ML20212A494

Ack Receipt of 860514 Response to IE Bulletin 85-003.Addl Info Requested within 30 Days on Action Item E Re Program to Accomplish Action Items b-d

ML20212A494
Person / Time
Site:	Waterford
Issue date:	07/23/1986
From:	Gagliardo J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:	Muench G LOUISIANA POWER & LIGHT CO.
References
IEB-85-003, IEB-85-3, NUDOCS 8607290023
Download: ML20212A494 (2)

Text

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F JUL 2 31986 C

In Reply Refer To:

Docket: 50-382 Louisiana Power & Light Company ATTN:

G. W. Muench, Director Nuclear Operations 317 Baronne Street P. O. Box 60340 New Orleans, Louisiana 70160 Gentlemen:

Thank you for your letter of May 14, 1986, in response to IE Bulletin No. 85-03 dated November 15,1985. - As a result of our preliminary review, we find that additional information is needed.

Specifically, Action Item e requests submittal of a program to accomplish Action Items b through d.

While your letter states tht Louisiana Power &' Light Company (LP&L) intends to accomplish the Action Items, it does not present a program to ensure completion.

Please provide the supplemental information within 30 days of the date of this letter.

" Original signed byJ U. E. GAG 13ARDQu J. E. Gagliardo, Chief Reactor Projects Branch cc:

Louisiana Power & Light Company ATTN:

G. E. Wuller, Onsite Licensing Coordinator P. O. Box B' Killona, Louisiana 70066 Louisiana Power & Light Company ATTN:

R. P. Barkhurst, Plant Manager P.

0.- Box B Killona, Louisiana 70066-Middle South Services ATTN: Mr. R. T. Lally P. O. Box 61000 New Orleans, Louisiana 70161 RIV:RPB/C M [ C:RPB/C @

C:RPB HFBundy:cnm GLConstable JEGag do 7//2f/86 7//pf86 7/M/p6j 8607290023 860723 PDR ADOCK 05000382

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ATTN:

K..,W. Cook, Nuclear Support and Licensing Manager 317 Baronne Street P. O. Box 60340 New Orleans, Louisiana 70160 Louisiana Radiation Control Program Director becto'DMB(IE01) bcc distrib. by RIV:

RPB D. Weiss, LFMB (AR-2015) l Resident Inspector R. D. Martin, RA SectionChief(RPB/C)

DRSP R&SPB RSB MIS SYSTEM D. Crutchfield, NRR RSTS Operator H. Bundy RIV File M. Jayne ELJordan RJKiessel, IE

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LOUISI ANA / 317 BARONNESTREET P. O. BOX 60340

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POWE R & LIG HT NEW ORLEANS, LOUISlANA 70160 (504) 595-3100

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NON rsNU May 14, 1986 W3P86-0076 A4.05 QA Mr. Robert D. Martin Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 1000 1

Arlington, TX 76011 i

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Subject:

Waterford SES Unit 3 i.

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d Docket No. 50-382 8

IE Bulletin 85-03

Dear Mr. Martin:

Through IE Bulletin 85-03, " Motor-Operated Valve Common Mode Failures During Plant Transients Due To Improper Switch Settings", the NRC requested certain information for motor-operated valves (MOVs) in the high pressure safety injection (HPSI) and emergency feedwater (EFW) systems.

In accordance with Item (a) of the Bulletin, LP&L has reviewed the design basis for operation of each MOV in the HPSI and EFW systems of Waterford 3.

The review results are documented in Attachments 1 and 2.

Included in the documentation is the maximum differential pressure expected during opening and closing of the valves for both normal and abnormal events, to the extent that these valve operations and events are present in the existing, approved design basis for Waterford 3.

Using the results of Attachments 1 and 2, LP&L intends to do the following:

1.

Switch settings (e.g. torque, torque bypass, etc.) for valve opening and closing will be reviewed and revised, as necessary.

This activity will include a program to review the methods for selecting and setting the switches. During the course of this project, should it be determined that a valve is inoperable, LP&L will document appropriate justification for continued operation in accordance with the applicable technical specification.

2.

Individual valve settings will be changed, as appropriate, to those established in item 1, above. Each valve identified in Attachments 1 and 2 will be demonstrated to be operable either by:

a.

testing the valve at the maximum differential pressure established in Attachments 1 and 2 (with the exception of testing a valve under simulated line break conditions), or

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"AN EQUAL OPPORTUNITY EMPLOYFR"

R.D. Martin W3P86-0076 Page 2 b.

providing appropriate justification for cases where testing under maximum differential pressure is not practicable.

