BNL Technical Report

A-3270R-2-90

mtTrt

Interim Report -

Aging Effects of Important Balance of Plant

Systems in Nuclear Power Plants

Prepared by:

A. Fresco and M. Subudhi

Engineering Technology Division

Departmrent of Nuclear Energy

Brookhaven National Laboratory

February 1990

FIN A-3270

Prepared for the U.S. Nuclear Regulatory Commission

Office of Nuclear Regulatory Research
I

CONTENTS

Page

ACKNOWLEDGMENTS ................................... v

EXECUTIVE SUMMARY ................................ I

1. Introduction ........................................ 3

1.1 Purpose ....................................... 3

1.2 Study Approach/Scope ............................... 3

2. Summary of Prior NRC-Sponsored Balance of Plant Activities ............. 6

2.1 Environmental Qualification of Class IE Equipment .......... 6

2.2 AEOD Studies of Engineered Safety Feature Actuations and Unplanned

Reactor Trips .................................... 7

2.2.1 Engineered Safety Feature Actuations ................... . 7

2.2.2 Unplanned Reactor Trips .......................... . 7

2.3 NRR-Sponsored Study by the MITRE Corporation Concerning Regulatory

Considerations for Balance-of-Plant ..... . . . . . . 8

2.4 June 9, 1985 Davis-Besse Loss of All Feedwater Event - NRC Lessons

Learned ........ . .. . . .. . .. 10

2.5 SECY-86-349 Policy Issue for Balance of Plant ..... . .. . . .. .. . . . . 11

2.6 NRC Inspection Procedures for Balance of Plant ..... . . . . . . . . . . . . 13

2.6.1 Temporary Instruction 2515/83 Balance of Plant Trial Inspection

Program (Feedwater System) ..... . . . . . . . . . . . . . . . . . . . 13

2.6.2 NRC Inspection Procedure 71500 Balance of Plant Inspection .... . . . 14

2.6.3 Temporary Instruction 2515/97 Maintenance Inspection .... . . . . . . 14

2.7 Recent AEOD Performance Indicator-Related Efforts.

2.7.1 Operating Experience Feedback Report-New Plants (NUREG-1275,

Volume 1) .................................. . 15

2.7.2 Operating Experience Feedback Report - Progress in Scram Reduction

(NUREG-1275, Volume 5) ......................... . 17

2.7.3 Development of Maintenance Effectiveness Indicator .... . . . . . . . 21

2.8 NRR Survey of Plant Design Strengths and Weaknesses .... . . . . . . . . . 22

3. Identifying Important BOP Systems ...... . . . . ......... . . . . . . . . 24

3.1 Assessment of Important BOP Systems From Prior NRC Sponsored

Activities ........ . . . .. . . . . .. . . . . .. . . . . .. . . . . .. . 24

3.2 Categorization Based Upon Probabilistic Risk Assessment (PRA) Insights . . . 25

4. Interim Recommendations and Future Efforts ..... . . . ....... . . . . . . . 36

5. References ........ . .. . . .. . . .......... . .. . . .. . . .. . . 38

LIST OF TABLES

3-1 Typical Boiling Water Reactor (BWR) Systems Expert Opinion Importance

Categorization.................................... . 28-30

3-2 Typical Pressurized Water Reactor (PWR) Systems Expert Opinion Importance

Categorization. . . . . . . . . . . . 31-33

3-3 Abnormal Events Involving BOP Systems in PWRs and BWRs .34

iii
ACKNOWLEDGEMENTS

Grateful appreciation is extended to our NRC Program Coordinator Jit Vora and Program

Manager Satish Aggarwal for their guidance in implementing this program. The authors wish

to thank their colleagues John Boccio, John Taylor and Richard Travis for their valuable comments and insights. In particular, James Higgins also led in developing the expert opinion survey through which Balance of Plant Systems were categorized. Certainly credit is due to the

staff of the Technical Publishing Center at Brookhaven National Laboratory for their excellent processing of this document.

v
EXECUTIVE SUMMARY

In any commercial nuclear power plant, there are a substantial number of Balance of

Plant (BOP) systems besides the commonly thought of main steam, feedwater, turbine-generator, condensate, etc.In recent years, BOP systems have arisen as major causes of plant transients, e.g. the June 9, 1985 Loss of All Feedwater event at Davis-Besse, and have received

increased attention from.the nuclear industry and the NRC.

This is an interim report describing the activities to date by Brookhaven National Laboratory (BNL) to support the Nuclear Plant Aging Research (NPAR) Program's study of Balance

of Plant (BOP) systems. This initial phase of the study provides a rationale for and a preliminary indication of those BOP systems which may warrant detailed study for the effects of aging. A brief discussion of the suggested approach to accomplish the overall objective of

identifiying the effects of aging of these BOP systems on nuclear plant safety is also presented.

In this study, BOP systems have been defined as all nonsafety-related systems, except for

those associated with the Nuclear Steam Supply Systems (NSSS) such as Control Rod Drive,

Pressurizer Power Operated Relief Valves (PORVs) and discharge piping, Pressurizer Relief

Tank (PRT), etc. for PWRs and Reactor Core Isolation Cooling (RCIC), Control Rod Drive

Hydraulic (CRDH), etc. for BWRs. Some of the nonsafety-related NSSS are being studied in

other phases of the NPAR program.-

From the results of the study, it was concluded that the frequency of unplanned reactor ad :~ O ' -. A< ~ . - J ' . i ! ''* -a. _, i ,i

scrams has been more frequently cited as an indicator of safety performance and that the fre- . . -.; , .~ . - , ,

quent contributors to unplanned reactor scrams caused by BOP systems are largely the power

conversion systems, i.e. the feedwater, main turbine, main generator, main steam (usually the

steam bypass to the main condenser) and the condensate systems. Other frequently contributing BOP systems are support systems such as the Electrical Distribution System, and less frequently the Circulating Water, Service/Instrument Air, Fire protection and HVAC Systems.

The Electrical Distribution System consists of the 120V AC Power, switchyard, large plant

1
loads, DC power and control centers. At a component level, the feedwater regulating valves

*and turbine-driven fedwater pumps are frequent conributors, as well as the main turbine

electro-hydraulic control (EHC)'subsystem. Failures in the 'main generators are also important.

These results coincide substantially with' the results from an alternative approach in

which important BOP systems were'categorized considering PRA insights.

Preliminary recommendations are as follows:

* The frequency of unplanned reactor scrams be considered the mosi important indicator of current, or potentially near term, safety problems.

* BOP systems that should be included within the NPAR program are those which significantld contribute to unplanned reactor scrams.

The next phase of this program will focus on several of the oldest plants. LERs in'volving

unplanned reactor scrams since the beginning of commercial operation until the present time

will be reviewed to determine if there is an increasing frequency of unplanned scrams caused

by the identified BOP systems. This group will include Monticello and Yankee Rowe, the pilot plants for the plant life extension study.

After completing the above, a group of intermediate age plants will be examined in the

same manner, followed by the youngest plants. These three groups will be compared and analyzed. The ultimate goal is to determine whether aging of the individual system components is

a significant factor affecting nuclear plant safety. This will be done in the same manner as has

been done for safety-related systems within the NPAR Program.

2
1. INTRODUCTION'

The U.S. Nuclear Regulatory Commission's Office of Nuclear Regulatory Research is

the prime sponsor of the ongoing Nuclear Plant Aging Research (NPAR) Program which focuses on the long-term safety implications of aging of nuclear plant systems and components.

In recent years, Balance of Plant (BOP) systems have arisen as major causes of plant transients, e.g. the June 9, 1985 Loss of All Feedwater event at Davis-Besse, and have received

increased attention from the nuclear industry and'the NRC. As such, these systems have also

been selected as targets for study in the NPAR Program.

1.1 Purpose

In any commercial nuclear power plant, there are a substantial number of BOP systems besides the commonly thought of manin steam, feedwater, turbine-generator, condensate,

etc. This is an interim report describing the activities to date by Brookhaven National Laboratory (BNL) to support the NPAR Program's study-of BOP systems. This initial phase of the

study provides a rationale for, and a preliminary indication of, those BOP systems which may

warrant detailed study for the effects of aging. A brief discussion of the suggested approach to

accomplish the overall objective of identifying the effects of aging of these BOP systems on

nuclear plant safety is also presented.

