ML20206J495
| ML20206J495 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 06/19/1986 |
| From: | Forney W, Wohld P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20206J433 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-2.F.1, TASK-2.F.2, TASK-TM 50-483-86-10, IEIN-86-003, IEIN-86-3, NUDOCS 8606270181 | |
| Download: ML20206J495 (18) | |
See also: IR 05000483/1986010
Text
. - -
.-
.
.
I
U. S. NUCLEAR REGULATORY COMMISSION
REGION III
L
Report No. 50-483/86010(DRP)
Docket No. 50-483
License No. NPF-30
!
Licensee: Union Electric Company
-
l
Post Office Box 149 - Mail Code 400
St. Louis, M0 63166
Facility Name: Callaway Plant, Unit 1
Inspection at: Callaway Site, Steedman, M0
f
Inspection Conducted: April 15 through May 31, 1986
Inspectors:
B. H. Little
!
'
C. H. Brown
M. R. h
O.L.
Wohld
4/n/n
U
Date
LO.L.
Approved by:
W. L.
orney,
lef
4/n/H,
Reactor Projects Section 1A
Date
Inspection Summary
Inspection on April 15 through May 31, 1986 (Report No. 50-483/860101',DRP))
Areas Inspected: A routine unannounced safety inspection by the resident
inspectors and one Region III inspector of licensee actions on previous
inspection findings, licensee event reports followup, followup on regional
requests TMI NUREG-0737 items closure, inspection of licensee events, monthly
surveillance, operational safety verification, monthly maintenance, Cycle 2
startup.
Results: Two unresolved items relating to EQ of Limitorque Valves discussed
in Paragraph 4.
Two violations were identified in Paragraph 6., failure to
maintain intermediate head safety injection operable, and failure to notify
the NRC within four hours of an event.
$$f2{ggj{ ha
g3
0
.
O
DETAILS
1.
Persons Contacted
D. F. Schnell, Vice President, Nuclear
S. E. Miltenberger, General Manager, Nuclear Operations
- G. L. Randolph, Manager, Callaway Plant
C. D. Naslund, Manager, Operations Support
A. P. Neuhalfen, Manager, Quality Assurance
- J. D. Blosser, Assistant Manager, Operations & Maintenance
- J. R. Peevy, Assistant Manager, Technical Services
P. T. Abbleby, Assistant Manager, Support Services
W. F. Powell, Assistant Manager, Materials
M. E. Taylor, Superintendent, Operations
D. E. Young, Superintendent, Maintenance
W. R. Robinson, Superintendent, I&C
R. R. Roselius, Superintendent, Health Physics
V. J. Shanks, Superintendent, Chemistry
J. A. Ridgel, Superintendent, Radwaste
G. J. Czecchin, Superintendent, Planning & Scheduling
W. H. Sheppard, Superintendent, Outages
J. M. Price, Superintendent, Training
G. R. Pendergraff, Superintendent, Security
J. E. Davis, Superintendent, Compliance
D. W. Capone, Manager, Nuclear Engineering
W. R. Campbell, Assistant Manager, Nuclear Engineering
A. C. Passwater, Superintendent, Licensing
T. H. McFarland, Superintendent, Design Control
R. D. Affolter, Superintendent, Systems Engineering
D. C. Poole, Consultant
W. H. Stahl, Supervisor, Engineering
- B. K. Stanfield, Assistant Engineer
- S. Petzel, Engineer
^W. R. Bledsoe, Engineer, Compliance
- Denotes those present at one or more exit interviews.
In addition, a number of equipment operators, reactor operators, senior
reactor operators, and other members of the quality control, operations,
maintenance, health physics and engineering staffs were contacted.
2.
Licensee Actions on Previous Inspection Findings (92701)
(Closed) Open Item (483/84-48-01(DRP)):
Licensee plans for modification
of the pressurizer power operated relief valves (PORV).
Based on problems
identified during testing of the PORVs at Wolf Creek Generating Station,
and the similarity of the Callaway and Wolf Creek Units, Westinghouse
issued a Field Change Notice (FCN) No. SCPM 10712 for Callaway to inspect
and record measurements of annular orifice gaps formed by the valve body
2
-
-
-
- -
-
-
-
-
-
- .
.
.
_
-
.
._
_
_
.
.
.
.
and valve cage.
Valvemodification(machiningofthecagerib)includedissubject
1
to inspection findings of inadequate clearances. The licensee
this FCN in Callaway Modification Package CMP-84-0651A, and accomplished
the work during the Cycle 1 refueling outage.
The licensee's inspection determined that machining of the valve's cage
and 54997. quired.The work was accomplished under Work Request Nos. 54996
rib was re
The inspector reviewed the applicable WRs including the
quality control inspection records.
The inspector determined that the
specified work was completed and the valves were successfully tested.
thelice)nseeagreed(483/85006-01(DRSS)):to calibrate their gaseous and
(Closed Open Item
During inspection 50-483/84-16,
,
with appro)riate gaseous and liguid sources during the first refueling
outage.
