ML17221A515
| ML17221A515 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 11/17/1987 |
| From: | Bibb H, Crlenjak R, Wilson B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML17221A513 | List: |
| References | |
| 50-335-87-21, 50-389-87-20, NUDOCS 8711230162 | |
| Download: ML17221A515 (17) | |
See also: IR 05000335/1987021
Text
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UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.W.
ATLANTA,GEORGIA 30323
Report Nos.:
50-335/87-21
and 50-389/87-20
Licensee:
Florida Power and Light Company
9250 West Flagler Street
Miami,
FL
33102
Docket Nos.:
50-335
and 50-389
Facility Name:
St.
Lucie 1 and
2
License Nos.:
DPR-67 and
Inspection
Conducted:
September
6 - October 31,
1987
Inspectors:
r en
a
,
en>or
es
nt
nspector
esi
ent
nspecto
Approved by:
c
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ect>on
>e
Division of Reactor Projects
JJ DV
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SUMMARY
Scope:
This inspection
involved
on site activities in the area of Technical
Specification
compliance,
operator
performance,
overall
plant operations,
quality
assurance
practices,
station
and
corporate
management
practices,
corrective
and
preventive
maintenance
activities, site security procedures,
radiation control activities, surveillance activities, refueling outage
review,
and plant events
review.
Results:
Of the
area
inspected,
two violations,
one with two examples,
were
identified (paragraphs
3 and 12).
87ii230162 8 000335
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REPORT DETAILS
Persons
Contacted
Licensee
Employees
K. Harris, St.
Lucie Vice President
G. J.
Boissy, Plant Manager
"R. Sipos,
Services
Manager
~"J ~
H. Barrow, Operations
Superintendent
T.
A. Diliard, Maintenance
Superintendent
J.
B. Harper,
gA Superintendent
"L.
W. Pearce,
Operations
Supervisor
"R. J. Frechette,
Chemistry Supervisor
"C.
F. Leppla,
I8C Supervisor
C.
A. Pell, Technical Staff Supervisor
E. J. Wunderlich, Reactor Engineering
Super v'isor
H.
F.
Buchanan,
Health Physics
Supervisor
W. White, Security Supervisor
~CD
L. Burton, Reliability and Support Supervisor
J.
Barrow, Fire Prevention Coordinator
R.
E.
Dawson, Assistant Plant Superintendent
-, Electrical
C. Wilson, Assistant Plant Superintendent
- Mechanical
"N.
G.
Roos, guality Control Supervisor
Other
licensee
employees
contacted
included
technicians,
operators,
mechanics,
security force members,
and office personnel.
"Attended exit interview.
Exit Interview
The
inspection
scope
and findings were
summarized
on
November 3,
1987,
with those
persons
indicated in paragraph
1 above.
The licensee
did not identify as proprietary any of the materials
provided
to or reviewed by the inspectors
during this inspection.
Licensee Action on Previous
Enforcement Matters
Withdrawn (Unit 1)
VIO 50-335/85-36-01:
Failure to Have
Two Shutdown
Cooling
Loops Operable
While In Mode With Reactor Coolant
Loops Drained.
A Notice of Violation was
issued
on February
10,
1985, for violating
Unit 1 technical specification (TS) 3.4. 1.4.2,
which requires
that
two
independent
trains of shutdown cooling
be operable
when in mode
5 with
reactor coolant loops not filled.
Specifically, the violation was written
because
the licensee
had taken
one train of component cooling water
(CCW),
which supports
out of service for maintenance.
By a
letter dated July 22,
1986,
the
licensee
stated
that its maintenance
of
the
CCW system
in mode
5 with reactor
coolant
loops not filled did not
violate
Unit 1
TS 3.4. 1.4.2.
By
memorandum
dated .November 7,
1986,
Region II asked
the Office of Nuclear Reactor Regulation
(NRR) to review
the licensee's
arguments
and confirm or provide
a regulatory position.
The inspection
report (50-335/85-36)
documented
that one of two required
shutdown cooling loops
were
because
the heat exchanger of its
respective
CCW,
CCW train
B
was
out of service
for repairs.
