NUREG/CR-5706

ORNL-6671

Potential Safety-Related

Pump Loss: An Assessment of

Industry Data

NRC Bulletin 88-04

Prepared by

:,Prepared by

: D. A. Casada

Oak Ridge National Laboratory

Prepared for

U.S. Nuclear Regulatory Commission
';

AVAILABILITY NOTICE

Availability of Reference Materials Cited in NRC Publications

Most documents cited In NRC publicatIons WMl be available from one of the following sources:

1. The NRC Public Document Room. 2120 L Street. NW. Lower Level, Washington. DC 20555

2., The Superintendent of Documents. U.S. Government Printing Office. P.O. Box 37082. Washington.

DC 20013-7082

3.; The National Technical Information Service, Springfield. VA 22161

Although the listing that follows represents the majority of documents cited In NRC publications. It Is not

Intended to be exhaustive.

Referenced documents available for Inspection and copying for a fee from the NRC Public Document Room

Include NRC correspondence and Intemal NRC memoranda; NRC Office of Inspectlon and Enforcement

bulletins. circulars. information notices, Inspection and Investigation notices; Ucensee Event Reports; vendor reports and correspondence: CommIsslon papers; and applicant and licensee documents and correspondence.-

The following documents in the NUREG series are available for purchase from the GPO Sales Program:

formal NRC staff and contractor reports. NRC-sponsored conference proceedings. and NRC booklets and

brochures. Also available are Regulatory Guides. NRC regulations In the Code of Federal Regulations, and

Nuclear Regulatory Commission Issuances.

Documents available from the National Technical Information Service include NUREG series reports and

technical reports prepared by other federal agencies and reports prepared by the Atomic Energy Commission. forerunner agency to the Nuclear Regulatory Coinmisslon. -

Documents available from public and special technical libraries Include all open literature Items, such as

books. joumal and periodical articles, and transactions. Federal Register notices, federal and state legislathn, and congressional reports can usualy be obtained from these libraries. -

Documents such as theses. dissertations, foreign reports and translations, and non-NRC conference proceedings are available for purchase from the organization sponsoring the publication cited.

Single copies of NRC draft reports are available free, to the extent of supply, upon written request to the

Office of Information Resources Management, Distribution Section. U.S. Nuclear Regulatory Commission.

Washington. DC 20555.

Copies of Industry codes and standards used in a substantive manner In the NRC regulatory process are

maintained at the NRC Ubrary. 7920 Norfolk Avenue. Bethesda. Maryland. and ire available there for reference use by the public. Codes and standards are usually copyrighted and may be purchased from the

originating organization or. If they are American National Standards, from the American National Standards

Institute. 1430 Broadway. New York, NY 10018.

I I

I~ I

DISCLAIMER NOTICE

This report was prepared as an account of work sponsored by an agency of the United States Goverment.

NeithertheUnited States Government nor any agencythereof, orany of their employees, makes anywarranty,

expresed or implied, or assumes any legal liability of responsibility for any third party's use, or the results of

such use, of any information, apparatus, product or process disclosed in this report, or represents that Its use

by such third party would not Infringe privately owned rights.
NUREG/CR-5706

ORNL-6671

RV

Potential Safety-Related

Pump Loss: An Assessment of

Industry Data

NRC Bulletin 88-04

Manuscript Completed: March 1991

Date Published: June 1991

Prepared by

D. A. Casada

Oak Ridge National Laboratory

Operated by Martin Marietta Energy Systems, Inc.

Oak Ridge National Laboratory

Oak Ridge, TN 37831

Prepared for

Division of Engineering

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

NRC FIN B0828

Under Contract No. DOE-ACO5-840R21400
iii

ABSTRACT

Nuclear utility plants are required to periodically test safety-related pumps to

demonstrate proper functioning of the pump. Historically, a substantial number of these

pumps have been routinely tested at the flow rate available through the pump's minimum

flow recirculation flow path, which in many cases was sized to avoid overheating only. It

has become more widely recognized that operation of a pump under low-flow conditions

can result in hydraulically unstable conditions that can damage the pump, even though the

rate of flow is adequate for heat removal.

Nuclear Regulatory Commission (NRC) Bulletin 88-04 required utilities to examine

(1) the potential for dead-heading of pumps due to parallel pump competition and (2) the

adequacy of the minimum flow rate provided for each safety-related pump. Utilities have

reviewed the currently recommended minimum flow rates with pump vendors and have

examined existing system design provisions, operating controls, and historical maintenance

experience.

Under the auspices of the NRC's Nuclear Plant Aging Research Program, Oak Ridge

National Laboratory has reviewed utility responses to Bulletin 88-04. An assessment of

the industry response and resultant conclusions and recommendations are presented.
v

CONTENTS

Page

LIST OF FIGURES ................ viu

LIST OF TABLES ................ ix

LIST OF ACRONYMS ....... .. xi

ACKNOWLEDGMENTS ................ xin

1. BACKGROUND AND SCOPE ...................................................... 1

2. DISCUSSIONS WITH PUMP MANUFACTURERS .......... .................. 5

3. ASSESSMENT OF PLANT RESPONSES ......................... 9

3.1 GENERAL DISCUSSION .......................................... ........... 9

3.2 PROCEDURAL AND DESIGN CHANGES ........... .................... 10

3.3 OTHER RESPONSE AC'I'IONS ..................... ......................... 14

3.4 RESPONSE LEVEL OF DETAIL .................... ......................... 14

4. DISTRIBUTION OF PUMPS BY TYPE .................. ......................... 17

5. ON-SITE REVIEWS OF PLANT DATA ........................................... 19

5.1 PLANTA ............................................................ 19

5.1.1 Strengths ............................................................ 19

5.1.2 Weaknesses ............................................................ 19

5.2 PLANTB ............................ ................................ 20

5.2.1 Strengths ............................................................ 20

5.2.2 Weaknesses ............................................................ 20

5.3 PLANT C ............................................................ 21

5.3.1 Strengths ............................................................ 21

5.3.2 Weaknesses ............................................................ 21

6. CONCLUSIONS .............................. .............................. 23

7. RECOMMENDATIONS ............................................................ 25

REFERENCES ............................................................ 27

APPENDIX A: NRC BULLETIN 88-04 ................................................ 29

APPENDIX B: PLANT C RESPONSE TO NRC BULLETIN 88-04 .... ........... 35
vi.

LIST OF FIGURES

Figure Page

1 Number of PWR units in which design and procedure

changes were made .................................. 1 1

2 Number of BWR units in which design and procedure

changes were made .11

3 Number of design changes by type of change 12

4 Number of procedure changes by type of procedure .13

5 Distribution of procedural or design change actions taken

in response to NRC Bulletin 88-04 .13

6 Distribution of BW;R plant ratings .16

7 Distribution of PWR plant ratings. 16
ix

LIST OF TABLES

Table Page

1 Systems included in assessment .................... 9

2 Distribution of procedural and design changes

by system. 10

3 Distribution of pump suppliers (pump suppliers

for specified system) .17
xi

LIST OF ACRONYMS

AFW auxiliary feedwater

ASME American Society of Mechanical Engineers

BEP best efficiency point

BWR boiling water reactor

CCP centrifugal charging pump

CCW component cooling water

Cs containment spray

ESW emergency service water

HPCI high-pressure coolant injection

HPCS high-pressure core spray

HPSI high-pressure safety injection

IST in-service testing

LOCA loss-of-coolant accident

LPCI low-pressure coolant injection

LPCS low-pressure core spray

LPSI low-pressure safety injection

NPRDS Nuclear Plant Reliability Data System

NPSH net positive suction head

NRC U. S. Nuclear Regulatory Commission

OEM original equipment manufacturer

ORNL Oak Ridge National Laboratory

PWR pressurized water reactor

RCIC reactor core isolation cooling

RHR residual heat removal

RWST refueling water storage tank
xiii

ACKNOWLEDGMENTS

A number of people have provided needed support for the completion of this study.

The contributions of Dave Eissenberg, Bill Greenstreet, Bill Farmer, Renee Li, and

Ted Sullivan, Jr., in providing guidance and organizational direction are appreciated.

Bob Clark undertook the difficult task of reviewing and providing an independent

assessment of all of the industry responses to the Bulletin. The insights he provided, as

well as those offered by Jimi Yerokun, were important "calibration" points.

The hospitality and cooperation of pump manufacturer experts, including Jeff

Bartholomew, Lloyd Hanson, and S. (Gopal) Gopalakrishnan of Byron-Jackson; Dennis

Bowman, Martin Prescott, Bill Marscher, and Bruno Schiavello of Dresser (Pacific &

Worthington); Philip Nagengast, Donald Sloteman, and Tim Wotring of Ingersoll-Rand;

and David Eddy, Earl Gordon, Stan Pace, and Donald Spencer of Sulzer-Bingham are

greatly appreciated.

