ML20153B391
| ML20153B391 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 03/08/1988 |
| From: | Linville J, Williams J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20153B359 | List: |
| References | |
| 50-352-87-31, IEB-87-002, IEB-87-2, NUDOCS 8803220117 | |
| Download: ML20153B391 (37) | |
See also: IR 05000352/1987031
Text
.
.
't
Q:
U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No. 87-31
Docket No.
50-352
License No. NPF-39
Licensee:
Philadelphia Electric Company
2301 Market Street
Philadelphia, Pa 19101
Facility Name:
Limerick Generating Station, Unit 1
Inspection Period: December 11, 1987 - January 25, 1988
Inspectors:
E. M. Kelly, Senior Resident Inspector
L. L. Scholl, Resident Inspector
A. E. Finkel, Reactor Engineer
C. H. Woodard, Reactor Engineer
A. Della Ratta, Safeguards Auditor
Reviewed by:
h.
-
S!f!ff
Da'e'
gH. Williams,ProjectEngineer
t
h
Approved by:
m s LinviTlWChief,
ects Section 2A
Date'
'
Summary:
Rout ne daytime (190 hour0.0022 days <br />0.0528 hours <br />3.141534e-4 weeks <br />7.2295e-5 months <br />
and backshift (10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> including
I
weekends) inspections of Unit 1 by the resident inspectors and regional
i
specialists consisting of (a) resolution of outstanding items including
l
response to NRC Bulletin 87-02 on fasteners; (b) walkdown of the emergency
service water system, plant tours, and observations of maintenance and
surveillance; and (c) review of Licensee Event Reports.
Events followed
included reactor enclosure isolations and a resin injection on December 24.
Meetings attended included routine PORC, and a drug training course.
l
l
l
8803220117 880308
ADOCK 05000352
G
l
-
_. ,
._ , _ -_.
. . _
__
.
_ _ _ _ _ _ _ , - -
.
.
.
.
a
.
No violations were identified; however, a self-identified violation involving
CREFAS operability is discussed in Detail 5.2.2.
Examples where the
licensee's activities are instrumental in the assurance of quality are
discussed in Detail 10, and include RCIC test restoration, RPS power supplies
and fitness for duty training.
We noted your management interest in the potential fire hazard associated with
Fairbanks Morse diesel engine lube oil leakage and your intention to promote
an industry owner's group to address and resolve this issue as discussed in
detail 6.2.
Another potential safety concern requiring management attention
concerns the potential for motor-operated valve overcurrent reversal problems,
initially identified during Unit 1 preoperational testing in July 1983 as
discussed in detail 8.1.
3
,
1
TABLE OF CONTENTS
Inspection 50-352/87-31
1.0 Principals Contacted...........................................
1
2.0 Followup on NRC
Findings.......................................
1
3.0 Plant Operations... ........
6
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
3.1 Summary of Events.......
..... ........................ ..
6
3.2 Operational Safety Verification................
..........
7
3.3 Station Tours.............................................
9
3.4 Safety System Operability......
9
.
. . . . . . . . . . . . . . . . . . . . . .
4.0 Onsite Followup of Events... ...... ... .......................
10
4.1 Secondary Containment Iso 1ations..........................
10
4.2 Resin Injection and Conductivity Transient.
11
. . . . . . . . . . . . .
4.3 SNM Shipment.......................
11
. . . . . . . . . . . . . . . . . .
4.4 Feedwater Level Transient.,
12
. . . . . . . . . . . . . . . . . . . . . . . . . . . . .
4.5 Worker Injured..... ........ ... . ................ ......
14
5.0 Licensee Reports..................
.............
15
. . . . . . . . . . . . . .
5.1 In-of fice Review of LER's. . . . . .
15
. . . . . . . . . . . . . . . . . . . . . . . .
5.2 Onsite Followup of LER's. ... . . .. ............. .......
16
5.3 Periodic or Special Reports....
20
. . .
. . . . . . . . . . . . . . . . . . .
5.4 Security Event Reports....
20
. . . . . . . . . . . . . . . . . . . . . . . . . .
6.0 Surveillance Testing......
23
.. ........................ .......
6.1 Test Observation....
23
....................................
6.2 Diesel Exhaust Leakage.............
24
. . . . . . . . . . . . . . . . . . . .
7.0
Maintenance.......................
28
. . . . . . . . . . . . . . . . . . . . . . . . . . .
7.1 Work Observation..
28
.
.... .. .... .......................
7.2 Battery Charger Mal function. . . . . . .
28
. . . . . . . . . . . . . . . . . . . . .
l
7.3 RCIC Overspeed Trip R3 set...................
28
. . . . .
. . . .
'
7.4 Diesel Control Switch..
29
. . . . . . . . . . . . . . . . . . . . . . . . . . . .
8.0 Common Unit Issues.
29
.........
. . . . . . . . . . . . . . . . . . . . .
. . . .
,
'
8.1 Valve Overcurrent Reversal.... . .
29
. . .
. .
. . . . .
8.2 DC MCC Wiring.............
......... .. .
30
. . . . . . . . . . . . .
l
9.0 Nuclear Review Board Organization.
31
. .
.
. .
. . . . . . . . . .
. . .
1
10.0 Assurance of Quality.
32
....... .. .............
. . . . . . . . . . . . . . .
10.1 Internal Panel Wiring Errors...
32
. . . . . . . . . . . . . . . . . .
10.2 Fitness for Duty Training.
32
. .
. . . . . . .
. . . . . . . . . . . .
10.3 RCIC Test Restoration.
33
...
. . . . . . . . . . . . . . . . . . . . . . . . . .
10.4 LER 87-23 Attention to Detail.. .
33
, . . . . . . . . . . . . . . . . .
.
13.0 Management Meetings........
34
. ..
.
. . . . . . .
. . . . . .
. . .
1
.
-. .. -
-. -_.
m
- d.
- /
9
DETAILS
1.0 Principals Contacted
4
Philadelphia Electric Company
J. Doering, Superintendent of Operations
R. Dubiel, Senior Health Physicist
G. Leitch, Vice President, Limerick
J. Grimes, Branch Engineer, Testing and Labs
J. Milito, Field Engineering Supervisor
D. Helwig, Manager, Quality Assurance
J. Spencer, Superintendent of Services
E. Sproat, Limerick Project Manager
E. Kistner, Nuclear Review Board Chairman
'
R. Scott, Modifications Superintendent
!
Also during this inspection period, the inspectors _ discussed plant status
and operations with other supervisors and engineers in the PECO, Bechtel
and General Electric organizations.
2.0 Followup on NRC Findings
2.1 NRC Bulletin No. 87-02; Fastener Testing
In response to NRC Bulletin 87-02 regarding fastener testing, the
'
licensee selected a sample of fasteners as required by action number
2 of the bulletin.
The inspectors witnessed the licensee's
selection on December 11 to ensure that a diverse sample of various
grades and sizes were randomly chosen from those in stock.
The
licensee performed chemical and mechanical testing however the test
,
results were not available as of the end of the insr.ection period
and will be included in a future inspection report.
During the selection of the fasteners, the inspectors noted that the
Unit 1 storeroom was well-organized with all components clearly
'
segregated and storeroom personnel were knowledgeable of procedures.
2.2 (Closed) Unresolved Item 87-05-02: CADD Drawing Errors
4
Conversion from manually drafted drawings to a computer aided design
drafting (CADD) system was initiated for Limerick Unit 1 piping and
instrumentation drawings (P& ids) in February 1986.
The CADD
drawings were originally issued on July 18-24, 1986, and drawings
were distributed onsite in August 1986.
,
,,.--,-,-r..
r
. --
_ .
i
2
s
Because of a number of errors discovered between September-December
1986, the licensee issued a Part 21 Report on March 16, 1987,
describing errors introduced in the drawings by the conversion from
the manually drafted drawings to the CADD system. A complete
quality reverification of the CADD generated drawings was completed,
and corrected drawings were in place at Category 1 drawing locations
by March 1987.
To correct the CADD system errors and assure that an effective
verification process is in place, the licensee issued an Engineering
Department Project Instruction (EDPI) No. 4.46.2, on January 8,
1988, titled, "Project Drawings Computer Aided Design and Drafting
(CADD)". The salient point in this procedure is a requirement for a
100's verification of any CADD generated drawing against the original
drawing before returning it to the project.
The licensee's QA organization performed surveillance SCR No.
ILC-100 on January 15, 1988, to verify that the corrective action
taken to correct the CADD problem has been effective.
The
surveillance report concluded that all issues bearing directly on
the quality of CADD produced drawings had been satisfactorily
addressed. An inspection of the present controls, per EDPI 4.46.2,
and a review of selected drawings by the inspector indicated that
the sunmary findings of the licensee's surveillance report No. SCR
ILC-100 were consistent and acceptable.
This item is closed.
2.3 (Closed) Unresolved Item 87-24-01; Rapid Plant Shutdown Procedures
'
The licensee revised and issued reactor engineering maneuvering
procedure RE-201 to provide instructions to operators on maintaining
a control sequence during a plant shutdown such that rod worth
minimizer (RWM) and rod sequence control systems (RSCS) will be
automatically latched-in at power levels below the low power
setpoint (22*4 power) without corresponding control rod
withdrawal / insert blocks.
The licensee also revised the RWM pull
sheet to ensure that RSCS Group 10 control rods (targeted deep) are
initially inserted and Group 9 rods (control cell core) are
subsequently inserted to quickly reduce power and maintain sequence.
The procedural changes were approved in PORC meeting 88-008.