The
licensee's
position was that:
1) two shutdown cooling loops were operable
because
a fully operable
system,
with
emergency
diesel
power
available,
was supplying both
shutdown cooling loops with cooling water,
2) because
CCW heat exchangers
are passive
rather than active fluid system
components,
they
do not fall under the single failure criterion for fluid
systems,
and
3)
TS
Amendment
56 incorporated
NRR's
long-term
shutdown
redundancy
requirements
expressed
in the letter dated
June ll, 1980,
from
the Director of the Division of Licensing to the licensee.
In
summary,
acknowledged
in
a
memorandum
to
Region II, dated
September
17,
1987,
that
the definition of "operability" in the
TS is
subject to a wide range of potential interpretations.
For plant operation
in
mode
5, Unit 1
TS 3.4. 1.4. 2 clearly requires
two redundant
shutdown
cooling loops;
however,
no Unit 1
TS clearly required
two redundant
cooling
loops.
Accordingly,
the
licensee
s
actions
in performing
maintenance
work
on
the
CCW system
heat
exchanger
in mode
5 does
not
constitute
a violation of Unit 1 TS.
This violation is withdrawn.
Closed
(Unit 2)
P2187-Ol:
Sorento
Electronics,
a subsidiary
of
GA
Technologies,
reported
in
a letter dated
February 23,
1987,
a possible
defective
co-axial
cable
used
in conjunction with the Regulatory
Guide
1.97,
Post-LOCA High Range Radiation Monitor.
This cable
was found to be
in
use
on St.
Lucie Unit 2.
The Licensee
conducted
a safety analysis.
(JPE-LR-018)
which concluded that the existing radiation monitors with
their
associated
coaxial
cables
are
acceptable
as
installed.
Addi-
tionally,
the radiation
monitors
have
been
shown to
meet the design
criteria for St.
Lucie Unit 2 and are not an unreviewed safety question.
Closed
(Unit 2)
UNR 50-389/87-17-01:
Containment
Not In
Accordance
With Regulatory
Guide
1.63.
A recent
inspection (ref.
Report 50-389/87-17
of Unit 2 identified
a
concern
with respect
to
conformance
with Regulatory
Guide 1.63,
Rev.
2, for the Maintenance
Hatch
Hoist Motor.
In resolving this concern,
the licensee
conducted
a review
of all electrical
circuits
routed
through containment
to
ensure
proper
disposition
in
accordance
with the
Regulatory
Guide.
Additionally, the
scope of the review was
expanded;
by the licensee,
to
include all associated
circuits powered
from safety related
power sources
to verify conformance
with Regulatory
Guide
1 ~ 75,
Rev.
2.
The following
Unit 2
energized
non-essential
circuits penetrating
containment
were
identified by the licensee
as not meeting their commitments to Regulatory
Guides
1.63 or 1.75:
a.
Maintenance
Hatch Monorail Hoist (BKR-41381),
b.
Reactor Building Jib Crane Receptacle
(BKR-42151),
c.
Power Receptacle
257,
261,
265 and 271 (BKR-41278),
d.
Power Receptacle
227,
232,
258,
262 and 266 (BKR-42044),
e.
Power Receptacle
259,
263,
267 and 270 (BKR-41379) and
f.
Power Receptacle
260,
264,
268 and 269 (BKR-42148)
g.
Reactor Building Telescoping
Crane
(BKR - 42152),
and
h.
Reactor Building Elevator Starter
(BKR - 42149)
il
Additionally, the
above
listed
energized
circuits
do not comply with
license conditions
2.C. 10 and 2.C. 11, which state:
Non-Safet
Loads
on Emer enc
Power Sources
Section 8.4.2
Prior to startup
following the first refueling outage,
the licensees
shall
implement the design modification to disconnect four-kilovolt
loads
on detection
of
a safety injection signal
and provide
two
isolation
devices
in series
for those
non-safety electrical
loads
that are
not disconnected
by
a safety injection signal
or loss of
offsite power.
Containment Electrical Penetration
(Section 8.4.3
SSER
3
Prior to startup
following the first refueling outage,
the licensees
shall
complete
the
design
modifications
to provide
independent
primary
and
backup fault protection for each electrical
conductor
penetrating
containment.
The failure to meet the license
conditions
2. C. 10
and
2. C. 11, for the
above
non-essential
circuits
penetrating
containment
is
a violation
(50-389/87-20-02).
