ML17342A155
| ML17342A155 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 07/05/1985 |
| From: | Brewer D, Elrod S, Peebles T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML17342A153 | List: |
| References | |
| 50-250-85-13, 50-251-85-13, IEB-79-18, NUDOCS 8508010782 | |
| Download: ML17342A155 (31) | |
See also: IR 05000250/1985013
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UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.W.
ATLANTA,GEORGIA 30323
Report Nos.:
50-250/85-13
and 50-251/85-13
Licensee:
Florida Power and Light Company
9250 West Flagler Street
Miami,
FL
33101
Docket Nos.:
50-250
and 50-251
Facility Name:
Turkey Point
3 and
4
License Nos.:
and
Inspection
Conducted:
April 8 - May 20
1985
Inspectors:
T. A.
Pee
s
enior
Re ident Inspe
6r
D.
R. Br
esi ent Inspector
Approved by:
p)
n A. Elrod, Section Chief
Divi ion of Reactor Projects
D
S gned
Da
S>gned
D
e
igned
SUMMARY
Scope:
This routine,
unannounced
inspection entailed
361 direct inspection
hours
at the site, including 91 hours0.00105 days <br />0.0253 hours <br />1.50463e-4 weeks <br />3.46255e-5 months <br />
on backshift,
in the areas
of licensee
action
on
previous inspection findings, Inspection
and Enforcement Bulletin ( IEB) followup,
annual
and
monthly surveillance,
maintenance
observations
and reviews,
opera-
tional safety verification, engineered
safety
features
walkdown, plant events,
preparation for refueling and independent
inspection.
Results:
Violations
Failure to implement
procedures
as required
by Technical Specification (TS) 6.8. 1; failure to implement the requirements
of
Appendix B, Criterion XII; failure to implement
the requirements
of
Appendix B, Criterion
XV; failure to implement the
requirements
of
Appendix B, Criterions VII; failure to establish
procedures
as
required
by
TS 6.8. 1;
and faiure to perform adequate
surveillance
per
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F
REPORT DETAILS
1.
Licensee
Employees
Contacted
C.
J.
- D
T.
- K
B.
"H.
D.
E.
D.
J.
"R.
"R.
J.
W.
F.
R.
E.
M. Wethy, Vice President - Turkey Point
J.
Baker, Plant Manager - Nuclear
P. Mendieta,
Services
Manager
Nuclear
D. Grandage,
Operations
Superintendent
Nuclear
A. Finn, Operations
Supervisor
L. Jones,
Technical
Department
Supervisor
A. Abrishami, Inservice Testing
( IST) Supervisor
E. Hartman,
Inservice Inspection (ISI) Supervisor
Tomaszewski,
Plant Engineering
Supervisor
A. Suarez,
Technical
Department
Engineer
A. Chancy,
Corporate
Licensing
Arias, Regulation
and Compliance Supervisor
L. Teuteberg,
Regulation
and Compliance
Engineer
Hart, Regulation
and Compliance
Engineer
W. Kappes,
Maintenance
Superintendent
Nuclear
R. Williams, Assistant Superintendent,
Electrical Maint
H. Southworth, Electrical
Department
Engineer
A. Longtemps, Assistant
Superintendent,
Mechanical
Main
F. Hayes, Assistant Superintendent,
Instrument
and
Control (I&C) Maintenance
enance
tenance
V. A. Kaminskas,
Reactor Engineering Supervisor
R.
G.
Mende,
Reactor
Engineer
R.
E. Garrett,
Plant Security Supervisor
- P.
W. Hughes,
Health Physics
(HP) Supervisor
and Acting Operations
Superintendent
R.
W.
J.
J.
L.
H.
R.
8W
J.
AG
T.
p.
B.
M. Brown, Assi stant
HP Supervi sor
C.
Mi1 1er, Training Supervi sor
J.
Baum, Assistant Training Supervisor
M. Donis, Site Engineering Supervisor
M. Mobray, Site Mechanical
Engineer
C. Huenniger,
Start-up Superintendent
T. Young, Project Site Manager
J. Crisler, Quality Control
(QC) Supervisor
H. Reinhardt,
QC Inspector
Bladow, Quality Assurance
(QA) Inspector
E.
Moaba,
Performance
Enhancement
Program
(PEP)
Manager
W. Hasse,
Safety Engineering
Group Chairman
M. Vaux, Safety Engineering
Group
C. Grozan,
Licensing Engineer
Pace,
Licensing Engineer
C. LaPira, Fire Protection Supervisor
~Attended exit interview
Other
licensee
employees
contacted
included
construction
craftsmen,
engineers,
technicians,
operators,
mechanics,
electricians
and
security
force members.
2.
Exit Interview
The
inspection
scope
and
findings
were
summarized
during
management
interviews
held
throughout
the
reporting
period with the
Plant
Manager-
Nuclear
and selected
members of his staff.
The exit meeting
was
held
on
May 10,
1985,
with the
persons
noted
in
paragraph
1.
The areas
requiring management
attention
were
reviewed'he
six items identified as potential violations were:
Failure to meet the requirements
of TS 6.8. 1, with several
examples
as
follows: failure to implement
procedures
in the
area
of contaminant
exclusion
(paragraph
7), radiation work permit requirements
(paragraphs
9 and ll) and housekeeping
(paragraph
7), (250, 251/85-13-01).