This
item is closed.
2.4 (Closed) Violation 87-19-01: As-built Drawing Update
On August 20, 1987, an incorrect revision of electrical drawing E-15
was removed from the control room and the blocking / permit
coordinator's office. On August 21, 1987, the As-Built Drawing
Update form (Appendix 7 to Procedure A-14) was revised to include a
change notice IDCN-006 to drawing E-15.
The redlining was
accomplished, and the drawings were distributed to Category 1
_ - _
._ _
.
>
locations.
In addition, all Category 1 drawings affected by plant
modifications made since a licensee modification audit on April 27,
1987 were reviewed for completeness.
The inspector reviewed the modification coordinator (MC) drawing
location checklist and verified licensee review of five Category 1
areas.
The MC staff audits the Category 1 locations biweekly to
verify current revision of drawings.
In addition, Administrative
Procedure A-14, Appendix 7, has been revised to implement these
changes. A specific time limit has been added to the A-14 procedure
requiring that verification of a modification change within two
working days after the change is received from reproduction. The
inspector, on a selected sampling basis, verified that the MC staff
personnel were performing this task within procedural criteria and
that the MC was performing an independent review of the redlining
process performed by MC personnel.
The inspector also verified that
the MC discussed the problem with responsible personnel and that
proper action was taken to resclve the problem. This item is
considered closed.
2.5 (0 pen) Unresolved Item 87-16-01; Construction Procedure Detail
During a plant piping modification the inlet water pipe to the RHR
oil cooler was inadvertantly loosened allowing cooling water to leak
into the RH9 pump motor. To insure that the Maintenance Request
Forms (MRFs) prepared by the Engineering and Research Department,
Construction Division, consider other areas than their specific
modification work, the licensee has taken the following steps:
Revised "Procedure for Installation of Mechanical Equipment, CD
-
5.7".
This procedure has added further guidance for the
Mechanical Site Lead and Construction Engineer as well as
additional planning and quality centrol inspection guidelines.
Procedure CD 5.3 "Procedure for Installation of Electrical
Equipment", was also updated to reflect similar changes as made
in CD 5.7 above.
-
Revised procedure for "Preparing Engineering Work Letters and
Construction Division Field Engineering, and Testing and
Laboratories Division Memoranda", ERDP 2.2, Revision 15.
'
The licensee has also prepared a check list to be used when
preparing the Construction Job Memorandum (CJM).
The inspector
reviewed two CJMs, Mod. 5133, "RHR Pemp Compartment Unit Cooler
l
Piping Change" and Mod. 5029, "Install Cooling Fans on SMVA Site
l
Services Transformer".
The procedures reflected subjects discussed
!
in the CJM.
Due to licensee organization changes being implemented at this time,
other procedures are being revised which were not reviewed during
this inspection period.
This item therefore remains open.
.
4
.-
2.6 (0 pen) Unresolved Item 85-30-01; Third Offsite Power Sources
The inspector reviewed the status of licensee progress in
implementing plant and procedure changes which would make a third
offsite power source available in the event that one of the two
offsite power sources required by technical specifications is lost.
Two items of concern regarding this power source are physical
independence from the 220 kV offsite source, and the capability to
place the source into service within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Plant modification
961 is being prepared to accomplish the following:
-
Install cabling in underground conduit
Refurbish equipment to be used (i.e., spare 33 kV circuit
-
breaker)
-
Install protective relaying for the circuit breaker.
The use of the underground conduit run is intended to provide
physical independence while the modification work will reduce the
time required to put this source in service upon failure of one of
the other two offsite power sources.
This item remains open pending
implementation of the necessary plant modifications.
2.7 (Closed) Unresolved Item 87-29-01: Station ALARA Review Committee
The inspector attended the Station ALARA Review Committee (SARC)
meeting chaired by the Station Manager on December 18, 1987.
The
meeting was conducted in accordance with Administrative Procedure
A-83 and a proper quorum was present, including first-line
supervision from construction and maintenance groups.
Items
discussed included:
-
Station exposure goals for 1988
-
The role of management and workers in ALARA
-
Specific post-job reviews
-
Individual ALARA concerns
The inspector also reviewed Revision 1 to procedure A-83 which was
approved in PORC Meeting 87-130 on December 30.
No unacceptable
conditions were noted, and based on the management support of the
SARC and the revision to A-83, this item was closed.
The inspector discussed proposed changes in SARC philosophy with
station management because of recent licensee reo ganization
activities.
Further NRC review of SARC activities and effectiveness
of the ALARA program will be reviewed in future NRC inspections.
- _ . .
. _ - -
- _ _ _ _
- .
.
._
_
_
/
= '
5
.-
2.8 (Closed) Inspector Followup Items 87-08-01 thru 04: Emergency
Procedures
'
.
Emergency operating procedures (EOPs) were reviewed in NRC
Inspection 50-352/87-08 for conformance with the BWR Owner's Group
program, specifically the General Technical Guidelines (GTG),
including documentation of bases.
The Limerick. Plant Specific lechnical Guidelines were approved on
January 11, 1988 and were based upon Revision 4 to the GTG.
The
revision clarified or resolved differences (noted during the NRC
inspection) between the Limerick-specific Transient Response
Implementation Trip (TRIP) procedures and the GTG. Also, a revision
to Administrative Procedure A-94, Preparation of TRIP Procedures,
.
was approved in PORC Meeting 87-117 on November 24, 1987, to
incorporate comments from NRC Inspection 50-352/87-08, and brought
the Limerick E0P program into full compliance with the GTG,
The inspector reviewed recent revisions to TRIP procedures T-101,
102, 111, 112, 116 and 117.
Revision 2 to each of these procedures
was approved in PORC Meeting 87-117 with minor comments noted, and
were issued for training and use on January 13, 1988.
These TRIP
procedures are in flow chart form for use in the main control room.
The procedure revisions were discussed with the responsible engineers,
operations group supervision, and licensed operators. All were
knowledgeable in the procedural steps changed, most notably on the
concept of re-entry into the TRIP procedures as new entry conditions
are identified.
Another finding from NRC Inspection 50-352/87-08 concerned bases
documentation for the TRIP procedures, which had not been maintained
as a controlled document nor maintained up to date. Since the TRIP
bases are used as a reference for training, and copies are kept in
the main control room for information, the licensee revised
Administrative Procedure A-94 to require controlled TRIP Bases and
issued Revision 0 to all T-101 series documents on December 9,1987.
The licensee also issued new TRIP procedures T-103 and 104 to
implement BWR owner's guidelines on secondary containment control
and radiation release controls.
The inspector reviewed the new
procedures, discussed comments related to T-103 and 104 in PORC
Meeting 87-110 (October 30, 1987) minutes, and ascertained that
verification and validation of the new procedures was performed by a
multi-disciplinary team en the Limerick simulator.
In addition,
Administrative Procedure A-94 was revised to formally require
simulator validation by a multi-disciplinary group for futu 2 TRIP
procedure changes.
Finally, additional comments by the NRC were expressed in a letter
to the licensee dated August 7, 1987 (W. Johnson to J. Gallager)
,
,
m._-c.,---
.--.c
.-
.
.
.
-
. _
__ .__ _
.
a
/
6
,
,
concerning inconsistencies in TRIP procedure T-116, Reactor Pressure
Vessel Flooding.
Specifically, the use of safety relief valves
(SRVs) as a vent path out of the top of the vessel or the head vent
or other alternate paths during vessel ficoding were questioned.
The NRC concerns were clarified by the licensee in stating that
there was_never any intention to establish equivalency between vent
methods (i.e., SRVs versus alternate paths) and that, if the
required number of SRVs could not be opened to allow use of the
minimum alternate flood pressure table (contained on the T-116 flow
,
chart), then other methods would be pursued.
The inspector had no
further concerns and these items are closed.
2.9 (Closed) Unresolved Item 85-30-05; Generator Brush Arm Stress
A Colt-Fairbanks diesel generator support arm on the rotor slip ring
brushes at Millstone Unit 3 failed by fatigue during preoperational
testing in August 1985.
The similarity in design to the Limerick
generators was reviewed in NRC Inspection 50-352/85-30 and found not
,
to be identical. The Limerick. brush support arms are one-half the
length of the Millstone design and, as such, are concluded to be
less susceptible to the fatigue failure. The licensee's Mechanical
Engineering representatives contacted the manufacturer of the
generators, Louis Allis Co., who confirmed that the shorter brush
arm on the Limerick generators has a much higher (eight times)
resonant frequency and that adequate wall thickness exists for the
brush holder assemblies.
,
The Millstone failures were stat w to be random, and resulted
from holes drilled too deep for the brush arms.
Evaluation of
Limerick installation drawings and x-ray pictures subsequently
provided by the licensee to Louis Allis showed adequate wall thick-
ness for the Limerick brush arm assemblies.
The Limerick stud
'
length of eight and one-half inches was concluded to greatly reduce
structural load factors in that area.
Based on the above, and on licensee QA Audit No. AL 86-41 which
addressed this issue, the item is closed.
,
!
3.0 Plant Operations
l
3.1 Summary of Events
Unit 1 operated at full power throughout the inspection period and
i
completed 121 consecutive days at power as of January 25, 1988.
No
.
further main turbine electrohydraulic control system vibration
!
concerns (the subject of extensive engineering analyses and control
l
system modifications following a scram on September 19, 1987) were
experienced in this period.
An injection of resin into feedwater
I
occurred on December 24, 1987 while placing a filter demineralizer
I
in service (Detail 4.2) causing a conductivity transient and reduced
power operation for four days.
l
i
- - . -
_ _ _ _ _
7
.