Each valve will be stroke tested, to the extent practical, to verify that the switch settings have been properly implemented.

3.

Procedures will be revised, or newly prepared as necessary, to ensure that correct switch settings are determined and maintained. Applicable industry recommendations will be considered.

Items 1-3 above will be completed by November 15, 1987.

It is anticipated that valve testing will be primarily conducted during the first refueling i

outage for Waterford 3, tentatively scheduled to begin in December, 1986.

A written report will be provided to the NRC discussing the items suggested in IE Bulletin 85-03 following completion of the program.

As requested in IE Bulletin 85-03, this transmittal is made under affidavit pursuant to Section 182a, Atomic Energy Act of 1954, as amended. Should you have any questions or comments on this matter please contact Mike Meisner at (504) 595-2832.

Yours ve t ruly, i

K.

Cook Nuclear Support & Licensing Manager KWC/MJM/ssf Attachments cc: NRC Document Control Desk, Washington, D.C.

(original)

NRC, Director, Office of II8E l

G.W. Knighton, NRC-NRR J.H. Wilson, NRC-NRR NRC Resident Inspectors Office i

l B.W. Churchill W.M. Stevenson

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the matter of

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Louisiana Power & Light Company

) Docket No. 50-382 Waterford 3 Steam Electric Station

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AFFIDAVIT R.M. Nelson, being duly sworn, hereby deposes and says that he is Licensing Manager of Louisiana Power & Light Company; that he is duly authorized to sign and act on behalf of K.W. Cook, Nuclear Support & Licensing Manager and file with the Nuclear Regulatory Commission the attached responses to IE Bulletin 85-03; that he is familiar with the content thereof; and that the matters set forth therein are true and correct to the best of his knowledge, information and belief.

Li @ censing Manager - Nuclear R.

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STATE OF LOUISIANA

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Subscribed and sworn to before me, a Notary Public and for the Parish and State above named this

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9 Rotary Publil' My Commission expires

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W3P86-0076 Sheet 1 of 3 Waterford Unit 3 Design Bases for Operation of HPSIS MOVs Postulated Postulated Valve Upstream Downstream Valve Function Tag. No Condition Condition Notes Manually open from SI-121B High: HPSI pump Low: RWSP @ atmos. - HPSIP mini flow Control Room SI-120B shutoff-head pressure recirculation SI-120A line isolation SI-121A valves Manually Close SI-121B High: HPSI Pump Low: RWSP @ atmos.

on RAS SI-120B shutoff head pressure SI-120A SI-121A Manually open from SI-506A High

HPSI pump Low: RCS at atmos.

Hot Leg Injection Control Room SI-506B shutoff head pressure isolation valves SI-502A SI-502B Manually close SI-506A High:HPSI pump Low: RCS at atmos.

F-11owing Hot Leg SI-506B shutoff head pressure ction SI-502A SI-502B Minually open from SI-219A High: HPSI pump Low: RCS at atmos. -

HPSI Train Control Room SI-219B shutoff head pressure Isolation Valves Minus11y close SI-219A High: HPSI pump Low: RCS at atmos.

from Control Room SI-219B shutoff head pressure to divart fraction of HPSI flow to hot leg Opsn on SIAS SI-225B High: HPSI pump Low: RCS at atmos.

HPSI isolation SI-225A shutoff head pressure valves SI-226B SI-226A SI-227B SI-227A SI-228B i

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! Manually close SI-225B High: HPSI pump Low: RCS at atmos.

fro 2 Control SI-225A shutoff head pressure

, Roan following SI-226B i Shutdown or due SI-226A 7tdvertent SI-227B ing SI-227A SI-228B SI-228A i

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W3P86-0076 Sheet 2 of 3 Waterford Unit 3 High Pressure Safety Injection System Design Basis Maximum Differential Pressure Valve Design Basis Design Basis Function Valve ID d P Open (psi) d P Close (psi)

HPSIP Mini Flow SI-121B 1502 1502 Recirculation Line SI-120B 1502 1502 Isolation HPSIP Mini Flow SI-120A 1502 1502 Recirculation SI-121A 1502-1502 Hot Leg Injection SI-506A 1502 1502 Isolation SI-506B 1502 1502 SI-502A 1502 1502 SI-502B 1502 1502 HPSI Train SI-219A 1499 1499 1

Isolation SI-219B 1498 1498 HPSI Isolation SI-2253 1502 1502 SI-225A 1502 1502 SI-226B 1502 1502 SI-226A 1502 1502 SI-227B 1502 1502 SI-227A 1502 1502 SI-228B 1502 1502 SI-228A 1502 1502

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Waterford Unit 3 Design Bases for Operation of EFWS MOVs Postulated Postulated Valve Upstream Downstream Valva Function Tag. No Condition Condition Notes Manually open MS401A High: main steam Low: atmos. press.