T1ese calibrations would be in addition to the solid source
calibrations currently performed.
The monitors involved include the
radwaste building monitor GH-RE-10B, the unit vent
and the liquid radwaste effluent monitor HB-RE-18. gas monitor GT-RE-21B,
.
Inspection in this matter included the review of Callaway vendor procedures
HTP-ZZ-04154,iological and Chemical Technology, Inc.HTP-ZZ-04155 and
issued by Rad
The licensee's calibration of the gaseous and liquid effluent monitors
has been completed.
I
(0 pen)UnresolvedItem(483/85007-01(DRS)):
Closure surveillance testing
of normally closed check valves that perform a safety function in the
,
closed position.
Per a )revious agreement
50-483/86012()RS))thelicenseep(asreportedinInspectionrovided
'
Report No.
a
related check valves at the Callaway Plant for inspector review.
A
preliminary, onsite review of the check valve list was done in an attempt
to categorize each valve as tested or not and to determine at least one
,
!
obvious closure requirement for each valve.
The results of this review
indicate that there are a number of check valves not being closure tested
j
that should be.
'
In the listing total of 225 safety related check valves, the review
results were as follows:
)
There are 143 check valves not tested for closure
l
8 - Prevent reverse flow through an idle pump in parallel pump
1
combinations.
4
22-Isolateseismic/non-seismicpipeboundariesordifferentpipeclasses
j
(suchasthosewhichdefinetheLOCA/non-LOCApipeboundaries).
8 - Direct auxiliary feedwater flow to the proper steam generator during
postulated accident conditions.
,
i
3
!
,
_ -
-
- . - .
-.
.
.
10 - Prevent overpressurization of auxiliary feedwater pump suction
piping.
4 - Prevent blowdown from one interconnected steam generator to another
during steam fault conditions.
23 - Had no apparent safety related closure requirement.
,
l
25 - Did not appear to be properly listed as safety related (apparent
listing error) or the v11ve is not used (internals removed).
43 - Were not identified either as having or not having a closure safety
function during the inspection because their function was not
apparent from a cursory drawing review.
There are 65 valves tested for closure
10 - Are normally open check valves which are tested per the ASME Code,
'
Section XI, closure stroke test requirements.
31 - Are tested to provide a high/ low piping interface isolation
(includes WASH 1400, Event V valves).
<
14 - Are leak tested for containment isolation.
10 - Are tested per the concerns raised by the inspector under this
unresolved item.
}
Others (17)
I
2 - Are visually inspected for " free swing" on a periodic basis.
13 - Are fire protection system check valves which are covered by
separate fire protection requirements and inspections.
2 - Are essential service water check valves that are not tested but the
i
failure of which would be readily identified by the affect on normal
j
system operation.
The inspector emphasized that the onsite check valve review was only a
" quick look" and that an in-depth evaluation is needed by qualified
system engineers.
The licensee agreed to perform and document a review
i
of the check valves on the list for their required closure functions and
!
closure test requirements.
The staff indicated that this would be
i
available for inspector review in approximately 90 days.
No violations or deviations were identified.
l
3.
Licensee Event Reports (LERs) Followup (92700)
Through direct observations, discussions with licensee personnel, and the
review of records, the following LERs were reviewed to determine that the
I
,
l
4
,
.
-
-
-
.
.
I
events were documented and evaluated, reportability requirements were
'
fulfilled, and appropriate corrective measures had been implemented.
(Closed) LER 85-010-00:
Reactor Trip on Partial Loss of Feedwater Flow.
On February 21, 1985, a reactor trip and associated actions occurred due
to low steam generator water 1e wis.
The low water levels occurred when
power was secured to a feedwater control panel which in turn resulted in
one main feedwater pump shutting down.
The remaining feedwater pump does
not have the capacity for 100% reactor power.
The loss of power to the
control panel was due to the failure of the supply transformer. The
transformer was replaced and a plant recovery was made.
The transformer
was found to contain foreign material and a procedure was issued to
inspect and clean the transformers on an 18-month :chedule.
The transformer
maintained the design safety function of isolation; therefore, a Part 21
report was not issued.
(Closed) LER 85-021-00:
Inadequate Seismic Qualification of Class IE
Batteries.
On April 4, 1985, the licensee was notified of a potential
problem relative to the spacing between the Class IE batteries and the
battery racks.
The immediate corrective action was to insert spacers
between the end cells and the battery rack end stringers.
The licensee
and contractor evaluated the corrective action and considered the spacers
a permanent fix.
(Closed) LER 85-023-00:
Inadvertent Engineered Safety Features Actuation.
At three different times, April 13, April 17, and May 6, 1985; inadvertent
containment purge isolation and control room ventilation isolation signals
were received.
The cause was a faulty vacuum transducer in a radiation
monitor. With a joint effort between the monitor vendor and several
plants that had experienced similar problems, a more reliable transducer
was developed and installed.
The modification has apparently solved the
problem.
(Closed) LER 85-025-02:
Intermediate Range Hi Flux Reactor Trip.