Failure to meet
the
requirements
of
Appendix
B, Criterion
XIII, Handling,
Storage
and
Shi pping
(paragr aphs
7
and
11),
(250,
251/85-13-02).
Failure to meet
the requirements
of
Appendix
B, Criterion
XV,
Control
of
Nonconforming
Materials
(paragraph
ll),
(250,
251/85-13-03).
Failure to meet
the requirements
of
Appendix
B, Criterion
VII, Packaging,
Shipping,
Receiving,
Storage
and Handling (paragraph
11), (250, 251/85-13"04).
Failure to implement maintenance
scheduling
and planning in accordance
with
ANSI
N18.7-1972Property "ANSI code" (as page type) with input value "ANSI</br></br>N18.7-1972" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process.,
section
5. 1.6.3,
as
required
by
(paragraph
7), (250, 251/85-13-05).
Failure to implement the requirements
of
10 CFR 50.55a.(g),
Inservice
Inspection
requirements
for Intake Cooling Water system (paragraph
6),
(250, 251/85-13-06).
Two unresolved
items
(URI) were identified:
adequacy
of actions
taken
as
required
by
IEB 79-18 (paragraph
5), this
was
presented
at the exit as
a
deviation but further inspection
was determined
to
be
necessary
(URI 250,
251/85-13-07);
and the validity of testing auxiliary feedwater
(AFW) check
valves
in
only
the
forward
flow direction
(paragraph
12),
(URI
250,
251/85-13-08).
One
inspector
followup
item (IFI) was identified:
followup corrective
actions for residual
heat
removal
(RHR) area
discrepancies
(paragraph
9),
(IFI 250, 251/85-13-10).
The licensee
acknowledged
the findings.
The licensee
did not identify as
proprietary any of the materials
provided to or reviewed
by the inspectors
during this inspection.
Another exit was held with the Plant Manager-Nuclear
on
May 17,
1985,
and
several
items
were discussed
dealing with the operation
of spent
fuel pit
systems differently than
the safety analysis
reports
(SAR).
The facility
license requires that operation of systems different than
the analysis
and
assumptions
of the
safety
analysis
be
reviewed prior to operation
for
unreviewed
safety questions
per
10 CFR 50 '9.
Operation
of the
systems
without
the
review is
an
unresolved
item,
(paragraph
11).
This
was
presented
as
a violation at the exit but was changed
to
an Unresolved
Item
(URI 250, 251/85-13-09)
pending the completion of special
Inspection
Report
(250, 251/85-23).
The
licensee
acknowledged
the finding.
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
Inspection
Findings (92702)
a.
Monthly update of Performance
Enhancement
Program
(PEP)
The
PEP
was
reviewed
to determine if commitments
were
being
met.
Status
was discussed
with the
PEP
Manager
and with other
members
of
management.
The facility upgrade
project
has
continued
pouring concrete
support
columns for the
new administration building .
The schedule for comple-
tion of the building including the third floor is the
end of December
1985,
but occupancy
is scheduled
for March 1986.
The
new
HP building
is progressing
toward its
May
1985 completion,
with move-in
scheduled
for July 1985.
The schedule
for the
PEP continues
to be met within acceptable
limits,
and all modifications
have
been cleared
by the Region.
b.
Previous
Inspection
Findings
(Closed) Violation 250, 251/84-11-01
(Closed) Violation 250/84-14-05
The inspectors
verified, through visual
observation,
that Administra-
tive
Procedure
(AP)
103.4,
In-Plant
Equipment
Clearance
Orders,
was
being properly and consistently
implemented.
The
use of caution
tags
has
been
implemented;
consequently,
personnel
no longer
use
"danger"
clearance
tags for unintended
purposes.
Indelible ink is used
on tags
exposed to the environment.
e'
(Closed)
IFI 250/84-18-05.
Fire pre-plans
have
been
implemented
and
are available in the control
room for use.
The pre-plans
are, routinely
utilized during fire drills by the shift supervisor (Plant Supervisor-
Nuclear [PSN]).
(Closed)
IFI 250,
251/83-05-01.
The
QC Inspector checklist for plant
work order
review has
been
substantially
incorporated into AP 190. 19,
Control of Maintenance
on Nuclear Safety
Related
and Fire Protection
Systems.
Those checklist
items
not included in the
procedure
are of
concern
to the
Department
as
a
review organization
and
do
not
specifically concern
other plant personnel.
(Closed)
IFI
250,
251/83-24-02.
Administrative controls
have
been
placed
on the
use of lead
seals
in accordance
with AP 103.5,
section
8.4.5, Administrative Control of Locks, Valves and Switches.
The lead
seals
are
only applied
by
a Senior
Reactor
Operator
(SRO) qualified
individual
who
has
been
designated
by the
PSN,
since
a
SRO qualified
individual is
required
to verify proper
valve
position
prior to
attaching
the seal.
The
seal
crimping tool can
be issued only by the
PSN.
The proper
method of positioning the
AFM valves prior to seal
installation is specified
in the procedure.
Independent
verification
is required.
(Open) Violation 250/84-18-07.
On May 15,
1985,
the inspector
reviewed
the
estimated
critical conditions
(ECC) calculations
performed for
Unit 3 reactor
startup
No. 296.