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NUCLEAR REGULATORY COMMISSION (i

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OFFICE OF INSPECTION AND ENFORCEMENT Washington, D.C. 20555 INSPECTION AND ENFORCEMENT MANUAL 1

DEPER TEMPORARY INSTRUCTION 2515/73 INSPECTION REQUIREMENTS FOR IE BULLETIN 85-03,

" MOTOR-0PERATED VALVE COMMON MODE FAILURES DURING PLANT TRANSIENTS DUE TO IMPROPER SWITCH SETTINGS" 2515/73-01 PURPOSE To provide guidance for performing the inspection followup of the licensees' activities taken in response to IE Bulletin 85-03,

" Motor-0perated Valve Common Mode Failures During Plant Transients Due to Improper Switch Settings."

2515/73-02 OBJECTIVE To ensure a minimum level of consistency in the inspection followup of the licensees' activities taken in response to IE Bulletin 85-03.

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2515/73-03 BACKGROUND On June 9, 1985, the Davis-Besse Plant experienced a complete loss of main and auxiliary feedwater which was caused in part by motor operated valve failures.

This event has been previously described in IE Information Notice 85-50, " Complete Loss of Main and Auxiliary Feedwater at a PWR Designed by Babcock & Wilcox," dated July 8, 1985, and NUREG-1154, " Loss of Main and Auxiliary Feedwater Event at the Davis-Besse Plant on June-9, 1985," dated July 1985.

As a result, it will not be described in detail in this TI.

To ensure that the switch settings on certain safety related motor-operated valves are set and maintained so as to accommodate the most severe loading expected during design basis events, IE Bulletin 85-03 was issued November 15, 1985.

This IE bulletin also briefly described (1) a recent valve failure at Sequoyah Nuclear Plant Unit 2 which was caused by an improperly set open limit switch, (2) the results of a recent study on the effective-ness of signature tracing techniques in determining the operational readi-ness of motor-operated valves, and (3) a summary of the previously issued generic communications pertaining to problems associated with motor operated valve switches.

A copy of IE Bulletin 85-03 is enclosed as

[

Appendix I.

2515/73-04 BASIC REQUIREMENTS For each facility with an operating license or construction permit:

Issue Date:

02/21/86

5 INSPECTION REQUIREMENTS FOR IE BULLETIN 85-03 2515/73-04.04 04.04 Monitoring of the setting of the switches on the selected sam-ple valves can also provide insight into the effectiveness of the licensee's maintenance program.

Some aspects to be given attention are:

a.

Are there signs of rust and/or moisture within the operator housing?

b.

Are the torque and limit switch contacts set, aligned, clean, and free of oxidation, corrosion, and pitting?

c.

Is the valve stem prop.erly lubricated?

d.

If possible, determine if the packing is of the proper

type, installation, and not excessively aged (damaged, feeling discolored).

e.

If possible, verify that the licensee determines if the valve operates properly with the manual handwheel.

f.

Has the licensee been faithful in following the past procedure?

For example, were as-found switch settings in accordance with the values specified in the previous procedure?

If not, the licensee's maintenance policy on other valves may be questionable.

04.05 IE, with NRR support, is responsible for review of the final

-response (IE Bulletin 85-03, item f) to the bulletin.

Any subsequent correspondence on this item will be directed thru the region to the licensee.

The completion of the review will be documented as a feeder report so that the appropriate region and resident can incorporate this information into an inspection report.

2515/73-05 REPORTING REQUIREMENTS Document the activities associated with this TI in the routine inspection reports as they are completed.

A copy of the inspection reports should be sent to Robert Baer, Chief, Engineering and Generic Communications Branch.

In addition to closure, the following milestones should also be reported:

05.01 Completion of review of the licensee's initial response to IE Bulletin 85-03 (based on IE HQ feeder reports to regions).