On
l
May 6, 1985, a reactor trip and associated actions occurred during a
l
reactor startup (power 0% and in Mode 2) due to a intermediate range high
flux signal.
The spurious signal was caused by a fuse blowing in the
neutron monitoring channel.
The vendor determined that a faulty switch
caused the fuse to blow.
The switch replacement appears to have solved
the problem, but the failure mode of the switch and other possible causes
that would overload the fuse are still being evaluated.
(Closed) LER 85-026-00:
High Negative Flux Rate Reactor Trip.
On June 7,
1985, a reactor trip occurred from 100% due to a high negative flux rate.
l
The negative rate was generated due to rod drop when four rod control
pnwer supplies failed (thyristor bank insulator failure) during trouble-
shooting an immovable control rod.
The failed equipment was replaced and
the review shows the failure to be an isolated occurrence and no further
I
action was to be taken.
The immovable control rod was due to a loose
terminal screw.
5
-
--
-
-
-
-
- - -
-
-
-
- - - -
.
.
(Closed) LER 85-042-00:
Inadvertent Reactor Trip.
On October 2, 1985, a
reactor trip from 100% power occurred due to personnel error.
Instrument
and Control personnel were performing test ISP-BB-0T002, "RTD Calibration
Verification" (a one time test) when an abnormal resistance reading was
found on the No. 4 RCS loop, the loop was in test condition, and trouble-
shooting was commenced.
Loop No.1 RTD terminal block test point was
erroneously taken from the prints. When the troubleshooting process
checked these terminals a signal was induced in loop No. 1 RCS protection
which completed the 2 out of 4 logic for the reactor trip from over
Temperature Delta T and Over Power Delta T signals.
The failure to notify
the shift supervisor before commencing troubleshooting was a failure to
follow plant policy procedures.
The error in reading the print contributed
to the event.
The test procedure did not allow troubleshooting if a
problem was located.
The test procedure also had an error which resulted
in the abnormal resistance reading.
The personnel involved were counseled
on procedure and policy compliance.
No citation was issued since under the Enforcement Policy this was
considered a Technical Specification violation of lesser severity which
was identified and satisfactorily corrected by the licensee, and no
further violations of a similar nature have occurred.
This item is
considered closed.
(Closed) LER 85-043-00:
Technical Specification Hourly Firewatch Patrol
Missed.
The Technical Specification 3.7.10.2 requires hourly firewatch
patrols to be established within one hour.
On October 3, 1985, the hourly
patrol was not established for one hour and twenty-five minutes in the
south electrical cable chase due to a misunderstanding of the firewatch
personnel.
The firewatch personnel were retrained on T/S requirements for
the firewatch patrols.
Also the operations personnel, if possible, verify
the patrol is established before taking a fire protection system out of
service.
No citation was issued since under the Enforcement Policy this was
considered a T/S violation of lesser severity which was identified and
satisfactorily corrected by the licensee, and no further violations of a
similar nature have occurred.
This item is considered closed.
(Closed) LER 85-045-00:
Technical Specification 3.7.10.2 Violation Due
to Personnel Error. On October 16, 1985, a portion of the sprinkler
system for the auxiliary building 2000 feet elevation cable trays was
not identified as inoperable during surveillance testing.
Therefore, a
continuous firewatch per Technical Specification 3.7.10.2 was not
established for about 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br />, although a hourly firewatch did patrol
these areas for this time period.
The delay was due to electricians and
engineers failing to communicate and recognize that sprinkler alarms were
also present on the multiplexer that was being worked.
The sprinV er
system was inoperable due to a failed " supervision actuation module"
which indicated the loss of the ability to actuate the pre-actuation
sprinkler system in these areas.
6
.
- - -
.
.
.
The surveillance procedure MSE-KC-FW001, " Technical Specification (T/S)
Fire Detection Functional and Supervisory Operability Test", has been
revised to include the alarms that should be actuated or cleared during
the applicable steps of the procedure.
The maintenance department was
provided a set of the electrical prints for the fire protection systems
and the electricians received training on the fire protection system.
The fire detection in these areas, which provide an alarm in the control
room, were functional during the time the above sprinkler system was
No citation was issued since under the Enforcement Policy this was
considered a Technical Specification violation of lesser severity which
was identified and satisfactorily corrected by the licensee, and no
further violations of a similar nature have occurred.
This item is
considered closed.
(Closed) LER 85-047-00: Operation with a Condition Prohibited by Technical
Specifications.
On October 18, 1985 at 1145 CST, the plant entered
Technical Specification 3.0.3 due to both centrifugal charging pumps
(CCPs) being inoperable.
"A" CCP had been taken out of service for
maintenance and later "B" CCP's room cooler fan was discovered to have
broken drive belts.
The "A" CCP was made operable in less than an hour
and preparations for a plant shutdown were suspended.
The belts were
replaced the following day.
This was considered to be caused by equipment
failure.