The
ECC calculation
was
based
on
an
average
temperature
of 525 degrees
F;
however,
the actual
temperature
was
535
degr ees
F.
This discrepancy
introduced
an
unnecessary
error
into the
ECC calculation of approximately
100 per cent
millirho (pcm).
Further review revealed that the licensee
had developed
no criteria by
which to determine if the difference
between calculated
temperature
and
actual
temperature
could adversely
affect
the
accuracy.
As
a
corrective
action,
the
licensee
plans
to
incorporate
a
maximum
differential
temperature
limit of five degrees
F.
An
additional
discrepancy
was identified in that the
ECC calculation
sheet
did not
record the date
and time of the shutdown.
This omission
increases
the
possibility that
improper
shutdown
times
could
be
used
in
making
required calculations.
The
licensee
has
developed
and
implemented
a
computer
program that
prints
a reactivity poison report at the time of a reactor trip.
Data
is available for xenon,
iodine and promethi um.
This informa-
tion will improve the
accuracy
of the
ECC calculations.
However, the
licensee
has yet to develop
a
mechanism
to alert
personnel
to the
possible
inaccuracies
in the computer
generated
poison reports that can
result
when the Digital Data Processing
System
(DDPS) computer
has
gone
off-line.
Personnel
are
aware that the computer generated
data
can
be
erroneous
after
a computer malfunction.
They
have
not
been
provided
with guidance
as
to
how
long the
poison
reports
remain
erroneous
following the event.
It is
up to individual discretion
as to whether
the computer
generated
data will be used.
This is unacceptable
because
it introduces
varying errors of unknown magnitude into the
ECC calcula-
tion process.
4.
Unresolved
Items
Unresolved
items
are matters
about
which more information is required to
determine
whether they are acceptable
or whether they may involve violations
or deviations.
Three unresolved
items were identified and are discussed
in
paragraphs
5,
11 and
12.
5
~
IE Bulletin (IEB) Followup (92703)
(Open)
The
inspector
reviewed
the requirements
of IEB 79-18,
Audibility Problems
Encountered
on Evacuation
of Personnel
From High-Noise
Areas,
which was issued
on August 7,
1979.
The results of a previous review
of this bulletin were
documented
in Inspection
Reports
250,
251/84-11
of
March 1984.
In that report the inspector
expressed
concern that
some alarms
and
announcements
were
not
audible
in certain
high
noise
areas.
The
licensee
agreed
to supply additional
information concerning
Plant
Change/
Modification (PC/M) 80-01 which was installed in January
1980 specifically
to address
the
IEB requirements.
The potential failure to fully implement
IEB 79-18 is
an
Unresolved
Item
(250, 251/85-13-07).
6.
Monthly and Annual Surveillance
Observation
(61726/61700)
The inspectors
observed
TS required surveillance
testing
and verified:
that
the test procedure
conformed to the requirements
of the TS; that testing
was
performed
in accordance
with adequate
procedures;
that test instrumentation
was calibrated; that limiting conditions for operation
(LCO) were met; that
test
results
met
acceptance
criteria
requirements
and
were
reviewed
by
personnel
other
than
the individual directing the test;
that deficiencies
were identified,
as appropriate,
and that any deficiencies identified during
the testing
were properly reviewed
and resolved
by management
personnel;
and
that
system
restoration
was
adequate.
For completed tests,
the inspector
veri.fied that testing
frequencies
were
met
and tests
were
performed
by
qualified individuals.
The inspectors
witnessed/reviewed
portions of the following test activities:
Daily Operability Verification of the
Power
Range Nuclear
Instrument
System
(EDG) Operability. Test
High Head Safety Injection
Pump Monthly Operability Test
Component
Cooling Water System Test
Rod Position Indication and Axial Flux Alarm Calibration
g
1
On
May 6,
1985,
a portion of the Unit 3 intake cooling water (ICW) system
piping
was disassembled
for inspection.
The concrete
lining was
in fair
condition with only the joints in need of repair.
The bolts
on the Unit 3
piping
were
being
replaced
due
to
age
and
cor rosion.
During this
inspection,
the
licensee
noted that
the
concrete
thrust
braces
for the
circulating water
pumps,
located in the
same area,
were
damaged
and began
an
inspection
and repair program.
The inspector
observed
that the bolting
on
the Unit 4
ICW piping was
more deteriorated
than Unit 3's
and notified the
licensee.
This
degraded
condition
was
not
licensee
identified
because
during
the ten year hydrostatic test
in
May 1984,
the section
of piping
which runs through the circulating water intake wells was considered
"buried
piping"; therefore,
no inspection for structural
distress
or corrosion
had
been
conducted
as required.
This is
a Violation (250,
251/85-13-06)
as
stated
below.
10 CFR 50.55a.(g)
requires
ASME Code testing for Class
3 components.
Code,
Section
XI (1974 Edition), article
IWD-1000 applies
the
requirements
of Section
IWD to Class
3 pressure-retaining
components.
IWD-2000 requires
inspection
of the
components
each inspection'nterval.
IWD-2600 requires
the visual examination to be conducted of the components
during system tests
for evidence of structural distress
or corrosion.
