ML20212Q001
| 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: |
| References | |
| 50-382-86-15, IEIN-86-003, IEIN-86-3, NUDOCS 8609030362 | |
| Download: ML20212Q001 (13) | |
See also: IR 05000382/1986015
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APPENDIX B
U. S. NUCLEAR REGULATORY COMMISSION
REGION IV
NRC Inspection Report:
50-382/86-15
License:
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Docket:
50-382
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Licensee:
Louisiana Powor & Light Company (LP&L)
4
317 Baronne Street
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P. O. Box 60340
New Orleans, Louisiana 70160
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Facility Name:
Waterford Steam Electric Station, Unit 3 (W3 SES)
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Inspection At:
Taft, Louisiana
Inspection Conducted: July 1-31, 1986
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Inspectors:
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E G. W Senior Resident Inspector
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T. T. Stater, Resident Inspector
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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:
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G.'%-GenstaMe7 Chief, Project Section C,
Date
Reactor Projects Branch
8609030362 860326
ADOCK 05000382
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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.
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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
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The inspection period began with the licensee progressing toward a. reactor.
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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
exceeded the 1.0 gpm specification of TS 3.4.5.2.
After adjusting various
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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.
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No violations or deviations were identified.
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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
to the violation response, Procedure MD-1-015, " Measuring and Test
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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.
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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.
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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.
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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,
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Breaker SI-EBKR-B1-DC-9-S was incorrectly identified in
Attachment 8.3 as SI-EBKR-1B-DC-9-S.
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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
completed by the end of the first refueling outage which is currently
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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
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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
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Environmental Monitoring
SA-W3-QA-86-22 - Environmental Monitoring (Non-rad)
No violations or deviations were identified.
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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~
As part of the review the following PORT meeting minutes were
examined:
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85-04
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86-11 through 26
c.
86-36 througa 46
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86-49 through 70
Based on the above reviews as well as reviews of selected plant operating
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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.
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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
PORC approval was appropriate for these particular circumstances,~ but
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would not normally be valid.
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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
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shift personnel relative to their awareness of the plant conditions.
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
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.
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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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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
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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.
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c.
OP-10-001, Revision 6, Attachment 8.8, "Cooldown to Hot Shutdown.
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(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.
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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
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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.
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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
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the NRC inspectors findings.
The licensee did not identify as proprietary
any of the material provided to or reviewed by the NRC inspectors during
this inspection.
.
_ - - _
.--