ML20212Q001

Insp Rept 50-382/86-15 During Jul 1986.Violations Noted: Failure to Follow Procedures for Calibr of Measuring & Test Equipment & to Make Prompt Notification.Unresolved Items Noted:Low Temp Overpressurization Protection

ML20212Q001
Person / Time
Site:	Waterford
Issue date:	08/26/1986
From:	Bundy H, Constable G, Luehman J, Staker T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20212P990	List: ML20212P988 ML20212P999 ML20212Q001
References
50-382-86-15, IEIN-86-003, IEIN-86-3, NUDOCS 8609030362
Download: ML20212Q001 (13)

See also: IR 05000382/1986015

Text

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APPENDIX B

U. S. NUCLEAR REGULATORY COMMISSION

REGION IV

NRC Inspection Report: 50-382/86-15 License: NPF-38

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Docket: 50-382 ,

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Licensee: Louisiana Powor & Light Company (LP&L)

317 Baronne Street -

P. O. Box 60340

New Orleans, Louisiana 70160 --

Facility Name: Waterford Steam Electric Station, Unit 3 (W3 SES)

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Inspection At: Taft, Louisiana

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Inspection Conducted: July 1-31, 1986

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Inspectors: _ 72( [ *

. E G. W Senior Resident Inspector . ' DatE

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T. T. Stater, Resident Inspector Date '

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H. F. Bundy, Project Inspector, Project

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Section C, Reactor Projects Branch

(Paragraphs 7, 10, 11, and 13)

Approved: 7

G.'%-GenstaMe7 Chief, Project Section C, Date

Reactor Projects Branch

8609030362 860326

PDR ADOCK 05000382

G PDR

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Inspection Summary

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Inspection Conducted' July 1-31, 1986 (Report 50-382/86-15)

Areas Inspected: , Routine, unahnounced inspection of: (1) Plant Status,

(2) Licensee Event Report (LER) Followup, (3) Monthly Surveillance, (4) Monthly

Maintenance, (5) Licensee Followup on Previously Identified Items, (6) ESF

Walkdown, (7) Offsite Review Committee, (8) Onsite Review Committee,

(9) Routine Inspection, (10) Low Temperature Overpressurization Protection,

'(11) Potential Generic Problems, and (12) Review of Special Reports.

Results: Within the areas inspected, the following violations were identified:

Paragraph 6 - Failure to follow procedures for calibration of M&TE

Paragraph 10 -' Failure to make prompt notification

In addition, two unresolved items were identified in paragraphs 12 and 13.

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DETAILS

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1. Persons Contacted

Principal Licensee Employees

G. W. Muench, Director, Nuclear Operations

• R. P. Barkhurst, Plant Manager, Nuclear

T. F. Gerrets,' Corporate QA Manager

S. A. Alleman, Assistant Plant Manager, Plant Technical Services

N..S. Carns, Assistant Plant Manager, Nuclear, Operations and Maintenance

J. N. Woods, QC Manager

A. S. Lockhart, Site Quality Manager

R. F. Burski, Engineering and Nuclear Safety Manager

K. L. Brewster, Onsite Licensing Engineer

G. E. Wuller, Onsite Licensing Coordinator

T. H. Smith, Maintenance Superintendent, Nuclear

P. V. Prasankumar, Technical Support Superintendent

• Present at exit interviews.

In addition to the above personnel, the NRC inspectors held discussions

with various operations, engineering, technical support, maintenance, and

administrative members of the licensee's staff.

- 2. Unresolved Items

An unresolved item is a matter about which more information is required to

determine whether it is acceptable or may involve' a violation or

deviation. Unresolved Item 8602-01 is discuss'ed in paragraph 10 of this

report. .

Two new unresolved items were identified during this. inspection period and.

are discussed in paragraphs-12 and 13.

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3. Plant Status --

The inspection period began with the licensee progressing toward a. reactor. -

startup. At 8:59 a.m. (CDT) on July 1,1986, the reactor was ytaken

critical. On July 8, 1986, because of concerns with the 2B Reactor

Coolant Pump (RCP) seal package, the licensee ordered a plant' shutdown to

replace both the 2B and 2A RCP seal packages. With the' plant.in Mode 5 on

July 10, 1986, the IB and 2B Safety Injection Tanks (SITS) were mistakenly

injected into the reactor coolant system (RCS) when their outlet valves

were inadvertently actuated during maintenance.