Contrary to the
above,
during the system inservice testing for the Class
3
ICW system
on Unit 3 in December
1983
and
on Unit 4 in
May 1984,
visual
inspections
for structural
di stress
or corrosion
were
not
conducted
for
piping
and
bolted
connections
which were
located
between
the
ICW
pump
discharge
check valve and the header isolation valves.
Maintenance
Observation
(62703
and 62700)
Station maintenance activities of safety-related
systems
and components
were
observed/reviewed
to ascertain
that they were
conducted
in accordance
with
approved
procedures,
regulatory guides,
industry codes
and standards
and in
conformance with TS.
The following items
were
considered
during this
review,
as appropriate:
LCOs
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;
procedures
used
were
adequate
to control
the activity; troubleshooting
activities were controlled,
and the repair record accurately reflected
what
actually took place;
functional testing
and/or calibrations
were performed
prior to returning
components
or systems
to service; quality'control records
were maintained; activities were
accompl.,i shed
by qualified personnel;
parts
and materials
used
were
properly certified;
radiological
controls
were
implemented;
gC
holdpoints
were
established
where
required
and
were
observed;
fire prevention
controls
were
implemented;
outside
contractor
force activities were controlled in accordance
with the approved
gA program;
and housekeeping
was actively pursued.
The following maintenance activities were observed
and/or reviewed:
RHR pump repair
Reactor vessel
head
stud removal
Hydrostatic testing of
RHR motor operated
(MOV) suction valves,
MOV-750 and
MOV-751
Reactor
upper internals split pin replacement
ICW flange bolt replacement
EDG preventive
maintenance
Unit 3 4160 volt bus breaker
maintenance
The inspectors
observed
the detensioning
of the reactor vessel
head closure
studs
in accordance
with Maintenance
Procedure
(MP) 1407.7.
The procedure
requires that tape
be placed
over the holes in the studs to preclude foreign
material
from entering.
The
work site for the
removal
of the
3A
pump
was
inspected.
Open
flanges in the safety related
component
cooling water
(CCW) supply to the
pump were
not protected
from foreign material
contaminants,
as required
by
section
8. 1. 1.5 of AP 0190. 10
when work was
not in progress
at the
work
site.
The failure to protect safety-related
'systems
and
components
from foreign
material
contamination
discussed
in
the
previous
two
paragraphs
is
a
Violation
(250,
251/85-13-01,
item a).
This violation is
a
repeat
of
previous violations as noted:
Violation 3.c of Inspection
Reports
250/84-34
and
251/84-35;
and
Violation 2.b
of Inspection
Reports
250/84-22
and
251/84-23.
Maintenance
work areas
were routinely observed
in the Units
3 and
4 cask
wash areas
and
on May 8, the Unit 3 new fuel storage
area.
In each location
an accumulation of waste
and debris
was not removed at the end of the work
shift
and
equipment
was
not properly
stored
as
requi red
by
AP 103. 11,
Housekeeping.
The failure to maintain housekeeping
standards
is
a Violation
(2SO,
251/85-13-01,
item c).
This violation is
a
repeat
of
a
previous
violation:
Violation 2.c of Inspection
Reports
2SO/84-22
and 251/84-23.
On
May 9,
1985,
during repair of the Unit 3
CCW system,
a frayed sling was
used
to hoist
a section of piping into position.
Quality Procedure
(QP)
13. 1,
section
5.4.2,
which
implements
Topical
Quality
Report
13.0
and
Criterion XIII of 10 CFR 50, Appendix B, specifies that frayed slings shall
not be used.
This
QP was not further implemented
on site.
The failure to
implement
Criterion XIII of
10 CFR'0,
Appendix
B
is
Violation
(250,
251/85-13-02,
item c) .
An additional
example of thi s violation is discussed
in paragraph
11.
On April 25,
1985, the
A emergency diesel
generator
was
removed
from service
at
a time when it was required
under
The resulting violation is
discussed
in paragraph
12.
8.
Operational
Safety Verification (71707)
The inspectors
observed
control
room operations,
reviewed applicable
logs,
conducted
discussions
with control
room operators
(CRO),
observed
shift
tur novers
and
confirmed operability of instrumentation.
The
inspectors
verified the operability of selected
emergency
systems,
reviewed
tagout
records,
verified compliance with TS limiting conditions of operation
and
verified the return to service of affected
components.
The
inspectors,
by observation
and direct
interviews verified that
the
physical security plan was being implemented
in accordance
with the station
security plan.
The inspectors
verified that maintenance
work orders
had been
submitted
as
required
and that followup and prioritization of work was accomplished.
The
inspectors
observed
plant
housekeeping/cleanliness
conditions
and
verified implementation of radiation protection control.
Tour s of the intake str ucture
and emergency diesel, auxiliary, control
and
turbine buildings
were
conducted
to observe
plant
equipment
conditions
including potential fire hazards,
fluid leaks
and excessive
vibrations.
The
inspectors
walked
down accessible
portions of the following safety-
related
systems
on Unit 3
and Unit 4 to verify operability
and
proper
valve/switch alignment:
Refueling Water Storage
Residual
Heat
Removal
Emergency Diesel
Generators
Pumps
Intake Cooling Water
4160 volt and
480 volt switchgear
Rod Position Indication Cabinets
Control
Room Vertical Panels
Nuclear Instrumentation
Drawers
High Head Safety Injection
Spent
Fuel Pit Cooling
On Nay 14,
1985, during
a tour of the Unit 3 charging
pump room,
an operator
was observed
to not meet the anti-contamination
clothing requirements
of
85-014.