Finally, the technical expertise and common sense approach to the study of pumps

provided by Mike Adams was instrumental to the project.
1. BACKGROUND AND SCOPE

Historically, minimum flow capacity for centrifugal pumps was based on ensuring that

the temperature rise through the pump was not excessive. As a general rule of thumb, the

minimum flow rate was specified so that the temperature rise through the pump would be

less than 15TF. It should be noted that this rule of thumb has not been universally applied

and that temperature rises greater than 50F have been used for some pump applications.

It has been recognized for many years that in higher energy density pumps at low-flow

operation, destructive hydraulic forces, not temperature rise, limit safe minimum flow.

Degradation can occur as the result of unsteady flow conditions within the pump, which

result in substantial radial and axial forces (static as well as dynamic) on both the stationary

and rotating parts. Damage can be manifested in a number of ways, including impeller or

diffuser breakage, thrust bearing and/or balance device failure due to excessive loading,

cavitation damage on suction stage impellers, increased seal leakage or failure, seal

injection -piping failure, shaft or coupling breakage, and rotating element seizure.I In

addition to the internal forces generated by unsteady flow within the pump itself, interaction

between the pump and the system at low-flow conditions can'result in substantial surging

and vibration that can affect not only the pump, but other system components and supports

as well.

As the effects of low-flow operation have become better understood by pump

technologists, design modifications that can reduce unsteady flow conditions have been

developed. Modifications to pump geometries have been demonstrated to allow operation

at lower flow rates with substantially'reduced impact.* Some pump original'equipment

manufacturers (OEMs) and non-OEM repair shops now offer design options or retrofits

that allow pumps to be operated acceptably at reduced minimum flow. However, a large

number of pumps remain in service which were not designed specifically to allow operation

under low-flow conditions and for which no modifications have been made.

- In May of 1988, the U.S. Nuclear Regulatory Commission (NRC) issued Bulletin

88-04, "Potential Safety-Related Pump Loss." The Bulletin addressed two general

concerns:

* The potential for dead-heading one of two pumps when operated in parallel.

* The adequacy of pump minimum flow protection provided by the installed minimum

flow (miniflow) lines.

With regard to the first concern, the Bulletin:specifically discussed the potential

problem of parallel pump operation during miniflow operadion, noting that the stronger

(e.g., higher head at same flow) of two pumps can dead-head the weaker pump. It was

also noted that the strong/weak pump situation is not a problem at moderate to high flow

conditions because of the shape of pump head-capacity curves in those regions. Relative to

the second item, the Bulletin noted that pump manufacturers now advise that desired

E. Makay and J. A. Barrett, "Changes in Hydraulic Component Geometries Greatly Increased Power

Plant Availability and Reduced Maintenance Costs: Case Histories," presented at the First International

Pump Symposium, Texas A&M University, May 1984.

l
2

minimum flow capacity is greater than was originally specified for some pumps. The

Bulletin required that all plants conduct a review of the safety-related pumps, including the

following specific requirements:

"All addressees are requested to do the following:

1. Promptly determine whether or not its facility has any safety-related system with a

pump and piping system configuration that does not preclude pump-to-pump

interaction during miniflow operation and could therefore result in dead-heading of one

or more of the pumps.

2. If the situation described in Item 1 exists, evaluate the system for flow division taking

into consideration (a) the actual line and component resistances for the as-built

configuration of the installed system; (b) the head versus flow characteristics of the

installed pumps, including actual test data for "strong" and "weak" pump flows; (c) the

effect of test instrument error and reading error, and (d) the worst case allowances for

deviation of pump test parameters as allowed by the American Society of Mechanical

Engineers Boiler and Pressure Vessel Code (ASME Code) Section XI, Paragraph

IWP-3 100.

3. Evaluate the adequacy of the minimum flow bypass lines for safety-related centrifugal

pumps with respect to damage resulting from operation and testing in the minimum

flow mode. This evaluation should include consideration of the effects of cumulative

operating hours in the minimum flow mode over the lifetime of the plant and during

the postulated accident scenario involving the largest time spent in this mode. The

evaluation should be based on best current estimates of potential pump damage from

operation of the specific pump models involved, derived from pertinent test data and

field experience on pump damage. The evaluation should also include verification

from the pump suppliers that current miniflow rates (or any proposed modifications to

miniflow systems) are sufficient to ensure that there will be no pump damage from low

flow operation. If the test data do not justify the existing capacity of the bypass lines

(e.g., if the data do not come from flows comparable to the current capacity) or if the

pump supplier does not verify the adequacy of the current miniflow capacity, the

licensee should provide a plan to obtain additional test data and/or modify the miniflow

capacity as needed."

A copy of Bulletin 88-04 is included as Appendix A.

Oak Ridge National Laboratory (ORNL), under the auspices of the NRC's Nuclear

Plant Aging Research Program, reviewed industry responses to the issues identified in the

Bulletin. The principal purposes of the study were to provide a general assessment of the

type and extent of actions taken in response to the Bulletin and to determine whether

additional research is needed to resolve the issues. The review consisted of several

elements:
3

* Discussions with representatives of several pump manufacturers.

* General review of all utility correspondence to the NRC responding to the Bulletin.

* Review of the distribution of pump suppliers whose pumps are used in selected

systems.

* Detailed on-site review at selected plants.

The results of this study are documented in this report. It should be noted that NRC

review of this issue is ongoing, in that individual site inspections will continue well into

1991. Site-specific reviews offer a superior means for acquiring a proper perspective on

exactly what the industry response has been. However, it is critical that the reviews be

conducted and reported in a consistent manner in order to achieve this benefit.
5

2. DISCUSSIONS WITH PUMP MANUFACTURERS

ORNL and consultant personnel met with representatives of four of the major

manufacturers of pumps used in safety-related service in U.S. plants. Firms visited

included Byron-Jackson, Dresser Pump Division (Pacific & Worthington), Ingersoll-Rand,

and Sulzer-Bingham. These four manufacturers together have furnished about 75% of the

pumps used in the safety-related systems of primary concern.

The manufacturers' representatives were very cooperative and interested in discussing

low-flow-related concerns. Each had developed responses to utility requests associated

with the Bulletin. The focus of the conversations was on what types of pumps (not on

specific models or plants) would be most susceptible to low-flow problems, how

degradation might be manifested, and the general approach taken by the manufacturers in

responding to requests from utilities.

The vendors, in general, did not have readily available for review historical failure data

that could provide new insights into failures and degradation related to operation at low

flow. There are few or no test data related to intermittent operation of the pumps. Several

of the vendors noted that although information such as replacement parts ordered for

specific pumps could be a useful indicator of the nature of historical problems and could

potentially provide some insights into the extent to which low-flow operation was a factor,

they either did not have the data compiled in a structure that would lend itself to easy use or

were naturally reluctant to share such information.

In response to a query as to what general types, of pump designs would be most

susceptible to low-flow degradation, most pointed to high-energy and high-suction specific

speed [high-flow, low-net positive suction head (NPSH) requirement] pumps. It was

noted that pump/system interaction can be a dominant source of damaging

pulsation/vibration.

Failure/degradation modes associated with low-flow operation that were most often

mentioned were seal failure, occasional shaft breakage, bearing failure, excessive wear of

wear rings, and cavitation damage. It is important to recognize that the OEM is not

involved in a comprehensive root cause analysis of every pump failure. Also, the OEM

may be contacted for a replacement or spare part by a customer other than the party that

experienced a failure (for example, where one utility buys a replacement part from another

utility in order to expedite pump repair, and the supplying utility repurchases another

spare). As a result, the OEM may not be made aware of the circumstances involved in the

failure. There is no data base of pump failures that have been caused by low-flow

degradation.

The issue of what would be appropriate means to qualify a pump for service at lowflow conditions was discussed. Although there was some diversity of how to practically

do this, there was fairly uniform agreement that what needed to be done was to measure the

forces present (e.g., radial thrust) in order to predict component life. The value of field

testing a pump to verify its ability to withstand certain conditions was discussed. There

was a general consensus that a successful test would only verify that pump's capability

under the test conditions. In other words, as-installed testing would demonstrate the
6

performance of the specific rotating element/stationary part configuration, whe6 operated at

the specific system test conditions. One manufacturer noted that tolerance stack-ups and

pretest service life could play major roles in the results of such testing. Another

manufacturer noted that in tests conducted by their company overseas, the hydraulically

induced forces associated with several pumps of the same model varied by a factor of 3 to

4.

The potential value of conducting a test program to better quantify the significance of

some of the important design and service factors to low-flow degradation was discussed.