On July 12, 1986, shortly before noon, the plant entered Mode 4 and

subsequently reached Mode 3 that evening. Because of increasing

temperatures on the 2A RCP the plant was again placed in Mode 5 on .

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July 13,1986. The plant experienced a loss of shutdown cooling on

July 14,1986. With the RCS partially drained, both low pressure safety

injection (LPSI) pumps lost suction. At 3:17 a.m. the B LPSI pump began

cavitating and was secured. Then, the A LPSI pump was started and also

began to. cavitate. At 5:35 a.m., with suction head restored and the pumps

vented, shutdown cooling was restored. The increase in loop temperature

during this period was negligible. The apparent cause of the loss of

LPSI. pump suction was a problem with the temporary reactor vessel level

~ indication. The licensee plans to submit licensee event reports (LERs) on

both the inadvertent SIT injection and the loss of shutdown cooling.

With the plant in Mode 3 the licensee declared a notification of unusual

event ,in accordance with the plant emergency plan at 5:45 a.m. on July 16,

1986, based on the fact 'that reactor coolant system unidentified leakage

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exceeded the 1.0 gpm specification of TS 3.4.5.2. After adjusting various

valves, the control rcom personnel verified that unidentified leakage had

been reduced to below 1.0 gpm (.75 gpm). At 10:11 a.m. on July 16, 1986,

the licensee exited the ACTION of TS 3.4.5.2 and terminated the unusual

event. Reactor criticality was again achieved at 5:30 p.m. on July 16,

1986, and Mode 1 was entered at 7:10 p.m. on that date.

The inadvertent SIT injections, noted above, were not promptly reported to

the NRC in accordance with 10 CFR Part 50.72 and are further discussed in

paragraph 10 below.

No other violations cr deviations were identified.

4. Licensee Event Report (LER) Followup

The following LERs were reviewed and closed. The NRC inspectors verified

that reporting requirements had been met, that causes had been identified,

that corrective actions appeared appropriate, that generic applicability

had been considered, and that the LER forms were complete. Additionally,

the NRC inspectors confirmed that no unreviewed safety questions were

involved and that violations of regulations or Technical

Specification (TS) conditions had been identified.

(Closed) LER 382/86-01 - Reactor Trip on Dropped Control Element Assembly.

The NRC inspector has reviewed the event and the report and has no

questions; only one comment about the report itself. The event was

attributed to a faulty automatic control element assembly calculator

timing module, yet no anomalies could be found within the card. In such a

case, either Block 13 cf the LER should be filled in or Block 14 should be

checked "yes" so that the actual cause of the problem is reported when

known for certain.

(Closed) LER 382/86-10 - Improperly Filed Scheduling Cards Results in

Exceeding the Surveillance Frequency for Secondary Activity Sample.

( No violations or deviations were identified.

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5.' Monthly Surveillance

The NRC inspectors observed / reviewed TS required testing and verified that

testing was performed in accordance with adequate procedures, that test

instrumentation was calibrated, that limiting conditions for

operation (LCO) were met, and that any deficiencies identified were

properly reviewed and resolved.

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Procedures witnessed, verified, or reviewed during this inspection period '

included OP-903-001, Revision 3, " Technical Specification Surveillance

Logs," Attachment 10.6, "Seven Day Surveillances," for the week of

Ju'y 28, 1986 and OP-903-053, Revision 3, " Fire Protection System Pump

Operability Test," which was performed July 31, 1986.

No violations or deviations were identified.

6. Monthly Maintenance

Station activities affecting safety-related systems and components were

observed / reviewed to ascertain that the activities were conducted in

accordance with approved procedures, regulatory guides and industry codes

or standards, and in conformance with TS.

During a routine check of M&TE in use and available for use in the plant

on July 2 and 3, 1986, the NRC inspectors identified a number of problems.

On July 2, 1986, the NRC inspectors noted MIPT 091.016 (calibration due

date June 19, 1986) about to be used in the B High Voltage Switchgear

Room. Discussion with the technician indicated he had been issued the

equipment just prior to coming to the switchgear room. The NRC inspectors

informed him the M&TE calibration period had expired and the technician

indicated that he would return it for calibration. While touring the

control room on July 2, 1986, M&TE MEET 025.037 (calibration due date

June 19, 1986) and MIES 020.007 (calibration due date June 5, 1986) were

noted by the NRC inspectors as being available for use. The control room

supervisor was advised of the past due equipment and he said that the

maintenance department would be informed.

Expired calibration M&TE in the plant is an apparent violation and is

identified as 382/8615-01.