7
..
Several secondary containment isolations occurred (Detail 4.1) due
to cold ambient air conditions and other control system
difficulties. Operating problems with the RCIC overspeed trip
mechanism (Detail 7.3) and emergency diesel engine exhaust (Detail
6.2) were experienced, as was a switch contact problem for the D14
diesel engine local / remote control switch (Detail 7.4).
The NRC issued a Temporary Waiver of Compliance on December 18,
1987, enabling the licensee to begin implementation of organizational
changes announced in November 1987, pending formal licer.se amend-
ments.
Effective January 1, 1988 the Limerick Station Manager was
transferred to Peach Bottom and the Vice President of Limerick Station
assumed the role of acting Station Manager pending training and
indoctrination of an individual selected to become Unit 1 Station
Manager.
3.2 Operational Safety Verification
3.2.1
Control Room Activities
The inspectors toured the control room daily to verify
proper manning, access control, adherence to procedures
and compliance with technical specifications.
The
inspectors reviewed shift superintendent, control room
supervision, and licensed operator logs and records
covering the entire inspection period.
On December 18,
1987 and January 25, 1988, backshif t inspections were
performed between the hours of 2:00 am and 6:00 am.
The inspectors reviewed logs and records for completeness,
abnormal conditions, and significant operating changes and
trends. Other records reviewed included:
reactor
engineer and shif t technical advisor (STA) books, night
orders, radiation work permits, the locked valve log,
maintenance request forms, temporary circuit alterations,
and ignition source control checklists.
The inspectors
also observed shift turnovers during the period.
Operations activities were observed to be in conformance
with Administrative Procedure A-7, Conduct of Plant
Operations.
On minuary 6, a reactor core isolation cooling (RCIC)
system area temperature instrument was found to be
inoperable during a review of Daily Surveillance Log
(ST-6-107-590-1) by shift supervision.
The inoperability
was a result of improper system restoration fol',,<ing
surveillance testing, as discussed in Detail 6.3.
.
_
_
_
_
.,
,
. _ , ,
k
- <
, .
8-
,
,
3.2.2
Security
During entry to and egress from the Unit 1 protected area
and vital areas, the inspecters observed that access
controls, security boundary integrity, search activities,
-
,
escorting and badging were 13 accordance with Security
'
Plan implementing procedures and guard force instructions.
The inspectors also observed the availability and
operability of. security systems such as search equipment,
perimeter detection devices, and security computer alarms.
The inspectors verified that the minimum number of armed
guards required by the Security Plan were present on
selected shifts by review of duty rosters, discussion with
licensee Shift Security Advisors, and observation of guard
,
'
force turnovers.
The inspector also reviewed the security procedures for
vehicles access via the North Gate. Post Order Number 7
e
was reviewed and discussed with the security force members
on duty. The inspector also observed the search and
entrance of a vehicle into the protected area, as well as
the controls imposed on emergency vehicles which require
immediate plant access. The guards displayed a thorough
,
knowledge of their duties as delineated on the post
,
orders. No violations were identified.
L
3.2.3
Fitness for Duty Investigations
On January 7, 1988, site access was terminated for a Unit
2 Bechtel construction worker when the licensee was made
aware (through police contacts) of his offsite illegal
drug activities.
The individual was interviewed by PECo
security management.
Site security reviewed the personnel
,
access records and determined that since the initial
'
implementation of the security plan, this individual never
had access to the Limerick Unit 1 protected area.
l
On January 8, 1988, plant management informed the resident
inspector that a former Peach Bottom employee, who is
currently being prosecuted for illegal drug activities,
had implicated other Peach Bottom employees as also being
involved in drug activities. Two of the individuals are
currently employed at Limerick and the licensee is con-
tinuing its investigation to determine if the allegation
,
is credible.
'
3.2.4
Radiological Controls
J
The inspectors observed the availability and use of radia-
tion monitoring equipment, including portal monitors and
portable friskers.
The inspectors also observed health
i
.
- -
- -
-
-
.
-
- - - -
-
-
-
-
-
.
.
.
-
-
9
'
physics (HP) supervision and technicians in plant
activities involving potentially significant radiological
conditions.
Radiological controls for posted radiation and
contaminated areas were assessed as part of the inspector
review. Radiological conditions were discussed with HP
technicians.
Proper locked high radiation area controls,
including appropriate and frequent surveys, were verified
to be employed.
The inspector had no concerns, and
identified no violations.
3.3 Station Tours
The inspectors toured accessible areas of the plant throughout the
inspection period, including:
the Unit I reactor and
turbine-auxiliary enclosures, the main control and auxiliary
equipment rooms; battery, emergency switchgear and cable spreading
rooms; the spray pond pumphouse; diesel generator cubicles and the
plant site perimeter.
During these tours, observations were made of
potential fire hazards, radiological conditions, housekeeping,
tagging of equipment, ongoing maintenance and surveillance, and the
availability of required equipment.
No unacceptable conditions were identified.
3.4 Safety System Operability
3.4.1
Emergency Service Water
The inspector performed a detailed walkdown of portions of
the emergency service water (ESW) system in order to
independently verify system operability.
The walkdown
included review of the following:
-
Technical Specifications 3/4.7.1.2, FSAR Section
9.2.2, P&ID M-11, and Licensed Operator Training
Plan 0680
Inspection of ESW equipment conditions
-
-
System check-off list S11.1.A (Col-1,
-2, -3) and
system operating procedures consistent with plant
drawings
-
Valves and switches properly aligned including
appropriate locking devices
Instrumentation properly valved-in and operable
-
-
- _ - _ _ _
,
10
.-
Satisfactory status of control indicators and
-
controls
Surveillance test procedures ST-011-203, 206, and 231
-
appropriately completed at the required interval
ESW Loop ' A' and 'B' Flow Balance procedures
-
RT-1-011-251 and 252 performed as part of the system
retest following piping modification work
Within the scope of the inspection, no unacceptable
conditions were noted.
The inspectors discussed recent
maintenance, modifications, and design concerns related to
the ESW system with responsible test engineers.
Proper
operation of the ESW System was also v2rified as part of
quarterly pump testing of the system as discussed in
detail 6.1.
No unacceptable conditions were noted.
4.0 Onsite Followup of Events
4.1 Secondary Containment Isolations
During the inspection period, reactor enclosure isolations occurred
on two dates. On December 29, two reactor enclosure isolations were
received within 30 minutes of each other due to low differential
pressure conditions in the secondary containment.
The loss of
building negative pressure occurred when the normal supply and
exhaust ventilation fans tripped on a low supply air temperature due
to extremely cold ambient conditions. A temperature control switch
was found to be defective. An additional set of heating coils was
placed in service and no additional problems were experienced.
On December 31, a reactor enclosure isolation occurred on a low
differential pressure signal which was caused by the tripping of the
reactor enclosure normal ventilation fans.
The ventilation fans
tripped on a low temperature signal but no problems with air
temperature or the heating coils were evident.
The inspectors confirmed proper system response during the
isolations, immediate corrective actions, and appropriate reporting
to the NRC.
No further reactor building isolations occurred from January 1
through the end of the report period, January 25.
The inspectors
will continue to follow corrective actions proposed by the licensee
to resolve ventilation problems.
..
11
.-
4.2 Resin Injection and Conductivity Transient
On December 24, approximately one pound of condensate filter
demineralizer (demin) resin was inadvertently injected into the
reactor vessel.
The injection occurred when 'E' condensate filter
demineralizer was placed in service.
Indications of the injection
were higher than normal main steam line radiation levels, increased
radiation levels at the steam jet air ejector, and a decrease in
reactor coolant pH.
'
The licensee investigation determined that there was no evidence of
failure of the demineralizer filter elements but that the most
likely cause of the injection was resin contamination of the
downstream side of the 'E' filter demineralizer elements during a
previous regeneration cycle.
Reactor coolant conductivity increased to a maximum of 3.62
micro-mhos per centimeter, and pH dropped to 5.42.
Both parameters
were restored to technical specification limits within six hours and
therefore no special actions were required.
The inspector verified
that the licensee has a program to monitor cumulative time of
operation with increased coolant conductivity levels. This time is
recorded in procedure RT-5-041-875-1.
Procedure changes were issued to filter-demineralizer regeneration
operating procedures to prevent the use of an improper drain path.
The licensee also developed a videotape training session which is
intended to reduce instances of personnel error.
This training was
a result of a special PORC meeting to review recent instances of
personnel error and possible corrective actions.
The inspector had
no further concerns.
4.3 Special Nuclear Material (SNM) Shipment
On December 9, 1987, the licensee shipped two packages to the
Commonwealth Edison Company's LaSalle County Nuclear Station. The
packages reportedly contained two fission counters, each containing
one gram of 93% enriched uranium. On December 14, 1987, the
licensee received a telephone call from a LaSalle Station
.
representative notifying the licensee that, upon receipt inspection
l
of the fission chamber shipping packages, the two fission chambers
'
containing the enriched uranium were not in the packages.
The
licensee immediately conducted a search of the fuel floor area where
the packaging was performed and found the fission chambers in a
yellow contamination control bag.
The fission chambers had been
inadvertently left out of the shipping packages,
i
l
l
t
I
-_
_
_-
.-
- _ _
- .
.__-
_-
..
12
The licensee's reactor engineer staff conducted an investigation and
presented a report to the PORC for evaluation on December 18.