In this study, BOP systems have been defined as all nonsafety-related systems, except

for those-associated with the Nuclear Sfiam Supply Systems (NSSS) such as Control Rod

Drive, Pressurizer Power Operated Relief Valves (PORVs) and discharge piping, Pressurizer

Relief Tank (PRT), etc.' for PWRs' and Reactor Core Isolation Cooling (RCIC), Control Rod

Drive Hydraulic (CRDH), etc. for BWRs. Some of the nonsafety-related NSSS are being studied in-other phases of the NPAR'program.

1;2 'Study Appi6ach/Scope -;

- In Section 2 of this report, a summary of prior NRC-sponsored BOP-related activities

and findings is priesented. This information is necessary to ensure that a consensus definition

3
of BOP systems is being applied in this study and is also very useful in identifying and

prioritizing the important BOP systems.

These activities include the following:

* Environmental qualification of Class IE Equipment.

* AEOD studies of engineered safety feature (ESF) actuations and unplanned reactor

scrams.

* NRR-sponsored study by the MITRE Corporation concerning regulatory considerations for Balance of Plant.

* June 9, 1985 Davis Besse Event - NRC Lessons learned

* SECY-86-349 Policy Issue for Balance of Plant

* NRC inspection procedures for Balance of Plant, including the temporary instruction

for maintenance team inspections (MTI).

* Recent AEOD performance indicator- related efforts such as NUREG-1275, operating experience feedback reports for new plants and also progess in scram reduction, as well as the ongoing development of a maintenance effectiveness indicator

(MET).

* NRR survey of plant design strengths and weaknesses.

Nearly all of the above-referenced documents particularly SECY-86-349, the staff

policy on BOP, convey the emphasis placed by the NRC to reduce the frequency of challenges

to plant safety systems. In turn, unplanned reactor scrams are the leading contributors to

challanges to safety systems, so that extensive NRC focus has been placed on reducing their

frequency.

In Section 3, an assessment is made of the important BOP systems based on their contribution to unplanned reactor scrams. An alternative approach in which BOP systems are categorized by expert opinion considering insights from probabilistic risk assessments (PRA) is

also presented for comparison purposes. In addition, the relevant aspects of a prior study by

4
BNL staff regarding reactor trips and their relationship to initiating 'events appearing in a PRA

is also presented. The study proposed utilizini-the frequency of abnormal events as a

risk-based performance indicator.

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5
2. SUMMARY OF PRIOR NRC-SPONSORED BOP ACTIVITIES

2.1 Environmental Qualification of Class IE Equipment

With the publication of IE Bulletin 79-01, "Environmental Qualification of Class LE

Equipment" on February 8, 1979, and subsequent revision through Bulletin 79-OlB, the NRC

established certain requirements for the environmental qualification of nonsafety-related

equipment important to safety. These requirements were formally establishedin 10CFR50.49,

"Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power

Plants" as follows:

"(b) Electric Equipment important to safety covered by this section is:

(1) .....

(2) Nonsafety-related electric equipment whose failure under postulated environmental conditions could prevent satisfactory accomplishment of safety

functions specified in subparagraphs (i) through (iii) of paragraph (b)(1) of

this section by the safety-related equipment."

According to paragraph (b)(1), safety-related electric equipment is that relied upon to

remain functional during and following design basis events to ensure:

(i) the integrity of the reactor coolant pressure boundary,

(ii) the capability to shutdown the reactor and maintain it in a safe shutdown condition, and

(iii) the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to 1OCFR100 quidelines.

Licensees were also required in paragraph (d) to prepare a list of equipment important

to safety required by Section 50.49.

Some licensees responded by separating the equipment into two categories: Balance

of Plant (Inside and Outside Containment) and NSSS (Inside Containment).1 However, this

6
definition appears to include primarily the traditional non-NSSS safety-related components as

BOP components. -

It should be noted that aging of BOP components is indirectly considered in environmental qualification as a significant cause of the postulated environmental conditions to

which the safety-related components would be subjected. That is, stress-related aging, corrosion or erosion of so-called high energy and moderate energy piping fluid systems both inside

and outside containment leading to rupture or cracking of the piping would in turn cause safety-related components to be subjected to pipe whip, jet impingement, moisture and/or flooding conditions.

2.2 -AEOD' Studies of Engineered Safety Feature Actuations and Unplanned Reactor

Trips'''-; '- ' ':

2.2.1 Engineered Safety Feature Actuations

In 1985 and 1986, the NRC's Office for the Analysis and-Evaluation of

Operation'al Data (AEOD) issued studies of engineered safety feature (ESF) actuations at nuclear power plants for the periods of January 1 through June 30, 1984 and July 1 through December 31,1984, respectively. 2' 3 In general, the BOP systems involved in the actuations

which'occurred were HVAC systems for the control building due to overly conservative radiation monitor settings or actual toxic gas infiltration into the control room at a limited number

of plants. For the most part, strictly defined nonsafety-related systems were not significant

contributors to ESF actuations.

2.2.2 Unplanned Reactor Trips

Also' in 1985 and 1986, AEOD issued trends and patterns reports of unplanned

reactor trips which occurred in 1984 and -1985, respectively.4 5 These studies provided the first

clear indication of the significant impact of BOPsystems and components on challenges to

safety systems, e'g. unplanned reactor trips.

'7
In both years, the majority of reactor trips occurred with the reactor powersabove 15%, in 1984 - 68% and in 1985 - 75%. In 1984 and 1985, 31% and 38%, respectively,

of all trips occurred while the plant was at 95% power or above.

Above 15% power, the major BOP systems, collectively referred to as the

Power Conversion Systems, i.e. feedwater, turbine, condensate, main steam

and main generator contributed to 59% and 52% of all reactor trips, in 1984

and 1985, respectively.

Hardware failures were the dominant cause of unplanned reactor trips above

15% power. In 1984 and 1985, hardware failure contributed 60% and 55% respectively of all trips above 15% power. The Feedwater Regulating Valves,

Main Steam Isolation Valves, Turbine Control Valves, and Turbine Stop

Valves were major contributors. These valves were responsible for 8% of all

trips above 15% power in 1985.

* The Electrical Distribution and Reactor Protection Systems were the major

non-power conversion system contributors in both 1984 and 1985, contributing 23% and 29% respectively of trips above 15% power. (Portions of the

Electrical Distribution System can be considered BOP since they are

nonsafety-related.)

2.3 NRR Sponsored Study by the MITRE Corporation Concerning Regulatory

Considerations for Balance-of-Plant

In October 1986, NUREG/CR-4783 was issued which is the final report by the MITRE

Corporation concerning certain regulatory and performance aspects of the conventional or

6

power conversion side of nuclear power plants, also referred to as the Balance-of-Plant. This

study was commissioned by the Office of Nuclear Reactor Regulation and essentially concurs

with the previously described AEOD results by focusing specifically on BOP systems only

8
and providing a more detailed analysis separating the data into different categories based on

reactor type (PWR or BWR) and age, amongst others.

In the report, MITRE characterizes and analyzes BOP failures and provides a perspective on the significance of the failures relative to overall plant safety. The report also

encompasses recommendations to'enhance the NRC regulatory approach to BOP performance

and reviews current BOP-related activities and identified gaps. However, the report was not

intended to focus exclusively, if at all, on the effects of aging of BOP systems. Many of the

failures identified pertained to operator errors.;

The findings'for the periods studied, i.e. 1984 and 1985, concur with the previous

AEOD studies in that for PWRs, condensate and feedwater accounted for 52% of BOP-caused

reactor trip's 'while 30% were due to the'turbogenerator. Similarly, for BWRs, 47% of trips

were caused by condensate and feedwater, and 26% were attributable to the turbogenerator.

Other significant conclusions were that:'

* BOP-related trips constitute 70% of all plant trips (for the time period studied).

*.There are large'unit-to- unit variatiohs in BOP performance: 60% of the mean BOP

trip frequency among'older plants and over 100%'among newer ones.

* Units that have above-average performance in the initial-years tend to stay better,

and vice versa.

* The plants which'were outliers'amongst the older unit group account for about twice

their proportionate share of BOP-related trips.'