NRC Inspection Report 382/85-33 documents a violation which resulted from

a failure to properly control M&TE and a further example is contained in

NRC Inspection Report 382/86-05. The licensee responded to

Violation 382/8633-02 in a letter dated February 21, 1986, and due in part

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to the violation response, Procedure MD-1-015, " Measuring and Test

Equipment Control," was revised.

The fact that M&TE remained in the plant past its due date indicates

weaknesses in the M&TE recall program described in paragraph 5.11 of

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MD-1-015, Revision 1. The issuance of M&TE which was past due indicates a

weakness in the implementation of MD-1-012, Revision 0, " Tool Control

Procedure."

The NRC inspector observed portions of the troubleshooting and repair of

the electronics for control element assembly (CEA 32)'under Condition

Identification Work Authorization (CIWA) 027604. This work was performed

on July 17, 1986, and the NRC inspector verified that all measuring and

test equipment used was properly calibrated in addition to verifying

proper procedures were used.

No other violations or deviations were observed.

7. ESF System Walkdown

The high pressure safety injection (HPSI) system was verified operable by

performing a walkdown of the accessible and essential portions of the

system on July 24 and 25, 1986.

The NRC inspector used the standby system valve lineups specified on

Attachment 8.1 and 8.2 and the breaker lineup specified on Attachment 8.3

of Procedure OP-9-008, Revision 5, in conjunction with the referenced

drawings. In addition, the NRC inspector reviewed the monthly lineup

check (0P-903-026) last performed on HPSI.

In performing the walkdown, both the NRC inspector and a licensee operator

incorrectly identified the positions of several main flow path valves. It

turned out that these valves had reverse acting stem travel. Prior to

completion of the inspection, the control room supervisor informed the NRC

inspector that appropriate operator aids had been placed to alert

personnel to the correct stem-to-valve position correlation for these

valves.

On completion of the walkdown, the NRC inspector advised the licensee of

the following observations:

a. Valves SI-113A and SI-113B are required to be closed on

Attachment 8.1, but are shown open on Drawing G167, Sheet 1. They

were found open during walkdown and exception was taken to the closed

position in the documentation generated for the lineup verification

performed on January 11, 1985.

b. There were numerous errors in both drawing sheet number and grid

location references for valves in Attachment 8.1.

c. Drawing G167, Sheet 1, erroneously depicts Valve SI-205B as a check

valve instead of a stop check valve,

d. Valves SI-1181A and SI-1182A were incorrectly identified on

Attachment 8.1 as SI-1181B and SI-11828. 'Although no exception was

taken to these mistakes in the documentation for the lineup

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verification performed on January 11, 1985, because of equipment

configuration it appears highly probable that the positions of these

valves were actually verified.

e. Valve SI-135A was erroneously indicated as being in the reactor

containment building instead of the reactor auxiliary building on

Attachment 8.2.

f. The breaker for Valve Motor SI-415A was erroneously labeled SI-129A.

g. The breaker panel labeling was incomplete for Valve Motors SI-120A,

SI-125A, and SI-135A.

h. The noun descript ons for numerous components in Attachment 8.3 were

erroneous.

i. Breakers SI-EBKR-60A-21-S and SI-EBKR-618-21-S for Valves SI-502A and

SI-506B motor space heaters were labeled as spares,

j. Breaker SI-EBKR-B1-DC-9-S was incorrectly identified in

Attachment 8.3 as SI-EBKR-1B-DC-9-S.

k. A number of breakers were not labeled with a component number.

1. The label for Breaker SI-EBKR-918-28-S indicated it would supply

control power to Valve Motor SI-506B as well as Valve Motor SI-302.

The component wiring diagram concurred with Attachment 8.3 which

stated that it supplied control power only for Valve Motor SI-302.

Although the NRC inspector identified a large number of nomenclature and

tagging errors, the HPSI system appeared to be correctly lined up and

verified. The operators resolved most identification errors on the spot.

As documented in NRC Inspectio1 Report (IR) 50-382/86-02, paragraph 8, and

updated as Unresolved Item 382/8606-01 in irs 50-382/86-06 and

50-382/86-11, the licensee has identified a program to upgrade all

safety-related checklists and properly tag all plant valves. The NRC

inspector was informed that this program also includes identification and

labeling of electrical breakers. No deficiencies directly impacting plant

safety have been identified. The upgrading program is scheduled to be

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completed by the end of the first refueling outage which is currently

scheduled to be completed by December 31, 1986. The upgrading program

completion and correction of the deficiencies identified in all four

reports will be inspected as a part of the followup to Unresolved

Item 382/8606-01.