Some of the pump manufacturers observed that because of the broad spectrum of pumps, it

would be very difficult to bound all concerns for all pumps with a testing program that

involved only a few pump designs; however, there was agreement that a relatively small,

but well controlled and monitored, test could contribute substantially to a clearer

understanding of the important parameters related to this issue. Several of the

manufacturers emphasized the need for such a test program to address intermittent

operation at low flow, in light of the fact that there are no objective data related to such

operation.

The subject of the suitability of current monitoring practices was discussed. There

was consistent agreement that testing pumps under miniflow conditions was of little value

from a hydraulic performance (head and flow) standpoint. The vendors observed that of

the means that are currently practicable, vibration monitoring and trending were the best

indicators of potentially damaging conditions. However, it was noted that monitoring

capability for some pumps (specifically, deep-well pumps) was limited. It was also noted

that the parameter that most needed to be measured was force on pump components, and

that could only be accurately monitored currently using intrusive means. Two changes to

current in-service testing (IST) practices were recommended:

1. periodically conduct flow, head, and vibration measurements when operating at close

to the pump's best efficiency point (BEP) in order to verify that pump performance has

not substantially degraded and

2. minimize or discontinue the practice of routinely testing pumps at minimum flow

conditions in order to demonstrate pump operability.

Some generalized methods for determining acceptable minimum flow rates have been

published by individuals, including employees of the pump manufacturers.2,* These

methods, which are empirical/analytical in nature, have been used to provide general

guidance on the establishment of acceptable flow regimes. The methods, however, are

geared toward continuous operation rather than the intermittent duty conditions under

which many of the safety-related pumps are used. Furthermore, one of the manufacturers

cautioned against using the generic correlation* for Bulletin 88-04 purposes.

The recommended minimum flow rates suggested by the manufacturers to utilities

appear, for the most part, to provide reasonable, conservative guidelines for safe operation

of the pumps. In most cases, the vendors notified utilities that more detailed assessment of

* S. Gopalakrishnan, "A New Method for Computing Minimum Flow," presented at the Fifth

International Pump Symposium, Texas A&M University, 1988.
7

a particular pump application could be made if the pump did not fall within the general

guidelines.

A variety of flow requirement structures was provided by the OEMs. Most typical of

these structures was a recommended minimum flow for continuous operation along with

lower flow requirements for briefer operating periods. An example is provided below.

Example

Pump BEP conditions: 4100 gal/min @ 325 ft head.

Minimum flow for continuous operation: 1500 gal/min.

Minimum flow - short-term operation (defined as less than 2 h in 24 h): 1100 gal/min.

In some cases, the manufacturers provided a very short term flow rate, principally for

start/stop operation (e.g., less than 15 min). The fraction of BEP flow recommended and

the structure of the recommendations depended on the methodology used by the OEM, the

pump design, and, to a certain extent, utility operating practices.
9

3. ASSESSMENT OF PLANT RESPONSES

3.1 GENERAL DISCUSSION

The correspondence from all plants to the NRC on Bulletin 88-04 was reviewed to

provide an indication of the range of actions taken in response to the Bulletin. The review

was made to evaluate the licensees' analyses and data for low-flow operation presented in

their response and to determine what actions, in terms of design changes, procedure

changes, special inspections, etc., have been or will be made.

The level of information provided in the correspondence varied substantially. For

some plants, there was a fairly detailed discussion of original and current minimum flow

recommendations and existing system configuration, as well as an identification of specific

design, procedural, or other changes made in response. There were also a number of

responses that provided only an indication that the issues had been reviewed, with little or

no systen/pump-specific information provided.

Prior to conducting the plant-by-plant review, several systems whose pumps were to

be included in the study were identified. Some of the systems with smaller safety-related

pumps, such as diesel fuel oil transfer and boric acid transfer pumps [pressurized water

reactors (PWRs)] were not included. The systems considered are listed in Table 1.

Table 1. Systems included in assessment

Systema Acronym

Auxiliary feedwater (PWR) AFW

Component cooling water (PWR) CCW

Containment spray (PWR) CS

Emergency service water (PWR and BWR) ESW

High-pressure coolant injection (BWR) HPCI

High-pressure core spray (BWR) HPCS

High-pressure safety injection/centrifugal charging pump (PWR) HPS1ICCP

Low-pressure coolant injection/residual heat removal (BWR) LPC1IRHR

Low-pressure core spray (BWR) LPCS

Low-pressure safety injection/residual heat removal (PWR) LPSI/RHR

Reactor core isolation cooling (BWR). RCIC

aPWR = pressurized water reactor, BWR = boiling water reactor.

The actual system designs vary from plant to plant. For example, some PWR plants

have combination charging/high-pressure safety injection pumps, while others have both a

centrifugal charging pump (often in addition to positive displacement pumps) and highpressure safety injection pumps. Thus, the individual plant systems that provide the

general functions associated with the generic systems listed in Table 1 were considered.
10

3.2 PROCEDURAL AND DESIGN CHANGES

The responses were reviewed to determine if the plant indicated that procedural

changes had been or were being implemented. There were 29 PWR units (at 18 sites) that

identified a total of 56 procedure changes and 38 design modifications made in response to

the Bulletin. There were 8 BWR units (at 6 sites) that identified a total of 17 procedure

changes and 6 design modifications made as a result of the Bulletin. Several of the units

that identified changes noted changes to design or procedures in several different systems.

The distribution of identified changes, by system, is provided in Table 2. Figures 1 and 2

depict this information graphically.

Table 2. Distribution of procedural and design changes by system

Number of plants Number of Number of plants

Plant changing procedures changing

type System procedures changeda design

PWR AFW 10 12 9

PWR CCW 2 2 2

PWR Cs 4 4 1

PWR ESW 2 2 0

PWR HPSI/CCP 5 5 9

PWR LPSVRHR 23 31 17

BWR ESW 0 0 0

BWR HPCI 0 0 0

BWR HPCS 2 2 0

BWR LPCIIRHR 6 7 0

BWR LPCS 6 7 6

BWR RCIC 1 1 0

Total 61 73 44

aMore than one type of procedure was changed at some plants.

It should be noted that the numbers of changes identified above are the numbers of

plants whose response identified that one or more procedure or design changes associated

with the particular system had been or would be made. It is likely that there were other

procedural or administrative actions taken that were not specifically identified in the

responses. Also, several plants noted procedural controls that were in existence prior to

issuance of the Bulletin.
11

25

20

15

10

5

ORNL-DWG 91-2742 ETD

A~~11 (Y'UI T~~77U7 T-TDVZT/OOfD I DQTIDLT7D JXA V, '-'.- TV %-ky A.LY TV LJA Li .. L L &JM/LALArL

* Design Procedure l M

Fig. 1. Number

were made.

of PWR units in which design and procedure changes

ORNLDWG 91-2743 ETD

6

5

4

3

2

1

0

ESW HPCS LPCUIRHR LPCS RCIC

l I I e I i . -P e I r

Fig. 2.

were made.

Number of BWR units in which design and procedure changes
12

In this respect, operating procedures for several systems from one plant (not one of the

three plants discussed in Sect. 5) were reviewed. It was found that all of the safety-related

system procedures reviewed included precautions relative to minimum allowable flow for

pump protection. The systems that are not safety related did not include such precautions,

indicating that the precautions had been placed in the procedures in response to the Bulletin

(or in response to some earlier operating experience feedback activity). In this utility's

response to the NRC, there was no mention of the procedural precautions. This is an

indication that the individuals responsible for the Bulletin response may not have been

intimately familiar with operating procedural requirements and precautions.*

Several different types of design changes were identified in the responses. These are

summarized in Figure 3. Most of the design changes involved either increasing the size of

the orifice in the miniflow line or otherwise modifying the minimum flow line. Eight plants

committed to install check valves downstream of pump discharge miniflow line connections

in order to ensure that pumps could not be dead-headed. Five commitments involved a

change to a control or alarm system related to pump minimum flow protection.

Figure 4 indicates the distribution of the types of procedures that were changed in

response to the issues discussed in the Bulletin. Note that some plants changed more than

one type of procedure for a given system; thus the total number of procedure changes

indicated by Figure 4 is greater than that indicated by Table 2 and Figs. 1 and 2.

ORNL-DWG 91-2744 ETD

25 -

20

15

10-

5-

[3 Changed control or alarm setpoint 0] Orifice/miniflow line change

| Installed check valve

Fig. 3. Number of design changes by type of change.

* In the author's experience, this is frequently a reflection of the fact that general office staff, as

opposed to operating plant staff, are often responsible for these types of responses (as was the situation in

this case).
13

-ORNL-DWG 91-2745 ETD

20

15

10

5

0

ElMaintenance X Emergency XUndesignated/other

Test procedures :3System operating

Fig. 4. Number of procedure ch anges -by type of procedure.