The licensee
has
completed
a safety evaluation for the non-conformances
discovered
during their reviews.
The safety evaluation
concluded that the
non-conformances
are
not
a substantial
safety hazard,
and as
such are not
reportable
under
Included
in the
licensee's
report,
are
recommended
actions for final disposition of these
non-conformances,
as
well
as
a safety evaluation
substantiating
these
recommendations.
The
recommended
actions
are to open the circuit breakers
during modes
1-4 for
those circuits identified in the safety evaluation.
Plant Engineering
(Juno) will provide
an engineering
package to modify the
pump circuit
to conform to the
Reg.
Guide 1.75,
Rev.
2.
4.
Plant Tours (Units
1 and 2)
The inspectors
conducted
plant tours periodically during the inspection
interval to verify that monitoring equipment
was recording
as required,
equipment
was properly tagged,
operations
personnel
were
aware of plant
conditions,
and plant housekeeping
efforts were adequate.
The inspectors
also
determined
that
appropriate
radiation
controls
were
properly
established,
critical clean
areas
were being controlled in accordance
with
procedures,
excess
equipment
or material
was
stored
and
combustible
materials
and debris
were
disposed
of expeditiously.
During tours,
the
inspectors
looked
for the
existence
of unusual
fluid leaks,
piping
vibrations,
pipe
hanger
and seismic restraint settings,
various valve and
, breaker positions,
equipment caution
and danger tags,
component positions,
adequacy
of fire fighting equipment,
and instrument calibration dates.
Some tours were conducted
on backshifts.
The
inspectors
routinely conducted
partial
walkdowns of
systems.
Valve,
breaker/switch
lineups
and
equipment
conditions
were
randomly
verified both locally
and in the control
room.
During the inspection
period
the
inspectors
conducted
a complete
walkdown in the accessible
areas
of the Units
1 and
2 component cooling water
(CCW), emergency diesel
generators
and
AC/DC electrical distribution systems
to verify that the
lineups
were in accordance
with licensee
requirements
for operability and
equipment material
conditions
were satisfactory.
Additionally, flow path
verifications
were
performed
on the following systems;
Units
1 and
2
chemical
volume control
(AFW).
The
NRR Project
Manager
(PM) for the St.
Lucie Plant visited the site
September
15-18,
1987.
Me reviewed
a reactor trip (subject of LER 335/
87-13),
external
missile protection,
several
plant change/ modification
packages,
mangrove restoration efforts and conducted
various plant tours.
The
PM reviewed
the circumstances
associated
with the Unit 1 reactor trip
of June
14,
1987.
The
1B main feedwater
pump tripped
and the reactor
subsequently
tripped
on
a high pressurizer
pressure.
The review included
the
licensee
event
report
(LER), 335/87-13,
submitted to the staff by
letter dated July 10,
1987;
the monthly report submitted to the staff by
letter
dated July 15,
1987;
and the post-trip review report
(completed
Operating
Procedure
No.
0030119,
Revision 4,
dated
June
14,
1986).
Additionally, the
event
was
discussed
with the licensee.
An in-depth
review of the event
was
performed to evaluate
the possibility that. the
reactor
tripped
on
low steam
generator
water level,
versus
high pres-
surizer pressure,
and the discrepancy of whether the power-operated relief
valves
(PORV's)
opened
during the event.
The sequence
of events
recorder
indicated that the reactor tripped
on high pressurizer
pressure
followed
five seconds
later by a low steam generator
water level trip.
Apparently,
the pressurizer
pressure
increase
was
dominant over the loss of steam
generator
level.
As
a result of discussions
with the licensee
personnel,
it was
determined
that the time difference
between
the signal
to
open
the
PORV's
and the signal
to close the
PORV's
was less
than two seconds.
Therefore, it would be difficult to determine if they opened
under these
conditions.
However,
the post-trip report stated
that the
PORVs opened
but the
LER stated that they did not.
The licensee
agreed
to resolve the
discrepancy
by either revising the
LER or correcting the monthly report.
The
PM observed
that
some safety-related
equipment,
located outside,
are
not fully protected
from postulated
tornado/hurricane
generated
external
missiles.