No citation was issued since under the Enforcement Policy this was
considered a Technical Specification violation of lesser severity which
was identified and satisfactorily corrected by the licensee, and no
further violations of a similar nature have occurred.
This item is
considered closed.
(Closed) LER 85-050-00:
Inadvertent Engineered Safety Features Actuation.
On November 27, 1985, a control room ventilation isolation and a
containment purge isolation occurred.
The cause appeared to be a fuse
failure at the microprocessor for the containment purge radiation monitor
and a tripped breaker supplying power to the monitor's flow pump.
The
containment was not being purged at the time and the redundant monitor
remained operable. Troubleshooting found no equipment damage.
The fuse
was replaced and the breaker closed with no further problem.
(Closed) LER 86-008-00:
Technical Specification Violation.
On April 3,
1986, while the plant was in Mode 5 (Cold Shutdown), the licensee
determined that Train "A" Control Room Emergency Ventilation System
(CREVS) had been inoperable since March 18, 1986.
The "A" Train became
inoperable when the air conditioning unit was deenergized to permit
inspection and repair of the CREVS fire dampers.
Technical Specification
(T/S) 3.7.6, Action Statement reouires that when in Mode 5 or 6; "With
'
one Control Room Emergency Ventilation System inoperable, restore the
inoperable system to OPERABLE status within 7 days or initiate and
maintain operation of the remaining OPERABLE Control Room Emergency
Ventilation System in the recirculation mode".
7
_ _ _ _ _ _ _ _ _ _ _ _ _ - _ _
,
_-.
.
.
On April 3, 1986, in response to a work request by maintenance personnel
to commence work on the fire dampers, control room personnel found that
the air conditioning unit ("A" Train) had been tagged out since March 18,
1986, without having placed the "B" Train CREVS in the recirculation
mode.
This action should have been taken by March 25, 1986.
Control
room personnel immediately placed the "B" Train CREVS in the recirculation
mode and documented the violation on Incident Report No.86-096.
The licensee's evaluation of the event determined the root cause to be
operations personnel's initial failure to correctly assess the
" operability" impact of tagging out the air conditioning unit.
Consequently, the "A" Train was not declared " inoperable" (no entry in
the equipment out of service logs).
To prevent recurrence, a procedural change was written to require an
independent review of WPAs, Equipment Out of Service Logs, Temporary
Modifications and Locked Components for each ascending mode change.
Also, personnel involved were re-instructed concerning T/S operability
requirements.
This event has been included in the licensed operator
requalifi stion program, " Lessons Learned".
The inspector determined that, once identified by the licensee, action
was promptly taken to correct the condition and report the violation.
During the period that the air conditioning unit was out of service,
control room temperatures were maintained below 84 degree F (the
temperature specified in T/S 4.7.6.a.).
The Train "B" CREVS and Train
"A" pressurization and filtration systems were operable and would have
,
protected control room personnel from airborne contamination if needed.
'
The event posed no threat to public health and safety.
i
No citation was issued since under the Enforcement Policy this was
considered a Technical Specification violation of lesser severity which
'
was identified and satisfactorily corrected by the licensee, and no
further violations of a similar nature have occurred. This item is
considered closed.
(Closed) LER 86-015-00:
Auxiliary Feedwater Actuation System (AFAS):
!
When PK51 Feeder Breaker Was Inadvertently Tripped.
On April 23, 1986,
while in Mode 1 with reactor power at 31%, the control room received
numerous alarms on the annunciator boards.
Letdown and makeup to the
volume control tank was lost, excess letdown and AFAS was initiated.
Plant conditions were determined to be stable, and no reactor trip
occurred.
l
The licensee control room personnel's initial investigation determined
that a non-vital power breaker PK51 had been manually opened.
Based on
no work having been authorized involving PK51, the event was considered
as possible tampering.
The control room personnel promptly notified the
shift security supervisor and the plant manager.
After hearing public
address (PA) instructions for the shift security officer to contact the
control room, the NRC resident inspector responded by going to the
control room.
4
8
- -
-
_
.
..
. - _
..
.
,
The inspector observed licensee's immediate response to the event which
included additional security measures being implemented.
The inspector
was advised of licensee's planned investigation in this matter.
On April 24, 1986, a radchem technician acknowledged to the licensee that
he had inadvertently opened breaker PK51 on April 23, 1986.
The
technician was a contract employee during the recent Cycle 1 refueling
outage and following the outage he was hired by Union Electric.
The NRC senior resident inspector and Region III security specialist
interviewed members of the licensee's staff, including the technician
involved in the event.
The inspectors were satisfied that the information
provided by the technician, with regard to operating switch PK51, was in
agreement with control room observations during the event.
The licensee's investigation determined that the event resulted from
unauthorized operation of breaker PK51.
The technician's action, although
well meaning, highlighted existing weaknesses in the licensee's "new hire"
indoctrination program.
Specifically, organizational interface,
departmental authority / responsibility, and administrative controls on work
and safety practices.
The licensee has implemented a radwaste " Indoctrination Checklist for New
Personnel", and revised the general employee training program to assure
responsibility / authority for equipment operation is stressed. This
incident will be included in the licensee's requalification training
program.