05.02 Completion of the licensees' tests and corrective actions in accordance with the approved program (based on Region's inspection).

05.03 Compile results of completed actions and issue status reports on an approximately semiannual schedule beginning mid-November 1986 (IE HQ contact responsibility).

! Issue Date:

02/21/86

SSINS No.:

6820 an.

OMB No.:

3150-0011 Expiration Date:

08/01/88 IEB 85-03 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, DC 20555 November 15, 1985 IE BULLETIN NO. 85-03:

MOTOR-OPERATED VALVE COMMON MODE FAILURES DURING PLANT TRANSIENTS DUE TO IMPROPER SWITCH SETTINGS Addressees:

All holders of nuclear power reactor operating licenses (OLs) or construction permits (cps) for action.

Purpose:

The purpose of this bulletin is to request licensees to develop and implement a program to ensure that switch settings on certain safety-related motor-operated valves are selected, set and maintained correctly to accommodate the maximum differential pressures expected on these valves during both normal and abnormal events within the design basis.

Description of Circumstances:

There have been two recent events, and a number of earlier events, during which motor-operated valves failed on demand, in a common mode, due to improper switch settings.

Event 1, Davis-Besse Plant - On June 9, 1985, the Davis-Besse Plant experienced a complete loss of main and auxiliary feedwater.

This event was described previously in IE Information Notice No. 85-50, " Complete Loss of Main and Auxiliary Feedwater at a PWR Designed by Babcock & Wilcox," and in NUREG-1154,

" Loss of Main and Auxiliary Feedwater Event at the Davis-Besse Plant on June 9, 1985." Normally open, Limitorque motor-operated auxiliary feedwater (AFW) gate valves failed to reopen on either an automatic or manual signal from the main control room after they were inadvertently closed during the event. While other failures also occurred in the AFW system, the failure of these two valves was itself enough to prevent AFW from reaching either steam generator.

During the recovery from this event, the valves were opened with the handwheels.

The results of licensee troubleshooting activities after the event led to the conclusion that the setting for the torque switch bypass limit (torque bypass) switch in each valve's control circuit had not been set to remain closed long enough to provide the necessary bypass function on vaive opening with differen-tial pressure conditions across the valve.

During the event, the valves experienced a high differential pressure after closing.

The torque bypass AC.11 12 5 Q.),[-

IEB 85-03 November 15, 1985 Page 2 of 6 switch on both valves was improperly set, causing the torque switches to become operable prematurely.

This condition stopped valve travel before the valve discs were fully off their seats.

The torque bypass switches were set to drop cut after the valves opened to 5%

full-stroke.

The 5% full-stroke setting was based on a number of handwheel turns.

In a 10 CFR 21 report, submitted subsequent to the event, Toledo Edison Company identified two reasons why the torque bypass switch settings were not adequate:

(1) the 5% full-stroke settings were not adequate for unseating the AFW system discharge valves with large differential pressures across the valves and (2) the procedure for setting this switch was defective in that the 5%

full-stroke was not specified to be in addition to the handwheel turns required to take up the motor-operator coast and backlash.

The torque bypass switch setting errors were revealed only when high differential pressure conditions across the valves caused higher loadings. The valve failures were reproduced during tests performed by the licensee with differential pressures applied across the valves.

During the tests, the valves operated properly when low differential pressures were applied across them, but failed to open when high differential pressures were applied.

The valves were instrumented during these tests to obtain signature traces of critical parameters.

Event 2, Sequoyah Plant Unit 2 - An event involving partial loss of main feedwater occurred on May 2, 1985, at Sequoyah Nuclear Plant Unit 2 while in Mode 2 and returning to power after a reactor trip.

Feedwater was being supplied through the main feedwater (MFW) system isolation valve bypass lines.

Operators attempted to open the MFW system isolation valves to supply water to the steam generators; however, two of the four MFW isolation valves would not open.

The startup was discontinued and the unit was returned to hot shutdown.

l During examination to determine the reason for the valve failures, the licensee discovered that both valve stems had sheared from their discs.

The discs were found in the closed positions within the valve seats. The stems had suffered 1' '

fracture failures through approximately three quarters of the diameter of the shafts, in addition to stress failures of the remaining quarter.

The Limitorque motor-operators on the valves use limit switches to control valve motion in the open direction. These MFW system isolation valves are large (18 inch diameter),

fast acting (154 inches per minute travel speed) valves.