No violations or deviations were identified.

8. Offsite Review Committee

During this inspection period the NRC inspectors began a review of the

i operations of the Safety Review Committee (SRC) against the requirements

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of TS 6.5.2. In the course of the review the NRC inspector examined the

minutes from SRC meetings 85-08 and 85-09 as well as 86-01 through 86-03.

To ensure the required nuclear operations quality assurance audits were

performed and documented under the oversight of the SRC in accordance with

TS 6.5.2.8 and 6.5.2.10, the NRC inspector reviewed the following audit

reports:

a. Plant Operations

SA-W3-QA-86-06 - Plant Operations

SA-W3-QA-86-18,- Inspection, Test and Operating Status

SA-W3-QA-86-10 - Technical Specification Administration

b. Training

SA-W3-QA-86-01 - Operator and STA Training and Qualification

SA-W3-QA-86-24 - Technical Training Qualification Administration

c. Environmental Monitoring

SA-W3-QA-86-22 - Environmental Monitoring (Non-rad)

No violations or deviations were identified. ,

9. Onsite Review Committee

During this inspection period the NRC inspectors-reviewed the operation of.

the Plant Operations Review Committee (PORC) against the requirements of~

TS 6.5.1. As part of the review the following PORT meeting minutes were

examined:

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a. 85-04

b. 86-11 through 26

c. 86-36 througa 46 -

d. 86-49 through 70

Based on the above reviews as well as reviews of selected plant operating '

logs, the NRC inspector had the following observations.

a. On May 5, 1986, Core Protection Calculator Channel B LO-DNBR and LPD

were placed in bypass. TS 6.5.1.6.K requires that the'PORC review

and document judgement concerning prolonged bypass operation. The

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PORC meeting minutes for meetings subsequent to the event do not

document the committee's judgement or even reference the bypass

operations.

b. On February 12 and 13, 1986, inadvertent releases of radioactivity

from the plant stack occurred while performing a test associated with

Station Modification (SM) 818. TS 6.5.1.6.m requires the PORC to

review such events and TS 6.5.1.8 requires documentation of the

review. As of July 31, 1986, Potential Reportable

Events (PRES) d6-29 and 86-31 had not been reviewed by the PORC.

c. Control Element Assembly (CEA) 1 dropped to the fully inserted

position with the plant at 40% power on January 8, 1986.

TS 6.5.1.6.g requires the PORC to review unit operations to detect

potential hazards to nuclear safety, yet no documentation of review

of this significant event exists in subsequent PORC meeting minutes.

In NRC Inspection Report 50-382/86-02 an unresolved item (382/8602-01) was

identified concerning the informal nature of the review and documentation

of occurrences that were not identified as PRES. It appears that the PRE

is a suitable document for identifying events reportable under

requirements such as 10 CFR Part 50.72 and 50.73 as well as TS. However,

the PORC has review responsibilities for a wider scope of events than just

those reportable to the NRC. Therefore, in order to ensure that the PORC

reviews all events falling within its responsibilities (such as

TS 6.5.1.6.g), either the scope of the applicability for the PRE needs to

be expanded or an additional documentation system appears to be necessary.

This problem will be inspected as part of Unresolved Item 382/8602-01.

The NRC inspector has discussed all of the above observations with the

regular PORC chairman.

On July 3,1986, the NRC inspector attended the regularly scheduled weekly

meeting of the PORC. A proper quorum of members was verified present and

other administrative requirements of the TS appeared to have been met.

One of the topics reviewed by the committee at this meeting was a number

of changes to TS that are to be made to support second cycle as well as

future core reload configurations. The changes were presented by licensee

licensing personnel and were not supported by any technical justifications

such as the results of revised accident analysis scenarios. A number of-

the PORC members raised questions about the fact that no supporting

documentation was provided. The licensing engineer making the

presentation explained that the analysis had not yet been done but the

changes were needed to complete the package for NRC review when the

analysis is ready. The PORC approved the changes; however, the minutes

were annotated to indicate that approval was contingent on satisfactory

analysis results. The NRC inspector has discussed the validity of

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conditional PORC approval of TS changes with licensee management and the

NRC licensing project manager. There was concurrence that conditional

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PORC approval was appropriate for these particular circumstances,~ but

would not normally be valid. ^

No violations or deviations were observed:

10. Routine Inspection

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By observation during the inspection period, the'NRC inspectors verified

that the control room manning requirements were;being met. In addition,

the NRC inspectors observed shift turnover to verify that continuity of

system status was maintained. The NRC inspectors periodically questioned

shift personnel relative to their awareness of the plant conditions.