The inspector determined that the licensee response was prompt and
thorough and that action has been taken to prevent recurrence.
No other violations were identified other than those noted above that
were identified, reported and corrected by the licensee.
4.
Followup on Regional Requests (92701)
a.
Temporary Instruction (TI) 2515/75, " Inspection of Limitorque Motor
Valve Operator Wiring"
An inspection was performed to ascertain the environmental qualifi-
cation (EQ) of wiring used in Limitorque Motor Valve Operators.
The
inspection included the following:
Physical inspection of Limitorque operator wiring to determine
what wiring is actually installed in the operators.
Review of licensee's environmental q0alification documentation
to ensure qualification of wiring is sdequately established.
Review of licensee's action relative to IE Information Notice (IN) 86-03, " Potential Deficiencies In Environmental
Qualification of Limitorque Motor Valve Operator Wiring".
9
"
-
_
-.
__ _
_ _ - _ _ _ _
.
.
t
Physical Inspection: On May 22, 1986, the inspector performed an
in plant inspection of four Limitorque Motor Valve Operators.
The
'
inspector selected the below listed valves based on a review of
'
SNUPPS Final Safety Analysis Report (FSAR), Table 3.11(B)-3,
" Identification of Safety-Related Equipment and Components".
Three
valves selected are located in the reactor containment building and
one is located in the lower piping penetration room of the reactor
auxiliary building.
The inspector was accompanied by representatives
from licensee's Quality Assurance, Quality Control, Engineering, and
Maintenance Departments.
VALVE NO.
DESCRIPTION
Safety Injection Tank Outlet Isolation Valve "B"
Safety Injection Tank Outlet Isolation Valve "C"
BB-HV-8037B
Pressurizer Relief Tank (PRT) Outlet Isolation Valve
EM-HV-8835
Safety Injection Discharge to Cold Leg Injection
Isolation Valve
Inspection Findings:
VALVE NO.
- TERMINAL WIRING
- FIELD WIRING
Raychem Flamtrol
G34
'
C02, C03, C04
G2A
Raychem Flamtrol
G34
CO2, C03, C04
G2A
BRAND Rex #43
BB-HV-8037-B
Rockbestos
G31
Firewall III
C07
EM-HV-8835
Raychem Flamtrol
G31
CO2, C04
- Each motor operator contained some terminal and field wires which
either lacked identification markings or with unreadable markings.
However, these wires were similar (size / color) to other wires
identified by markings.
In addition, limit switch space heater
wiring was unmarked, approximately 20 AWG size.
SNUPPS Report of
Independent Review of EQ Programs (Response to NUREG 0588) states:
"In all cases, the limit switch space heater is connected in a
Class IE circuit.
Since the heater failure mode will result in
an open circuit, it is considered that the heaters need not be
qualified.
However, Limitorque has performed an accident test
10
.
.
.
on a heater to demonstrate that the heater remains operative following
seismic aging and a simulated LOCA (i.e., it would not fail in a
manner detrimental to plant safety.)
Review of Licensee's EQ Documentation
The inspector reviewed the below listed EQ Test Reports and determined
that the operator wiring in the four cperators inspected had been
environmentally qualified.
EQ TEST REPORT NOS.
TERMINAL WIRING
'
Franklin Institute Research
Raychem Flamtrol
Laboratories Test Nos. E-031.2, E-031.3
Rockbestos Test Report Nos. E-057-020-03,
Rockbestos
E-057-021-06, E-057-036-02, E-057-050-02
Firewell III
FIELD WIRING
Rockbestos Test Report Nos.
CO2, C03, C04,
E-057-020-03, E-057-021-06,
C07
E-057-036-02, E-057-050-02
ANACONDA Test Report No.
G2A, G31, G34
E-58-0005-03
BRAND Rex Test Report No.
BRAND Rex
E-0578-0014-02
The inspector reviewed licensee plant walkdown sheets to determine
if licensee's identification of operator wiring was in agreement
with the inspectors findings for the four operators inspected. The
licensee's quality records of plant walkdowns included inspections
performed through Startup Work Requests (SWRs) and Quality Control
(QC) checklists.
The licensee records were in agreement with the
inspector's findings.
Licensee's Response to IE Information Notice 86-03
In January 1986, the licensee performed a Quality Assurance (QA)
surveillance on EQ of Limitorque Motor Valve Operator wiring (QA
Surveillance Report No. P8601-12).
The surveillance determined that
each operator was field inspected prior to initial plant startup
using a Startup Work Request (SWR).
QA sample inspection of
approximately 10% of the SWRs determined that the operator wiring
was environmentally qualified.
Union Electric Nuclear Engineering (UENE) in response to Information Notice 86-03 performed a review of all SWRs relating to the field
inspection of Limitorque Valve Operators. This review identified
six valves as having suspect internal wiring.
Nuclear Engineering
11
. -.
--
.-
-.
.- -
. - _ __ .