Because of the high speed of these valves and the large mass of the discs, the selection of the limit switch setpoint needs to account for the large momentum of the disc and its continued motion after the limit switch deenergizes the valve motor-operator.

The set point was not correctly established and the disc impacted the backseat during opening.

The failure mechanism of these valves was identified by the licensee to be impact loading of the stem on the opening stroke as a result of the disc impacting the backseat, combined with a stress failure of the remaining portion of the stem on the opening stroke. Main feedwater valves are not included in the actions requested by this bulletin.

The NRC is, however, continuing to evaluate the Sequoyah event.

NRC Field Evaluation - As a part of the resolution of Generic Issue II.E.6, "In-Situ Testing of Valves," the NRC contracted with the Oak Ridge National Laboratory in 1984 to perform a limited study to determine the effectiveness of signature tracing techniques in determining the operational readiness of

IEB 85-03 November 15, 1985 Page 3 of 6 safety related motor-operated valves.

It was hoped also that this study could provide some insight as to current conditions of valve switch settings at nuclear power plants.

Signature traces of motor current, torque and limit switch actuations and axial motion of the worm gear (an indication of operator torque) were obtained from 36 motor-operated valves at 4 nuclear plant sites.

Although the formal technical report [NUREG/CR-4380 " Evaluation of the Motor-Operated Valve Analysis and Test System (M0 VATS) to Detect Degradation, Incorrect Adjustments, and Other Abnormalities in Motor-0perated Valves"] has not been issued, the current draft of the report indicates that (1) this inspection method can be used to improve current ASME methods and (2) there were abnormalities with nearly every valve tested.

Table 1 contains a summary of the study's findings with respect to switch setting abnormalities.

Of particular interest with respect to the events described above is the finding that 75% of the valves had improperly set torque bypass switches (56% of the valves had the close-to-open torque bypass switch set so that it was opening before the valve fully unseated) and 8% of the valves were unintentionally backseating.

The abnormalities in Table 1 have not been fully evaluated at this time, and they should not be interpreted to mean that any abnormality resulted in an inoperable valve.

Background:

The NRC has previously identified common mode failures, on demand, of valves.

IE Circular No. 77-01, " Malfunction of Limitorque Valve Operators," reported that on October 28, 1976, two motor-operated (Limitorque) valves located between the refueling water storage tank and the charging pump suction at the Trojan Nuclear Plant failed to open in response to a spurious safety injection (SI) signal.

The malfunction in both valves resulted from the torque switch in the opening circuit becoming activated before the valves were fully off their seats.

The valves also were equipped with a torque bypass switch.

Each of the valves that malfunctioned was found to have its torque bypass switch adjusted such that it allowed the torque switch to be operable in the circuit before the valve was moved from its seat.

The licensee's investigation revealed that in each case the valve had been manually closed hard on its seat following a maintenance operation.

Examination by the licensee revealed similar improper j

adjustments of the torque bypass switches on several other motor-operated i

valves in safety-related systems.

IE Information Notice No. 81-31, " Failure of Safety Injection Valves To Operate Against Differential Pressure," reported on September 3, 1981, that both trains i

j of the San Onofre Unit 1 safety injection system were found to be inoperable when challenged to operate against differential pressure.

Improperly set 4

switches were the principal cause of these failures.

There were no adverse consequences in this particular event because there was no accident that required safety injection.

The reactor pressure remained above the safety injection pump's shutoff head; therefore, no actual injection of water would have occurred if the valves had opened.

However, had reactor pressure decreased and actual injection been required, injection flow would not have been automa-tically available as designed.

These valves had been regularly tested at each l

IEB 85-03 November 15, 1985 Page 4 of 6 refueling outage, but the tests were not required to be performed with differential pressure across these valves.

Florida Power Corporation reported an event at Crystal River Unit 3 in LER 77-9.

During plant cooldown with the unit in hot shutdown, decay heat removal valves in the decay heat removal pump suction would not open with remote actuation.

These failures were caused by pressure acting on the gate valve discs.

The valves were opened manually with the handwheels.

The torque switches were reset.

In addition to common mode valve failures on demand, there have been numerous common mode failures discovered during testing or as a result of investigating a single failure.

NUREG/CR-2270, " Common Cause Fault Rates for Valves,"

February 1983, contains reports of 99 common cause valve fault events from 1976 through 1980.