For general
area
entry,
the
RWP required
that
shoe
covers
and
gloves
be worn.
The operator
entered
the
room without wearing
any gloves.
Health
Physics
Administrative
(HPA)
Procedure
0-HPA 002
requires
that
personnel
read
and
comply with all instructions,
requirements
and
remarks
listed
on an applicable
RWP.
Failure to comply with posted
RWP requi rements
is Violation (250, 251/85-13-01,
item b.(2)).
An additional
example of this
violation is discussed
in paragraph
11.
9.
Engineered
Safety Features
Walkdown (71710)
The inspector s verified the operability of the Units 3 and
4
RHR systems
by
performing
a complete
walkdown of the accessible
portions of the
systems.
The following items
were specifically reviewed
and/or
observed
as
appro-
priate:
a.
that the licensee's
system lineup procedures
matched plant drawings
and
the as-built configuration;
b.
that the
equipment
conditions
were satisfactory
and
items that might
degrade
performance
were identified and evaluated,
e.g.,
hangers
and
supports
were operable,
housekeeping
was adequate;
c.
that instrumentation
was properly
valved in and functioning
and that
calibration dates
were not exceeded;
d.
that
valves
were
in proper position,
breaker
alignment
was correct,
power was available,
and valves were locked/lockwired
as required;
e.
local
and
remote position indication
was
compared,
and remote instru-
mentation
was functional;
and
f.
breakers
and
instrumentation
cabinets
were
inspected
to verify that
they were free of damage
and interference.
The following discrepancies
were identified and initial licensee corrective
actions
have
been
taken:
RHR valve 4-758 air line was crushed;
with extra plastic
hoses;
4B
pump
mechanical
had
a missing
cotter pin; Unit 4
RHR heat
exchanger
support
had
one
loose bolt;
wooden
wedges
were found between
and solidified mortar was found in a
sump.
Close
out of these
items will occur at
a later walkdown (IFI 250,
251/85-13-10).
Within this area,
no violations or deviations
were iden'tified.
10.
Plant Events
(93702)
An independent
review was conducted of the following events.
On April 11,
1985,
at
3:45 a.m.
a Unit 4 pressurizer
steam
sample
containment isolation valve did not close
on
command,
and
a report of
loss of containment integrity was prepared
since the
TS have
no action-
time allowed.
A leak at
a swagelok connection outside the containment
between
the
containment
and
the
outside
automatic
isolation
valve
caused
the attempt to close the inside automatic valve from the control
room;
however,
dual
indication indicating 'intermediate
position
was
received.
A containment
entry
was
made
and
the
inside
valve
was
verified to close
when the fuses
were pulled.
The inside manual
valve
was also closed
and the leak stopped at 4:30 a.m.
The leaking fitting
was replaced
and the inside valve will remain closed.
10
On April 25,
1985,
at
1:00 a.m.
the
A
EDG was
removed
from service
which placed Unit 4 in a
TS action
statement
which was
not recognized
prior to entering the condition.
Further discussion
is in paragraph
12
on Independent
Inspection.
The action statement
was cleared within the
time allowed.
On April 29,
1985,
at
2:45 p.m.
the
Unit 3
A bus
was
de-energized
during replacement
of a cover
on the Unit 3 main generator
relay panel.
Control
room lighting
was lost
as
was annunciator
power-with annun-
ciator
power restored
in
20
seconds
and full lighting within
10
minutes.
A lockout of the
main
transformer
occurred
and", with the
unit's start-up transformer
and
3B bus undergoing
maintenance,
the only
source
of power to Unit 3
was the
A EDG which
had auto-started.
The
reactor
core
had
been
off-loaded,
and
the
temperature
was
137
degrees
F with an increase
of five degrees
F per hour predicted without
cooling.
The
SFP cooling
pump did not start since it is not connected
to the safety bus.
At 4:50 p.m.
an Unusual
Event (UE) was declared
due
to the length of time required to repower
Unit 3 from normal
sources.
The
SFP cooling
pump was started after the non-safety
bus was loaded
on
the
EDG around 5:00 p.m.
At 8:20 p.m.
power to the
3A bus
was fed from
the Unit 4 start-up transformer
through
a maintenance
feed connection.
At 2:35 a.m.
power from the Unit 3 start-up transformer
was restored to
the
3A and
3B buses
and the
UE terminated.
On May 1, the inspector verified that:
the Unit 3
SFP cooling
pump was
aligned via the
lower suction path which allows about
900 gallons
per
minute (gpm) instead of the 2200
gpm when aligned via the upper suction
path;
the
temperature
of 134 degrees
F was representative
of all
surface points;
CCW flow was being verified to be less
than
2250 gpm-
being
2100
gpm
instead
of the
2800
gpm
design
flow;
surface
radiation
readings
were
120 mi llirem per
hour
(mrem/hr.)
and at five
feet
above
the
SFP,
without fuel
movement
on
the
bridge
crane,
the
readings
were
10 to
15 mrem/hr.;
and
make-up
to the
was
2400 to
3000 gallons per day.