- =
.__
__
__
_
-
. _ . _. _ . _ _ _ . .
__.
.
.
memorandum No. 558, dated March 11, 1986, requested site engineering
to perform a field inspection to identify the installed wiring. The
results of the field inspection was as follows:
VALVE NOS.
WIRING
EQ STATUS
BB-HV-8000A
Rockbestos
Qualified
BB-HV-8000B
Raychem
Qualified
EJ-HV-8701A
Raychem
Jualified
j
EJ-HV-8716A,B
Not Identifiable
Questionable
EJ-HV-8809B
Techbestos 14 AWG
Questionable
4
600-V
,
The licensee has replaced the operator wiring associated with
EJ-HV-8716A,B and EJ-HV-88098 with environmentally qualified wiring
l
and has requested Westinghouse response regarding the EQ status of
the wiring removed.
The inspector determined that the licensee was responsive to IE
4
Information Notice 86-03 and took prompt corrective action to ensure
,
that the installed operator wiring is environmentally qualified.
i
-
The licensee's evaluation of the EQ status of the wiring replaced is
1
in progress. This matter is unresolved pending further NRC review.
j
Unresolved Item No. 483/86610-01(DRP)
Motor Operated Valves (MOV) Conduit Seals
On May 28, 1986, the licensee advised the inspector that an
+
'
engineering department review of construction documentation was
unable to establish that all required containment MOV conduit seals
were in place.
The licensee performed the review-in response to
4
!
conduit seal deficiencies identified at the Wolf Crc.
Plant.
The
'
licensee determined that although work authorizing documents had
been issued, there was no sign offs for work accomplishment.
,
!
On May 23, 1986, while the plant was in Mode 5 (Cold Shutdown), the
licensee issued Work Request Nos. 60511 through 60519 to install
'
containment MOV conduit seals in accordance with Bechtel Drawing
M-2Y007 (Conduit Seals for Containment MOVs). This action was taken
,
to assure existing EQ status of the MOVs prior to pending plant
startup.
The licensee stated that external visual inspection could
not readily verify conduit realing in accordance with the design
drawing, as drawing M-2Y007 requires or does not' require the use of
i
'
sealant depending on the actual field routing of the conduit.
The
licensee's evaluation of this matter is continuing and plans to
,'
perform a field inspection during the next shutdown.
The licensee reviewed Bechtel Drawing M-2Y007, SNUPPS Report of
,
Independent Review of EQ Programs, and WR Nos. 60511 through 60519.
The inspector also interviewed licensee maintenance personnel that
-
performed the conduit sealing work.
Based on this review,.the
inspector determined that the containment H0Vs conduit seals have
been installed in accordance with Bechtel Drawirg M-2Y007.
However,
12'
,
-- -
---
-, e
-, ,,
+--r
-.n
,
e-,o
~
..-~,r.
-,-
n
,r
-,--,-m.-es,-,e-,--+--v
cn-r
,
~ - , , , - - -
.
.
the acceptability of conduit seals prior to May 23, 1986, could not
be determined.
This matter is unresolved pending further NRC
review.
Unresolved Item No. 483/86010-02(DRP)
b.
Temporary Instruction (TI) 2515/77, " Licensee Response to Selected
Safety Issues (Biofouling of Cooling Water Heat Exchangers)
An inspection was performed to assess licensee's programs for
detection and prevention of biofouling of cooling water heat
exchangers.
The inspection included a review of applicable
procedures and interviews with licensee's maintenance, chemistry,
engineering, and operations personnel.
Prior to the initial startup, the licensee detected tube damage due
to biofouling in the main generator hydrogen coolers which was
attributed to stagnant water conditions prior to plant startup.
No
additional biofouling has been experienced. The licensee has
implemented procedures; ETP-ZZ-03002, " Performance Testing of Plant
Heat Exchangers" and ETP-ZZ-03003, " Monitoring of Plant Heat
Exchangers".
The licensee also maintains 1 PPM chlorine in the
service water system as a preventive measure.
The licensee provides
procedures and operator training relating to degraded heat exchanger
performance.
No violations or deviations were identified.
5.
TMI NUREG-0737 Items Closure (92705)
The following TMI NUREG-0737 line items are considered to be closed:
II.B.3.3
II.B.3.4
II.F.1
II.F.2A
II.F.2B
II.F.2C
A review of Inspection Report Nos. 84-10(DRMSP), 84-16(DRMSP),
86004(DRSS), and others was made and discussions were held with the
applicable inspectors to verify that these line items were ccmpleted.
The ittm identification is included here as a correlation for NUREG-0737
tracking system as the previous closeouts were for the Safety Evaluation
Report tracking system or other numbering systems.
j
No violations or deviations were identified.
6.
Inspection of Licensee Events-Inoperable Intermediate Head Safety
Injection (IHSI) System (92700)
4
a.
Background
l
13
{
-
..
-
.
.
.
_ - -
,.
.
.
.