Based
on the licensee's preliminary findings, the following changes were
initiated for incorporation in Administrative Procedure A-44:
-
The Reactor Engineeer shall have responsibility for arranging
the packaging and shipment of SNM.
QC shall verify the packaging of all SNM for shipment.
-
-
All fission chambers checked out of the storeroom will be
tagged with highly visible tags that identify the items as
SNM until the item is installed, returned to the storeroom or
shipped.
The inspector had no further concerns, and identified no violations.
The controls for SNM will be reviewed in future NRC inspections.
The inspector reviewed Upset Report (UR)-034 prepared by the
operations staff and reviewed and issued by the PORC on January 14,
1988.
The report covered a reactor water level transient due to a
less of feedwater control power that occurred on November 19, 1987,
and was followed-up and documented in detail 4.2 of NRC Inspection
Report 50-352/87-28.
The UR contained a detailed sequence of events for the
transient, an analysis of selected equipment failures and associated
plant response, and three distinct recommendations. Also attached
to the UR, and reviewed by the inspector, were feedwater control
logic diagrams, strip chart records, selected emergency display
system plots and process computer alarm typer printout.
The inspector discussed the immediate operator response to the
transient with licensed personnel on shift at the time of the
The reactor operator quickly recognized a feedwater
control circuit failure and took manual control and began feed pump
turbine runbacks within 20 seconds by reducing motor speed changer
settings.
Reactor vessel level increased to +52.8 inches in 23
seconds (one inch below the high level main turbine trip setting
which would have resulted in a reactor scram) before the combination
of an automatic reactor recirculation pump runback and the
operator's actions stopped the level increase.
A rapid level
decrease was then experienced which reached a minimum of +18.4
inches in 43 seconds (six inches above the scram setpoint) before
additional operator actions to increase feed pump speeds terminated
the water level drop.
Operator manual control of reactor water level was then complicated
by several factors during the next four to six minutes.
The 'A'
feed pump was running at low flow, but with no minimum flow
protection because its recirculation line had been previously
..
.
_ - . - .
.
_
_ .
_ _ -
_ _ - _ _ - - _ _
__
_- __ _ _ _ _ _ _ _ . _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ ___ -
13
.
'
,
-1
l
isolated because of leakage.
The 'A' pump was therefore. tripped in
five_ minutes. -The 'C' feed pump turbine speed had been originally
locked-in at 3690 rpm and continued to feed at a mass flow rate of
approximately 3.9 million 1b/hr.
However, because flow indication
had been lost, the reactor operator thought that the 'C' pump was not
feeding and tripped the pump five and one half minutes into the
transient. This caused another rapid water level decrease from a
normal condition of +34 inches to a minimum of +18.0 inches before
the reactor operator took manual control of the 'B' feed pump (the
only pump feeding at this point) to again terminate the level drop.
Normal reactor vessel level was stabilized in 12 minutes by
reopening the ' A' feed pump miriimum flow line and increasing speed
of the 'A'
and 'B'
pumps.
- Attempts to insert control rods (to increase thermal margins) were
prevented by RWM insert blocks caused by a loss of steam flow
indication and a resultant false low power setpoint condition.
Reactor power remained at approximately 57% with reactor
recirculation pump speed at the 28% limiter setting. Approximately
one hour after initiation of the transient, operators had identified
the cause of the failures to be an inadvertantly opened circuit
breaker in nonsafety related panel 10Y201. An additional
,
!
complication arose when the open circuit was re-energized which
caused both reactor recirculation pumps to increase speed.
'A'
loop
,
flow increased from 6,000 to 16,750 gpm and 'B' loop flow increased
from 7,500 to 21,000 gpm.
The additional flow caused power to
increase to the APRM upscale alarm, but remained below the 75% flow
limiter setting at which a high flux scram may have occurred.
>
Recirculation pump speeds were then manually dacreased and adjusted
to develop a core flow of 50%. With the RWM no longer enforcing
after restoration of the open control circuit, a symmetrical rod
pattern was established, process computer thermal limit calculations
were obtained and were verified to be within limits.
Normal reactor
water level was m;intained with the two operating feed pumps.
The cause of the transient was the inadvertant de-energization of
circuit 11 in electrical panel 10Y201 oy a nonlicensed operator
whose hand had slipped while attempting to open an adjacent
,
auxiliary boiler circuit brecker in the same panel. The effect of
.
the loss of circuit 11, which powered various nonsafety-related
J
circuits, included:
1
i
Loss of 'C' feed pump flow indication (downscale) and lockup of
--
the turbine motor gear unit.
-
Feedwater master controller failure upscale, causing the 'A'
and 'B' feed pumps to increase speed and flow.
Reactor vessel
i
level correspondingly increased and reactor power increased
'
from 85 to 89% in 16 seconds,
l
4
1
a
,
._wn
.-
.-n-v
,...__n
__,.v,__-
m
,- - , . , , . . . , .
, - , _ . , _ _ _ . , . , , , - , , , , _ - . . - . _ - , , , - . , - - -
.
.
.
.
.
.
.
.-.
.
.
5
,
,
14
,
Reactor recirculation pump automatic runback to 28% rated speed
-
occurred, reducing power to 45% in 18 seconds. The runback was
due to the loss of circuit 11 power to feedwater flow summer
C32-K615 whose output dropped below 20% of rated feedwater flow
for greater than a 15 second time delay.
-
Steam flow / feed flow indication power was lost and, because-
-
sensed steam flow indication dropped below 20%, the rod worth
minimizer (RWM) began enforcing control rod insert and withdraw
blocks.
An important feature of the reactor recirculation pump runback
circuitry was later identified as part of the post-transient review.
The automatic reset of the runback logic (such as occurred when
circuit 11 was re-energized) can result in an unexpected increase in
pump speed when a low feedwater flow condition clears. The
increased pump speed returns to the value that the manual controller
had been demanding.
In this event, the shift superintendent had
recognized this automatic reset feature and had directed that the
,
'
manual controllers'be reduced to below the 28% speed limiter. The
controller demands were subsequently reduced by the reactor operator
'
but not to a setting below the limiter.
The licensee prepared a
request to engineering (EPE-1191) on December 1, 1987 for modifi-
cation of the runback logic.
Seal-in of the speed limitation on low
feedwater flow, thereby requiring a manual reset, will prevent the
possibility of high flux scrams by eliminating the automatic reset of
the recirculation pump runback logic and will simplify transient
response and recovery.
The inspector concluded that the investigation conducted by
operations as documented in UR-034 was thorough and that
recommendations for improved breaker blocking techniques would be
adequately tracked as PORC commitments from PORC meeting 88-01. No
further concerns were identified.
i
4.5 Worker Injured in Decontamination Tank
On January 16, at approximately 9:30 a.m., a contractor became
unconscious and stopped breathing while working in a decontamination
tank.
The worker was immediately discovered by a health physics
technician who successfully administered cardiopulmonary
resuscitation (CPR). The worker stopped breathing a second time and
again was revived.
The individual regained consciousness and was
(
transported to the Pottstown Memorial Hospital where he was examined
and released by noon the same day.
The worker was not contaminated.
'
The worker was employed by Bartlett Nuclear Inc. and was attempting
to clean a drain in a 4'x4'x8' decontamination tank (located in a
-
trailer outside of the Radwaste Enclosure) which had been used two
l
to three days earlier.
However, the lower section of the tank still
contained an atmosphere of freon that had not been fully purged.
,
r
i
i
- ~ - -
-
o
15
..
The licensee has stopped further high pressure freon decontamination
activities and is alerting the industry of this potential generic
safety concern.
5.0 Licensee Report
5.1
In-Office Review of Licensee Event Reports
The inspector reviewed Unit 1 LERs submitted to the NRC Region I
office to verify that details of the event were clearly reported,
including the accuracy of the description of the cause and the
adequacy of the corrective action. Where multiple causes are
suspected, or may be different than reported in the LER, this is
,
indicated below.
The inspector determined whethei further
information was required from the licensee, whether generic
implications were involved, and whether the event warranted on-site
followup. The following LERs were reviewed:
LER Number
and Date
Subject
Root Cause
85-74
Valve isolations
Blown fuses
Revision 1
1/15/88
87-23
Battery charger
Failed electronics
Revision 1
f r.i l u re
12/10/87
87-48
EHC pressure loss and
EHC control system
Revision 1
reactor scram
instabilities
1/5/88
87-49
Startup of recircula-
Incorrect temperature
Revision 1
tion pump without
reference in test
12/3/87
verifying temperatures
procedure
87-61
CREFAS inoperability
Licensed operator error and
12/16/87
for 1 1/2 hours
inadequate procedure
87-62
Nonlicensed operator error
12/21/87
isolation and initia-
in using A&C exhaust fan
combination
87-63
Fire suppression water
Licensed operator error
12/24/87
header isolation
in failing to adhere to
administrative proceduros
.__ _
- _. -
- - _____ __ _____
3
16
. . .
87-64
Auxiliary boiler trip and
12/21/87
isolation and SGTS/RERS loss of auxiliary heating
initiation
steam to supply fans
87-65
Low intet air temperature
12/21/87
isolation and SGTS/RERS trip of normal ventilation
I
initiation
supply fans
87-66
Blockage of drain port in
1/7/88
Malfunction
reset mechanism
'
The events described in LER Nos. 87-62, 64 and 65 were previously
addressed in Detail 4 of Inspection Report 50-352/87-28, and
corrective actions are addressed in Detail 5.2.1 of this report.
LER 87-61 is addressed in Detail 5.2.2 and LER 87-66 is addressed in
Detail 5.2.3.
