs) are basically small boiler feed

pumps as used in small capacity, mostly older fossil fuel electric generating

plants. Thus they retain most of the design features, plus the potential

operating and reliability problems inherent in the feedwater pumps.

Plants have traditionally used AUXFPs per PWR. In most plants, one of

these is steam turbine driven and the other is electric motor driven. The

steam turbine driven unit has the advantage of variable speed and the maximum

design operating speeds are generally in the 4000 and 5000 rpm range. The

motor driven unit is usually designed for two-pole induction motor speed,

and powered from either one of two standby diesel driven generators. Newer

plants may also have a third AUXFP per reactor which is called the "backup", and is electric motor driven either from main line power or switchable

to the diesel driven generator bus.

The overall approach which now is emerging in the newest plants is to

have the "back-up" AUXFP be a non-safety related, non-nuclear class (non4-68
class) pump (thus less expensive) and to use it for all normal plant startup, shut-down and non-emergency service.

2.0 Summary of Applications

AUXFPs are used to supply feedwater to the steam generators under plant

startup, shutdown, and emergency conditions. These pumps have a best-efficiency-point (BEP) flow at any operating speed and if used in a continuous

operating mode, normal delivered flow is between approximately 50% and 120%

of the BEP flow. However, the AUXFP application is by definition a transient

operational mode, i.e., startup, shutdown, and emergency. Thus, one could

say that there are basically two operating regimes: (a) standby and (b)

normal. Normal includes any flow from shut off and bypass flow to full

run-out flow. Alternately, if the term "normal operation" pertains to that

regime in which the pump resides most of the time, then normal operation

would be the standby condition and any condition under which the AUXFP is

pumping would be considered off-normal.

Nearly all AUXFPs presently in-service are installed with little or none

of the specific parameter monitoring devices which are commonly integrated

into continuously running main power cycle equipment. Most auxiliary feedwater systems are not presently configured to facilitate regular periodic

testing of AUXFPs over the complete range of flow rates that would simulate

the various emergency scenarios for which AUXFPs are installed. The present

surveillance practice as required by plant technical specifications (TS),

which consists primarily of starting each AUXFP once every one to three

months for a short duration test at bypass flow, is inadequate to verify the

pump's operational readiness. The Current TS surveillance requirements do

4-69
not require the collection of adequate data for performance trending to

establish a correlation between AUXFP operating parameters (e.g., vibration,

bearing temperatures, etc.) and wear and aging. Consequently, with present

typical installations, aging and service wear monitoring would have to be

based primarily upon a detailed and thorough inspection of the pump in the

disassembled state. Unless specific operating problems arose with a particular AUXFP, such a detailed inspection would be practical only during a

scheduled plant shutdown, such as at refueling time.

3.0 Results/Findings

Three failure modes apply to auxiliary feedwater pumps: (a) failure to

operate, b) failure to operate as required, and (c) external leakage.

Failure causes corresponding to these failure modes are identified in general

terms and subsequently described in most specific terms in a section on

failure cause analysis. The single most important factor relevant to auxiliary feedwater pump potential failures is the presence of large hydraulic

forces within such pumps, particularly at low flow rates substantially different than the best-efficiency flow. In most plants the AUXFP bypass line

is a single line sized to pass 5-15% of best efficiency flow. Thus, this

type of testing could be the main contributor to wear and aging of various

AUXFP components. In some newer plants (e.g., Palo Verde) an additional

full-flow bypass line is provided to allow testing of the AUXFPs over the

full operating flow range, even when the plant is operating in a normal

generating mode.

4-70
SUMMARY OF AUXFP-TYPE FAILURES

REPORTED IN LERS (1973 THROUGH 1983)

AUXFP-TYPE FAILURES REPORTED IN

NPRDS DATA BASE (1974 THORUGH 1985)

Failed component

Bearings

Packing and seal

Casing

Internal components

Impeller

Capacity

Shaft

Other

48

30

4

4

2

2

2

8

Methods of detection

Testing

Operation

Maintenance

Not stated

42

29

6

23

Failed component

Packing/gasket

Bearings

Internal components

Shaft

Methods of detection

Incidental observation

Surveillance testing

Routine observation

Audiovisual alarm

Operational abnormality

Special inspection

Maintenance action

Repair/replace

Modify

Failure cause

Wear

Lubrication

Binding

Aging

Abnormal stress

50

38

6

6

33

25

12

12

12

6

Maintenance action

Replacement

Repair

Modification

No repair required

94

6

67

25

6

2'