A majority of the procedural and design changes indicated in'the responses were made

by a relatively small number of plants. Figure 5 demonstrates that all of the indicated

changes were made by just less than half of the plants. Half of the corrective actions

discussed in the plant responses were made by about 10% of the plants. Note that this plot

is structured such that Plant 1 made the most changes, Plant 2 the second most, and so

forth. About half of the plants did not identify changes that had been or would be made.

...... ORNL-DWG 91-2746 ETD

W

bo U

'4-

0

Von

V

U

100

80

60

40

20

0

Plant Number (in descending order of number of changes made)

Fig. 5. Distribution of procedural or design actions taken in response

to NRC Bulletin 88-04.
14

3.3 OTHER RESPONSE ACTIONS

In addition to design or procedure changes, special tests and/or special analyses

performed in conjunction with the Bulletin were identified. A total of 44 special tests or

inspections that either had been conducted or would be conducted to monitor pump

condition were identified. About two-thirds of these tests were associated with either

LPCS, LPC1ILPSIIRHR, or CS systems.

A number of types of special analyses were identified. Most of these involved parallel

pump competition. A total of 48 such analyses were specified. Almost half of these

involved the LPCTILPSIIRHR systems. Twenty-one of the analyses were minimum

acceptable flow calculations, based on either a published correlation or on other

undesignated bases. It is important to note that these analyses were performed by the

utility, not by the pump manufacturer.

Fifteen of the analyses involved hydraulic flow balance-type calculations performed to

verify that pump competition was not a concern. While there were many examples where

there was an assertion that pump competition was not a concern because of individual

miniflow orifices, the 15 analyses indicated a more in-depth review. However, it should

be noted that one of the plants conducting such a review subsequently found that the pumps

for which the review had been performed did, in fact, compete to the extent that one of the

pumps was dead-headed during actual operation.

Three analyses involved static/dynamic force calculations (see discussion on this in

Sect. 5). The remainder were categorized as "other/undesignated."

3.4 RESPONSE LEVEL OF DETAIL

The level of detail included and the general completeness of the information included in

the individual plant responses were assessed. Four categories were considered:

Category Description

a Extent to which pertinent systems were specifically identified

b Level of detail provided concerning the pumps and systems

c Extent to which contact with the pump manufacturers was verified

d Extent to which conformance to the pump manufacturer recommendations

was clearly identified
15

The following rating levels were established:

Rating D...r. onf

1 Extensive review, which included all appropriate information. For example, for

Category a, if all appropriate systems were noted as having been reviewed, this

rating would be given.

2 Marginal to generally adequate response, which included most appropriate

information. For example, for Category a, if all but one or two systems' were

specifically identified as having been reviewed, this rating would be given.

3 Insufficient data provided. For example, for Category a, if only one or two

systems were specifically identified as having been reviewed, this rating would

be given.

Each of the four categories was assessed for each plant. While the rating of the level

of information provided in a response is not necessarily a perfect indicator of what was

actually done during the plant's review, it does provide a secondary measure of the extent

to which the Bulletin concerns were addressed.

Most plants either identified at least those systems noted in Table 1 or at least the

majority of them. Many plants did not specifically address systems such as ESW or CCW,

which for some plants do not include miniflow lines in the design but rather depend upon

system alignment to always ensure proper minimum flow.

A substantial variation between BWR and PWR responses was noted in the areas of

contact with the pump vendors and compliance with pump vendor recommendations. It

appears that many of the BWR plants relied primarily upon the BWR Owner's Group

generic response and either did not contact the pump vendors or at least did not indicate

such contact. However, it should also be noted that, in general, BWR plants include

provision for full-flow testing of at least the pumps provided under the nuclear steam

supply system scope of supply and would be less likely to incur pump degradation

associated with routine low-flow operation of these pumps.

Figures 6 and 7 graphically provide the results of this portion of the assessment.
16

ORNL-DWG 91-2747 ETD

to

45

to

cS

c)

100

90

80

70

60

50

40

30

20

10

0

Fig. 6. Distribution of BWR plant ratings. Number of plants receiving

specified response rating by category (see categories below).

Categories

a: Extent to which pertinent systems were specifically identified.

b: Level of detail provided concerning the pumps and systems.

c: Extent to which contact with the pump manufacturers was verified.

d: Extent to which conformance to the pump manufacturer recommendations

was clearly identified.

ORNL-DWG 91-2748 ETD

bo

4-4

0

iS

to

c

._

04

c

100

90

80

70

60

50

40

30

20

10

0

Fig. 7. Distribution of PWR plant ratings. Number of plants receiving

specified response rating by category (see categories above).
....

17

4. DISTRIBUTION OF PUMPS BY TYPE

A review was conducted of Nuclear Plant Reliability Data System (NPRDS)

information on pumps in use in safety-related systems considered. According to the

NPRDS data base, there are 2061 Safety Class 1, 2, or 3 centrifugal pumps in service in

the systems considered in this review. A total of 30 different manufacturers are identified

as suppliers; however, almost three-fourths of the pumps were supplied by 5

manufacturers. It should be noted that a few of the pumps included in the above figures are

small, auxiliary support pumps.

Specific pump models may be used in different applications. For example, ByronJackson type DVMX pumps (horizontal, multistage pumps) are used in AFW, HPCI, and

HPSI systems. The individual DVMX pump ratings vary substantially (e.g., there may be

an order of magnitude difference in BEP flow), depending upon number of stages, impeller

size, etc. For this reason, as well as the fact that there may be considerable differences in

behavior under low-flow conditions, even for essentially identical pumps (as noted in Sect.

2), a review of failure experience by model was deemed inappropriate for this generic

study.

A tabulation of the single leading and the top five suppliers of pumps (by number of

pumps in service) for the systems considered is provided in Table 3. Because of the

similarity in design service conditions for certain systems (e.g., AFW and RCIC), some of

the systems were considered jointly with other systems.

Table 3. Distribution of pump suppliers (pump suppliers for specified

system)

Fraction of

Number of pumps supplied Fraction of

pump by leading supplier pumps supplied

System(s) suppliers for system by top five suppliers

AFW/RCIC 9 0.36 0.94

CCW 15. 0.39 0.83

Cs 11 0.36 0.89

ESW 19 0.16 0.66

RHR/LPCI/LPSILPCS 11 0.41 0.93

HPSVHPCVHPCS 18 0.42 0.84

All above systems 30 0.25 0.73

For several of the pump suppliers, there are also a number of styles of pumps in

service within a given type of system. Furthermore, other features of the system and

general plant design and operating procedures can substantially affect how pumps are

operated within given systems. Thus, it is extremely difficult to generalize about historical

problems for pumps of a certain style or pumps used within a general class of system.
19

5. ON-SITE REVIEWS OF PLANT DATA

To gain a better understanding of the issue of low-flow concerns addressed in the

Bulletin, on-site reviews at three PWR plants were conducted. One of the reviews was

made in conjunction with a one-week NRC inspection of the utility's actions taken relative

to the Bulletin; another was a shorter duration, smaller scope review that was not

performed in conjunction with an inspection activity; anrd the third was associated with a

special inspection of a specific pump. All of the reviews proved to be beneficial in terms of

providing detailed insights as to the level of attention focused on this subject at these plants.

Some examples of strengths and weaknesses noted during these reviews are provided in

the discussion below.

Normal and emergency operating procedures, as well as maintenance and surveillance

procedures, were reviewed at all three of the plants visited. Considerable variations in

operating and surveillance practices relative to pump minimum flow provisions were

found.

5.1 PLANT A

5.1.1 Strengths

Plant A conducted most of its normal surveillance testing near the BEP and had

conducted special testing of several pumps. The vibration monitoring program appeared to

be thorough, and some of the plant personnel contacted demonstrated a good understanding

of the concerns over low-flow operation.

In general, this plant's systems were well designed from the standpoint of testability -

that is, normal testing is conducted near the BEP. This should help reduce long-term

degradation. It appeared that the plant had done a thorough job of reviewing normal

operating procedures to ensure that pumps are not operated in a damaging mode during

routine operations.

5.1.2 Weaknesses

An analysis performed for pumps in one of the systems did not provide an adequate

basis for conclusions reached. The pumps in the affected system may be operated at

minimum flow conditions following certain design-basis accident conditions. As a result,

special testing of these pumps was conducted. The pumps experienced significant

vibration during the tests, which were of short duration and were performed under

minimum flow conditions specifically for the purpose of collecting vibration data. The

utility appropriately identified the concern associated with the high vibration level, notified

the NRC, and initiated a review of possible corrective measures. However, a nuclear

industry consulting organization subsequently performed a simplified analysis of the pump

and determined that it was satisfactory for long-term operation, even under the high

vibration levels. The utility accepted this analysis and used it as the basis for resolving the
20

concern. The analysis was judged by ORNL not to be adequate for its intended-purpose, in

that it translated a complex dynamic problem into a simplistic static problem and drew

inappropriate conclusions from the results.