LER No. 87-63 is addressed in Detail 3.2.1 of
Inspection Report No. 50-352/87-28.
The inspector noted that three.LER's during this inspection period
(LER Nos. 87-61, 62 and 63) were submitted late by an average of
eight days beyond the 30-day requirement. While not a concern for
the quality and accuracy of the reports, this lateness suggests a
management inadequacy in planning and compiling pertinent data and
enlisting appropriate support among various organizations and work
groups involved in the events reported.
These comments were
discussed with the Limerick Licensing Engineer and at the exit
meeting for this report.
5.2 Onsite Followup of Licensee Event Reports
For those LERs selected for onsite followup, the inspector verified
that the reporting requirements of 10 CFR 50.73 and the Technical
Specifications had been met, that appropriate corrective action had
been taken, that the event was appropriately reviewed by the
licensee, and that continued operation of facility was conducted in
accordance with technical specification limits.
5.2.1
LER Nos. 87-62, 64 and 65; Reactor Enclosure Isolations
As a result of the ongoing problem with reactor enclosure
isolations, the licensee has accomplished the following:
the normal reactor building ventilation supply fan
-
low temperature trip function was taken out of
service, using a temporary circuit alteration to
prevent spurious actuaticns. Air inlet temperature is
monitored by plant operators when outside air
+emperature is below 35 degrees F.
.
_
>
,.
17
.-
the importance of notifying control room personnel
-
before changing equipment operating lineups was
stressed to all operating personnel in turnover
meetings and by shift night orders.
investigation of the inability of the 'A'
and 'C'
fan
-
lineup to maintain adequate negative pressure in
the secondary containment is continuing, An Operator
Aid has been posted to warn operators of the problem
with using this fan combination.
an engineering evaluation is being performed to
-
review several methods of improving the reliability
of the auxiliary boilers.
The status of this review
is to be reported to the Nuclear Review Board (NRB)
at the next scheduled meeting. Also, the Test
Engineering Group prcvided some additional guidance to
shift supervision on how to operate the boilers to
improve reliability under varying steam demand
conditions.
The inspectors will continue to follow these issues.
5.2.2
LER 87-61; CREFAS Inoperability
LER 87-61 described the inoperability of both trains of
control room emergency fresh air supply (CREFAS) for
approximately three and one half hours on November 7,
1987, because of the lack of an available return flow
path. The loss of a flow path was the result of operators
placing the CREFAS return fan operating switches in the
off position because of previous overcurrent trips of the
fans.
The fan trips resulted after operators had entered
Off-Normal procedure ON-115, Loss of Control Enclosure
Cooling, following failure of the 'A' control enclosure
chiller due to a faulty relay.
The 'B' chiller was
out of service for maintenance at the time. After one
hour without control enclosure cooling, and as directed by
procedure ON-115, operators placed control and auxiliary
equipment room ventilation systems in a simultaneous purge
mode of operation to maintain room temperatures below 78
degrees F.
However, the control room ventilation return
fans then tripped (15 minutes after beginning purge
operation) on overcurrent conditions because purging both
spaces (control and auxiliary equipment rooms)
simultaneously is outside of the design capacity of the
duct.
The return fan breakers were reset and, to
prevent further fan trips and potential damage, the
- .-
_ -_
-
-_
_.
_ _ _ - - _ _ _ _ _ _
_ _ _ _ _ - _ _ _ _ _ _
_ _ _ _ _ _ _ _ _ _ _ .
. _ _ _ _ _ _ _ _ _ _ _ _ _ _
a.
18
.
.-
.
1
.
control switches were placed in the off position.
Purging
was continued for the next three hours by opening control
room doors for an exhaust path.
- Operators were unaware that'CREFAS operability required
,
automatic fan control switch positions, since CREFAS
.
initiation will terminate any purge operational alignment
'
(and the one in which the return fans were experiencing
tripping).
Recurrent return fan trips would have been
experienced with the control switches left in automatic.
.The fans remained available for manual operation during
'
the three and one half hours in which CREFAS was
inoperable. After this time, shift supervision discussed
.
the operational problems experienced during purge
operation with station management and concluded that-
-
continued operation with the return fan operating switches
in an off position did not meet technical specification
requirements for CREFAS operability (i.e. toxic gas or
.
radiation initiating conditions).
The inspector verified
that control room temperatures remained below 78 degrees
,
during subsequent purge operations. The 'A' chiller was
'
e
returned to service, following relay replacement,
!
approximately 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> after cooling was lost.
'
The licensee issued a memorandum to shift personnel and
revised procedure ON-115 on November 10, 1987, containing
appropriate instructions on purge operation. At PORC
meeting number 88-02 held on January 22, 1988, control
room ventilation operability questions were discussed
,
including ON-115 instructions, purge operation and
!
attendant concerns for electronic equipment reliability at
increased temperature conditions.
The PORC recommended
engineering analysis of CREFAS initiation response times
and isolation functions during purge operation when a
-
chiller is not available.
I
The inspector concluded that no violation would be issued
for this licensee-identified problem (50-352/87-31-01)
'
since it existed for a relatively short time and was a
>
'
unique operational problem that has not been recurrent,
i
j
Appropriate reporting and corrective actions were
!
accomplished. The inspector had no further concerns.
1
i
5.2.3
LER No.87-66; HPCI Overspeed Trip Malfunction
i
Licensee Event Report 87-66 reported the inoperability of
the High Pressure Coolant Injection (HPCI) system due to
l
,
'
l
erratic operation of the HPCI turbine stop valve during
surveillance testing.
During the test, the hydraulic
.
mechanical overspeed trip mechanism cycled between the
l
l
tripped and normal positions, causing the turbine stop
l
i
.
..
19
.
'
valve to close and open several times.
The cause was a
blockage of an internal port in the overspeed trip reset
'
plunger assembly which subsequently cleared during
operation of the auxiliary oil pump.
The overspeed trips
could not be duplicated following the initial occurrence.
The HPCI turbine lube oil was sampled for particulates and
found acceptable.
The HPCI pump and turbine were operated
for 100 minutes during the performance of the quarterly
flow test (ST-6-055-340-1), with no recurrence of the
problem. The licensee also has increased the operating
time of the auxiliary oil pump during the performance of
the monthly HPCI turbine overspeed trip check
(RT-1-055-330-1) in an attempt to reproduce the problem.
No repeat problems have occurred.
The licensee is tracking this problem by use of the Plant
Incident Tracking (PIT) System as Item No. 87-12,
5.2.4
LER No. 87-48, (Revision 1); Scram Due to EHC Weld Failure
The licensee revised LER 87-48 to describe adjustments to
the electrohydraulic control (EHC) system steamline
resonance compensator to remove signal oscillations
experienced because of a modification to change from full
arc to partial-arc turbine steam admission.
The EHC fluid
pressure loss which resulted in a September 19, 1987 scram
was concluded to be vibration-induced failure of an
improperly bonded weld, and therefore an isolated event.
All other EHC welds were inspected visually and with dye
penetrant and no evidence of cracking was found,
A modified steamline resonance compensator (SLRC) was
installed which effectively removed the control signal
oscillations at power levels above 80% as reported in NRC
Inspection 50-352/87-28.
Full power operation was
achieved by November 21, 1987.
The temporary circuit
alteration (TCA) which enabled installation of a second
resonance compensator in the 'A' channel regulator is
planned to be made into perrcanent modification number
87-5731.
The inspector observed the on-line installation
of the new resonance compensation filters, reviewed the
safety evaluation for the modification, and based upon
stable servo-currents for the main turbine control valves
over the past two months concluded that the 2.55 Hertz
third harmonic frequency of the main steam lines was
eliminated.
._
_ _ _ .
- ,,
20
- *
The inspector also reviewed other associated concerns with
turbine partial-are admission operation as outlined in a
corporate engineering memorandum to the Limerick Station
Manager (PM0M-1502) dated December 7,1987. The concerns
"
are for turbine control valve opening setpoints, waterhammer
potential and operation with control valve number 3 against
the open stop.--Although LER 87-48 did not discuss these
additional concerns, the inspector noted that steps are
'
being taken to pursue modifications to correct the problems
at the next Unit 1 outage.
5.3 Periodic or Special Reports
Periodic or_ special reports submitted by the licensee were reviewed
by the inspector.
The reports were reviewed for the required
information, that results and/or supporting information were
l
i
consistent with design predictions and performance specifications,
and whether any information in the report should be classified as an
abnormal occurrence.
l
The following reports were reviewed:
Monthly operating report for December 1987
--
PECO letter to NRC (Fogarty to Russell) dated November 30,
--
1987; Summary Reports for ISI and ASME Repairs
PECO letter to NRC (Gallagher to Lazarus) dated December 23,
--
1987; Objectives of April 1988 Emergency Exercise
1
,
Annual Report of Plant Modifications (Alden to Russell) dated
--
December 8, 1987
Cycle 2 Startup Report (Alden to Russell) dated November 24,
'
--
1987
<
Annual Report of Safety Valve Challenges dated January 7, 1988
--
The inspector had no questions about the reports.
.
5.4 Security Event Reports
The inspector reviewed the following reports made by the licensee in
accordance with the security requirements of 10 CFR Section 73.71.
i
Details of the reports were withheld from public disclosure as
1
i
safeguards or confidential information.
{
I
Report 87-503 dated December 7, 1987
--
Report 87-504 dated January 5,1988
--
The details were found to be accurate, appropriately reported and
4
contained appropriate corrective actions.
Followup of the reports
i
l
will be pursued in future NRC Regional security specialist
i
inspections. No violations were identified,
.