The NRC has previously identified other problems with motor-operated valve switches in Bulletin No. 72-3, "Limitorque Valve Operator Failures"; IE Information Notice No. 79-03, "Limitorque Valve Geared Limit Switch Lubricant";

Circular No. 81-13, " Torque Switch Electrical Bypass Circuit for Safeguard Service Valve Motors"; Information Notice No. 82-10 "Following Up Symptomatic Repairs To Assure Resolution of the Problem"; and Information Notice No.84-10,

" Motor-operated Valve Torque Switches Set Below the Manufacturer's Recommended Value."

The failure and potential failure of Westinghouse Electro-Mechanical Division motor-operated gate valves to close are discussed in IE Bulletin No. 81-02 and IE Bulletin No. 81-02, Supplement 1, " Failure of Gate Type Valves To Close Against Differential Pressure."

Copies of the above referenced NRC bulletins, circulars and information notices can be obtained from your local public document room.

Actions for All Holders of Operating Licenses or Construction Permits:

For motor-operated valves in the high pressure coolant injection / core spray and emergency feedwater systems (RCIC for BWRs) that are required to be tested for operational readiness in accordance with 10 CFR 50.55a(g), develop and implement a program to ensure that valve operator switches are selected, set and maintained properly.

This should include the following components:

a.

Review and document the design basis for the operation of each valve.

This documentation should include the maximum differential pressure expected during both opening and closing the valve for both normal and abnormal events to the extent that these valve operations and events are included in the existing, approved design basis, (i.e., the design basis documented in pertinent licensee submittals such as FSAR analyses and fully-approved operating and emergency procedures, etc).

When determining the maximum differential pressure, those single equipment failures and inadvertent equipment operations (such as inadvertent valve closures or openings) that are within the plant design basis should be assumed.

IEB 85-03 November 15, 1985 Page 5 of 6 b.

Using the results from item a above, establish the correct switch settings.

This shall include a program to review and revise, as necessary, the methods for selecting and setting all switches (i.e., torque, torque bypass, position limit, overload) for each valve operation (opening and closing).

If the licensee determines that a valve is inoperable, the licensee shall also make an appropriate justification for continued operation in accordance with the applicable technical specification.

Individual valve settings shall be changed, as appropriate, to those c.

established in item b, above. Whether the valve setting is changed or not, the valve will be demonstrated to be operable by testing the valve at the maximum differential pressure determined in item a above with the excep-tion that testing motor-operated valves under conditions simulating a break in the line containing the valve is not required.

Otherwise, justi-fication should be provided for any cases where testing with the maximum differential pressure cannot practicably be performed.

This justification should include the alternative to maximum differential pressure testing which will be used to verify the correct settings.

Note: This bulletin is not intended to establish a requirement for valve testing for the condition simulating a break in the line containing the valve.

However, to the extent that such valve operation is relied upon in the design basis, a break in the line containing the valve should be considered in the analyses prescribed in items a and b above.

The resulting switch settings for pipe break conditions should be verified, to the extent practical, by the same methods that would be used to verify other settings (if any) that are not tested at the maximum differential pressure.

Each valve shall be stroke tested, to the extent practical, to verify that the settings defined in item b above have been properly implemented even if testing with differential pressure can not be performed.

d.

Prepare or revise procedures to ensure that correct switch settings are determined and maintained throughout the life of the plant.* Ensure that applicable industry recommendations are considered in the preparation of these procedures.

e.

Within 180 days of the date of this bulletin, submit a written report to the NRC that:

(1) reports the results of item a and (2) contains the program to accomplish items b through d above including a schedule for completion of these items.

• This item is intended to be completely consistent with action item 3.2, " Post-Maintenance Testing (All Other Safety-Related Components)," of Generic Letter 83-28, " Required Actions Based on Generic Implications of Salem ATWS Events."

These procedures should include provisions to monitor valve performance to ensure the switch settings are correct.

This is particularly important if the torque or torque bypass switch setting has been significantly raised above that required.

IEB 85-03 November 15, 1985 Page 6 of 6 1.

For plants with an OL, the schedule shall ensure that these items are completed as soon as practical and within two years from the date of this bulletin.

2.

For plants with a CP, this schedule shall ensure that these items are completed before the scheduled date for OL issuance or within two years from the date of this bulletin, whichever is later.

f.

Provide a written report on completion of the above program.