A second weakness identified in Plant A was that utility personnel had apparently not

as carefully considered operation under the guidance of their emergency procedures as they

had operation under normal operating procedures. A review of a separate system's pumps

at Plant A indicated that there are certain operating conditions [e.g., following a small-break

loss-of-coolant accident (LOCA)] where the HPSI pumps could potentially be dead-headed

because of their minimum flow line to the refueling water storage tank (RWST) being

isolated following switchover to the containment sump. The system realignment for this

mode of operation had not been considered by the personnel reviewing system procedural

controls in conjunction with the Bulletin response.

5.2 PLANT B

5.2.1 Strengths

Plant B appeared to have done a thorough review of both their normal, test, and

emergency procedures relative to pump low-flow operation. A potential problem with one

pump (RHR) was identified, and a pump test and inspection program was developed in an

attempt to qualify the pump's ability to operate satisfactorily under low-flow conditions.

This test involved running the pump at low flow for what was then estimated to be a

conservative period of time (relative to the time that the pump would have to operate

following an accident) and subsequently disassembling and inspecting the pump.

Vibration, flow, and head data were collected during the test. As far as is discernible from

the review of all plant responses, this is the only plant to conduct this level of both testing

and inspection on a particular pump.

Plant B appeared to have carefully reviewed all their pump applications and had taken

actions (both procedural and design change) to alleviate areas of concern.

5.2.2 Weaknesses

The vibration data collected from the special test conducted by Plant B were developed

and reviewed in a format that provided little useful information. There were considerable

uncertainties associated with the flow measurements during the test. Subsequent reviews

of emergency operating procedures by ORNL indicated that the test duration (and other

parameters) may not have bounded the limiting postaccident requirements. As a result, an

analysis involving simplified assumptions was performed by a nuclear industry consulting

organization to attempt to address the duration issue and other factors.

A review of the analysis (performed for Plant B by the same organization used at Plant

A) by ORNL found that it attempted to reduce complicated dynamic problems to simplified

static conditions for evaluation. It was concluded that the method used by the consultant

was inadequate and inappropriate for the intended purpose. This conclusion was validated

during subsequent discussions with pump vendors, including the manufacturer of the RHR

pump used at Plant B.
21

5.3 PLANT C

The review of the Plant C response was completed in conjunction with an NRC

inspection. The review focused on four systems - HPSI, RHR, CS, and AFW. A

summary of the observations follows. A more detailed discussion of the review results is

provided in Appendix B.

5.3.1 Strengths

Plant C had implemented a program to provide main control board indication when

pumps were being operated under conditions that were below continuous acceptable

minimum flow (the indication consisted of labels placed on pump motor current indicators).

While pump motor running current indication is only a crude indication of pump operating

conditions, it was a consistent type of data available for all affected pumps (most of the

pumps did not have individual pump flow indication).

5.3.2 Weaknesses

This plant had several pumps with relatively low minimum flow rates, and testing is

normally conducted at or near minimum flow.

For one system (RHR), the minimum flow rates were substantially less than current

vendor recommendations. The utility's evaluation essentially relied on the fact that the

good operating experience for the pumps in question indicated that the pumps were not

being adversely affected. The corrective maintenance on the RHR pumps had, in fact, been

minimal. However, the periodic testing for these pumps was done under miniflow

conditions and thus would not have been capable of detecting hydraulic degradation.

The minimum flow lines for RHR and CS at this plant shared a common return header

downstream of the miniflow orifices. The utility had not performed an analysis of the

common line frictional head loss (which, in turn, affected the individual pump flow).

Preliminary calculations performed during this visit indicated that the common line loss was

not negligible (from the standpoint of its impact upon individual pump flow rate) if all the

pumps were operated in parallel. The utility plans to perform detailed analyses on this

subject.

In another system (AFW), the 20-gal/min flow rate available through the miniflow line

met the vendor recommendation for very short term, start/stop type of operation (defined

by the vendor as 15 min or less). However, there were no procedural or administrative

controls in place to ensure that the pump was operated within this criterion nor within the

vendor-defined minimum flow requirement of 55 gal/min for short-term operation (3 h or

less). Disassembly and inspection of the same model pumps, used in the same service, and

with similar minimum flow provisions at another utility revealed considerable wear of a

nature that indicated low-flow degradation (scoring of the balance drum, cracked diffusers,

excessive wear of wear rings, and thrust bearing damage). Plant C was not aware of the

results of the other plant's inspections.
22

For the HPSUCCP system, the condition under which the system was operated was

depended upon to maintain flow above vendor recommendations. However, a review of

procedures revealed that when the chemical and volume control system was operated

according to plant procedures in the alternate letdown mode, the total charging flow would

be less than vendor recommendations.
23

6. CONCLUSIONS

Generally speaking, the response of the industry to the low-flow degradation issues

'discussed in Bulletin 88-04 appears to be relatively superficial. -There are some exceptions

to this general observation. Furthermore, a superficial response for a specific system and

pump does not necessarily equate to the existence of a problem for that system/pump.

Given the facts that (1) low-flow degradation is, to a large extent, a common-mode failure

concern and (2) that failure of a pump cannot be readily overcome by manual operator

action (as is the case for many valve/valve operator failures, for example), it is important

that more detailed attention be focused on low-flow concerns.

Two specific conclusions can be drawn. First, there are no generic guidelines for

determining the acceptability of a pump for operation under the various modes and times

required in support of both normal and emergency conditions. Currently, ASME pump

testing requirements are used by the industry to verify pump operational readiness.

However, the ASME requirements are not geared toward demonstrating satisfactory lowflow operating capability.

Second, the low-flow issue was not adequately addressed by all plants. It is judged

that the very fact that there has been no generic guidance on how to assess the suitability of

pumps under given conditions effectively guaranteed that this would be the case, since

utilities were essentially on their own to determine how to qualify individual pumps that

were questionable.
25

7. RECOMMENDATIONS

The following recommendations are based on the review of the issue discussed in

previous sections. During 1991, ORNL will be involved in additional activities related to

Bulletin 88-04. These recommendations are thus preliminary in nature and are subject to

modification as additional insights are gained.

1. It is recommended that the industry be encouraged to develop pump qualification

criteria and processes. The program should provide a generic procedure for validating

the satisfactory application of individual pumps for all potential operating modes.

2. Consideration should be given to the development and/or implementation of new

diagnostic techniques, such as motor current signature analysis (employing advanced

signal analysis techniques), which would provide more meaningful information about

important pump operating parameters than is currently available (outside of the use of

intrusive monitoring involving elaborate test configurations).

3. A small-scale pump test program should be considered. The purpose of the program

should be to identify and quantify, at least relatively, the important parameters

associated with low-flow degradation of at least one pump model. The testing would

be partially destructive in nature since it should provide for accelerated aging of the

pump parts.

4. Additional utility review of pumps should be encouraged but should be preceded by

the completion of tasks recommended in items 1 and 2.

Recommendations 1, 2, and 3 are not unrelated. If better means of readily assessing

pump condition existed, the task of developing qualification criteria would be simplified.

Qualification criteria and process would help focus diagnostic development and application.

Test program results would help identify the parameters that are most important to monitor.

Thus, these areas should be worked in parallel and coordinated. The overall goal of these

activities should be to provide the basis for determining the acceptability of a given pump

for the conditions at which it may be required to operate.
27

REFERENCES

1. M. L. Adams and E. Makay, Martin Marietta Energy Systems, Inc., Oak Ridge Natl.

Lab., "Aging and Service Wear of Auxiliary Feedwater Pumps for PWR Nuclear

Power Plants," USNRC Report NUREG/CR-4597, Vol. 1 (ORNL-6282/V1), July

1986.*

2. C. C. Heald and R. Palgrave, "Backflow Control Improves Pump Performance," Oil

and Gas Journal, pp. 96-105 (February 25, 1985).t

* Available from the National Technical Information Service, Springfield, VA 22161.

t Available in public technical libraries.
29

Appendix A

NRC BULLETIN 88-04
31

OMB No.: 3150-0011

NRCB 88-04

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C. 20555

May 5, 1988

NRC BULLETIN NO. 88-04: POTENTIAL SAFETY-RELATED PUMP LOSS

Addressees:

All holders of operating licenses or construction permits for nuclear power

reactors.