Identified cause

Lack of lubrication or

cooling

Maintenance error

Wear/end of life

Design error

Crud

Operator error

Other

Not stated

Unknown

58

12

12

12

23 6

17

15

6

4

2

4

19

10

4-71
SUMMARY OF IMPORTANT AUXFP PART FAILURE ASSESSMENT

Part Materials Significant tresorea; and Measurable parameters failure causes

Shaft and 400-series S.S. (a) echanical; breakage (a), b), and (c) vibration,

fasteners (b) hydraulic; breakage and wear bearing temperature, appear-

(c) tribological; wear and ance, transmitted toque

seizing

Impellers CrNl alloy steels, (a) mechnnical; breakage (a) and (b) rotor unbalance

17-4Ph (b) hydraulic; breakage and wear vibration, appearance, de-

(c) tribological; wear-surface livered flow

wear (c) rotor vibration, appearance

Thrust 400-series S.S. (a) mechanical, breakage, seizing (a), (b) and (c) transmitted

runners (b) hydraulic; breakage, eizing toraue, rotational speed,

(c) triboloxical; rubhing, lub- rotor axial position

ricant dirt and breakdown

Stationary 4 00-series S.S. (a) hydraulic; breakage (a) delivered flow, appearvanes (dif- ance, vibration

fuser or

volute)

Wear rings 400-series S.S. (a) mechanical; seizine (a), (b) and (c) vbration,

(b) hydraulic; seizing transmitted toraue, delivered

(c) tribological; wear-surface flow, aoearance, clearance

wear

Thrust 400-aeries S.S. (a) mechanical; breakage, seizing (a), (h) and (c) rotor axial

balancers (b) hydraulic; breakage, seizing position, transmitted toroue,

(c) tribological; wear rotational speed, appearance

Thrust Rolling contact (a) mechanical; breakage (a), (b) and (c) rotor axial

bearincs elements (b) hydraulic; breakage poaltion, transmitted torcue,

(Specialty steels) (c) tribologlcal; wear rotational seed, ppearance,

clearance

Radial Bearing white (a) mechanical; breakage, seizing (a), (b) and C) rotor vibrabearings metal (typically (b) hydraulic; breakage, seizing ton, bearing temperature,

tin-base babbitt) (c) triboloxical; seizing, wear transmitted toraue, apearance

Shaft Stuffing-box or (a) mechanical; breakage, wear (a), (h) and (c) seal leakage

seals mechanical type (b) hydraulic; breakage, wear rate, rotor vibration, local

(c) tribologlcal; wear shaft, temperature

Coupling Gear type (a) mechanical; breakage, wear (a). (b) and (c) rotor vbra-

(usually) (b) hydraulic; breakage, wear tion, transmitted toraue,

(c) tribological; breakage wear appearance

4-72
4.0 Utilization of Research Results in the Regulatory Process

The results of this and subsequently planned studies will be useful to

the regulatory process and can be factored into this process through the

following vehicles:

a. Regulatory Guidelines: 1.147 Inservice Inspection Code Case Acceptability - ASME Section X1, Division 1

b. Generic Issues: Item C-ll Assessment of Failure and Reliability of

Pumps and Valves

c. Technical Specifications: The present technical specification and

surveillance testing requirements should be reconsidered to ensure

the operational readiness for all AUXFPs after each time the pump

surveillance requirements are satisfied.

d. ASME Boiler and Pressure Vessel Code Section X1 - Rules for Inservice

Inspection of Nuclear Power Plant Components

e. ASME Operation and Maintenance Committee Standards

OM-2 Requirements for Performance Testing of Nuclear Power Plant

Closed Cooling Water Systems

OM-6 Requirements for Performance Testing of Pumps in Light-Water

Cooled Nuclear Power Plants

OM-14 Requirements for Vibration Monitoring of Rotating Equipment

OM-15 Requirements for Performance Testing of Nuclear Power Plant

Emergency Core Cooling Systems

4-73
4.9 Snubbers NUREG/CR-4279 February 1986 PNL/FIN No. B2865

Aging and Service Wear of Hydraulic and Mechanical Snubbers Used on

Safety- Related Piping and Components of Nuclear Power Plants.

1.0 Background

This report presents and overview of hydraulic and mechanical snubbers

used on nuclear piping systems and components, based on information from the

literature and other sources.

The initial effort was to review the applicability of snubbers, evaluate

the benefits of reductions in snubber population, and provide input to the

preparation of appropriate regulatory guides. Therefore, this report does

not cover all of the Phase I elements of aging and wear assessment for snubbers. .It is expected that the Phase II effort will be directed at aging and

defect characterization and to develop recommendations for inspection, surveillance and maintenance. 