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Through log review and plant tours, the NRC inspectors verified compliance

with selected TS and limiting conditions for operations.

During the cours"e of the inspection, observations relative to protected

and vital area security were made including access controls, boundary

integrity, search, escort, and badging.

On a regular basis, radiation work permits (RWPs) were reviewed and the

specific work activity was monitored to assure the activities were being

conducted per the RWPs. Selected radiation protection instruments were

periodically checked and equipment operability and calibration frequency

were verified.

The NRC inspectors kept informed on a daily basis of overall status of

plant and of any significant safety matter related to plant operations.

Discussions were held with plant management and various members of the

operations staff on a regular basis. Selected portions of operating logs

and data sheets were reviewed daily.

The NRC inspectors conducted various plant tours and made frequent visits

to the control room. Observations included: witnessing work activities

in progress; verifying the status of operating and standby safety systems

and equipment; confirming valve positions, instrument and recorder

readings, annunciator alarms; and housekeeping.

The NRC inspector discussed the inadvertent injection of the SITS,

discussed in paragraph 3 of this report, with the plant operations

superintendent. After the sequence of events was explained the NRC

inspector inquired about the reporting of the event. The operations

superintendent responded that the event had not been classified as a

notification under 10 CFR Part 50.72 but it was still under consideration

for a possible report under 10 CFR Part 50.73. The NRC inspector

re-reviewed the event, consulted NRC Region IV, and provided licensee

management with excerpts from NUREG 1022 and Supplements 1 and 2 to

NUREG 1022. After receiving the information, the licensee still felt that

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the event had been properly classified as not reportable under

10 CFR 50.72 because only the SIT isolation valves had opened and a full

ESF system logic actuation had not occurred. <

After conferring with the NRC Office for Analysis and Evaluation'of

Operational Data (AEOD), it was determined that a 10 CFR 50.72

notification should have been made. The actuation of the SIT isolation

valves is considered an actuation of an ESF and therefore a notification

under 50.72(b)(2)(ii) and an LER under 50.73(2)(iv)'are required.

The NRC inspector reviewed the licensee's Potentially' Reportable

Event (PRE)86-043 which discusses the discharge of liquid via an effluent

pathway with an inoperable radioactive liquid effluent ' monitor. The

licensee had concluded that according to TS 3.3.3.10 the event described-

in this PRE is only reportable in the next semiannual radioactive effluent -

release report. However, in this case the PRE should also be the. basis

for an LER under 10 CFR 50.73a(2)(i)(B), a condition prohibited by plant:

TS. Not only was the monitor inoperable for a time period greater than

14 days (thus requiring the report in the effluent report) but discharge

via the pathway continued after the 14 days which is prohibited by the

applicable TS ACTION (ACTION 28). This ACTION states, "With the number of

channels OPERABLE less than required minimum channels OPERABLE

requirement," (In this case there was only one channel'and one channel was

the minimum channels OPERABLE) " effluent releases via this pathway may

continue for up to 14 days provided that prior to initiating a release:

a. At least two independent samples are analyzed in accordance with

Specification 4.11.1.1.1, and

b. At least two technically qualified members of the facility staff

independently verify the release rate calculations and discharge line

valving;

Otherwise, suspend release of radioactive effluents via this pathway."

The licensee evaluated this ACTION and felt that the intent of the

requirement was not to prohibit release via the pathway but rather to

ensure that the sampling and verification requirements were met and that

this applies after the 14 days of inoperability as well.

Failure to make the proper 10 CFR 50.72 notification and 10 CFR 50.73

report is an apparent violation and is identified as 382/8615-02.

11. Review of Special Reports

The NRC inspector reviewed Special Report SR-86-004-00, " Emergency Diesel

Generator A Malfunction Due to Failure of Power Voltage Dropping

Resistor." It appears that this event was properly resolved by the

licensee and reported in accordance with TS 4.8.1.1.3 and 6.9.2. The NRC

inspector also reviewed Special Report SR-86-003-00, " Faulty Sample Line

Check Valve Rendered Wide Range Gas Monitors Inoperable Greater than Seven

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Days." It appears the licensee took appropriate corrective actions and

properly reported this event in accordance with TS 3.3.3.1, ACTION 27 and

6.9.2.