* Personnel errors' account ffr about 50% of the known causes of BOP trips.

* About 40%'of the BOP trips duetto human error occur during maintenance, test, repair and inspection activities.

* Root causes in more than 33% of "plant component failures are listed as unknown.

* Single component failures within the turbogenerator, condensate and related support

systems account for about 50% of BOP trips.

.9
2.4 June 9, 1985 Davis-Besse Loss of All Feedwater Event - NRC Lessons Learned

On November 26, 1985, William J. Dircks, the Executive Director for Operations, issued an internal NRC memorandum concerning the Lessons Learned from the June 9, 1985

loss of all feedwater event at Davis-Besse.7 The first lesson concerned more timely identification and completion of safety issues, the second concerned broader consideration of positive

and potentially negative safety impacts of regulatory actions, while the third reads as follows:

"3. Increased Emphasis on Balance of Plant Equipment

The paramount importance of proper maintenance in maintaining levels

of reliability assumed in the Safety analyses that form the licensing basis

for operating plants has been accorded greater recognition and increased

emphasis and attention by both NRC and utility management in the aftermath of the TMI accident. However, it appears from the circumstances noted in the review of the June 9 Davis-Besse event that an inappropriate, artificial distinction (alluded to in 2. above) between the

importance of safety-related vs nonsafety-related plant features may

have led some licensees to place inadequate emphasis on proper maintenance of all equipment necessary to assure proper facility operations.

Some balance-of-plant systems may actually have equal or perhaps

greater safety importance (cumulatively) than equipment classified as

safety-related because their too-frequent failure can needlessly challenge the safety-related systems, and their failure can also aggravate

conditions under which the safety-related systems must respond. We

need to give increased attention to assuring that the attention of licensee

management is focused properly on this important aspect of plant operations, and that important balance-of-plant systems and equipment receive adequate attention in the overall maintenance picture. We should

10
also consider 's'ilusly, in the context of finalizing our improvement

plans, whether this requires significantly increased commitment of regulatory attention to balance-of-plant areas within our licensing review

and inspection program." :'

2.5 SECY-86-349 Policy Issue for Balance Plant

As a result'of the Davis-Besse memo and the concurrent AEOD and NRR reports, the

NRC staff provided to th&e Commissioners 'an Information Policy Issue for:Balance of Plant,

designated SECY-86-349, dated November 21, 1986.8 This SECY document was intended to

describe' the BOP programs currently underway and those planned for the future.

With respect to the inspection process, the 'document states'that licensee activities relating to the'design', maintenance, operation and testing of BOP systems and equipment may

be examined to the degree they impact on safe'operations.' v

Furthieimore, it's'tates -that inspections of steam and power conversion BOP systems

are mostly raictive ih'nature. That is, where there are trips and transients or recurring performance problems with BOP systems and equipment which present significant challenges to

safety systems,;in'spectors' will, to varying degrees; 'explore causes and corrective actions with

the licensee. For'example, post-trip reviews-may lead to investigations and evaluations of

BOP where reactor trips are initiated by BOP equipment failures. Where events such as

feedwait&e pump'trips recur'andthe licensee is not identifying and dealing with the underlying

cause, the inspection process may result in discussions with the licensee and requests for appropriate licensee actions.

The document summarizes some of the specific approaches being taken by the Regional Offices in addition to the reactive BOP inspections, as follows:

Some Regional Offices address BOP performance and its impact on operational performance in SALP reports. In one case, statistics have been kept on several important

performance indicators such as the number of reactor scrams initiated by BOP equipI-11
ment. Licensee BOP performance is assessed in terms of this information and underlying problems are identified where they are known.

* Inspections are performed on modifications and testing of certain selected and reviewed non-safety-related or BOP modifications which are judged to rank high in

importance to overall plant safety.

* Resident Inspeciors in the normal course of their plant walkthroughs observe material conditions, housekeeping and work ongoing in BOP areas. Where problems are

observed, these are brought to the licensee's attention.

* In light of available information from PRA studies, IEt and the Regions are developing trial programs for inspections to focus more on major risk contributors irrespective of safety classification. In some cases this is significant in that BOP systems

have interconnections with safety-related systems such that their availability and/or

reliability is affected by the BOP portion. Consequently, the staff is developing tabulations from selected existing PRAs that would identify risk-significant systems and

components, associated failure mode(s), and identify areas warranting more frequent system walkdowns. Existing PRAs are being utilized to develop information

which can be applied to similar plants not covered by a plant-specific PRA.

Regarding future plans, the document mentions that a trial inspection program of a selected important BOP system such as the feedwater system is being developed. This is discussed in the next section.

Sic incorporated into NRR.

12
2.6; NRC Inspection Proc'edures for Balance of Plant

2.6.1 Temporary Instruction 2515/83 Balance of Plant Trial-Inspection-Program

(Feedwater System)

As a result of the preceding activities concerning BOP, and as stated in

SECY-86-349, the NRC issued Temporary-Instruction 2515/83 effective February 26, 1987.9

The purpose of the instruction was to implement a limited trial program to inspect a genieric

BOP system, i.e. the Feedwater System, that has been shown to be a significant contributor to

unplanned challenges to reactor' safety systems, as well as a significant contributor to overall

plant'risk from a'probabilistic risk assessment (PRA) perspective. The results of the trial inspection program were'used to assess how inspections of BOP equipment can be effectively

incorporated into the routine inspection program and to assist in determining whether greater

emphasis should be placed on inspecting BOP systems. The basic guidance specified in the

procedure is t6 assess licensee responsiveness to and adequacy of corrective action as an indicator of the'licensee's commitment to' safety. -The general categories for which specific inspection s s are given are openitions,-maintenanice, modifications, design, and management support.

Aging of components is not directly cited, but rather as an indirect cause of

equipment failure which should be evaluated for input to the preventive maintenance program

and for possible design modifications. The corrective maintenance program is inspected to ensure that corrective maintenance is documenied in maintenance history records, appropriately

trended, analyzed, a'nd'applied to the'preventive maintenance program through procedural revisionsordesign 'changes.'' -

Under operations; 'the inspection covers general conditions of equipment, e.g.

no excess corrosioi, 'no evidefice of steam, water or oilleaks, as;well as verification that

Feedwater System corrosion control measures are in place. - - -- '

; 13
It would appear that aging of components would not ordinarily be directly cited

as a cause of feedwater equipment failure by the NRC inspectors.

2.6.2 NRC Inspection procedure 71500 - Balance of Plant Inspection

Upon considering the results of the temporary. instruction discussed above, the

NRC determined the need exists for a permanent inspection procedure for BOP, and which

was not limited to the Feedwater System. Such a procedure was issued as No. 71500 on Septemiber 30, 1988.10

Inspection Procedure 71500 differs from Temporary Instruction 2515/83 in the

following ways. Design is no longer a separate category but has been combined with modifications as a single category. Also, a new category has been added entitled Root Cause.

The procedure relates only indirectly to aging of BOP components, in the same

manner as the previously described temporary instruction, for maintenance and operations. A

more direct relationship to aging of components may result from the procedure's instructions

for Root Cause. The inspection guidance states that at least four components that have a history of unreliability or have caused or complicated recovery from plant trips or transients

should be selected. The root cause for each component failure or event should be identified by

reviewing the licensee's root cause analysis, if performed. The analysis should implicate

broad programmatic areas such as maintenance, operator training, etc. that are contributors to

the component failure or event, and that are in need of upgrade.

2.6.3 Temporary Instruction 2515/97 Maintenance Inspection

In the last few years, the NRC has intensified its emphasis on maintenance,

most importantly by publication in the Federal Register of proposed rules to ensure the effectiveness of maintenance programs for nuclear power plants on November 28, 1988. The proposed rule states that it is the objective of the Commission that all components, systems and

structures of nuclear power plants be effectively maintained so that plant equipment will per14
form its intended function whe'n required.`-The scope of the p'roposed rule is intended to cover

all systems, structures and components including those in the Balance of Plant (BOP).