.1
. _ _
_ _ _ _ _ _ _ _ _ _ . . _ _ - _ . . - _ _ _ _ _ _ . _ _ _ , _ - _ .
. - _ , , . , , , ,
a.
21
,
,
5.5 Part 21 Reports
The following reports made under 10 CFR Part 21 were reviewed by the
inspectors for accuracy, corrective action and progress of
resolution.
5.5.1
Brown Boveri Undervoltage Relays
The manufacturer of an undervoltage (UV) relay used for
the reactor protection system power supply breakers
reported a potential for misoperation to the NRC in a
letter dated December 22, 1987. Brown Eoveri Model 27N
relays (Catalog No. 211T4175-HF) with a harmonic filter
were identified to be susceptible to spurious trips when
DC control voltage is reapplied after having been off for
a period of time.
Relay diodes under this condition
became reverse-biased (in an off state), unstabilized, and
depending upon wiring configuration are capable of
coupling a small amount of the AC input signal being
monitored.
The coupling induced allows partial
energization of certain integrated circuits within the
relay, causing them to experience indeterminant states.
When the DC coil voltage is restored, the relay circuits
do not consistently power-up in a nornal fashion, and a
false trip signal is generated in about 30 milliseconds.
Following another 30 milliseconds, the trip condition
clears.
The licensee discovered the spurious tripping as part of
followup testing described in Revision 1 to LER 87-23.
The failure of the UV relay occurred when the Division 1
battery charger was lost which resulted in a loss of DC
coil power.
Subsequent re-energization of the coil caused
the 'A' channel RPS power supply breaker to open.
The
unexplained tripping of the breaker by the UV relay
was subsequently resolved by site testing of three spare
UV relays modified with the harmonic distortion module.
Each relay was tested 100 times (i.e. DC coil power
removed and then re-energized) with the following results:
Relay
% of failures (trips when re-energized)
1
none
2
30%
3
50%
The licensee sent the relays to the manufacturer for
more testing to confirm the design problem with the
relays.
-
-
-
-
-
. _ _ . _______ _ ____ - .
e_
i
-l
4
22
. . .
L
The inspectors verified that the spurious trips of RPS
'
breakers by UV relays do not pose an operational safety
concern since the failure is safe and occurs only upon loss
of and restoration of DC coil power. The licensee committed
.
to a supplement to LER 87-23 to update the status of
replacement of the UV relays.
The inspector had no further
concerns.
5.5.2
Containment Penetration Overcurrent Protection
!
On August 14, 1987 the licensee notified the NRC per 10 CFR Part 21 that redundant overcurrent protective devices
were not provided for several conductors that pass through
,
primary containment penetrations.
This requirement for a
second circuit breaker was identified in design
,
calculations, but was not later embodied in all design and
f
construction documents resulting in only single protection
5
on several penetrations.
.
,
Plant design modification 5573 was implemented during the
first refuel outage to provide N d up overcurrent
i.
protection for the affected circuits. The inspector had
'
'
no further questions on this item.
5.5.3
Post-LOCA Radiation Monitor Cable Resistance
,
!
The manufacturer of the Unit 1 po n-accident radiation
'
monitors, General Atomic (GA) Technologies, Inc., reported
!
a deficiency in the Rockbestos cable insulation resistance
,
at elevated temperatures for the subject moritors
,
installed in the drywell. The deficiency was reported to
'
the NRC on February 23, 1987 under 10 CFR Part 21. The
i
monitor's signal coaxial cable located inside primary
[
containment and supplied by Sorrento Electronics (a
i
'
subsidiary of GA), was found to have insufficient
,
insulation resistance (three megohms per 1000 lineal feet)
-
.
at elevated temperatures.
'
!
accuracy requirements specify a minimum cable insulation
resistance of 500 megohms.
j
i
!
!
The licensee evaluated the four Unit 1 monitor configurations
'
and, using the longest cable run, analyzed the worst-case
error introduced for post-accident condition radiation
monitor readings because of high leakage current due to
cable dielectric losses.
In a memorandum to the Limerick
Station Manager from Electrical Engineering dated August
t
17, 1987, the results of design analysis number 5506 were
l
'
presented which indicated the following:
j
,
,
!
!
!
.
!
I
!
-
- - -
-
-
-
-
-
---
- - -
.
23
'
-
The worst case error caused by signal cable leakage
-
current is 31%
-
The Limerick Emergency Plan Alert setpoint for
post-accident containment radiation levels is 100
rad / hour; a 31% error at high temperature would be
within the required factor of two accuracy
At temperatures below 340 degrees but above 245
-
degrees,"the error introduced is less than 30%.
Below 245 degrees, the -nnired accuracy is
-
maintained in all cas
monitor indications are
valid for all radiati.
1es.
Based on the conclusions o
- > analysis, the licensee
justified continued operativa with the monitors until the
cable could be replaced during the next outage. Operator
Aid 26-24 was approved by the PORC and issued on August
20, 1987 at main control room panel number 10C600.
TM
aid graphically depit.ts the conditions under which
containment radiation readings may be invalid (i.e.
ce
radiation fields up to 35 rad / hour conincident with
drywell temperatures between 240 and 340 degrees F). The
inspectors noted that these conditions would not be
theoretically expected to be present except for the
initial two to ten minutes of a design basis event
(reference FSAR Section 6.2).
The inspector had no further questions.
6.0 Surveillance Testino
6.1 Test Observations
The inspector observed the performance or reviewed the results of
the following tests:
ST-3-107-790-1; Control Rod Scram Timing
--
ST-6-107-590-1; Daily Surveillance Log
--
ST-6-011-203-0; 'A'
Loop ESW Valve Test
--
ST-6-011-206-0; 'B'
Loop ESW Valve Test
--
ST-2-049-613-1; Nuclear Steam Supply Division 1 Functional
--
--
ST-6-092-311-1; Monthly D-11 Diesel Run
ST-6-011-231-0; Loop A ESW Pump, Valve and Flow Test
--
The tests were observed to determine that surveillance procedures
conformed with technical specification requirements; testing was being
performed in accordance with Administrative Procedures A-43 and 47;
proper administrative controls and tagouts were obtained prior to
testing; testing was performed by qualified personnel in accordance
with approved procedures and calibrated instrumentation; test data
and results were accurate and withir, techn' cal specification limits;
and equipment was properly returned to service following testing.
.
24
...
No unacceptable conditions were noted.
6.2 Diesel Exhaust Leakage
The inspector witnessed the licensee's conduct of the monthly
surveillance test for the D-11 emergency diesel generator.
Successful conduct of this test includes:
pre-lube of the engine
for two to three minutes; starting and reaching rated speed and
voltage within ten seconds after the cranking cycle is initiated;
and then loading the generator to at least rated load in no more
than 200 seconds after synchronizing with the grid.
Engine
operation is then continued at rated load for not less than one
hour.
6.2.1
Initial Condition
Prior to conducting the test the engine was inspected for
evidence of oil leakage at the flanges of the transition
piping section from the engine exhaust gas lines to their
respective turbochargers since oil seepage from the
flanges has caused smoke in the diesel cubicles when the
exhaust lines become hot.
Routine smoke alarms from diesel
exhaust manifcid leakage when an engine is initially started
were stated to be a concern since these could mask a true
fire smoke alarm.
There have never been any actual fires
in diesel cubicles at Unit 1.
The exhaust flanges showed evidence of oil darpness on the
engine side but no accumulation of oil was evident.
There
was no evidence of oil on the engine except for some minor
amounts of oil in the engine cy Mcder galley in the
vicinity of the fuel injectors.
This was considered
normal.
6.2.2
Engine Operation
During the pre-lube cycle and prior to engine start, there
was no evidence of oil. Within the first minute of
starting the engine, oil leakage began at the lower
transition flange and continued at a rate of approximately
one drop per second for about five minutes.
The flanges
on the opposite side showed no evidence of leaking;
however, the upper flange was covered with loose asbestos
wrap.
Small amounts of oil were observed to drop from the
air start distributor onto the turbocharger piping.
._.
..
.-
-
h
..
25
. . .
>
During this time the exhaust piping began to get hot and
the-oil at the flanges and on the exhaust piping began to
smoke' heavily.
The exhaust piping reaches temperatures of
up to 900 degrees F.
Leakage from le flanges stopped
within about ten minutes of engine operation. - Large
-amounts of oil vaporized from both sides of the engine
during the first five to ten minutes of loaded operation.
After about 15 minutes of operation the smoking stopped,
when all of the oil in the exhaust line had evaperated or
was being :onsumed within the exhaust.
6.2.3
Smoke Ef fects and Fire Hazard
The licensee routinely experiences smoke alarms from the
diesel oil leakage.
TSe smoke in the diesel cubicle
created a hazy cloud throughout the room.
The inspector
confirmed that the pre-lube pump came on, as demanded, and
also shut off properly such that pre-lube was discontinued
automatically along with an engine start.
The inspector noted substantial oil leakage in the hot
manifold area (in the vicinity of the Woodward Governor and
the cylinder galley) could cause a fire.
6.2.4
Vendor Recommendations
The inspector reviewed the manufacturer's operating
instruction manual and operating instructions. Operating
Procedure S-92.1.0 was reviewed and the inspector found
t ht.
in all matters relating to pre-lube, starting,
running, loading and unloading of the diesel, the
licensee's procedures were in agreement with the
manufacturer's recommendations.
6.2.5
Engine Repairs
Because of the oil leakage from the exhaust line, the
licensee elected to replace the gasket. All bolts were
verified to have been torqued properly in the previous
installation.