No violations or deviations were identified.

12. Low Temperature Overpressurization Protection (LTOP)

Because Waterford 3 is a latar Combustion Engineering, Inc. (CE) design of

pressurized water reactor, it does not have a power operated relief

valve (PORV) and in low temperature conditions must rely on relief valves

attached to the shutdown cooling system (SDCS) for LTOP.

Paragraph 5.28.3a of the Waterford 3 FSAR states that the most limiting

low temperature overpressure transients initiated by a single operator

error or equipment failure are either an inadvertent safety injection (two

high pressure safety injection-and three charging pumps) or an RCP start

when a positive steam generator to reactor vessel delta T exists.

Further, paragraph 5.28.3d of the FSAR states that " administrative

controls necessary to provide LTOP are limited to those controls that open

the SDCS isolation valves."

Section 6.3.2.5.1 of the FSAR requires safety injection tank (SIT)

pressure to be reduced to 377 psig when reactor coolant system (RCS)

pressure reaches 650 psig and subsequently requires the SIT isolation

valves to be closed and states the reason for doing so is "to prevent

accidental overpressurization of the shutdown cooling system . . . ." The

-above requirements are not automatic and therefore must be done using

administrative controls. The use of additional administiative controls

conflicts with the statement of paragraph 5.28.3d and raises the question

of whether the discharge of a pressurized SIT under low temperature

renditions needs additional analysis. This is identified below as an

unresolved item pending lic nsee evaluation of the apparently conflicting

FSAR statements and the ability of LTOP protection to handle an SIT

discharge.

The NRC inspector's further review of requirements associated with low-

temperature overpressurization protection revealed the following:

a. The note accompanying the APPLICABILITY section of the TS 3.5.1 is

not consistent with the requirements of FSAR Section 6.3.2.5.1.

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. According to the note the pressure of the SITS may be maintained at

pressures up to 625 psig when the RCS pressure is less than

1750 psia. In the FSAR it is required that when RCS pressure reaches

650 psig, SIT pressure must be reduced to 377 psig.

b. The TS note allows the SIT isolation valves to be, closed'in Mode 4

when the FSAR implies they will be closed.

I c. OP-10-001, Revision 6, Attachment 8.8, "Cooldown to Hot Shutdown.

(Mode 3 to Mode 4)," does require the isolation of'the SITS as

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specified in the FSAR and does require SIT depressurization, but does

not specifically require the depressurization of the SITS at 650 psig

as discussed in FSAR Section 6.3.2.5.1. Rather step 8.8.16 of

Attachment 8.8 instructs the operator to begin SIT depressurization

to 235-300 psig when RCS pressure is less than 1750 psig. All

apparent conflicts and inconsistencies associated with LTOP will be

tracked as Unresolved Item 382/8615-03. l

No violations or deviations were identified.

13. Potential Generic Problems

In response to IE Information Notice 86-03, " Potential Deficiencies in

Environmental Qualification of Limitorque Motor Valve Operator Wiring,"

the licensee inspected all 64 operators of this type installed in

safety related systems in the plant during inspections conducted in March

and June 1986. On July 16, 1986, the licensee informed the NRC inspector

that environmental qualification (EQ) for some wiring in 20 of the

operators was found to be indeterminate. These wires were replaced with

EQ wire for 18 of the operators in March and the remaining 2 in June.

(For further information about the two operators repaired in June 1986,

see NRC Inspection Report 50-382/86-13, paragraph 8.) The removed wires

were disposed of without further analysis.

On July 17, 1986, the NRC project inspector informed the W3 SES plant

manager that non-EQ grease had been discovered in valve motor operators

installed in another plant. The licensee completed inspections of EQ

operators for installation of the correct grease on July 24, 1986, and

reported the results to the NRC inspector. Four operators were found to '

have non-EQ grease and the grease in two others was indeterminate.

Replacement of the grease in all 6 operators was completed July 29, 1986.

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No other limitorque motor valve operator problems were identified.

The NRC inspectors are following up and potential enforcement action will

be tracked as Unresolved Item 382/8615-04. .

14. Exit Interview

The inspection scope and findings were summarized on August 1, 1986, with

those persor.s indicated in paragraph 1 above. The licensee acknowledged

the NRC inspectors findings. The licensee did not identify as proprietary

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any of the material provided to or reviewed by the NRC inspectors during

this inspection.

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