On April 13, 1986, the licensee advised the senior resident
inspector that on April 12, 1986, while in Mode 3 (Hot Standby), the
plant was placed in a condition prohibited by Technical
Specification (T/S) 3.5.2, when the IHSI system was inadvertently
isolated.
The inspector was given a copy of Incident Report No.
86.109 which documented the violation.
The inspector was also
briefed on the event, the cause and immediate action taken, and of
licensee's planned investigation in this matter.
An inspection was performed to assess the event, root
cause/ contributing factors, and licensee corrective measures.
The
inspection included a review of event reports, operating logs,
administrative and surveillance procedures, personnel interviews,
and meetings with licensee management.
b.
Inspection Findings
On April 12, 1986, at 0402 CST, the safety injection (SI) cold leg
isolation valve EM-HV-8835 was closed to perform surveillance test
OSP-EP-V0003 (Section XI Accumulator Safety Injection Valve
Operability). Technical Specification (T/S) Limiting Condition for
Operation 3.5.2 specified that the IHSI system be operable in Mode
3.
T/S Surveillance Requirements 4.5.2 specifies that EM-HV-8835
(Safety Injection Cold Leg Isolation Valve) be open.
EM-HV-8835,
being closed, isolated the common discharge path from both SI pumps
to the cold leg injection, putting the plant in a condition
prohibited by Technical Specifications.
On April 12, 1986, at 1010 CST, the reactor operator, while taking
the required daily leg readings, observed that EM-HV-8835 was
closed.
The reactor operator immediately informed the shift
supervisor (S/S).
The S/S declared both SI Trains inoperable and
entered T/S 3.0.3 and had valve EM-HV-8835 opened.
The S/S issued
Incident Report No.86-109 documenting the violation.
On May 7, 1986, the inspector met with the licensee to assess
licensee's investigation, evaluaticn of cause, and corrective
actions regarding the IHSI system isolation.
The licensee discussed
their findings of root cause and contributing factors and of
corrective action taken and planned.
The cause of the event was
attributed to personnel scheduling and performance errors as
follows:
Scheduling
Scheduling Personnel - Scheduled the performance of OSP-EP-V0003,
"as required in Mode 3 prior to RCS pressure reaching 1000 psig".
Compliance Personnel - Identified OSP-EP-V0003 on an attachment to
the Mode 3 Change Letter, "to be performed in Mode 3 as conditions
permit".
I
14
. - _ _ _
_ _ _
.
.
The task performance review considered the operability requirement
of the safety injection accumulators but failed to recognize the
Surveillance Task Sheet (STS) task performance mode requirements,
which specified Mode 4 only.
c.
Performance
Operations personnel erroneously authorized and performed
OSP-EP-V0003 in Mode 3.
Several errors were made regarding the
authorization and performance as follows:
(1) OSP-EP-V0003 - Initial condition specified performance of the
test in Mode 4.
Operations personnel issued a Temporary Change
Notice (TCN) changing Mode 4 to Mode 3.
APA-ZZ-00101
(Preparation, Review, Approval and Control of Plant Procedures)
provides for temporary procedure changes, "which clearly does
not change the intent".
Management Directive U0 86-69 issued
March 4,1986, reemphasized (s) control of TCNs and identified
that "significant changes to initial conditions are changes
which are changing the intent".
(2) Operations personnel changed the Surveillance Task Sheet No.
ST-00070 Task Performance Mode from " Mode 4 only" to Mode 3.
This change did not receive the required review and approval as
specified in APA-ZZ-00340 (Surveillance Program
Administration).
(3) Operations personnel's failure to be cognizant of the overall
plant effect of closing EM-HV-8835.
(The isolation of a
required safety system).
(4)
In addition to issuing an Incident Report, the licensee classified
the event as a 30 day Licensee Event Report (10 CFR 50.73).
However, a four hour report to NRC Operations Center should
also have been made in accordance with 10 CFR 50.72(b)(2)(iii).
This report was not made.
Failure to notify the NRC within four.
hours is a violation of 10 50.72(b)(2)(iii).
No. 483/86010-03(DRP).
d.
Licensee's Corrective Action to Pravent Recurrence Included:
(1) For future outages, outage scheduling will schedule
OSP-EP-V0003 in Mode 4 as a Mode 3 restraint.
(2) Progressive discipline has been initiated for appropriate
outage personnel.
Outage Planning and Scheduling personnel
,
have been advised concerning outages involvement in this event.
(3) An outage procedure currently in draft form will specifically
address use of the STS " Task Performance Mode" for scheduling
surveillances.
(4) Future mode change letters will reflect only required task
performance conditions and T/S requirements for mode changes.
(5) The TCN that allowed performance of the OSP in Mode 3 was voided.
15
._
--
- -
-
.
.
(6) Progressive discipline has been initiated for operations
personnel involved in this event and the necessity to comply
with programmatic controls has been reemphasized
(7) Management will reemphasize the existing administrative controls
for revisions to task sheets and surveillance procedures to
appropriate plant personnel.
(8) Appropriate personnel will receive guidance concerning reporting
requirements of 10 CFR 50.72.