To'provide the background to the proposed rule, the NRC has had in effect a

Temporary Instruction, 2515/97, titled "Maintenance Inspection."12 It states that the inspection team will determine what failures of significant equipment (probabilistic risk assessment

(PRA)-identified, safety-related, orBOP that affects safety-related and special interest items)

have occurred and will inspect the licensee's trending and maintenance activities to schedule,

repair, and prevent further failure of that equipment. Other selective examinations of equipment failures attributed to maintenance will be examined to determine the adequacy of licensee corrective'actions and root-cause determinations of the failures.

By' means of the temporary instruction, the NRC has been conducting maintenance team inspection which will cover all US .'nuclear power plants and are targeted for comple'tion by April '1991.: It appears that the inspection reports arising from this effort may provide'v'aluable insights into aging of BOP components.

2.7 'Recent AEOD Perfor mance Indicator - Related Efforts

'In the'years since the first AEOD studies of Engineered Safety Feature Actuations and

Unplanned Reactor Trips, as described previously in Paragraphs 2.2.1 and 2.2.2, efforts have

been directed-generally within the NRC to establish various indicators of plant performance.

Particular emphasis has been placed on reductions in the frequency of unplanned reactor trips

(scrams), and also on the'effectiveness of utility maintenance programs as' a means to reduce

the scram frequency. AEOD's efforts in these areas are described below.

2.7.1 Operating Experience Feedback Report- New Plants (NUREG-1275, Volume

1)13

In this report, mature plants are defined as those which were licensed before

January 1, 1983 (76 plants). The study scope included the operational experience for 22 new

plants frpm January 1983 through June 1986. The data for mature plants, which were exam15
ined for contrast, were obtained from licensee event reports (LERs) submitted for Calander

Year 1985. With respect to BOP, the report concluded that:

* On a percentage basis, the causes of scrams at new plants are very similar to

those of mature plants. The primary causes are associated with the BOP systems, with the Feedwater System dominating.

* Focusing on the BOP prior to operation and early in plant life appears to provide a high return regarding the reduction of unplanned scrams and ESF actuations.

* In particular, additional reviews of feedwater and turbine control and bypass

systems should be conducted to identify sensitivities and plant-specific characteristics that could contribute to transients, or the (in)ability of the system to

cope with expected transients.

* Conducting systematic reviews of equipment-protective logics and setpoints

on components such as pumps (suction trip, time delay, vibration trip) or

power supplies to identify areas where a time delay or additional channels for

coincidence could reduce the potential for unnecessary transients or spurious

actuations. Special attention should be paid to first-of-a-kind features not incorporated in earlier designs, such as the main steam reheater drain high level

trip and other turbine protective trips.

* More focused attention should be given to BOP operations, e.g. the ability of

the operators to survive feedwater transients and load rejections, to reduce the

frequency of such transients.

* 29% of unplanned scrams during early operation originate in the Main

Feedwater (MFW) System.

16
* Specifically', for both new and mature plainit's equipment problems in the

MFW System originate most often with the MFW pumps, control valves and

their associated cohtrol systems.

eEquipment problems accounted for 52% of all unplanned MFW-initiated

scrams while human error accounted for 31% of such scrams.'These scrams

IIoccur most'ftei whil attenmpting to change plani power level (36%);,followed by sieady'stat&eoperation (23%) and t~eting (19%).

* The'most prevalent MFW problems involved- '

' -flow control during startup and at low power levels.'

-regulating valve control systenms.

-'pump control systems.

* The turbine system causes about 15% of all scrams for'new plants, and 10%

for mature plants.

* Specifically, the turbine, Main Steam and ste'a'm dump (or bypass) iystems ac-'

count for roughly 18% of the new plant scrams.

* Equipment failure acoiounts'for'56%' and human error accounts for 24% of

such Turbine/Main Steam-initiated scrams." Testing contributes in 46% of

such cases, followed by 26% during power changes. Turbine stop valve testing and MSIV surveillance testing are largely responsible for the testing contribution.

2.7.2 Operating Experience Feedback Reports Prbgrress in Scram Reduction

(NUREG-1275, Volume 5)14.15

As a followup to the previous Operating Experience Feedbkck Report described .. ~~~~~ - ou~e. !.f .bkG174 Mar .................

in 2.7.1 above, AEOD issued Volume 5-of NUREG-1275 in March 1989. As in the previous

report, new' plants are" defined 'as those plaints which had received their operating licenses

within the past 2 years. There were 34 different plants which were within two years of receiv17
ing an operating license for at least part of the time between January 1, 1984 and December 31,

1987, the study period., A total of 1885 unplanned reactor scrams occurred during the

four-year period 1984-1987. Of these, 1319 occurred at plants with more than two years of

operational experience and 566 were experienced by plants with two years or less of

operational experience. A "scram" is defined as any RPS actuation, either automatic or manual, that resulted in contr6l rod motion. RPS actuations without control rod motion, which incidentally occur in large numbers.while the reactor is shut down, are not included. If a given

plant was in its first two licensed years for any part of the period from 1984 through 1987,

scrams experienced during that period were pooled, averaged, and analyzed as "New Plant"

experience. For the same plant, scrams which occurred after completion of two years of licensed operation were analyzed as part of the "Mature Plant" experience. Thus, plants were

phased into the mature plant statistics over the period analyzed.

The results are provided on both a generic overall basis as well as by each NSSS

vendor, and by new plant versus mature plant. With respect to BOP systems, the following is

noted:

* In general, scrams at new plants arise from the same sources as scrams at mature plants, i.e. the Main Feedwater, Main Turbine and Reactor Protection

Systems. The new plants experienced scrams attributable to those systems at a

higher rate than mature plants.

* The power conversion systems, i.e. Feedwater, Main Turbine, Main Generator, Main Steam and Condensate Systems are the dominant causes of unplanned reactor scrams at new plants.

* At new plants, typical Feedwater System equipment malfunctions originated

in piece parts, such as printed circuit cards, switches and controllers, that in

turn led to malfunctions of feedwater pumps and of system valves, e.g.

18
feedwaterre'gulating valves, feedwater system isolation valves, and feedwater

bypass valves. - -

* For mature plants, the power conversion systems accounted for 58% of the unplanned scrams for the entire period of 1984-1987.

* Similarly, support systems accounted for 17% of the 1984-1987 unplanned

scrams at mature plants, with-electrical distribution, i.e. 120V AC, switchyard, large plant loads, DC power and control centers, accounting for 13% of

the 17% figure. (The remaining 26% are attributable to the NSSS.)

* Equipment failures related to feedwater pumps, which typically occur at

higher power levels and`primarily to the turbine-driven designs, and

feedwater regulating valves were caused by the same kinds of faults across the

industry (problems with valve operators, controller cards, control oil, and lube

oil).

* Specifically for feedwater valves, a significant number of valve operator failures caused by system- or valve-induced vibration have occurred, as well as

by oil, moisture, and/or rust or foreign particles in the instrument air system.

Contained leakage problems'resulting from damaged valve trims (plug and

cage or sets) and improperly adjusted valve operators have resulted in steam

generator overfill and subsequent reactor trips. Correctivemeasures have

been identified for all of these problems.

- For motor-driven feedwater pumps, premature bearing and seal failures have

occurred due to poor pump-to-motor alignment, thrusting or-pump instability

problems, excessive pump vibration, water contamination of the lube oil, intrusion of foreign material such as dirt or particles into the seal water, seal design weaknesses and poor installation. Shaft or impeller failures have occurred due to cyclic fatigue from normal to excessive vibration or other

19
- - l -~~~~~~~~~~~~~~~~~~~~~~~~~~

loadings. Casing erosion/corrosion problems due to long-term wear or corrosion have also been observed. Corrective measures have also been identified

for all of these problems.

* For turbine-driven feedwater pumps, lube and control oil contamination has

been caused by water getting into lube oil from steam leaking past turbine

steam and lube oil seals. Lube and control oil leaks have been caused by broken or cracked hoses or piping and gasket problems. Governor or control failures have occurred due to dirty control oil, oil leaks, mechanical failure or aging. Premature turbine, pump bearing, or rotating element failures have been

caused by improper alignment and/or rotating element balancing. Pump or

turbine trips due to low oil pressure have been caused by sudden decreases in

lube oil pressure during pump switchover. As before, corrective measures

have been identified.