Purpose:

The purpose of this bulletin is to request all licensees to investigate and

correct as applicable two miniflow design concerns. The first concern involves the potential for the dead-heading of one or more pumps in safety-related

systems that have a miniflow line common to two or more pumps or other piping

configurations that do not preclude pump-to-pump interaction during miniflow

operation. A second concern Is whether or not the installed miniflow capacity

is adequate for even a single pump In operation.

Description of Circumstances:

Westinghouse Electric Corporation recently notified all utilities with

Westinghouse-designed nuclear steam supply systems (NSSS) of the two concerns noted.'above. NRC Information Notice 87-59 forwarded a summary of

these concerns to all holders of operating licenses or construction permits

for nuclear power reactors and indicated that further staff evaluation might

result in a request for specific licensee actions. Several licensees have

confirmed the existence of these concerns in their plants (Turkey'Point,

H. B. Robinson, Vermont Yankee). This bulletin is the result of the staff's

evaluation.

Discussion:

When two centrifugal pumps operate In parallel and one of the pumps is stronger

than the other (i.e., has a higher developed head for the same flow), the weaker

pump may be dead-headed when the pumps.are operating in the minimum flow mode.

The phenomenon-is manifested at low flow rates because of the flatness of the

pump characteristic curve in this range.' The head difference is not a problem

at moderate to high flow conditions because of the shape of the pump charac-

-teristic curve in-these regions.

Traditionally, the required miniflow for the'se'pumps'was established solely on

the basis of pumped fluid temperature rise. 'Today, however, it is generally

8804290177
32

NRCB 88-04

May 5, 1988

Pace 2 of 4

understood that temperature rise is not the only factor influencing safe continuous minimum flow operation. Centrifugal pumps will demonstrate a flow

condition that has been described as hydraulic instability or Impeller recirculation at some point below the best efficiency point (BEP) on their characteristic curve. These unsteady flow phenomena become progressively more pronounced

as the flow is further decreased and can result in pump damage from pump vibration, excessive forces on the impeller, and cavitation. It is now generally

recommended that the limitations associated with these hydraulic phenomena be

considered when specifying minimum flow capacity.

The first potential problem involves parallel pump operation with both pumps

recirculating through a common miniflow recirculation line or with a pipina

configuration that does not preclude pump-to-pump interaction during miniflow

operation. The problem was identified on a plant whose licensee requested that

Westinghouse determine if parallel operation while on miniflow is acceptable.

Westinghouse reviewed the plant's residual heat removal (RHR) system configuration. The review determined that the potential exists for the stronger pump

to dead-head the weaker pump during low flow, parallel pump operating conditions while on miniflow only. In addition, it was determined that even without

pump interaction the recirculation flow available was not adequate to ensure continuous operation of even a single RHR pump on miniflow. Although these issues

are based on an evaluation of RHR pumps at a particular plant, the first

concern may exist at other plants cor- -ured with a common pump recirculation

flow path and the second concern may also exist at other plants independent of

whether or not there is a common recirculation pump flow path.

The NRC staff believes that these issues may be relevant to all water-cooled

reactor designs, regardless of the pump application or the NSSS manufacturer.

This is based on the belief that miniflow lines have traditionally been designed for only 5% to 15% of pump design flow. Some pump manufacturers now

are advising that their pumps should have minimum flow capacities of 5% to

more than 50% of best efficiency flow for extended operation to protect acainst

hydraulic instability or impeller recirculation problems.

Actions Requested:

All addressees are requested to do the following:

1. Promptly determine whether or not its facility has any safety-related

system with a pump and piping system configuration that does not preclude pump-to-pump interaction during miniflow operation and could

therefore result in dead-heading of one or more of the pumps.

2. If the situation described in Item 1 exists, evaluate the system for flow

division taking into consideration (a) the actual line and component resistances for the as-built configuration of the identified system; (b) the

head versus flow characteristics of the installed pumps, including actual

test data for "strong" and "weak" pump flows; (c) the effect of test instrument error and reading error; and (d) the worst case allowances for

deviation of pump test parameters as allowed by the American Society of

Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Section Xt,

Paragraph IWP-3100.
33

NRCB 88-04

May 5, 1988

Page 3 of 4

3. Evaluate the adequacy of the minimum flow bypass lines for safetyrelated centrifugal pumps with respect to damage resulting from

operation and testing in the minimum flow mode. This evaluation

should include consideration of the effects of cumulative operating

hours in the minimum flow mode over the lifetime of the plant and

during the postulated accident scenario involving the largest time

spent in this mode. The evaluation-should be based on best current

estimates of potential pump damage from operation of the specific

pump models involved, derived from pertinent test data and field

experience on pump damage. The evaluation should also include

verification from the pump suppliers that current miniflow rates

(or any proposed modifications to miniflow systems) are sufficient

to ensure that there will be no pump damage from low flow operation.

If the test data do not justify the existing capacity of the bypass

lines (e.g., if the data do not come from flows comparable to the

current capacity) or if the pump supplier does not verify-the adequacy

of the current miniflow capacity, the licensee should provide a plan

to obtain additional test data and/or modify the miniflow capacity as

needed.

4. Within 60 days of receipt of this bulletin, provide a written response

that (a) summarizes the problems and the systems affected, (b) identifies

the short-term and long-term modifications to plant operating procedures

or hardware that have been or are being Implemented to ensure safe plant

operations, (c) identifies an appropriate schedule for long-term resolution of this and/or other significant problems that are identified as a

result of this bulletin, and (di provides justification for continued

operation particularly with regard to General Design Criterion 35 of

Appendix A to Title 10 of the Code of Federal Regulations (10 CFR 50),

"Emergency Core Cooling" and 10 CFR 50.46, "Acceptance Criteria for

Emergency Core Cooling System for Light Water Nuclear Power Reactors."

5. Within 30 days of completion of the long-term resolution actions, provide

a written response describing the actions taken.

6. An evaluation of vour actions in response to this bulletin should he

documented and maintained at the plant site for a minimum of two (2)

years. That evaluation should, as a minimum, address the piping system

configuration in accordance with Item 1 above, each of the four factors

discussed in Item 2, pertinent test data and field experience on minimum

flow operation, and verification of the adequacy of current miniflow

capacity by the pump manufacturer.

The written reports, required above, shall be addressed to the U.S. Nuclear

Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555,

under oath or affirmation under the provisions of Section 182a, Atomic Energy

Act of 1954, as amended. In addition, a copy shall be submitted to the

appropriate Regional Administrator.
34

NRCB 88-04

May 5, 1988

Paae 4 of 4

This requirement for Information was approved by the Office of Manaoement

and Budget under clearance number 3150-0011. Comments on burden and duplication should be directed to the Office of Management and Budget, Reports

Management, Room 3208, New Executive Office Building, Washington, D.C. 20503.

If you have any questions about this matter, please contact the technical

contact listed below or the appropriate ERR project manager.

charles E. Rossi, Director

Division of Operational Events Assessment

Office of Nuclear Reactor Regulation

Technical Contact: T. Collins, NPR

(301) 492-0897

Attachment: List of Recently Issued NRC Bulletins
35

Appendix B

PLANT C RESPONSE TO NRC BULLETIN 88-04

During the on-site visit at Plant C, design features; operating, testing and maintenance

procedures; and plant analysis relative to Bulletin 88-04 were reviewed. This review was

done in conjunction with an NRC inspection of the plant response to the Bulletin.

Appropriate excerpts from the report prepared are provided below.

B.1 GENERAL OBSERVATIONS

B .1 .1 Written Response Evaluation

Essentially -all of the utility responses to NRC Bulletin 88-04 have been reviewed by

ORNL personnel. The range of responses has been'broad, both in the context of the

quantity and quality of useful information contained in the responses, as well as in the

indicated actions taken to address the concern.

The response of Plant C provided design information, simplified drawings of the

recirculation systems, and estimated times at minimum'flow in various procedures for

HPSI, RHR, CS, and AFW pumps. In -that regard, the response provided more

substantive information than most responses.

With the exception of the RHR pumps, the Plant C response did not discuss current

vendor recommendations. The RHR pump'discussion was included because the minimum

flow provided was less than vendor recommendations. The plant basically relied on lack of

historical failure and observed degradation' as the basis for'system adequacy. This basis

has been used in other responses, in -the context that utilities have argued that they have not

experienced failures or degradation due to low-flow operation.

-Thus, the Plant C response, at face value, and without looking at particular pumps,

corresponded to an average response to NRC Bulletin 88-04.