On
May 2,
1985,
at
12: 19 a.m.
the Unit 3
CRO was
removing the
3A bus
from service
and the
A EDG auto-started.
The procedure
which was in
use
had
an
error
as it called for the
disabling
of the
3B
sequencer.
The A EDG was stopped,
and the
3A bus properly removed
from
service.
The
SFP temperature
reached
-153 degrees
F in four hours while
the
SFP cooling was not in service.
On
May 15,
1985,
at 1:07 p.m.
a construction
worker
bumped
a relay in
the Unit 4 safeguards
relay rack while working one to two inches
away
preparing
to pull
a cable.
This relay actuation initiated
a turbine
trip on "hi-hi steam
generator
level" in the
C steam
generator
(SG)
which secured
both
main feedwater
(MFW)
pumps
and isolated
the
C SG.
All
systems
responded
normally.
Pressurizer
pressure
reduced
to
1990 psig.
The unit returned to operation that evening.
On
May 17,
1985,
an
UE was decl ar ed at 12: 10 p.m.
when Unit 4 tripped
due to
a loss of off-site power.
The
B
MFM pump tripped
when
the
de-energized
due
to
a fire, North of Miami, affecting
a
500
kV transmission
line.
The
C
SG low level coincident with feedwater
flow less
than
steam
flow tripped
the
reactor
which tripped
the
turbine.
Pressurizer
pressure
reached
2325 psig.
Both
auto-
started
and
natural
circulation
was
established;
the
6B feedwater
safety valve lifted, and the
main
steam isolation valves
closed.
All
other
systems
responded
normally.
At 2:07 p.m.
the
UE terminated
as
off-site
power
was
restored
and
forced
reactor
circulation
was
initiated for Unit 4.
The Unit 3
UE terminated at 2:23 p.m.
No violations or deviations
were identified in this section.
11.
Preparation
For/And Refueling (60705/60710)
a,
On April 29,
1985,
the inspector
observed
the receipt
inspection of
burnable
poison
inserts
in the
Unit 3
new fuel
room.
The
receipt inspection
was performed in accordance
with Operating
Procedure
(OP)
16009. 11,
On-Site
Unpacking,
Inspection
and
Manual
Loading of
Vessel
Flux
Depression
Assemblies.
The
licensee's
Quality
Assurance
Topical
Report
incorporates
the
requirements
of
ANSI
N45.2.2-1972Property "ANSI code" (as page type) with input value "ANSI</br></br>N45.2.2-1972" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process.,
Packaging,
Shipping; Receiving,
Storage
and Handling of
Items
for Nuclear
Power
Plants.
QP 7. 1
which
implements
Topical
Quality
Requirements
and
Appendix B,
Criterion VII,
specifies that receipt inspections
of nuclear fuel will be performed in
accordance
with site specific procedures.
The licensee
considers
fuel
inserts
and rod control cluster
(RCC) assemblies
to also
be covered
by
specific
receipt
inspections
and
has
developed
the
following
procedures:
OP 16009.1
OP 16009.6
OP 16009. 10
OP 16009.11
Receipt
and Handling of New Fuel Containers
On-Site Unpacking,
Inspection
and Manual
Loading
of Burnable
Poison
Rod Assemblies
On-Site Unpacking,
Inspection
and Manual
Loading
of Rod Control Cluster Assemblies
On-Site Unpacking,
Inspection
and Manual
Loading
of Hafnium Vessel
Flux Depression
Assemblies
A review of OP 16009. 11 revealed that it inadequate
implemented
TQR 7.0
and
7. 1
requirements
which
inappropriate
ANSI
N45.2.2-1972Property "ANSI code" (as page type) with input value "ANSI</br></br>N45.2.2-1972" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process.,
requirements
in that:
( 1)
The receipt inspection did not verify that the purchased
material
conformed
to the
procurement
documents.
The purchase
documents
available
during
the receipt
inspection
did not specify
which
quality standards
were required to
be
met
by the vendor.
A QA
product certification supplied
by the vendor could not be verified
to conform with prearranged
mutually acceptable
purchase criteria
section
5 '.2.2).
12
(2)
Specific
guidance
was
not
available
with which
the
receipt
inspector
could
determine
that
damage
had
not occurred
to the
inserts
and
that
the
inserts
were
sufficiently clean
(ANSI
N45.2.2-1972Property "ANSI code" (as page type) with input value "ANSI</br></br>N45.2.2-1972" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process.,
sections
5.2.2.7
and 5.2.2.8).
(3)
A preliminary visual inspection of the shipping containers
was not
required.
Criteria for determining that
damage
had not occurred
due to fire, exposure,
rough handling
and tie
down failure were
not procedurally
incorporated.
The preliminary inspection prior
to
unpacking
was
not
documented
(ANSI
N45.2.2-1972Property "ANSI code" (as page type) with input value "ANSI</br></br>N45.2.2-1972" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process.,
section
5.2.1).
10 CFR 50, Appendix B, Criterion VII, as
implemented
by the
FPL Topical
Quality Assurance
Report,
Revision 6,
TQR 7.0,
Control of Purchased
Items
and
Services,
requires
that
measures
shall
be
established
to
assure
that
purchased
material
conforms
to
the
requirements
of
applicable
procurement
documents.