* The main turbine system was the second leading contributor to unplanned

scrams, at an essentially constant rate during 1984-1987. Of these, turbine

surveillance testing caused the largest number of scrams. Equipment failure-initiated scrams involved switches, relays, circuit cards, fuses and voltage

regulators. Main turbine lubrication subsystem problems generally involved

failures, strainers, unloaders, motor operated valves, and piping. Turbine

blade failure did occur but was rare.

* One half. of all main turbine equipment-initiated scrams came from

electro-hydraulic control (EHC) subsystems. The electronic failures consisted mainly of circuit cards relays and switches. Hydraulic system failures

involved contaminated (dirty) oil, sticking valves, fluid leaks and pressure

losses.

20
The ~ ~ 'or e | In

* The main generator failed'pri narily by loss of excitation voltage, leading to

loss of generator load, turbine trip, and unplanned reactor scram. This is an example of a single point'failure that invariably led to a scram.

* For support system-initiated scrams, the electrical distribution system was the

dominant contributor. Other contributing systems included the circulating

water, service/instrument air, fire protection and HVAC systems. The 120V

AC instrument system was th&' dominant contributor to electric distribution

system-initiated scraffis.'Equipmentffailure was the dominant cause of electri-

-cal distribution-initiated scras. Failure often occurred off-site and involved

either transmission lines or substation equipment. They usually occurred during steady state operation at or near full power. Generally, no other contributing activity such as testing or maintenance was in progress. Maintenance errors such as operation of the' wrong breaker, or improper testing or trouble

shooting, were the dominant human error in the electrical distribution system.

2.7.3 Development of Maintenance Effectiveness Indicator -

AEOD has sponsored a' pro'gram under FIN L-1345 for the'development of a

Maintenance Effectiveness Indicator and a preliminary report was issued in October 1989.16

The study was conduictedby analyzing 3881 'Nuclear Plant Reliability Data Systems

(NPRDS)17 component failure data for preselected systems and signals any increase in the

failure rate that exceeds a predetermined value. The number and frequency of the flagged failure rate increases is then trended forill systems considered over the study' Period to obtain a

measure of the level of 'maintenance effectiveness at a plant. Based on a review of the NPRDS

narrative descriptions, the cause of 6eachfailure was assigned to one of five distinct categories.

The categories were analyzed t6iassess th'reriative contribution of ineffective maintenance to

equipment failures. The five categories are Ineffective Maintenance,'Randomr Design/

Installation/Construction, Normal Aging/Wearout/End of Life, and Unknown.

21
.w V

In the report, reference is made to AEOD/S804B18 in which it is noted that

NPRDS does not currently include certain BOP systems and components that have historically been significant contributors to plant outages, such as the turbine-generator and associated support systems, the condenser, the circulating water system, non-nuclear portions of the

service water and closed cooling water systems, the instrument air system, and the service air

system. In December 1988, May 1989, and June 1989, official steps were taken by the NPRDS

User's Group to include the main generator, main turbine, and condenser in the NPRDS reporting scope. Thus, for the purposes of the current study, it does not appear that the Maintenance Effectiveness Indicator can provide any significant input to studying the effects of aging. on BOP systems at this time. However, as the program develops further, with the

accumulation of data on BOP systems one would expect this program to provide valuable insights into the aging of BOP systems.

2.8 NRR Survey of Plant Design Strengths and Weaknesses

On September 20, 1988, an internal NRC memorandum19 ,

20 .

21 was issued from Mr.

Dennis M. Crutchfield, Acting Associate Director for Projects, NRR, to all Project Managers

(PM) concerning the strengths and weaknesses of plant design. Each PM was asked to identify

the strengths and weaknesses of plant hardware and design, separate from operational performance. As part of the guidance given to the PMs, the memorandum incorporates two tables,

Table 1 entitled "Reactor Safety Functions and Risk Significant Systems for PWRS" and similarly Table 2, "Reactor Safety Functions and Risk Significant Systems for BWRs."

Of the PWR systems or components identified as risk significant in Table 1, only the

Main Feedwater, Offsite AC Power, Fire Protection and Accident Monitoring Instrumentation

Systems can be considered to be BOP systems, and all are nonsafety-related, except for the

latter. Only the Main Feedwater and Offsite AC Power Systems fall within the minimum set of

systems of components that should be considered in the PM's review.

22
For BWRs, the Feedwater System,' the Steani By-Pass Capacity, Offsiie AC Power,

Fire Protection; and the'Accident Monitoring Instrumentation Systems are the only BOP systems listed as risk'significant in Table 2, and with the exception of the Accident Monitoring

Instrumentation, all arfenonsafety-ielated. '

-. . i - . . -., :- .A: ... * ..-... 

.

. .. ..

tr . ..

,.

C ' -

.

- I . . I .

I

.w 'I

23
3. IDENTIFYING IMPORTANT BOP SYSTEMS

In any commercial nuclear power plant, there are numerous BOP, systems besides the

commonly thought of main steam, feedwater, turbine-generator, condensate, etc. The key objective of this initial phase of the study is to identify and prioritize only, the most important

BOP systems, and consequently to focus the aging study only on those systems. The following

methods are potential means to identify important BOP systems.

3.1 Assessment of Important BOP Systems From Prior NRC-Sponsored Activities

In Section 2 of this report, a summary of the NRC-Sponsored BOP activities which

have occurred was presented. From nearly all of the references cited, a clear picture emerges

that the most significant contributors to unplanned reactor scrams caused by BOP systems are

the power conversion systems, i.e. the feedwater, main turbine, main generator, main steam

(usually the steam bypass to the main condenser) and the condensate systems. Other significant BOP systems are support systems such as the Electrical Distribution System, and less significantly the Circulating Water, Service/Instrument Air, Fire protection and HVAC Systems.

The Electrical Distribution System consists of the 120V AC Power, switchyard, large plant

loads, DC power and control centers. At a component level, the feedwater regulating valves

and turbine-driven feedwater pumps are significant, as well as the main turbine electro-hydraulic (EHC) control subsystem. Failures in the main generators are also important.

In reviewing the NRC-sponsored BOP activities described in Section 2, the preponderant safety criterion against which important BOP systems are identified is in reduction in

challenges to safety systems. Typically this has been represented by reduction in unplanned

reactor scrams. 4.5,13,14 15 According to 10CFR50.73, licensees are required to report several

other plant conditions besides those causing an unplanned reactor scram, among which are

Engineered Safety Feature (ESF) Actuation Signals, Technical Specification violations, deviations or required shutdowns, loss of system safety function, or excessive airborne radioactivity or liquid effluent release.

24
ESF Actuation Signals and Technical Specification violations or loss of system safety

function have been studied.2 9:13 Generily sipeakirig, the data show that, except for certain:

BOP systems'which may be nonsafety-related but subject to Technical Specification requirements such as Fire Protection or Radiation Mdnitoring, BOP systems'are not significant 'contributors to those challenges to"safety systems'

It should be noted that manyniiore incidents occur at nuclearplants than are required to

be reported as LERs. If one were to examine the individual plant records of these less serious

incidents, additional or different BOP systemis might be identified as' showing a high failure

rate'and Who-se continued poor operating perf6rmance might eventually cause additional challenges to safety systems.

3.2 Categorization Based Up6n Probabilistic Risk Assessment (PRA) Insights

As just mentioned, identification of important BOP systems within the NRC-sponsored BOP activities has relied primarily upon measuring against contributions to unplanned

reactor scrams. The focus there is on prevention'of accidents, a'clearly beneficial result.:

In the field of PRA', obviously a more globa approach is taken A Level-I PRA, in

which the overall core'danage freq.iuency is'calciilited, goes Well beyond the initial challenge

to safety systems represented b'ythe un lanned reactor scram.-It must include subsequent loss

of decay heat remnoval capability, either short or'long terim,'represented by failures of mitigating systemis. BOP systemssuch as feedwater, main steam, steam bypass to the main condenser,'and condensate normally figure prominently as mitigating systems, as well as means of

preventing a reactor scram: During normal plant operationr,'those systems are serving their intended purpose of removing fission product heat and converting it to electricity via the turbine-generator.'As already noted, those'system's a'e'prim-e contributors to unplanned reactor

scrarns.'However,'subsequent to a scrami,' those systems are the preferred means of decay heat

removal by -means of the' steamr bypass tb'theiiiain'condenser. Thus loss of those systems

25
. af

would normally force a reactor scram and a very near term reliance on the standby, or safety-related, decay heat removal systems, as well as high or low pressure safety injection, etc.