The postulated
scenario
involves
a severe
weather condition
at the site,
a missile
being generated
and propelled toward the safety-
related
equipment,
the missile breaking
the
boundary of the equipment,
and the
subsequent
inability of the plant to achieve
cold shutdown, if
required.
Additional assumptions
include
no on-site electrical
power and
multiple missiles.
Equipment
observed
to have
a potential for being not
fully protected,
under certain
conditions,
are
the Unit 1 condensate
storage
tank,
two Unit 1 diesel
fuel oil storage
tanks,
two Unit, 1
component
cooling water
system
heat
exchangers,
and the refueling water
taken for each unit.
After detailed discussion with the licensee,
the
was satisfied with the licensee's
response
and prior actions.
l
The technical
specifications
for both units require that the licensee
conduct
a
beach
survey and mangrove photographic
survey at least
once per
year.
These
surveys
are associated
with flood protection
measures
for the
site.
The results
of the
beach
survey
were submitted to the staff by a
letter dated July 7,
1987.
The results
indicated that the present
dune
condition
is
acceptable.
The
results
of the
mangrove
survey
were
submitted to the staff by letters
dated
March
2 and July 1,
1987.
The
results indicated that there
has
been
some deterioration of the mangroves.
As
a result,
the
licensee
performed
an
engineering
evaluation.
The
licensee
concluded that the mangroves
are not required to maintain design
basis
of the
St.
Lucie site
to protect safety-related
structures
and
equipment
from probable
maximum hurricane
surge
and erosion
damage.
Thus,
the licensee
determined
that the deterioration did not create
a condition
of any safety significance.
The licensee
has
embarked
on
a program to
rejuvenate
the
50
acre
mangrove
tract.
The licensee
has installed
a
piping system
to water the tract.
The piping system
draws water from the
intake
canal
and
pumps it through
PVC pipe to the northern
edge of the
50 acre tract,
where the water is released
along the edge of the tract.
Pooled water in the tract is eventually released
back to the intake canal.
The licensee
believes that much of the 50 acre tract will be restored
in a
few years.
The
PM believes that, although under
no requirement to do so,
the licensee is acting in a responsible
manner to restore
the tract.
A number of Unit 1 plant change/modification
packages
were
reviewed
by
the
PM.
Emphasis
was placed
on the safety evaluation
conducted
by the
licensee.
These
were St.
Lucie Unit 1/Unit 2 security systems
(141-81);
electrical
E-4 nozzle (003-184);
steam line radiation monitor
weather
enclosure
(024-184);
and
environmental
qual ification
update
(077-186).
Unit 2 plant change/modification
packages
were reviewed during
a prior 1987 site visit.
All safety evaluations
were adequate.
It should
be
noted
that
the security
system
plant change/modification
package
containing
safeguards
information was stored in a locked file cabinet,
as
required.
Additionally, the
environmental
qualification
update
plant
-change/modification
package
is
a
good
example of a
change
made
under
10 CFR 50.59 that is not hardware-orientated.
It demonstrates
that the
licensee
envelopes
all
changes
described
in the
FSAR into their change
program,
not just hardware
changes.
The
PM conducted
various tours.
Particular
emphasis
was
placed
on the
reactor auxiliary building (RAB) for each unit.
The
PM toured all levels
inside the
RAB's and the ground level outside the
RAB's and fuel handling
buildings.
All outside
doors
to the RAB's were closed
as required.
All
outside
doors
to the fuel handling buildings were closed
as required,
except for one
door
on the Unit 2 fuel'andling building, which was
open
because
new fuel
was just received on-site
and
was just
moved into the
building.
A guard
was posted
because
the door was
open
and personnel
were
working in the
immediate area.
The licensee
has
had
some problem, in the
past,
keeping
some exterior
RAB doors closed.
The reason for keeping the
doors
closed is if an airborne radiological release
occurred in the
RAB,
an open door would represent
an unmonitored
and unfiltered
escape
pathway,
partially bypassing
the
RAB ventilation cleanup
and filtration system,
Significant improvement
has
been
noted
and housekeeping
was excellent in
both RAB's.
The
PM noted
a possible
discrepancy
in the shield building wall penetra-
tion (no. 57)
associated
with the
containment
mini-purge
system line
(Unit 2).
Apparently,
the piping run went through the penetration,
but
the areas
between
the outside of the line and the inside of the penetra-
tion was
not sealed.