The functions and functional requirements of snubbers are discussed.

The real versus perceived need for snubbers is reviewed, based primarily on

studies conducted by a Pressure Vessel Research Committee. Tests conducted

to qualify snubbers, to accept them on a case-by case basis, and to establish

their fitness for continued operation are reviewed. A conclusion from this

report is that many snubbers installed in nuclear power plants may be unnecessary and could be removed. Work outside the scope of this report has confirmed that the removal of many snubbers can be justified. An approved

approach to evaluate snubber removal has been incorporated into ASME III and

approved by the U.S. Nuclear Regulatory Commission (NRC) on a case-by-case

basis.

4-74
-

As more snubbers were used, several operating problems arose; for example, degradation and leaking of seals on hydraulic snubbers and functional

failures of both hydraulic and mechanical snubbers. These problems led to

increased qualification and testing requirements. Thus, the original cost

of a snubber represents only a small fraction of the overall cost of qualification, installation, maintenance, and testing. In addition, maintenance

and testing result in substantial radiation exposure in older plants.

Another problem, that was not recognized initially, was the limitation

on in-service inspection (ISI) resulting from the large numbers of snubbers

and supports that prevent access to many welds in piping systems. A further

problem was the concern that stiff piping systems may be inherently more

susceptible to overloading and possible failure than flexible systems.

As a result of these concerns, Technical and Steering Committees on

Piping were organized under the Pressure Vessel Research Committee PVRC)

with active industry and NRC participation. In the past two years, these

groups have developed a more relaxed interim position on seismic damping, a

modified and less conservative position on spectra broadening, and a document on industry practice related to design approaches leading to fewer

snubbers. These positions have been accepted by the NRC on a case-by-case

basis, and portions have been incorporated into Appendix N of ASME III (the

reactor construction code). A task group on seismic design under an NRC

Piping Review Committee has recommended that the case-by-case status be

converted to generic positions. The NRC Executive Director for Operations

has issued a directive to develop such generic positions as cited in NUREG--

1061.

4-75
2.0 Summary of Applications

Snubbers of the type discussed in this report have two functions: 1)

they should move freely at low accelerations and 2). they should lock up at

higher accelerations. The valid use of snubbers is in locations of limited

clearance and possible high thermal expansion.

Several organizations have reported that a typical 1100-MWe capacity

boiling-water reactor (BWR) can have 9,000 to 10,000 supports on seismic

Category I piping (as many as 800 spring hangers and 1500 snubbers). An

1100-MWe pressurized-water reactor (PWR) could have 7,000 to 10,000 supports

(200 spring hangers and 950 snubbers).

Engineering, fabrication, construction, and hardware costs will be

lower with fewer supports; however, when the cost of items such as analysis,

computer time, and reconciliation are considered, the cost difference in the

two approaches may not be significant. The difference is heavily influenced

by two factors: 1) the installed cost per support and 2) the total life (40

years) cost per support. The second factor refers to inspection and maintenance costs associated with snubbers--typically $5,000 to $10,000 per snubber

on small-bore piping.

The table below provides a basis for developing a program on older

hydraulic and mechanical snubbers. It contains the parameters cited in the

Draft Regulatory Guide, ASME M4, and ASME XI. The table is a qualitative

presentation of aging factors such as wear, corrosion, and contamination. A

selection of both hydraulic and mechanical snubbers could have a series of

tests covering spring rates, release rates, dead band, drag, etc., that

would note deviations from normal. The snubbers could then be disassembled

to ascertain the causes for the deviations. Any quantification will depend

4-76
on the snubbers that are selected. ASME M4 is suggested as a benchmark for

such testing.

Common Causes for Degradation in Snubber Operating Characteristics

Measured Parameter(a)

Stiffness Activation Release Dead Friction

(spring rate) Force Rate Band (drag/breal

_ _ - M M

Cause

Wear

kaway)

Corrosion

Viscosity

Temperature

Entrapped air

Contamination

H, M

H

H

H

H

H

H

H H

H,M

(a) M - mechanical

H - hydraulic

The parameters in the above table, such as corrosion and viscosity are

strongly influenced by temperature and irradiation. Thus, times to failure

may differ markedly from one portion of the plant to another. This is of

concern because an inherent assumption in testing a small sample is the

homogeneity of the population. Snubbers taken from one region of a reactor

(for example, a cooler region) could display a markedly different failure

history than ones removed from another region (for example, a BWR dry well).