B .1.2 On-Site Inspection Activities - General Observations

The site review indicated that the level of scrutiny provided, from an overall context,

was minimal. This observation is based on several diverse indicators:

1. A review of the operating procedure for the chemical and volume control system

indicated that the procedure allowed continuous operation under certain conditions

(alternate letdown) that would result in flow less than the vendor recommendations for

continuous operation,- even though the utility had asserted that the pumps 'would be

operated at greater than the vendor-recommended flow during normal operation.

2. A review of recent full-flow test data for the RHR and CS pumps indicated that the

pumps appear to be operating substantially below the results from the vendor and

preoperational testing. While the data may be misleading because of instrumentation
36

inaccurately depicting pump condition, the plant had not compared test results to the

original data. This is an important point because the plant largely depended upon a

lack of historical failure/degradation as indication that low-flow degradation has not

occurred, and this type of testing provides the best indication about whether the

hydraulic performance of the pump has degraded or not.

3. There is a common recirculation line for the RHR and CS miniflow lines downstream

of the individual pump's minimum flow orifices. The backpressure from parallel

operation had not been considered by the plant. The utility is now in the process of

calculating the effect. The impact may turn out to be relatively small (on the order of

10% reduction in individual pump flow rate, compared with the flow rate when

operated alone); however, it had not been considered prior to the NRC inspection.

4. Plant personnel were unaware that another plant (Plant D) using the same model

pumps as those used in Plant C's AFW system had found substantial damage

associated with low-flow operation. The damage was found when the pumps were

disassembled for inspection as part of the Plant D response to Bulletin 88-04.

It is important to note that Plant C depended heavily upon successful test results and

the lack of calls for corrective maintenance required for the AFW pumps in assessing pump

suitability. Plant D noted that performance tests had not identified the damage; rather, it

had been discovered only when the pumps were disassembled.

These observations led to the conclusion that a relatively cursory review had been

performed.

The plant had taken one notable action to help avoid operation at low flow. Each

pump considered had pump motor current indication provided on the main control board.

The plant had implemented a program to "band" the motor current indicators to provide

operators with an indication of the suitability of the flow regime in which the pumps were

being operated. Motor current was selected instead of pump flow because the plant did not

have individual flow instruments dedicated to each of the pumps.

As a footnote to these general observations, a concern about the turbine-driven AFW

pump was identified. This pump has - 20-galVmin minimum flow. About half of the flow

goes back to the condensate storage tank, while the other half goes through lube oil and

turbine bearing coolers and then back to pump suction. This would result in a continuing

increase in suction temperature with time. Operation of this pump for a protracted period in

minimum flow would create overheating problems (note that the temperature rise on a

single pass through the pump is estimated to be 500 F).

B.2 OBSERVATIONS CONCERNING MAINTENANCE AT PLANT C

Maintenance files for several of the safety-related pumps at Plant C were reviewed

during the inspection. Failure reports for the Plant C pumps, as well as failure reports for

the same model pumps in service at other plants that have been provided to the Institute of

Nuclear Power Operation's NPRDS data base, were also reviewed.
37

B.2.1 RHR Pumps

RHR pump A was disassembled and inspected for the first time in 1989. The only

wear noted was in bushings and sleeves on the intermediate portion of the shaft.

RHR pump B was disassembled just prior to the NRC inspection. Both the lower and

upper shaft sections (there are a total of three shaft sections) were noted to be about 37 mils

out of true, their bushings were worn, and the first-stage wear ring clearances were found

to have increased by about 50 mils, based on verbal communications with maintenance

personnel. Based on subsequent discussions with utility personnel, no further damage was

noted.-

B.2.2 Containment Spray Pumps

No maintenance involving pump disassembly was noted.

B.2.3 AFW Pumps

AFW pump A failed to develop required discharge pressure during periodic testing in

late 1983. The pump manufacturer, who was called in to investigate, found excessive

thrust clearance. Damage to the seventh- and eighth-stage impellers was found, the balance

piston was badly scored, and the thrust bearing had begun to wipe.' The utility attributed

this to the original shaft positioning (balance drum setting)..

It should be noted that Plant D, which uses the same model pump as is used for the

motor-driven pumps at Plant C, disassembly and inspection of the AFW pumps found

extensive indication of low-flow degradation, including wiped thrust bearings, balance

drum grooving, and excessive wear ring clearances, even though the pumps had passed

performance tests. The nature of the tests is unknown. Plant D plans to install a full-flow

test line for pump testing and extended operation.

B.2.4 HPSI Pumps

HPSI pump C was found with an eroded fifth stage in 1982. During 1985, a

mechanic found the B pump shaft to be bound. Internal damage was noted. At the time of

the inspection, the C pump rotating element had been removed as a result of some

performance degradation noted by in-service testing. Preliminary inspection noted erosion

on the suction side of the first stage. This led maintenance personnel to question whether

the erosion on the fifth stage noted in the earlier event was really on the first stage, which

seems to be a valid observation. Note that this pump has opposed impellers, with the fifth

stage at one end and the first stage at the other.

B.2.5 Discussion

Plant C has experienced a relatively low rate of failures with the above-mentioned

pumps. While there is some indication of degradation due to operation at low flow for the
38

HPSI pumps, it is inconclusive. In this context, however, it is important to note that Plant

D did not know of degradation in their AFW pumps until they disassembled and inspected

them.

The maintenance personnel at Plant C were very forthright and cooperative. The

experienced personnel we met with appeared to be extremely competent.

B.3 OBSERVATIONS FOR RHR AND CS PUMP DESIGN AND

TESTING AT PLANT C

During the on-site review at Plant C, it was noted that the RHR and CS pump data

collected during testing that is conducted on a refueling frequency appeared anomalous. In

addition, it was noted that these pumps share a 6-in. common recirculation line.

Background information and discussions follow.

B.3.1 Design and Test Results

B.3.1.1 Pump and miniflow design

The RHR pumps are Vendor A, two-stage vertical can pumps, Model Y, with 400-hp,

1760-rpm motors, which were furnished by Vendor A as part of the pump package. The

pump BEP is - 4100 gal/min at 325 ft head. The miniflow orifices were originally sized to

provide 350 gal/min at 325-ft differential pressure. With a single pump operating,

measured flow rate for each RHR pump operating under miniflow only has been - 400

gal/min (note that the orifice differential pressure under minimum flow is - 365 ft).

The CS pumps are Vendor B, two-stage vertical can pumps, Model Z, with 350-hp,

1780-rpm motors. The pump BEP is - 4750 gal/min at 265 ft head. With a single CS

pump operating, measured flow rate for each pump operating under miniflow only has

been - 300 gal/min.

A portion of the minimum flow line for the RHR and CS pumps is common to all

RHR and CS pumps. Each pump has its own miniflow restriction orifice upstream of the

common portion of the line.

B.3.1.2 Current vendor-recommended minimum flow

RHR Pumps

Continuous: 1500 gal/min (-37% BEP); Short term (_• 2 h in 24 h): 1100 gal/min. Vendor

A specifically warned that at 350-galmin minimum flow, the following could occur:

* Considerable internal vibration and hydraulic instability could occur.

* Pump bushings could be damaged.

* Sustained vibration during an incident could result in significant damage or failure of

the bushings, with extensive damage to the pumping element.

CS Pumps

Continuous: 2300 gal/min (-48% BEP); Intermittent: 1600 gal/min (60 to 1500 h/year);

Short term (_ 60 h/year): 200 gal/min.
39

B.3.1.3 Test Data

Preoperational test data for both RHR pumps were reviewed. In the case of RHR

pump A, the data were found to be consistent with the vendor-supplied test curve. The

preoperational test curve for RHR pump B was below the vendor curve, particularly at the

higher flow end of the curve. Current IST data for the RHR pumps were reviewed and

found to be below both curves for both pumps. The IST data are acquired with

recirculation flow only; at a measured flow rate of 400 gal/min.. It should be noted that the

400 gal/min is in excess of the 350 originally specified; however, it appears to be consistent

with the installed orifices.

During each refueling outage, a test (STI) is conducted in which flow is delivered to

the core at flow rates in excess of the pump BEP. This test has recorded flow rate,

discharge pressure, and suction pressure for parallel pumps. Using the test data from

1990, it appears that the A and B pumps' performance deviates from vendor/preoperational

test data in the range at which the test is conducted (4550 to 4900 gal/min). The noted

deviation is estimated to be as follows:

Pump Pressure deviation at test flow (%) Rlow deviation at test head (%)

A 27/14 1218

B 2117 11/4

The above estimates are relative to the preoperational test results. The first entry in

each column assumes no head loss from pump discharge to the pressure instrument that is

used to measure discharge pressure and no pump suction pressure depression when the

pump is running. The second entry is based on measured head plus 15 psi in an effort to

approximate head loss from pump discharge to the discharge pressure instrument and

suction pressure depression when the pump is running.