This discrepancy
was discussed
with the licensee.
The licensee
demonstrated,
by utilizing drawings,
that the
seal
between
the outside of the pipe and the inside of the penetration
was
made at the
inside of the shield building wall. 'dditionally, the licensee
stated
that leak tightness
of the annulus
area is periodically checked to ensure
that
there
is
no
unacceptable
leakage.
The
licensee's
explanation
resolved the
PM's concern.
Plant Operations
Review (Units 1 and 2)
The inspectors,
periodically during the
inspection
interval,
reviewed
shift logs
and
operations
records,
including data
sheets,
instrument
traces,
and records'f
equipment
malfunctions.
This
review included
control
room logs
and auxiliary logs, operating orders,
standing orders,
jumper
logs
and
equipment
tagout
records.
The
inspectors
routinely
observed
operator
alertness
and
demeanor
during plant tours.
During
routine
operations,
operator
performance
and
response
actions
were
observed
and
evaluated.
The
inspectors
conducted
random off-hours
inspections
during the reporting interval to assure
that operations
and
security
remained
at
an acceptable
level.
Shift turnovers
were observed
to verify that they were
conducted
in accordance
with approved licensee
procedures.
The inspectors
performed
an in-depth review of the following
safety-related
tagouts
(clearances):
Unit 1
1-9-64
1-9-94
1-10-7
1-10-8
Unit 2
PM On Reactor Trip Circuit Breakers
Pressure
Test - lA Charging
Pump
Pressure
Test -
1C Charging
Pump
Pressure
Test - 1B Charging
Pump
Tag Open Breakers
To Comply With
RG 1.63 and 1. 75
Repair
LCV-2110P
2C Charging
Pump - Adjust Accumulator Pressure
for Mode Change
Technical Specification
Compliance (Units 1 and 2)
During this reporting interval,
the inspectors
verified compliance with
limiting conditions
for operations
(LCO's)
and results
of selected
surveillance
tests.
These
verifications
were
accomplished
by direct
observation
of monitoring
instrumentation,
valve positions,
switch
positions,
and
review of completed
logs
and records.
The licensee's
compliance
with
LCO
action
statements
were
reviewed
on
selected
occurrences
as they happened.
Maintenance
Observation
Station
maintenance
activities of selected
safety-related
system
and
components
were observed/reviewed
to ascertain that they were conducted
in
accordance
with requirements.
The following items were considered
during
this review; limiting conditions for operations
were met, activities were
accomplished
using
approved
procedures,
functional tests
and/or calibra-
tions were performed prior to returning components
or systems
to service;
quality control
records
were maintained;
activities were -accomplished
by
qualified personnel;
parts
and materials
used
were properly certified;
and radiological
controls
were
implemented
as
required.
Work requests
were
reviewed
to determine
status
of outstanding
jobs
and to assure
the priority was
assigned
to safety-related
equipment.
The inspectors
observed portions of the following maintenance activities;
Unit 1
PWO 7608
Pre/Post
Test Calibration of IST Test Gages
Unit 2
PWO 5885
Auto Load Sequence
Test
Review of Nonroutine Events
Reported
by the Licensee
(Units
1 and 2)
The following Licensee
Event Reported
(LER's) were reviewed for potential
generic
impact,
to detect
trends,
and to determine
whether corrective
actions
appeared
appropriate.
Events which were reported
immediately were
also
reviewed
as
they occurred to determine that technical specifications
were
being
met
and that
the public health
and safety
were of upmost
consideration.
The following LER's are considered
closed:
Unit 1
87-13
87-14
Unit 2
Reactor
Trip on High Pressure,
Loss of
1B S/G
Feedpump
(ref.
paragraph
4, this report
and
IE Report
No. 335/87-14,
dated
July 28,
1987)
Unidentified
Leakage
Greater
Than
TS Limits (ref. paragraph
l2, this report).
87-06
2A and
2B Diesel Generator
Automatic Load Sequence
Relays Missed
Surveillance
On September
17,
1987,
the licensee's
gA organization
discovered that the
surveillance
for the
2A and
2B Emergency
Diesel
Generator
(DG) 12 month
test of the automatic
load sequence
relays
had not been properly completed
when
last
performed
(May 4,
1987).