This report should provide (1) a verification of completion of the requested program, (2) a summary of the findings as to valve operability prior to any adjust-ments as a result of this bulletin, and (3) a summary of data in accordance, with Table 2, Suggested Data Summary Format.

The NRC staff intends to use -

this data to assist in the resolution of Generic Issue II.E.6.1.

This report shall be submitted to the NRC within 60 days of completion of the program.

Table 2 should be expanded, if appropriate, to include a summary of all data required to evaluate the response to this bulletin.

The written reports shall be submitted to the appropriate Regional Administrator under oath or affirmation under provisions of Section 182a, Atomic Energy Act of 1954, as amended.

Also, the original copy of the cover letters and a copy of the reports shall be transmitted to the U.S. Nuclear Regulatory Commis'sion, Document Control Desk, Washington, DC 20555 for reproduction and distribution.

This request for information was approved by the Office of Management and Budget under a blanket clearance number 3150-0011.

Comments on burden and duplication may be directed to the Office of Management and Budget, Reports Management, Room 3208, New Executive Office Building, Washington, DC 20503.

Although no specific request or requirement is intended, the time required to complete each action item above would be helpful to the NRC in evaluating the cost of this bulletin.

If you have any questions regarding this matter, please contact the Regional Administrator of the appropriate NRC regional office or the technical contact listed below.

James M. Taylor, Director Office of Inspection and Enforcement Technical Contacts: H. A. Bailey, IE R. J. Kiessel, IE (301) 492-9006 (301) 492-8119 Attachments:

1.

Table 1 2.

Table 2 3.

List of Recently Issued IE Bulletins

Attachmtnt 1 IEB 85-03 November 15, 1985 TABLE 1 Summary of Significant MOV Abnormalities Bypass switch improperly set 75*

Incorrect thrust 50 Unbalanced torque switch 33 Valve backseating 8

High motor current 3

Torque switch abnormalities 2

Miscellaneous abnormalities 33 l

• Percent of valves experiencing abnormality.

The total does not equal 100 percent as most valves had more than one abnormality.

1 t

I

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' Attachment 2 IEB 85-03 Page 1 of 2 November 15, 1985 l

l l

Table 2 i

Suggested Data Summary Format a

Switch Settings Final Switch i

Design Prior to Adjustments Settings in Re-I Valve Basis AP Test AP as a Result of Bulletin sponse to Bulletin Valve Valve Operator Function Open/Close Open/Close Open/Close Open/Close

}

Component ID, Manufacturer, (e.g.

i Manufacturer, Model, AFW pump i

Type, Model, Motor RPM, discharge i

Size, Rating Output Speed isolation (RPM) valve) 1 4

I I

4 I

I J

1

Attichment 2 IEB 85-03 Page 2 of 2 November 15, 1985 Table 2 (Continued)

Suggested Data Summary Format Valve Component As Found Valve ID Operability Test Method Description / Justification (e.g. justification should include the values of parameters or data that demonstrate that the valve will operate properly at its design basis)

Attachm:nt 3 IEB 85-03 November 15, 1985 LIST OF RECENTLY ISSUED IE BULLETINS Bulletin Date of No.

Subject Issue Issued to 85-02 Undervoltage Trip Attachments 11/5/85 All power reactor Of Westinghouse 08-50 Type licensees and Reactor Trip Breakers applicants 85-01 Steam Binding Of Auxiliary 10/29/85 All power reactor Feedwater Pumps facility licensees and CP holders listed in Attachment 1 for action; all other power reactor facilities for information 84-03 Refueling Cavity Water Seal 8/24/84 All power reactor facilities holding an OL or CP except Fort St. Vrain 84-02 Failures of General Electric 3/12/84 All power reactor Type HFA Relays In Use In facilities holding Class 1E Safety System an OL or CP 84-01 Cracks In Boiling Water 2/3/84 All BWR facilities Reactor Mark I Containment with Mark I contain-Vent Headers ment and currently in cold shutdown with an OL for Action and All other BWRs with an OL or CP for information 83-08 Electrical Circuit Breakers 12/28/83 All power reactor With An Undervoltage Trip facilities holding Feature In Use In Safety-an OL or CP Related Applications Other Than The Reactor Trip System 83-07 Apparently Fraudulent 12/09/83 Same as IEB 83-07 Sup. 2 Products Sold By Ray Miller, Inc.

OL = Operating License CP = Construction Permit

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