The
Quality
Assurance
Program
incorporates
the
requirements
The failure to
meet
the
requirements
of
ANSI
N45.2.2-1972Property "ANSI code" (as page type) with input value "ANSI</br></br>N45.2.2-1972" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process.
is
a
Violation
(250,
251/85-13-04).
During the receipt
inspection
a
QC Inspector
observed
a
condit,ion
adverse
to quality,
and
he did not ensure that corrective actions were
implemented
in
accordance
with
approved
procedures.
One
burnable
poison
insert
was
inadvertently
contaminated
with
grease
from
an
overhead hoist and consequently
was not initially acceptable
for use in
the reactor core.
The nonconformance
was not documented
as required
by
AP 190. 13, Corrective Action for Conditions
Adverse to Quality.
Any
subsequent
measures
used to clean the insert or subsequent
reinspection
of the insert
were
not documented.
The receipt
inspection
documen-
tation did not mention the discrepancy.
The insert
was
placed
in the
spent fuel pool without evaluation.
The failure to implement corrective actions for conditions
adverse
to
quality in
accordance
with approved
procedures
is
a violation of
Appendix B,
Criterion XV,
as
implemented
by
FPL Topical
Quality
Assurance
Report,
Revision 6,
TQR 15.0,
Nonconforming
Materials,
Parts
or
Components,
QP 15.2
and
AP 190. 13
(250,
, 251/85-13"03).
During
poison
insert
handling
the
inspector
observed
the
use, of
handling
equipment
which did not meet the requirements
of 10 CFR 50,
Appendix B,
Criterion XIII, as
implemented
by
Topical
Quality
Assurance
Report,
Revision 6,
TQR 13.0, Handling, Storage
and Shipping;
QP 13. 1,
and ANSI N45.2.2-1972,
Packaging,
Shipping,
Receiving,
Storage
and
Handling of Items for Nuclear
Power
Plants.
QP 13. 1,
section
5.4.2,
Inspection of Equipment,
requires
that, prior to use,
handling
equipment
be
inspected
for acceptability.
ANSI N45.2.2 requires that
equipment
not be used if it fails to meet manufacturer s specifications,
if it is frayed or deteriorated,
or if it contains
contaminants
that
would be detrimental
to the material
being handled.
'P(
13
On April 29,
1985,
a nylon rope was knotted
and fashioned into a sling
and
used to hoist hafnium burnable
poison assemblies.
The knotted rope
did not meet
any manufacturer's
specifications.
Factory manufactured
slings were available for use
and were not used.
On April 29,
1985,
an electric hoist in the Unit 3
new fuel
storage
room
was
used
to lift hafnium burnable
poison
assemblies.
The hoist
contained
contaminants
in the form of grease
which dripped
on
a poison
assembly.
Failure to meet
equipment
handling
requirements
is
a Violation (250,
251/85-13-02,
items
a and b).
An additional
instance of this violation
is discussed
in paragraph
8.
On April 30,
1985,
the inspector
observed
the change
out of
RCC in the
Unit 3
SFP.
Personnel
were
required
to
comply with the clothing
protection criteria of
RWP 85-500 which specified that
persons
using
telephone
headsets
wear
a full hood.
The full hood provides better
anti-contamination
protection
than the more frequently
used
cap because
it covers the entire
back of the neck.
This is the location where the
telephone
headset
cord tends to rub.
The
HP Department
was
supplying
continuous
coverage
and assistance
for the
change
out evolution.
The inspector
observed
one person wearing
a telephone
headset
who was
not wearing
a full hood.
The
HP technician
on the
scene,
like the
telephone talker, did not realize that the
RWP required
a full hood.
HPA
procedure
O-HPA-002,
Requirements
for Entry
and
Work in
RCA,
requires
in section 5.2. 15 that all personnel
working in an area
where
a
is
required
shall
read
and
comply with all
instructions,
requirements
and remarks listed
on the
RWP.
Failure
to
comply
with
requirements
is
a
Violation
(250,
251/85-13-01,
item b.(1)).
An additional
example of this violation is
discussed
in paragraph
9.
In Inspection
Reports
250, 251/84-18
and 250/84-35
and 251/84-36,
which
covered the
May and
November
1984 periods,
respectively,
the inspectors
addressed
several
aspects
of the operation of SFP
systems that did not
correspond
to that, described
in the
SAR.
The modification of these
systems
had
not received
a
review for
unreviewed
safety
question
determinat'ion
prior to
the
modification
of the
system
from that
described
in the
SAR.
The facility operating
licenses
permit
the
modification of systems
only after
a
review
has
been
conducted
in
accordance
with
As
several
of
these
reviews
for
unreviewed safety question determination
have yet to be
made
and as the
systems
are
being
operated differently than
as described
in the
SAR,
these
items constitute
an Unresolved
Item (250, 251/85-13-09).
A
%I
I
1
Independent
Inspection
(92706)
During the report
period the inspectors
routinely attended
meetings with
licensee
management
and monitored shift turnovers
between
PSN, shift foreman
(Nuclear Match Engineers
[NME]), and
CROs.
These
meetings
provided
a daily
status of plant operating
and testing activities in progress
as well
as
a
discussion
of significant
problems
or incidents.