The
review of the
surveillance
procedure
indicated that only the components
on the 4160 volt 2A3 and
2B3
switchgear
were tested.
The test did not include the components
on the
4160 volt
2AB switchgear
and
the
480 volt load centers/motor
control
centers.
Apparently,
the
May 1987
reviews
by individual(s) implementing
the
maintenance
surveillance
procedure
as
well
as
quality control
personnel
given charge of reviewing the plant work order and the surveil-
lance
maintenance
procedure
for'ompleteness
and
compliance
to the
surveillance
requirements
did not identify the incomplete surveillance
procedure.
The
immediate
and corrective actions
implemented
by the
licensee
were;
(1) electrical
maintenance
personnel
satisfactorily performed the required
surveillance
in accordance
with the approved procedure,
(2) plant manage-
ment
have
reemphasized
the
importance of adequate
review of plant work
orders
and surveillance
procedures,
and (3) the surveillance
procedure
was
revised for clarity.
Unit 2 technical specification (TS) 4.8. 1. 1.2.d
requires
that while
operating
in modes
1,
2,
3,
and
4 each diesel
generator
shall
be demon-
strate
by verifying at least
once
per
12
months
that
the
automatic
load
sequence
times
are
The last
12
month
surveillance
of the
automatic
load
sequence
relays
was satisfactorily
performed in May 1986.
As an .immediate action the surveillance
procedure
was performed satisfactorily
on September
18,
1987.
This is approximately
18 days
past
the
TS allowable
25K extension of the testing interval.
The
surveillance
was
completed,
with all
components
testing satisfactorily,
approximately
22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> after discovery of the
incomplete
status
of the
previous surveillance of May 4,
1987.
In summary,
the failure to properly complete
the scheduled
surveillance
of the diesel
generator
on .May 4,
1987
and the subsequent
exceeding
the
allowed
25K extension
is
a violation of Unit 2
TS.
However,
since it
was identified by the licensee,
considered
to be severity level IV or V,
reported,
corrected
within a reasonable
length of time,
and
was not
a
violation which could be expected
to have
been prevented
by the licensee's
previous
corrective
actions,
in accordance
with 10 CFR 2,
Appendix
C,
V.A., a Notice of Violation will not be issued.
Physical
Protection (Units 1 and 2)
The inspectors verified by observation
and interviews during the reporting
interval
that
measures
taken to assure
the physical
protection of the
facility met current requirements.
Areas inspected
included the organiza-
tion of the security force,
the establishment
and maintenance
of gates,
doors
and isolation
zones
in the proper conditions,
that access
control
and badging
was proper,
and procedures
were followed.
Surveillance
Observations
During the
inspection period,
the inspectors
verified plant operations
in compliance with selected
technical
specifications
(TS) requirements.
Typical of these
were confirmation of compliance with the
TS for reactor
coolant
chemistry,
refueling water, tank,
containment
pressure,
control
room ventilation
and
and
DC electrical
sources.
The
inspectors
verified that
testing
was
performed
in
accordance
with
adequate
procedures,
test instrumentation
was calibrated,
limiting conditions for
operations
were
met,
removal
and restoration
of the affected
components
were
accomplished,
test
results
met requirements
and
were
reviewed
by
personnel
other
than
the individual directing the test,
and that
any
deficiencies
identified during the testing
were properly reviewed
and
resolved
by appropriate
management
personnel.
The inspectors
observed
portions of the following surveillance(s):
10
Unit 1
1-0110050 Control Element Assembly Periodic Exercise
1-0010125
Check Sheets
1, 2, 3, 4, 6,
and 10;
and Data Sheets
1, 3,
4,'nd
9
HP-4
Attachment B-Daily Area Surveys
Unit 2
HP-4.2
Health Physics
Weekly Checks
Boric Acid/Safety Injection Tanks - Activity/Concentration
Refueling Outage
Review
During the inspection period,
the inspector
observed
certain
aspects
of
the licensee's
preparations
for refueling of Unit 2 (scheduled,
October
3
thru November 1).
The following objectives
were addressed:
b.
Ascertain
the
adequacy
of licensee
procedures
for the conduct of
refueling operations,
Ascertain
the
adequacy
of the licensee's
administrative
requirements
for control of:
(1)
Refueling operations,
and
(2)
Plant conditions during refueling,
and
c.