In a Level II PRA, since the probability of containment failure is the calculated end

result, containment heat removal and pressure reducing systems are incorporated, into the

analysis. These are all safety-related systems and thus do not involve the BOP systems. In the

final type of PRA, a Level III, the probability and extent of early and latent health effects to the

surrounding population, assuming containment failure, are calculated. Thus no additional insights regarding BOP systems would normally ensue.

In most PRAs, the number of systems modeled in detail by means of fault trees is generally about twelve to fifteen, and typically these are safety-related non-BOP systems. In any

given commercial nuclear power plant, the total number of mechanical and electrical systems

is usually on the order of three to four times that number.

Since the BOP systems already identified as important with respect to causing unplanned reactor scrams are only a small fraction of the total number of plant systems, an expert

opinion survey was conducted using BNL staff familiar with both PRA and power plant design in order to determine which systems they would identify as important with respect to the

more global perspective of PRA, i.e. not simply limited to unplanned reactor scrams. The survey was divided between BWRs and PWRs, with the system titles described in most cases according to the Sequence Coding and Search System (SCSS) code listings for systems.22 From

the master list of systems, the individual expert placed the various systems into the appropriate columns, 1 through 6. The objective was to sort all of the plant systems into a logical

framework.

Expert opinion was particularly required in assigning systems to Columns 3 and 4. In

these columns, systems were located based on whether or not the expert judged them to be potentially important, even though they, are not normally included in a typical PRA.

26
... ..  1. , I , 1 7 , . .'.,

For both BWRs and PWRs (see Tables 3-1 and 3-2), the nonsafety-related BOP

(non-NSSS) systems which were identified as being typically included in PRAs, i.e. Column

2, coincided very closely with the BOP systems identified as important in causing unplanned

reactor scrams. That is, the typically'included systems were the electrical systems such as

Normal AC, High Voltage AC, Medium Yoltage AC and the power conversion systems such

as Feedwater and Condensate, Turbine (or Steam) Bypass, Main Condenser, etc. The Circulating Water and Station Service Water Systems were also identified as appearing in some PRAs.

There were a substantial number of systems not normally included in PRAs which

were considered to be potentially important, column 3, such as miscellaneous electrical systems, the turbine building closed cooling water (BWRs) or service water-(PWRs), the cooling

towers, miscellaneous HVAC systems, monitoring, detection andcontrol systems and'other

systems such as security and communications.-Even without the nonsafety-related NSSS, Column 6, included as BOP systems, the total numberof BOP systems is extensive, Coiumns 2, 3

and 4, yet the vast majority of these systems do not appear as significant contributors to unplanned reactor scrams.

It should not be too surprising that the BOP systems typically included in PRAs'coincide quite closely with the previously mentioned important BOP systems according to unplanned reactor scrams. The PRAs by definition must accurately reflect and emphasize the

frequency of initiating events, and so their data sources are essentially the same as the LER

data base, although plant-specific failures below the severity level of an LER issuance may be

used as well.

In a separate, unrelated effort, BNL staff members developed a study of initiating

events with special emphasis on reactor trips.2 This effort w*as part of a program to develop

risk-based performance indicators (RBFI). The study notes that in PRAs, reactor trips are categorized into four initiating type events:

27
TABLE 3-1 'ITYPICAL BOILING WATER REACTOR (1WR) SYSTENIS

- IMP'OlRlANCE CAT(EGORIZATION

- NON-NSSS NSSS

NON*SAFETY-RELATED

Not In PIL Not in PKA

1~~~~~~~ypically includcd (rotcntially # (Not

(T)Safcty-Related Ty in PRA Important) I) (Important) ()Safety-Related (JNon-.Safcty-Relaed

00

Elecrical Systems

- Norm. Aux. AC

Powcr (NAACP)

* Standby AC Powcr

(SACP)

* Essen. AC Power

(I3ACP)

* DC Power (DCP)

* SwltchgeariLoad

Centers/MCCs

* Power/Control

Cables

Cooline Water Systems

* Standby/Emerg.

Service Water

- RIIR Service

Watcr

Safety W ReeL Chilled

Water (e.g..

RIBSVS, CRAC)

* Fucl Pool Cooling

& Cleanup

lcEilinz. Vent.. & Air

Cond. (I IVAC) Systems

- -x. Bldg. Standby

Vent. (KIISVS)

* Control Room Air

Cond. (CRAC)

* Sec. Cont.

(Standby Gas

Treatment)

- DrywelVroruS

- Kx. Aux. Bldg.

* iDG 3Bldg.

Electrical Systems

- Normal AC . Turbogenerator -

* I ligh Voltage ACL

- Med. Voltage ACt

Coolin2 Water Systems

* Component Cooling

Water (CCW)" 2

* Circulating Water 4

* Station Service

Water4

Fire Protection4

Power Conversion

System

- Main Steam

- Cond. & 1IW

* Turbine Bypass

Elecirkal Systems

* Station Transformers

(NST & RST)

- Isolated Phase Bus

- Non Scq. Busses

* Substations

* AC Instrumnent

power

* Main Power Transfer

* Emergency Lighting

- Elect. Heat Tracing

- lighting & Taxed

Motive Power

- Grounding &

Cathodic Protection

(otling Water Svslems

* Circuling Water

* Turb. Bldg. Closed

Loop Cooling Water

- Reactor Enc. Cooling

Water

- Screenwell Canals

* Cooling Tower

System

Communication System

Misc. I IVAC Systems

- Primary Containment

Atmos. Control

* Aux. Equipment

- Rx. Bldg. Normal

Ventilation

* Main Chilled Water

- Drywcll Chilied

Water

Misc. Electrical

- Grounding System

- Unit rrot. h

Metering

- lighting (yardibidgs.)

* Cathodic Protection

Vents & Drains

- Plant Buildings

i Equipment

* Roof

Domestic Water/Saint.

Sewaae

Radw aste

* Gaseous

* Liquid

* Solid

* Decont. System

* Steam Seal &

Radwaste Steam

I IVAC

- Fuel building

. Waste Manage. bldg.

* Turbine bldg.

lest & Serviciny Aids

- Storage'

- Turning Gear - Vac. Priming & Air

Removal

* Sample System

* Primary Cont. Int.

Leak Rate Test

* Chlorination System

* Cooling Tower

iBlowxown

Reactor Vessel and

Internals

Fuel and Control Rod

Control Rod Drive

(CRD)&

Reactor Vessel

Instrumentation

- Rx Water Level

Control

* FW Control

Containment Systems

* Primary Containment

(Suppression Pool)

* Sccondary

Containment

Nuclear Steam Supply

Ssltms (NSSSSI -MSlVsL

Neutron Monitorinr 3

- Intermediate Range

Monitoring

- Local Power Range

Monitoring

- Average Power Range

Monitoring

Reactivity Control

- Reactor Protection

(RPS)

* Standby Liquid

Control (SLC)

- Alt. Rod Injection (for

ATWS)

Recirculationt

Main Steam

Condensate and

FeedwaterI.

Reactor Core Isolation

Coolint (RCIC)

Reactor Water Cleanup

(RWCU}

Process Instrumentation

and Control

* Electro Hydraulic

Control (EIIC)

- Feedwater Control

Standby Gas Treatment

Neutron Monitoring

- Source Range

Monitoring

- Traversing Incore

Probe

Computer System

(Process)

* Performance

Monitoring

* Display

Reactivity Control

* Rod Control & Info.

* Recire. Ilow Control

- Reactor Manual

Control (RMC)

* Rod Sequence

Control System

(KSCS)

I Rod Worth

Minimizer (RWM)

K Rod Illock Monitor

(ItUM)

0
NON-NSSS NSSS

5 ~ ~ ~ ~~~~~NONNA-TYREL11: I ,, . W ~~ ~- : . :11 ~~~Not In PRA -- n Not in P'RA, 1

i-_pically included PoSAtentially ':" (Not I

4)Sateiy.Related I In PRAs I) mportant) important) I) ( Sarety-Related Non.Satety.Rclted

Structures & Iluildings

- Primary Contain.