The inspector determined that the violation, once identified by the
licensee, was promptly corrected, documented, and received a high level
of attention.
Based on the short duration of the violation, plant
conditions of low temperature and pressure with low' stored heat energy
and the availability of backup emergency core cooling systems, the event
posed no significant threat to the public or plant safety.
However, the
event highlighted significant performance errors.
These errors included;
inadequate task reviews, failure to adhere to licensee administrative
procedures, and failure to be cognizant of the overall plant effect
resulting from surveillance testing.
The licensee's failure to maintain the IHSI system " operable" while in
Mode 1, 2, and 3 is a violation of Callaway Plant Technical Specifications
Limiting Condition for Operating 3.5.2.
No. 483/86010-04(DRP).
7.
Monthly Surveillance (61726)
The inspectors reviewed or observed selected portions of Technical
Specification required surveillance testing during power operations and
prior to mode changes relative to the startup from the refueling outage.
Items which were considered during the inspections included whether
adequate procedures were used to perform the testing, test instrumentation
was calibrated, test results conformed with Technical Specifications and
procedural requirements, and the test was performed within the required
time limits.
The inspector determined that the test results ware reviewed
by someone other than the personnel involved with the performance of the
test, and that any deficiencies identified during the testing were reviewed
and resolved by appropriate management personnel.
No violations or deviations were identified.
8.
Operational Safety Verification (71707)
,
The inspectors observed control room operations, reviewed applicable
logs, and conducted discussions with control room operators throughout
the inspection period. The inspector verified the operability of selected
safety related systems, reviewed tagout records, and verified proper
return to service of affected components.
Tours of the reactor, auxiliary,
and turbine buildings were conducted.
During these tours, observations
were made relative to plant equipment conditions, fire hazards, fire
protection, adherence to procedures, radiological control and conditions,
16
__
_ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _
.
s
housekeeping, security, tagging of equipment, ongoing maintenance and
surveillance, containment integrity, and availability of safety related
equipment.
No violations or deviations were identified.
9.
Monthly Maintenance (62703)
Selected portions of the plant maintenance activities on safety related
systems and components were observed or reviewed to ascertain that the
activities were performed in accordance with approved procedures,
regulatory guides, industry codes and standards, and that the performance
of the activities conformed to the Technical Specifications.
The following items were considered during these inspections: the limiting
conditions for operation were met while components or systems were removed
from service; approvals were obtained prior to initiating the work;
activities were accomplished using approved procedures and were. inspected
as applicable; functional testing and/or calibrating were performed prior
to returning the components or systems to service; parts and materials
that were used were properly certified; radiological controls were
implemented as necessary; and, fire prevention controls were implemented.
1
No violations or deviations were identified.
10.
Cycle 2 Startup (61702, 61705 through 61710)
The initial criticality of the Cycle 2 core was observed by the inspector
on April 15, 1986.
The startup was performed per ETP-ZZ-ST002, " Engineering
Test for Initial Criticality".
Selected portions of the following tests /
procedures were observed during-their performance and the results were
reviewed after the evaluation of data was completed.
ETP-ZZ-00007
Reactimeter Dynamic Checkout
ETP-ZZ-ST004
All Rod Out Baron Endpoint
ETP-SR-ST001
All Rods Out Flux Map
i
Moderator Temperature Coefficient Measurement
j
Reactor Coolant Flow Measurements
Incore/Excore Calibration
ETP-ZZ-ST005
Rod Bank Worth Measurement
The moderator temperature coefficient was slightly positive for all rods
out. The rod withdrawl restriction will continue for 4000 MWD /MTV burnup
4
of the core.
The other tests indicated the results were about where they
,
were expected.
The mode changes were observed and requirement check
sheets were reviewed.
Selected requirements for mode changes were
verified to have been performed.
No violations or deviations were identified.
17
l
._
e
..
11.
Unresolved Items
Unresolved items are matters about which more information is required in
order to ascertain.wheter they are acceptable items, violations, or
deviations.
Two unresolved items disclosed during the inspection are
discussed in Paragraph 4.
12.
Exit Interview
The inspector met with licensee representatives (denoted under Persons
Contacted) at intervals during the inspection period.
The inspector
summarized the scope and findings of the inspection.
The licensee
representatives acknowledged the findings as reported herein.
The
inspector also discussed the likely informational content of thi
inspection report with regard to documents or processes reviewed by the
inspector during the inspection.
The licensee did not identify any such
documents / processes as proprietary.
13.
Enforcement Conference
An Enforcement Conference was held on June 3, 1986, at the NRC Region III
office, Glen Ellyn, Illinois between Mr. D. F. Schnell and members of the
NRC Region III staff.
During the meeting the Licensee presented facts
relative to the event on April 12, 1986, discussed in Paragraph 6 above.
The Licensee presented background information, corrective action to
prevent recurrence, and potential mitigating facts which the NRC will use
to determine the appropriate escalated enforcement action.
l
t
i
i
18
i
i
.
-
-
-.
,.
_
.--