Ascertain
the
adequacy
of the licensee's
implementation of controls
for items 2a.
and b.,
above,
The
inspector
reviewed
the following refueling related
procedures
to
ensure
that
the
licensee
had
implemented
controls for the
conduct of
refueling operations
and for establishing
and maintaining control of plant
conditions in accordance
with technical specifications:
~0eratione
2-0030127
2-0030128
2-011022
2-0120021
2-1600022
2-1600023
2-1610020
2-1630021
2-1630022
2-1630023
2-1630024
2-1630025
2-1630028
Reactor Plant Cooldown - Hot Standby to Cold Shutdown
Reactor
Shutdown
Coupling and Uncoupling of Control Element Assembly
(CEA)
Extension Shafts
Draining the Rector Coolant System
Unit 2 'efueling Operation
Refueling Sequencing
Guidelines
Receipt
and Handling of New Fuel
New Fuel Elevator Operation
.
Spent
Fuel Handling Machine Operation
Fuel Transfer System Operation
Refueling Machine Operation
CEA Change Fixture Operation
New Fuel Handling Crane Operation
11
Administration
0005746
Outage
Management
0010438
Control of Heavy
Load Lifts
Maintenance
2-M-0036
Reactor
Vessel
Maintenance - Sequence
of Operations
Instrument'and
Control
2-120054
Low Temperature
Overpressure
Protection Setpoint Verification
~ll
1th Ph
HP-23
Health
Physics Activities in the Reactor
Containment
Building
During Shutdown
HP-40A
Receipt of Radioactive Material
Review of Plant Events
On October 8,
1987, with Unit 1 in mode
1 at
100K power, during
a leak
rate
calculation,
the
licensee
discovered
that unidentified
reactor
coolant
leakage
was
1.08 gpm, greater
than the
TS limit of 1 gpm.
An
unusual
event
was declared
at 5:29 a.m.
and
the
NRC notified (ref.
PN).
Operations
personnel
entered
containment
at 5:30 a.m. to investigate
and
possibly identify the
source
of the leakage.
The containment entry team
reported
leaks in, several
reactor
coolant
pump
(RCP) areas.
At 6: 12 a.m.,
a reactor
shutdown
was
commenced
so that detailed
investigations
could be
conducted
in containment.
At 9: 13
a.m.
the turbine was
removed from the
grl d.
At. 9: 30 a. m.,
another
containment
entry
was
conducted
by
a
team of
operators
and
maintenance
personnel
to reassess
the leaks.
From this
investigation it was discovered that the
1A1
RCP vapor seal
was slinging a
mist of water.
Additionally, the
1A2
RCP vapor seal
was showing signs of
leakage
along with the 1Bl charging
to the
1B1 loop
and safety
injection line to the 182 loop.
Because
the exact sources
of the leakage
could not
be identified the licensee
decided
to remain in an
unusual
event.
The licensee
assigned
a
team to evaluate
the event
and prepare
a plan to
positively identify the leak sources
and plan for necessary
repairs.
At
12:05 p. m.,
the unit was
shut
down and in mode 3.
The unusual
event
was
terminated
at 2:05 p.m.
because
the
leaks
were
now identified with
total identified leakage
less than the
TS limit of 10 gpm.
12
The following sources
of leakage
were identified and repaired:
a+
b.
C.
d.
e.
lAl RCP seal
leak from a cracked weld joint on a seal
nozzle flange,
1B1 loop 2-inch charging line check valve cover plate gasket
leak,
1B2 loop 12-inch safety injection line check valve hinge pin gasket
leak,
2-inch pressurizer auxiliary spray line check valve gasket
leak,
1B1 loop 3-inch pressurizer
main spray line check valve gasket
leak,
and
a minor
PORV discharge
leak.
Maximum unidentified
leakage
recorded
during the event
based
on letdown
system mismatch
was 4.9
gpm.
The failure of the lA1 RCP seal
cracked weld
was probably responsible
for the leak rate going from less
than to greater
than
1
gpm unidentified over
a short period of time (between
normal leak
rate calculations).
All leaks
were repaired
and the unit returned
to
service
on October 18,
1987.