There are two bolts holding each of the
four-inch flanged exhaust lines to the manifold.
Each
exhaust line is four to eight feet long, with a built-in
expansion bellows.
The pipes are curved to provide for
fit-up from the cylinder exhausts to the turbocharger
- _ _ _ _ _ _ _ _ _ _ _ _ _ . _ - _ - _ _ _ _ - _ - - _ _ - _ _ _ _ . . - _ .
. _ _ _ _
_ -
.__ _ _ _ _ _ _ _ _ _ _ _ _ _
. _ _ _ _ -
.
26
..
The pipe with the leaking flange was removed and
approximately one-half pint of lube oil was found in the
piping.
Examination of both the pipe and manifold flanges
showed about a three-inch arc length where the leak had
occurred.
The gasket showed obvious evidence of crushing.
in the arc area where the leak occurred.
The licensee's analysis of the gasket failure will. include
the amount of force (against the gasket) to ensure proper
bolt torque. When the exhaust pipe heats up from ambient
to operating temperature it expands, causing additional
pressure on the gasket (in spite of the bellows
compression). The pressure would be uniform all around
the flange and gasket if the force on the pipe were
perpendicular to the flange / gasket face. However, the
configuration of these exhaust lines appear to cause the
force to be concentrated on one side of the gasket which
would eventually destroy the gasket and permit leakage.
~
The other section of exhaust pipe was removed and no
evidence of leakage or gasket damage was evident.
The
section of the exhaust pipe was angled from the cylinder
exhaust and did not have any oil accumulation. Upstream
of the connecting piping, there was approximately one inch
of oil accumulation.
6.2.6
Conclusions
It appears that, during engine shutdown, lubricating oil
from the upper crankshaft accumulates at the cylinders
(between the pistons) and then below to the exhaust lines
where it lays until it either weeps out through the
gaskets or until the engine is.run again and it burns out.
Lube oil in the exhaust line would not be a problem if itL
did not weep out since after the engine exhaust heats up
the lube oil is burned away.
The licensee is considering manually turning-over the engine
or air-rolling it after shutdown, such that the lubricating
oil pumped to the upper engine is drained to the crankcase
rather than being allowed to stand in the upper engine and
drain into the exhaust manifold.
The problem is applicable
to the Fairbanks Morse opposed piston design and is not
unique to Limerick.
The licensee is initiating the formation
of a Fairbanks Horse nuclear diesel owners' group and has
proposed addressing the issue of exhaust manifold flange
leakage. This item is unresolved pending resolution of the
exhaust leakage problem at the turbocharger inlet flanges
(50-352/87-31-02).
__ ____ _____ _____ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _
_ _ ___ _ _ _ _ _ _
. _ _ _
_ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ .
. _ _ _ _ _ _ _ _ _
.
27
..
.
6.3 Improper RCIC Restoration
The inspector reviewed the circumstances which resulted in a
violation of technical specifications which require that inoperable
isolation actuation instruments be placed in the trip condition
within one hour. On January 15, an instrument and control (I&C)
technician performed surveillance test ST-2-049-613-1 which, in
part, functionally tests the Reactor Core Isolation Cooling (RCIC)
system pipe routing area high temperature isolation instrumentation.
During restoration from this test, thermocouple leads were
terminated incorrectly (the positive and negative leads were
reversed) and an independent verification also failed to identify
this condition.
On the two shifts following the I&C surveillance test, the
operations personnel recognized a change in the indicated
temperature for the affected instrument during the performance of
the daily surveillance log, procedure ST-6-107-590-1.
The
instrument continued to provide an on-scale indication, however,
with the leads reversed, the temperature reading was approximately
30 degrees lower than actual area temperature.
The third shift
questioned the change in the temperature reading and initiated
troubleshooting which discovered and corrected the thermocouple
Test procedure ST-2-049-613-1 was reviewed and found to have
adequate directions for restoration and verification of the leads
which are lifted to perform the test.
The cause of the occurrence
appears to have been inattention to detail on the part of the I&C
technicians.
The independent verification of restoration program at
Limerick has been good in the past, and this appears to be an
isolated incident.
The daily surveillance log is utilized to perform instrument channel
checks. A channel check is the qualitative assessment of channel
behavior during operation by observation; however, specific
quantitative acceptanc e criteria are not provided in the procedure.
The inspectors concluded that additional guidance in the procedure
would be beneficial to the operators, and allow an easier and more
reliable comparison of similar readings when attempting to determine
t
equipment operability during the performance of a channel check.
The licensee is evaluating other utility methods and procedures used
to accomplish channel checks.
The inspectors will review the
proposed corrective action for this issue in LER No. 88-01.
,
L
_
.-. __.
.
_ - .
_ -_
_
_ _ _ _ _
...
28
... .
7.0 Maintenance
The inspector observed selected maintenance activities on safety related
equipment to ascertain that:
the work was conducted in accordance with
Administrative Procedures A-25, 26 and 27 using r.pproved work
instructions or. procedures; proper equipment permits and tagging were
applied; craft performing the work were appropriately qualified and
supported; and return-to-service of equipment included adequate
post-maintenance testing and operational verification.
7.1 Work Observation
Portions of the following work activities were observed or reviewed:
--
MRF No. 8701102; HPCI Turbine Lube Oil and Filter Change
--
MRF No. 8780076; HPCI Lube Oil System Inspection
--
PMQ No. 056-022; HPCI 011 Tank Clean and Refill
PMQ No. 056-028; HPCI Lube Oil Filter Replacement
--
No unacceptable conditions were noted.
7.2 Battery Charger Malfunction
During observation of maintenance repairs to the Division 3 battery
ch&rger on December 4, 1987 reported in NRC Inspection Report
50-352/87-28, the inspector discussed the potential use of a spare
charger should battery inoperability exceed eight hours.
Limerick
FSAR Section 8.3.2 describes two spare battery chargers which are
provided to replace any of the Class 1E chargers as direct
replacements.
The spare chargers are briefly noted in the FSAR description of DC
system independence, as addressed by NRC Regulatory Guide 1.6,
'
because there are no design provisions for transferring loads
between redundant DC systems.
The inspector noted that, in an
argument similar to the potential use of a third offsite AC power
source (refer to NRC unresolved item 85-30-01), the use of a spare
charger as a qualified DC power source should involve consideration
of seismic, security, fire and other design issues.
The licensee
developed a request to engineering on January 7, 1988 to redesign an
existing modification (number 412) which will install a spare
400-amp charger for DC channels A and B, and a spare 75-amp charger
.for DC channels C and D.
The requested redesign will evaluate
seismic mounting, channel separation and AC supply independence.
The inspector had no further concerns.
7.3 RCIC Overspeed Trip Reset
During the performance of RCIC turbine trip test, ST-2-050-606-1, on
January 20, the turbine trip could not be reset.
Test engineering
investigation revealed that the overspeed '. rip mechanism had
.
. _ _ _ _ _
_
,
_ __._ . ,-.__._-~- _ _ __ _ _
, _ _ , _ _ _ . _ -. _ _ .
_.
...
29
,. .
actuated.
The trip was reset, however, during efforts to repeat the
incident, trip / throttle valve HV-50-112 could not be remotely reset
from the control room and had to be manually closed locally.
This was the result of a worn component on the trip / throttle valve.
As of the end of the inspection period, maintenance engineers were
developing a repair method.
The suspected cause of the initial overspeed trip is that the trip
mechanism linkage may have traveled a portion of the way towards the
trip position (possibly as a result of being bumped by someone
working in the area).
The subsequent mechanical. shock associated
with the electrical trip caused the mechanical overspeed trip
linkage to fully activate.
7.4 Diesel Control Switch
During the performance of the weekly diesel D-14 surveillance test
ST-6-092-314-1 on January 19 a. lack of governor control was
.
experienced from the main control room.
The D-14 engine was
'
declared inoperable pending an engineering evaluation.
The test was
satisfactorily completed after the contacts were cleaned on the
multi-contact local / remote switch.
On January 22, temporary circuit
alteration (TCA)-1234 was applied to ensure governor control from
the control room pending switch replacement. However, if local
governor control was required the TCA would have to be removed.
Therefore, a change to the safe shutdown procedure SE-8, Appendix D,
was issued which directs removal of this TCA if this safe shutdown
methed is needed. An operator aid was posted at the local / remote
switch to reference the change to SE-8 should it become necessary to
i
remove the TCA.
8.0 Common Unit Issues
'
The inspectors addressed the following issues for applicability and
effect upon Unit 1.
The issues were identified by either NRC or licensee
inspections of Unit 2.
The issues include potential significant
!
construction deficiences currently under review for Unit 2.
The
inspectors ascertained that:
the licensee's QA programs are addressing
,
Unit 1 applicability; Unit 1 station management and operating staff are
!
aware of the potential concerns; and the safety significance and proposed
resolution are appropriately determined.
8.1 Valve Overcurrent Reversal
!
The inspector reviewed Unit l's potential significant deficiency
report (SDR) number 85 identified in September 1983 by the licensee
but subsequently determined to be not reportable on January 3, 1984.
,
l
l
,
{
--
-.
.
_
.-
. .
_. .
.
.*.
30
,.
,
The deficiency was associated with preoperational testing of motor
operated valves (M0V's) in the reactor water cleanup (RWCU) system.
The RWCU MOVs being tested were containment isolation valves which,
when travelling in the open direction when an automatic closure
signal was simulated, tripped and failed in a partially open
condition.