- Secondary Contain.

Emergency Generator

(Dilesel)

* Lube Oil

- Fuel

- Starting

- Cooling

* I&C

Comrnressed Air1Gns

- Instrument Air

- Prim. Contain. Inst.

Gas

Containment

* Isolalion

- Leakage Control

- Spray

- Press. Supp.

Makeup

* Comb. Oas Control

* Vacuum Reliet

* Cont. Hteat

Removal (CFIR)

Main Steam

- MSIV Leakage

Control

t~0

Monitorinz Systems

- Hearing Temp. &

Vibrtildro' ''

* Radition' Area

- Process Radiation

- Envir'nment - Plant

- Leak

Service h landlling

* Fuel

- RVes lel - Retueling

* In Vessel & Under

Vessel ' ' ' '

- Startup Equipment

Detection & Control

Svstem:

- Leak Detection

* Fire Detection

- Post'Aeident

Sampling

- FW Chemistry

Control ' '

Structures. Panels &

Stornges

* Main CR' Panels

Loc6l Panels &-

Racks '

* Aumt Control Panels

* Inst'' Pipini

* Scieenwell/Canals

* Cooling Tower'

* Fuel Storage

- Conduit & Cable

Trays ;'

. Plant Stack

Matl. & rEuipment

I landling

- Seal Water

Aux. Steam

Emergency Core

Cooling Systems (ECCS)

* Iigh Pressure Core

Spray (IIPCS)

* Low Pressure Core

Spray (LPCS)

* Aux. Depress. System

(ADS)

- Res. Ileat Removal

(RIIR - LPCI)

* Iligh Press. Core Inj.

(IiPCI)

- Cond. Storage Tank

(CST)

- Isolation Condenser

Ncudear Boiler

Overpressure

Prolection (SRVs)

ESF Actuation

i

Turbine Generator

- Main Turbine

* Main Generator

* Inst. & Control
l'lON*NSSS NSSS

NONWSAFETY-RELKIED

Not in PRA Not in PRA

Typically included - (Potentially (Not

2) Safety-Related I In lIRAs ) Important) ()l Important) )Saftly-Rcalaed )Non-Safely-Rclatcd

0

(;enerator Suoorm

System -

- Lube Oil (Gen. .

Turb.)

- Stator Cooling

* Hydrogen Seal

* CO;I Purge'

* Excitation

* I&C

Power Conversion

systems (PCS)

* Moist. Sep. &

Reheat. Drains

- Turbine Steam

Scaling

- Steam Extraction

Radwaste Offeas

* Offgas recombiner

* Noncondensible Gas

Extraction

* Radwaste I leading

Compressed Air1'

2

Securitl

Remote Shuldown

Panel 4

Makeup & Purification

* Demin. & Makeup

Water

Cond. Demineralizes

- Cond. Transfer

Notes:

I Partially safety-related.

2 Safety-related at somc plints.

3 Reactor protection - saftiy related; Monitoring portion - non-sactly related.

4 Included in some PIUs.

.4 .
.I

TABLE 3.2 TYPICAL PRESSURIZED WATER REACTOR (PMR) SYSTEMS

IMPORTANCE CArEGORIZATION

_NON-NSSS I NSSS

NON oSAFETY REAIotPED _

Not in PRA Not In PIU

Typically Included (Potentially (Not 1

) Safety-Related I in PRAs JI.. 3 mportant) kV Important) j Satety-Related )Non-Safety-Related

RectviI onro

Electrical Systems

- Norm. Aux AC

Power (NAACP)

* Standby AC Power

(SACP) _

- Essen. AC Power

(EACP)'

- DC Power (DCP) Switchgear/Load

Centers/MCCs

Power/Control

Cables/AC Vital

Powver."--

Coolint Water Systems

-. Standby/Enmerg.

'Service Water

- Component Cooling

Water :

Fuel Pool Cooling

- -h . eanup

* Safety Rel. Chilled

;Water (e.g.,.

CRLNC)

Ileatini. Veni..hAr

Cond. (I IVAG) Systems

1i. lidg. IIVAC

Rx. Aux. Dldg.

lIVAC

* Control Bldg.

I IVAC

* Emerg. Gen. Bldg.

IIVAC .-

Structures & luildinsn

- Primary Contain.

Secondary Contain.

Electrical Systems

- Normal AC . Turbogenerator

I ligh Voltage AC'

Med. Voltage AC'

tCooling Water Snlems

. Circulating Water

. Station Service

Water4

Power Conversion

System

. Main Steam . Condensate

F- eedwater - Feedwater Control

- Turbine Byps

.* Main Condenser

- Non-nuclear

Instrtimentation

tire P'rotection

Electrical Systems-

* Station Transformers

(NST & RST)

- Isolated Phase Bus

- Non Seq. Busses - Substations

AC Instrument

Power

Main Power Ttansfer

' -!lectric I eat Tracing

Emerg. Ughting &

Taxed Motive Power

'Grounding &'

Cathodic Protection - Conduit & Cable

"Trays

CTolinc Water Systems

- Turb. Bldg. Service

Water - Cooling Tower - Chilled Water

Power Conversion

System ,

* Turbogenerator

* Tarbogenerator

Steam Sealing

- Turbogenerator

(I&C):

Piower Conversion

System _ ^

- Turbine

. Steam Sealing

- Lube Oil

* I&C'

Misc. Electrical

- Grounding System

- Unit Prot. &

Metering

- Ughting (Yard/Bldgs.); - Cathodic Protection

11VAC

- Fuel Building

. Waste Management

Bldg.

- Turbine Building

Auxiliary Steam

Vents & Drains

- Plant Buildings

. Equipment

' Root

Potable & Sanitary

Water

Radwaste

* Gaseous

i Uquid

- Solid

Miscellaneous

- Plant Stack

- Material & Equip.

I landling

* Turning Gear

- Primary Containment

Integrated Leak

Rate Test

- Chlorination System

- Cooling Tower

Blowdown

Reactor

* Reactor Vessel

- Reactor Core

* Fuel & Control Rods

- Control Rod Drive

* Primary Coolant

- Pressurizer

* Safety & Relief Valves

Instrumentation &

Control

* Neutron Monitoring3

-Low Power Range

-Intermediate Power

Range

-Average Power

Range

* Reactivity Control

-Reactor Protection

-ATWS Mods. (Alt.

Rx Trip)

* Solid State Protection/

CbntrolEngineered Safetj

Features Actuation

Emergency Core

CoolinfIfloration

-Itigh Pressure Injection

-Intermediate Pressure

Injection

-Upper'Ilead Injection

-Low Pressure

Injection1esIdual

lleat Removal

't-Borated/Refueling

Waier Storage 'ranks.,

Reactivitr Condrol

* Control Rod Drive

* Control Rod Drive

Cooling Water

- Rod Sequencing

- Chemical A Volume

Control '

* Boron Recovery

- Neutron Monitoring3

Process Cornmputer

- Performance

Monitoring

* Display
NON.NSSS NSSS

NON.SAF-RTY-RIjLATIIED

Not in PRA Not in PRA

:Typically included (PotcriLally Q (Not

@ Sarely-Related 2 inn IRAs k Important) (:Important) a Safcty-Relaicd V) Non-Safety-Relatcd

Emergency Generntor

ft)icsell

- Lube Oil

- Fuel - Starting

- Cooling

. I&C '.'

Compressed Air/Gas

SystCms -

* Instrument Air1' 2

- Prim. Contain. Inst.

Gas

Conlainment

* Isolation - Leakage Control

- Comb. Gas Control

* Ice Condenser

* Vacuum Rellef

- lleat Removal

Secondary Cooling

* Main Steam

* Main Steam

Pressure Rclict

Fecedwater

- Condensatel.