This issue could warrant consideration in selecting the sample to be tested.

4-77
3.0 Results/Findings

A review of available data reveals the large number of failure mechanisms, many of which are related to aging/degradation. In the case of hydraulic snubbers, several of the failure mechanisms result in a loss of

fluid and a consequent loss of function. This occurrence will be a fail-safe

mode in that there should be no lockup and overstressing of the pipe. Based

on early arguments on flexible versus stiff piping, the lack of operation of

hydraulic snubbers during a seismic event should have limited effects.

Whether the same can be said concerning more severe dynamic loads depends on

analysis or testing. It can be concluded that the loss of function of some

hydraulic snubbers should have a limited effect on the probability of failure

of piping during a seismic event.

With mechanical snubbers, the situation is not as clear. One obvious

aging/degradation mechanism--corrosion of internals over a period of time--has

been known to lead to loss of function and lockup. In this instance, the

situation is not fail-safe; and, mechanical snubbers located in regions of

high thermal expansion of piping during heat-up and cool-down may severely

stress snubbers or piping, or both. If sufficiently high stresses occur, the

snubber may be torn from the wall, or the pipe may be damaged. In regions

of lower thermal expansion, the locked up snubber may exist for a substantial

time. If a dynamic load occurs, the snubbers will behave as rigid members,

and they or the pipe may be damaged. Mechanical snubbers have locked up

without .being detected for several days or weeks.

Several utilities have extensive snubber replacement programs, replacing

one form of hydraulic for another or replacing mechanical for hydraulic.

Thus, the age factor cannot be quantified using an assumption that the same

4-78
snubbers were originally in the system. The same can be said for seals and

seal materials in hydraulic snubbers where several changes have occurred in

original materials as well as replacement materials.

Obvious degradation mechanisms include seal aging in hydraulic snubbers

and vibrations in mechanical snubbers. Several load conditions, such as

water and steam hammer or valve closure, can render snubbers inoperable.

Under these conditions, loss of function can occur whether the snubbers are

new or old.

If the Draft Regulatory Guide were converted to an active Regulatory

Guide, the aspect of an inspection program should be resolved. While it

could be argued that further modifications could result in improvements, it

appears that the draft Regulatory guide covers the salient aspects pertinent

to an inspection program.

Environmental qualification has been handled on a case-by-case basis,

and there is no assurance that a generic program exists. Furthermore, there

are several aspects for such a program. Under normal operating conditions,

an ocean site will require a different environmental qualification program

than an inland site. In addition, a high-temperature dry well of a RWR will

be markedly different from reactor regions near ambient temperature. In

essence, the areas of environmental qualification are not well defined and

specific criteria are needed.

With regard to ISI and testing, there should be a reassessment of existing requirements to establish what modifications may be necessary. The

existing visual examination in ASME XI may not be adequate, and the bench

testing program probably requires review and modification. Currently, only

smaller snubbers are covered by ASME XI; expansion will be required due to

4-79
the failures of larger snubbers, particularly hydraulic ones. It may be

necessary to test in-situ in recognition of the massive size of these larger

snubbers. In this case, a simpler test may be necessary to determine the

functionality of these snubbers. ASME 'M4 represents a positive step toward

improving testing; however, it may not be totally adequate.

Example of Failure Modes and Causes for Power Station Snubbers(a)

a) Hydraulic Snubber Failure Modes(a)

Mode

Low activation in tension

Low activation in compression

Low bleed in tension

Low bleed in compression

High activation in tension

High activation in compression

High bleed in tension

High bleed in compression

No. of

Failed Snubbers

13

3

11

10

17

10

6

3

% of Snubbers % of Snubbers

Tested Failed(b)

8 2

2 6

7 23

6 21

11 35

6 21

4 13

2 6

b) Hydraulic Snubber Failure Causes

No.

Mode Failed

No observable defect

Degraded EP seals

Polyurethane piston seals

Poppet upside down

Debris in fluid

Poppet stuck

Activation adjustment screw broken

Piston/cylinder scoring

of

Snubbers

6

3

7

3

1

1

1

2

% of Snubbers of Snubbers

Tested Failed_

4 26

2 13

4 30

2 13

1 4

1 4

1 4

1 9

4-80
Example of Failure Modes and Causes for Power Station Snubbers (Cont.)

c) Mechanical Snubber Failure Modes (c)

Mode

High drag

Locked up

High acceleration

Locked up in compression

No. of

Failed Snubbers

14

5

1

1

% of Snubbers of Snubbers

Tested Failed_

21 66

8 24

2 5

2 5

(a)

(b)

(c)

A total of 160 hydraulic snubbers were

failed.