On
October 25,
1987,
while
reassembling
the
Unit 2
reactor
vessel
internals
and performing
a liftof the in-core instrument (ICI) plate onto
the work platform,
the plate
was
deformed
due to excessive
force being
applied
using the polar crane.
The procedure,
2-M-0036,
Reactor
Vessel
Maintenance - Sequence
of Operations,
paragraph
9. 10.36,
requires that the
load cell readout
be continuously
observed
during the lift and that the
free
hanging weight not
be
exceeded
by more than
600 lbs.
Oue to the
confusion caused
by the shift change
and the use of the crane, without the
load cell, for other lifts in containment
immediately following the shift
change,
the
load cell
was
not reinstalled prior to continuing procedure
2-M-0036.
The failure to continuously observe
the load cell readout
when
proceeding
to step
37 of the procedure
resulted in the deformation of the
instrument
plate.
This is
a violation of Unit 2
TS 6. 8. 1, failure to
establish/implement
procedures, first example
(50-389/87-20-01).
As of the
end of the reporting period, the licensee
had completed taking
measurements
to determine
the extent of the ICI plate deformation
and
inspections
of the plate welds.
The plate
was
found to have
a maximum
deformation
on the outer edge or rim of 1 7/8 inches.
The manufacturer's
specification
is
a
maximum
1/2
inch deformation.
Additionally, the
licensee
was in the process
of fabrication the tools, in accordance
with
Combustion
Engineering
recommendations,
necessary
for straightening
the
plate.
The licensee
has
completed
a review of other lifts which had been
performed
previously during the
outage
to ensure'hat
all lifts were
conducted
properly.
No further problems
were identified.
The outage
was
expected to be delayed approximately
seven
days.
On October 28,
1987, the licensee
had removed the Unit 1 turbine generator
from the grid to balance
the generator exciter.
The exciter
had developed
a vibration
problem earlier in the
month.
On October 29,
1987, after
13
successfully
completing balancing of the exciter and returning the turbine
generator
to service,
the turbine operator inadvertently stopped
the only
operating
condensate
pump,
the
1A pump.
Subsequently,
a trip of the main
feed
pump occurred
and at 3:35 a.m., at 20K power,
and the reactor tripped
on low steam generator
water level.
All system functioned
as designed.
The
1A condensate
pump was stopped while electric realigning the operating
condensate
pumps
(from lA and
1B to
1A and
1C operating).
After
completing
the
pump electrical
realignment,
using procedure
1-0700020,
Condensate
and
the operator
encountered
problems
removing one
(of three)
of the interlock keys required for changing
the disconnect
lineup.
The control
room
was
informed of the
problems
and sent
an
electrical
supervisor
to the disconnect
area to provide assistance
to the
operator.
At this point, for reasons
undetermined,
the operator inserted
one of the three
keys into the
1A pump electrical
disconnect
l.ock.
The
interlock was
then activated
which tripped the remaining,
lA condensate
pump.
The interlock prevents
either
opening
an energized
disconnect
or
aligning the
1C
condensate
pump to two independent
power sources.
In
summary,
the
procedure,
1-070020,
was
inadequate
because it does
not
specify
how
or. under what conditions
the interlock keys
may
be removed
after
the
disconnect
lineup is
completed.
Additionally, the operator
failed to follow procedures,
~ in that,
upon
encountering
problems
in
removing
the last
remaining
key,
he inserted
another
key without the
utilization of the
procedure.
This is
a violation of Unit 1
TS 6.8. 1,
failure to establish/implement
procedures,
second
example,
(50-335/87-
21-01).
While recovering
from the event described
above,
on October 29,
1987, with
statup
operations
in progress,
shutdown
rods withdrawn
and Unit 1 not
critical at
5:42 a.m.,
the unit tripped
on
a spurious
channel
B high
startup
rate
(SUR).
All equipment functioned
as designed.
Channel
C
SUR
was tripped
because
of previous
problems with the instrument.
When the
spurious
channel
B
SUR occurred the required
2 out of 4 logic was
made
up,
initiating the
The licensee
repaired
channel
C, placed
channel
B in bypass
and
commenced
a reactor startup.
The reactor
was
critical at 11: 10 a.m.
and the unit placed on-line at 1: 10 p.m. the
same
day.
No further problems
were experienced.