The failure was a result of the instantaneous reversal
which overrides the open signal and causes a high current that trips
the 480 volt circuit breaker feeding the valve operator. The
licensee's conclusion that this defic!ency was not significant was
based on a redundant isolation valve 11 series with the affected
MOV, and on the low probability that both valves would be opened
simultaneously. While not considered reportable, the licensee did
replace the RWCU isolation valve breakers with a larger breaker and
changed breaker overcurrent settings to twice the locked-rotor
current of the motors.
RWCU valves HV44-1F001 and 004 have 7.8 HP
-
motors which normally draw 11.4 full-load amps and have locked-rotor
ratings of 94.3 amperes.
The adjustable circuit breaker overcurrent
settings were changed to 190 amps and the adjustments were
documented on Startup Field Report Nos. 61A-13 and 41 approved on
April 30, 1984.
The Startup Field Reports suggested that the ramifications of the
overcu,' rent reversal problem be evaluated for other high speea
isolation valves. While that recommendation was addressed for the
other RWCU valves, the licensee could not confirm whether or not the
issue was expanded to include other Unit 1 isolation-valves.
The inspector discussed the issue of overcurrent trip of MOV's with
,
the Limerick Project Manager and Supervising Field Engineer.
Selected Unit 1 MOV data for the locked rotor current and actual
,
breaker instantaneous overcurrent settings were being made available
at the end of the inspection period.
This item 1s unresolved
pending further review (50-352/87-31-03).
8.2 DC Motor Control Centers
The licensee reported a potential Unit 2 significant construction
deficiency to the NRC Region I office on January 22, 1988.
Internal
wiring on three direct current (DC) motor control centers (MCC's)
was found to be different from the Westinghouse wiring diagrams.
The licensee corrected a number of wiring errors during Unit 1
preoperational testing (specifically ' blue-tag' testing).
The
inspector reviewed selected field modification control (FMC)
packages and Unit 1 Rework Notices for DC equipment, and concluded
that the problem, while not reported as a Unit 1 construction
deficiency, nad been recognized and corrected prior to Unit 1
operation which began in December 1984.
In fact, because of the
'
Unit 1 DC MCC experiences, the subject potentiai Unit 2 deficiency
was initiated. Also, because of the relative simplicity and limited
. .
, _ _
____-_______ __ _ _______ -
, , ,
4
31
.
,
.
number of DC MCC circuits, the wiring errors were easier to identify
and correct than analogous AC wiring errors.
The inspector had no
further concerns, and identified no violations.
9.0 Nuclecr Review Board Organization
The Nuclear Review Board (NRB) is organized as an independent advisory
group dealing with reactor safety at the licensee's nuclear stations.
The NRB reports to a Senior Vice President on a regular basis and to the
Office of the Chief Executive Offi.:er on a quarterly basis, as a minimum.
In addition, the Chairman of the NRB meets directly with the Chairman of
the Nuclear Committee of the Board of Directors (NCB) and reports to the
Board at least once annually.
The inspector noted that the licensee's Nuclear Review Board charter
(Revision 10) was recently revised.
To support NRB activities, the
licensee has issued the following NRB procedures:
-
NRB-1, Review Practice
NRB-2, Determination of an Unreviewed Safety Question
-
-
NRB-3, Audits
NRB-4, Teleconferencing
-
-
NRB-5, Use of Consultants
The inspector reviewed those procedures and selected the following NRB
meeting notes for evaluation:
-
No. 8302L, Plant Manager's Report
-
No. 8415L, Report on Radiation Protection
-
No. 8414L, Special Items
-
No. 8507L, Review of OEAC Minutes
-
No. 8403L, Review of Item List
-
No. 8501L, PORC Review
'
-
No. 8506L, Current Licensing Issues
-
No. 8701L, Review of Licensee Event Reports
No. 8702L, Review of Violations
-
No. 8413L, Quality Assurance
-
-
No. 8505L, Review of Safety Evaluation
I
The inspector selected meeting notes 8501L, 8201L and 8702L for detailed
l
review.
The reports were well written.
It appears that the NRB members
posed in-depth questions to arrive at the stated conclusions.
Items that
require further action are documented in an open items status list which
is maintained by the NRB Chairman's assistant.
This list is planned to
!
be added to a computer tracking system in the near future.
!
l-
NRB agenda items and the results of NRB meetings appear to be well
documented, with historical data available.
The NRB Audit Program for
l
1988 is presently in a management review cycle and was not available
j
during this inspection period.
l
l
The inspector had no further questions at this time.
l
__
_ _
_
-1
..
1
-
...
32
,. .
10.0 Assurance of Quality
This area of evaluation, recently added to the NRC's Systematic
Assessment of Licensee Performance (SALP) process, assesses activities at
all leveh of the licensee's organization that are integral in assuring
the quality of Unit 1 operation.
Inspection findings that are notable
examples wherein quality work is being either maintained or hampered, as
found in previous details of this inspection report, are summarized
below.
10.1 hternal Panel Wiring Errors
The licensee completed an inspection of the remainder of the Power
Generation Control Complex (PGCC) panels after an initial inspection
had identified several deficiencies in which inactive leads were not
properly insulated and secured.
The initial inspaction was
parformed to determine if any miswirings, similar to those
ident1fied on Unit 2, were present in the Unit 1 panels.
The li.;ensee's latest inspection was expanded to include 1352
connections and identified minor discrepancies.
However, no actual
incorrect wirings have been found during Unit 1 inspections.
NRC
Inspectio- n'oort 50-352/87-28 had stated that the wiring problems
in the Unis > PGCC panels were identified on August 31; however, a
review of records indicates the findings actually occurred on
September 23, 1987.
Changes have been made to the Temporary Circuit Alteration (TCA) and
Troubleshooting Control Form (TCF) procedures, A-41.1 and A-42, to
add caution and sign-off steps which ensure all available
documentation (i.e. wiring diagrams and schematics) as well visual
inspect.,ns of as-built wiring configurations are used to identify
any labeling discrepancies prior to implementing the TCA on the TCF.
The licensee has not made a determination on what further actions
will be taken to resolve the wire labeling problems which were
identified on Unit 1.
The item is being tracked as a PORC
commitment and the inspectors will follow the licensee's resolution
in a future inspection.
10.2 Fitness for Duty Training
The inspector observed a portion of an Industrial Substance Abuse
Supervisory Awareness Program which was presented to Bechtel Company
supervisors.
fhe training was presented by a consultant under
contract to PECo. A wide range of topics were covered including the
physical effects of drugs and alcohol on the body, how to recognize
individuals using drugs or alcohol, and how to proceed if you
suspect a worker is coming to work under the influence of drugs or
-
. - . - - -
. - - - , - , _ - , , _ _ - - - _ _ _ - _ - _ - -
.\\
, . ,
33
alcohol. A demonstration of dogs trained in drug detection was also
given.
The drug dogs were onsite to perfcrm a sweep of the Unit 1
and 2 areas as a method of detecting and deterring onsite drug.use
or possession.
These actions demonstrate an aggressive, multifaceted program
designed to eliminate onsite drug use at Limerick.
10.3 RCIC Test Restoration
As discussed in Detail 6.3, there was an apparent breakdown in the
independent verification of restoration procedures during the
performance of a RCIC surveillance test.
Proper system restorations
are an important quality attribute and, although this has not been a
recurrent problem, the verification was not adequately performed.
It is noted that the licensee's policy is to avoid lifting leads
wherever possible but the inspector also noted minimal, if any,
Quality Control involvement in observation or surveillance of such
activities.
10.4 LER 87-23 Attention to Detail
The original investigation of LER 87-23 included the separate
question as to why the RPS channe) A power supply breaker tripped
open due to an undervoltage (UV) relay shunt trip signal during.
re-energization of Division 1 DC power on June 11, 1987. Although
not the primary reason for reportability in LER-87-23, the
licensee's field engineers did pursue the cause of the spurious UV
relay operation with unsuccessful attempts to duplicate the trip.
Subsequent testing of spare relays in July 1987 resulted in
repeatable shunt trips upon re-application of DC control voltage,
and the licensee then contacted the relay manufacturer to determine
root cause.
Vendor testing in November 1987 found a subtle coupling
effect caused by wiring configurations which will be corrected with
a minor modification.
The recent test data was reported in a
revision to LER 87-23 and a Part 21 report (see Detail 5.5.1) to the
NRC on December 22, 1987.
The status of replacement of the relays
will be reported in a future supplement to LER 87-23.
The attention to detail and pursuit of root cause in this case is a
good example of the licensee's field engineering expertise and
management.
The technical understanding and level of attention to
the RPS power supplies has continued to be very high, and has
ensured good reliability for the RPS.
.
,.
.
-.
._.
_.
___
.
__
__
.._
.__
_-
$
..\\.
34
,.
111.0 Management Meetings
11.1. Preliminary Inspection Findings-
The NRC resident inspectors discussed the issues in this report
throughout the inspection period, and summarized.the findings at an
exit meeting held with the Superintendant of Operations on January
25, 1988. At the meeting, the licensee's representatives indicated
that the items discussed in this report did not involve proprietary
information.
No written inspection material was provided to
licensee representatives during the inspection period.
11.2 Regional Specialist Inspection
The inspectors attended an exit meeting for the following NRC
specialist inspection conducted at Unit 1 during the period:
Dates
-Subject
Inspection No.
Lead Inspector
January 1?-21
Emergency
.
88-01
C. Gordon
Preparedness
l
!
.-
-
. . . - . - - . - - - - -
. . - . - - - - - . . - . - . - . . - - - . . . . - - - . -