* Condensate Storage

* Auxiliary Fecdwater

* Steam Generators

- Feedwater Chemistry

Control '

- Condensate

Dcmincralizer

* Dcmineralized Waler

* Chemical Additive

Injcction

* Non-condensible

Gasses Extraction

* Steam Generator

Blowdown

- Generator

- Stator Cooling

* Lube Oil

- Hydrogen Seal

. CO; Purge . Excitation

* I&C - Moisture Separators

& Reheater Drains

* Extraction Steam

Miscellaneous IIVAC

Systems -

* Turbine Bldg. IsVAC

* Pumping Sta: IIVAC

Monitoring Systems

- Environ. Monitoring

* Plant Monitoring

Leak Monitoring

- Sampling

* Fire Detection

C(omnpreimed Air/t is

Systems

* Instrument Air' 2

* Control/Service Air

- Core Flooding

Accumulators

. Chemical & Volume

Conlrol Systeml

Emergency Boration

Containment

-Spray

* Isolation

Radiation Mtonitoring

4
I.

NON-NSSS NSSS

NON-SAFETY-Rl IHAllD

Not in lI A Not in 1PRA

lypicnily included (Iotentilnly m (Not

Safety-Rlalted (5 In lRAs ( Important) %.L; Important) 'b' Non-Satety.Reiated

Miscellaneous Fluids

* Insulating Oil

* Lube Oil

* IFucl Storage Service & Iandling

* Fuel-

* RIx Vessel/elead

Flange

* Refueling

- InVessel &

UnderVessel

* Startup Equipment

Structures. Panels

Main Control Room

Panels

- Local Pancis &

Racks

- Aux. Control Panels

- instrument

PlpingtTubing

Conduits & Cable

Trays

Securitl

Communications

Note: -

Hti, - .

t Partially sarety-related.

2 ~ Safety-retated at some plants.

3 Reactor protection - sarety related; Monitoring portion - non-safety related.

4 Included In some PRAs.

.11
* Reactor trips with no expected impact on the operability of feedwater and/or power

conversion systems (e.g. manual scram, loss of load, etc.).

* Reactor trips with expected impact on the operability of feedwater and power conversion systems (e.g. loss of condenser vacuum, etc.).

* Reactor trip with expected impact on feedwater and power convesion systems as

well as the mitigating systems (e.g. loss of offsite power, loss of instrument air, etc.).

* Reactor trips generated by various sizes of loss of coolant events (small, medium, or

large LOCAs caused by cascade-failure of RCP seals, inadvertent opening of

PORVs, etc.).

In this study, several abnormal events are cited as important indicators which may lead

to iniating events such as reactor trip. Other outcomes can be return to normal power operation

or normal manual shutdown. It is suggested that the leading RBFIs can be identified by tracking the frequency and duration of these abnormal conditions. The abnormal events cited for

either PWRs or BWRs which involve the nonsafety-related BOP systems as defined in this report are showvn in Table 3-3.

Table 3-3 Abnormal Events Involving'BOP Systems in PWRs and BWRs

PWRs

Failure of Steam Dump to Open

Loss of Instrument Air

Failure of Turbine to Runback Automatically and Manually

Failure of Impulse Pressure Transmitter (Low) (Main Steam Upstream of Turbine Stop Valves

Loss of Condenser Circulating Pump

Loss of One Main Feedwater Pump at High Power

Spontaneous Opening of the Main Generator Output Breakers

Partial Loss of Offsite Power

BWRs

Master Feedwater Controller Failure

Condensate or Condensate Booster Pump Trip

Loss of Feedwater Heater Extraction Steam

Stator Cooling Water Pump Trip (Main Generator)

Steam Jet Air Ejector Malfunction (Main Condenser)

Electrohydraulic Control (EHC) Faults (Main Turbine)

34
As is evident in Table' 3-3, excepi for the Loss'ofInstarueint Air and Partial Loss of

Offsite Power,'all of the abnormal events cited involve the power conversion systems such as

main'steafmifeedwater, concensate, main condenser and turbogenerator. The partial loss of

offsite power coincides with the electrical distribution system mentioned in Paragraph 3.1 as

important with respect to unplanned reactor scrams. This list coincides quite closely with the

systemns identified previously and thus provides further substantiation of the choice of important BOP sysems. - ' 

'' ! .: , I

I . - ; ' I. . '.

. , I i . :

*I' '. ' '. l e

I.:.

... ,, 1 , , -

. .. I ::

35
4. INTERIM RECOMMENDATIONS AND FUTURE EFFORTS

From a comparison of the means to identify important systems, and the respective results,

discussed in Section 3, the data show that the power conversion systems are the prime source

of unplanned reactor scrams; followed by the electrical distribution systems and the Circulating Water, Service/Instrument Air, Fire Protection and HVAC systems.

The frequency of unplanned reactor scrams as a performance indicator is the most frequently cited measure and the results appear to coincide with the results obtained with the

more global approach inherent in the development of a PRA. Failures of the power conversion

and electrical distribution systems can appear as both transient initiators and failures of mitigating systems in a PRA, hence they are of particular importance.

In the next phase of this program, several of the oldest plants will be considered as a

group and their LERs involving unplanned reactor scrams since the beginning of commercial

operation until the present time will be reviewed to determine if there is an increasing frequency of unplanned scrams caused by or involving, the previously cited BOP systems. This

group of oldest plants will include the ones proposed by industry as the pilot plants for the

plant life extension study, Monticello and Yankee Rowe.

A group of plants of intermediate age, as well as a group of the newest plants, will also be

examined in the same manner. All three groups will be analyzed and compared to each other to

identify any trends or patterns.

The ultimate goal is to examine the individual system component failure mechanisms to

determine whether aging of BOP systems is a significant factor affecting nuclear plant safety.

This will be done in the same manner as has been done for safety-related systems within the

NPAR Programs, such as the Component Cooling Water and Residual Heat Removal

Systems.23 24 In conducting these phases of the study, primary emphasis will be placed on using the NRC-sponsored Sequence Coding Search System (SCSS)25 for LERs and also the Nuclear Power Experience (NPE)26 data base. NPRDS will not yield much useful information be36
cause BOP systems (limited to turbine, generator and condenser) were only recently included

in the data base. . .*- -

e...< l_

*4. X *. . ,

I . I I -.

.I . . . i . I . .

. . . . I .. ; . I . t.

! . , . , i : , ;A  ,

I i ,-

. :7 ' I I

. I I I

I I .1.

I I 4 .. ,

I -  , , . . . I .  I .,

. I : , - , --- .... ..

i '- :,., 

I

:',I- -, . .1 I * I

I ~ I., I . , .; ,: .1 , .

37
5. REFERENCES

1. Stone & Webster Engineering Corp., "Summary of IE Bulletin 79-01 Review -

Balance of Plant (Inside & Outside Containment) and NSSS (Inside Containment) for

Class IE Equipment - North Anna Power Station Units 1 & 2 - Virginia Electric and

Power Company", February 25, 1980, (ACN No. 800260731).

2. T.R. Wolf, M.R. Harper, K. Higgins, "Engineered Safety Feature Actuations at

Commercial United States Nuclear Power Reictors - January 1 through June 30,

1984", AEOD/P503, August 1985.

3. M.R. Harper, T.R. Wolf, K. Higgins, "Engineered Safety Feature Actuations at

Commercial United States Nuclear Power Reactors - July 1 through December 31,

1984", AEOD/P603, August 1986.

4. L. Bell, P. O'Reilly, M. Harper, "Trends and Patterns Report of Unplanned Reactor

Trips at U.S. Light Water Reactors in 1984", AEOD/P504, August 1985.

5. L. Bell, "Trends and Patterns Report of Unplanned Reactor Trips at U.S. Light Water

Reactors in 1985", AEOD/P602, August 1986.

6. R. Lay et al., "Regulatory Considerations for the Balance-of-Plant", NUREG/

CR-4783, MTR-86 W00213, MITRE Corporation, Final Report, October 1986.

7. NRC Memorandum from William J. Dircks, Executive Director for Operations,

"Davis-Besse Event - NRC Lessons Learned", November 26, 1985.

8. NRC Policy Issue (Information) for the Commissioners, from Victor Stello, Jr.,

Executive Director for Operations, "Balance of Plant", SECY-86-349, November 21,

1986.

9. NRC Inspection and Enforcement Manual, Temporary Instruction 2515/83, "Balance

of Plant Trial Inspection Program (Feedwater System)", Issue Date February 26, 1987.

10. NRC Inspection Manual, Inspection Procedure 71500, "Balance of Plant Inspection",

Issue Date September 30, 1988.

38