Failures total more than 100%, because

than one failure mode or cause.

A total of 66 mechanical snubbers were

failed.

tested; 48 snubbers (or 30%)

some snubbers exhibited more

tested; 21 snubbers (or 32%)

Number.and Proportion of Snubber Failures by Manufacturer for Data Set A 1)

.Manufacturer

Unknown

Bergen-Paterson

Blow Knox

ITT-Grinnell Corp.

Pacific Scientific

Anchor-Oolth

International

Nuclear Safeguards

Corp.

ITT Barton

McDowelL elton

Power Piping Co.

Unknown

22 (1%)

3 (9%)

Failure Cause

InstallDesign atlon 

104 (6%) 168 (10%)

425 (63%) 46 (7%)

1 (25%) 3 (75%)

256 (51S) 33 (7%)

9 (27%) 6 (18%)

91 -

Operation

1450 (83%)

207 (31%)

217 (43%)

15 (45%)

3

Unknown

45 (3%)

1 (1%)

3 (9%)

Failure Mbde

Out of Wou

Frozen Tolerance LO

150 (9%) 648 (37%) 901

1 (1%) 57 (8%) 619

- - 4

- 215 (42%) 291

13 (39%) 3 (9%) 14

- - 3

91 -

- -- 1

- - 1

Id Not

(52%)

(91%)

(1 00%)

(58%)

(42%)

I

I

I __

II _

(1) Developed from detailed computer search of LER abstracts (1973-1983).

4-81
Cost of Reducing the Number of Snubbers in Existing Plants

For backfit applications, the cost of redesigning and implementing a

more flexible system must be estimated and then subtracted from the predicted cost savings. The cost of backfitting plants to reduce the number of

snubbers has been estimated to be from $820,000 to $890,000 per plant. This

cost does not include the increased employee exposure that would result from

repiping and removing unnecessary snubbers. However, this exposure rate

might be expected to be on the same order as that incurred from inspecting

and maintaining a snubber (estimated to be 2.5 man-rem/snubber). The $1,000,

man-rem rule of thumb was used to quantify this cost. The cost savings

would be $2,500/snubber for each snubber eliminated.

In a plant containing 1,000 snubbers, a 30% reduction in the number of

snubbers could be expected to result in a total cost reduction of form

$526,000 to $3,500,000 (See table below). All of the previous caveats regarding the impact of hidden costs such as increased plant downtime or damage to

piping still apply. The exposure/implementation costs were obtained by

multiplying the unit cost ($2,500) by the reduction in snubbers (300).

Estimated Cost Savings Resulting from Reducing the Number of Snubbers

in an Existing Plant (a)

Total Costs. S

2% Dscount 5% Discount 10% Discount

Cost Cateqory Cost. S Timlnq Rate Rate Rate

CaOIta 0 - 0 0 0

Imolementatlon 855.000 0 855.000 855.000 855.000

Insoectlonfmaintenance (115,100) Annual (2,577,832) (769,369) 1.085.038)

Exposure/imp ementation 750,000 0 750,000 750,000 750.000

Exposure savinas (111,000) Annual (2.486,007) (1.706.342) (1,046.388)

Total (3,458,859) (1,870,711) (526,426)

(a) The followinq assumptions were used: 30-year rwmnlnina plant life; 2% eal

discount rate: 300 unit reduction In snubbers (and a representative

vocu latIon).

4-82
- -; 4 -hi

4.0 Utilization of Research Results in the Regulatory Process

The results of this and subsequent planned studies will be useful in

the regulatory process and can be factored into this process through the

following vehicles:

a. Regulatory GuideS: (1) NRC Standard Review Plan, Section 3.9.3 cites

the criteria applied for snubbers. Subsection II.3b,1-7 deals specifically with snubbers. Section 3.9.2 deals with piping preoperational

vibration and plant start-up test programs. (2) Draft RG SC 708-4,

Rev.1, February 1981 deals with qualification testing, acceptance testing and functional specification.

b. Generic Issues: Generic Letter 84-13, May 3, 1984, Technical Specifications for Snubbers.

c. ASME: () PVP-45, Criteria for Nuclear Safety Related Piping and Component Support Snubbers (1980) discussed acceptance testing. (2) Preservice and in-service testing are addressed in ASME XI and ANSI/ASME

OM4. (3) ASMEIII, Section NF-3000 covers design rules for supports.

4-83