The preoperational testing of the CS pumps did not measure flow. During each

refueling outage, flow is measured within the same procedure used for RHR flow

measurement (ST1). These data indicate that all three of the pumps are performing

substantially below the originally supplied vendor curves (indicated head is only about twothirds that measured at the vendor test facility for the same flow). Even if it is assumed that

there is a 15-psi difference between pump actual head and that implied by data from

procedure STI, the indicated operating point for pumps would still be more than 20% less

(in either flow or head) than the vendor test curve indicates.

For both the RHR and CS pumps, recent test data appear anomalous. The cause(s) of

the anomalous data is unclear and may very well be explainable by instrumentation

accuracy, measurement points, etc.' Plant C indicated that further analysis and review

would be pursued.

* It appears that the anomalous data are the result of something other than the pumps, since all the

pumps seem to show the same general trend. However, this needs to be borne out by further analysis.
40

B.3.2 RHR/Containment Spray Common Return Line Review

During the review of system designs and operation, it was noted that the recirculation

line for the CS and RHR pumps is a common line downstream of the orifice for each of the

pumps. Isometric drawings for a portion of the common return line were reviewed during

the visit. A complete trace of the common line through isometric drawings was not made

because of time constraints. Based on estimated flow rates for each of the pumps when

operated under minimum flow conditions, the backpressure associated with frictional and

elevation head losses of just the reviewed portion of the common line when all pumps were

operated simultaneously was sufficient to cause a 5% reduction in individual pump flow,

assuming clean pipe. If allowances for some fouling and the additional pipe and fittings

were made, the flow reduction associated with multiple pump operation in parallel would

be more significant, perhaps as high as 25%.

Plant C had not considered common line losses prior to this visit but indicated that they

planned to perform a detailed analysis.
41

NUREG/CR-5706

ORNL6671

Dist. Category RV

Internal Distribution

1. .J.M.Asher 41. K. H.Luk

2-32. D. A. Casada 42. J. C. Moyers

33. R. L. Clark 43. G. A. Murphy

34. D. F. Cox 44. C. E. Pugh

35. D. M. Eissenberg 45. ORNL Patent Office

36. E. C. Fox - 46. Central Research Library

37. R. H. Greene 47. Document Reference Section

38. H. D.-Haynes 48-49. Laboratory Records Dept.

39. J. E. Jones Jr. .50. Laboratory Records (RC)

40. R. C. Kryter

External Distribution

51. M. L. Adams, Department of Mechanical and Aerospace Engineering, Case

Western Reserve University, Cleveland, OH 44106

52. Y. C. Li, U.S. Nuclear-Regulatory Commission, MS-7E23, Washington, DC

20555

53. E. J. Sullivan, Jr., U.S. Nuclear Regulatory Commission, MS-7E23,

Washington, DC 20555

54. S. Gopalakrishnan, Vice President and Chief Engineer, BW/lP International,

Inc., 2300 East Vernor Avenue, Vernon, CA 90058

55. M. R. Prescott, Chief Engineer, Dresser Industries, Pacific Pumps Division,

5717 Bickett Street, Huntington Park, CA 90255

56. W. D. Marscher, Engineering Manager, Dresser Pump Division, 401

Worthington Avenue, Harrison, NJ 07029

57. D. P. Sloteman, Manager - Pump R&D, Ingersoll-Rand Company, 942

Memorial Parkway, Phillipsburg, NJ 08865

58. D. L. Eddy, Manager of Field Engineering, Sulzer Bingham Company, 2800

Northwest Front Avenue, P. 0. Box 10247, Portland OR 97210

59. G. Sliter, Electric Power Research Institute, P. 0. 10412, Palo Alto, CA

94303

60. J. W. Tills, Institute for Nuclear Power Operations, 1100 Circle 75 Parkway,

Atlanta, GA 30339-3064

61. J. H. Taylor, Brookhaven National Laboratory, Bldg. 130, Upton, NY

11973
42

62. A. B. Johnson, Battelle-PNL, MS P8-10, P. 0. Box 999, Richland, WA

99352

63. J. P. Vora, U.S. Nuclear Regulatory Commission, Office of Nuclear

Regulatory Research, Electrical and Mechanical Engineering Branch, 5650

Nicholson Lane, Rockville, MD 20852

64. G. H. Weidenhamer, U.S. Nuclear Regulatory Commission, Office of

Nuclear Regulatory Research, Electrical and Mechanical Engineering Branch,

5650 Nicholson Lane, Rockville, MD 20852

65. E. J. Brown, U.S. Nuclear Regulatory Commission, Office for Analysis and

Evaluation of Operational Data, Reactor Operations Analysis Branch,

Maryland National Bank Building, 7735 Old Georgetown Road, Bethesda,

MD 20814

66. C. Michelson, Advisory Committee on Reactor Safeguards, 20 Argonne

Plaza, Suite 365, Oak Ridge, TN 37830

67. M. Vagins, U.S. Nuclear Regulatory Commission, Office of Nuclear

Regulatory Research, Chief, Electrical and Mechanical Engineering Branch,

Division of Engineering, 5650 Nicholson Lane, Rockville, MD 20852

68. W. S. Farmer, U.S. Nuclear Regulatory Commission, Office of Nuclear

Regulatory Research, Electrical and Mechnical Engineering Branch, 5650

Nicholson Lane, Rockville, MD 20852

69. M. J. Jacobus, Sandia National Laboratories, P.O. Box 5800, Division 6447,

Albuquerque, NM 87185

70. H. L. Magleby, Idaho National Engineering Laboratory, MS 2406, P. 0. Box

1625, Idaho Falls, ID 83415

71. Office of Assistant Manager for Energy Research and Development,

Department of Energy, Oak Ridge Operations Office, Oak Ridge, TN 37831

72-73. Office of Scientific and Technical Information, P.O. Box 62, Oak Ridge, TN

37831

74-403. Given distribution as shown in NRC category RV (10 - NTIS)
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

129J91 lu. A"d by NRC. Aod Vol., S......................... SHOD.. mRe.. NRCFA 1102. dr~~~~~~~~~~~~~~~~~~~~~~~~~~~~~wd Ado."Vg. N.-w esr. Al &,V.)

3201.3202 BIBLIOGRAPHIC DATA SHEET NUREG/CR-5706

{See instructions on the reverse] ORNL667 1

2. TITLE AND SUBTITLE

Potential Safety-Related Pump Loss - An Assessment of Industry Data 3. DATE REPORT PUBLISHED

. ~~MONTH Y EAR

NRC Bulletin 88-04 Jume 1991

4. FIN W01iNUMBER

S. AUTHOR IS) 6. TYPE OF REPORT

D. A. Casada Technical

7. PERIOD COVERED li/ctsur.o.e sw

E. PERFORMING ORGANIZATION - NAME AND ADDRESS t1f NRC. Pow;* Dihkon0ft. Olfrct efifp4. U.S Nookee Aluaito C>- noWs Ed Tier dddU: it Cot rape

,wt a m.JiiV .dd'ru

Oak Ridge National Laboratory

Oak Ridge, TN 37831

S. SPONSORING ORGANIZATION -NAME AND ADDR ESS I" N#iRC. .s oypt :fA'roruroo. NRC O'er Office of Region. U.& Nucear Ruguartory Comm;";a.

aed mailiw"r vm.d)YL

Division of Engineering

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

10. SUPPLEMENTARY NOTES

11. ABSTRACT I2O wo,* or oJ

This report documents the results of a study of the nuclear industry's response to NRC Bulletin

88-04. The work was conducted for the U.S. Nuclear Regulatory Commission (NRC) Nuclear

Plant Aging Research Program. All written correspondence between utilities and the NRC was

reviewed and classified. Major pump vendors were interviewed to discuss their perspectives on

low-flow degradation of pumps. Individual sites were visited to review the details of system

design and procedural controls relative to the Bulletin issues.

12. KEY WORDS/DESCR PTORS IL tha, t assist .hv - IOCJiM^ low ferot. 1I. 13e,,. AVAILABILITY STATEMSENT

pumps, low-flow, surveillance, testing, inspection, NPAR, residual Unlimited

heat removal, auxiliary feedwater 14. SECURITY CLASSIO ICAT7ON

(This leaeei

Unclassified IThas #leoo~t

Unclassified

15. NUMBER Of PAGES

16 PRICE

NRC FORM 335 12.891
THIS DOCUMENT WAS PRINTED USING RECYCLED PAPER
UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555

SPECIAL FOURTH-CLASS RATE I POSTAGE & FEES PAID

USNRC

1

PERMIT No. G-67

zC

C.r.

-40

0s

OFFICIAL BUSINESS

PENALTY FOR PRIVATE USE, $300