Based
on these
discus-
sions,
the
inspectors
reviewed
potential
problem
areas
to
independently
assess:
their importance to safety;
the proposed
solutions;
improvement
and
progress;
and
adequacy
of corrective actions.
The inspector's
reviews of
these
matters
were
not restricted
to
the
defined
inspection
program.
Independent
inspection efforts were conducted
in the following areas:
Estimated Critical Conditions Calculations
Fire Drills
Check Valve Testing
Reactor
Vessel
Head Stud Reference
Transition Temperature
Residual
Heat
Removal
Suction Valve Testing
On April 25,
1985, at approximately
1:00 a.m.
the licensee
placed the
A EDG
out of service to perform routine preventive
maintenance.
Unit 4 was oper-
ating at
100 percent
power.
Unit 3 was in refueling shutdown with all fuel
removed
from the vessel.
The
3B 4160 volt electrical
bus
was de-energized
to facilitate routine
breaker
inspections;
consequently,
the
3B safety
injection
pump was inoperable.
The
licensee
determined
that
three
of four safety
injection
pumps
were
as
allowed
by
TS 3.4. l.b.2.
The
PSN's
log contained
an
entry
stating that
two safety injection
pumps would remain operable
in the event
of a blackout.
This entry was incorrect.
On
a loss of offsite power with
the
A EDG out of service,
the
B
EDG would be the sole
source of emergency
power.
The
B
EDG would normally supply power to the Unit 3 and Unit 4 4160
volt B buses.
However, with the
3B 4160 volt bus de-energized,
only the
4B
4160 volt bus
would receive
emergency
power.
Consequently
only the
4B
safety
injection
pump would
be available.
The
3A and
4A pumps
would be
unavailable
due to loss
of offsite
power in conjunction with the
A EDG
clearance.
The
3B pump would be unavailable
due to the de-energized
bus.
At approximately 4:30 a.m. the licensee
determined that
a potential
problem
existed concerning
removal of the
A EDG from service with the
3B 4160 volt
bus de-energized.
A decision
was
made to restore
the
A EDG to service
as
soon
as
possible.
The
was
declared
in service
at
7:30 a.m.
after
successfully
completing
an operability run.
The licensee
was concerned that
the
3A and
4A safety injection pumps might not be considered
under
While this
TS specifies
that
components
need
not
be considered
due to the inoperability of their emergency
power supplies, it
also requires
that
redundant
components
be operable
with their
normal
or
emergency
power supplies
available.
The
concern
was that
the
3B safety
injection
pump was redundant
to the other
pumps
and its normal
and emergency
15
power supplies
were not available.
Afte~ restoring
the
A EDG to service,
the licensee
referred
the matter to the licensing department
for clarifi-
cation.
states
that
power operation
may continue if one
EOG is out of
service
provided that the
remaining
EDG is tested daily and its associated
engineered
safety features
are operable.
If one
EDG is inoperable
and the
other
EOG's engineered
safety features
are not fully operable
then
TS 3.0. 1
applies
and the affected reactor
must
be shut
down within seven
hours.
The
licensee
did not initially realize
that
applied
to the
EDG/safety
injection
pump
problem
of April 25.
However,
the
licensee
restored
the
A EDG to operation
before
exceeding
the
seven
hour
shutdown
limit imposed
by
TS 3.0. 1
through
consequently,
neither
TS 3.7.2.b nor TS 3.0. 1 were violated.
The
was
implementing
a preplanned
maintenance
department activity when
he authorized
the removal of the
A EDG from service.
The clearance
tags
had
been
prepared well in advance
by other
CROs.
Maintenance
personnel
had
been
. selected
for the task
and were waiting to begin their work.
The
B
EDG was
tested
to verify that it was
subsequent
to receiving
routine
maintenance
of the type planned for the
A EDG.
Shortly after the
B
EDG was
declared
the
A EDG
was
removed
from service.
Apparently,
the
ability to
remove
the
B
EOG from service without creating
a
TS limiting
condition of operation resulted
in an inadequate
analysis
of the effect of
removing the
A EDG from service.
TS 6.8. 1 requires
that written procedures
and administrative policies
be
established,
implemented
and maintained that meet or exceed 'the requirements
and recommendations
of sections
5. 1 and 5.3 of ANSI 18.7-1972
and Appendix A
of USNRC Regulatory Guide 1.33.
section
5. 1.6.3,
Scheduling
of Maintenance,
requires that
maintenance
be
scheduled
and
planned
so
as not to jeopardize
the safety of
the reactor.
Planning
shall
consider
the possible
safety
consequences
of
concurrent
or
sequential
maintenance,
testing
or operating
activities.
Equipment required to be operable for the
mode in which the reactor exists
shall
be available,
and
maintenance
shall
be performed in manner
such that
the license limits are not violated.
The failure to preplan
a maintenance
activity so
as
not to jeopardize
a
safety
system
is
contrary
to
the
requirements
of
ANSI
N18.7-1972Property "ANSI code" (as page type) with input value "ANSI</br></br>N18.7-1972" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process.,
section
5. 1 and fails to meet the requirements
of TS 6.8. 1; hence, this is
a
Violation (250, 251/85-13-05).
An inspection
was conducted of actions
taken
by the licensee
to address
the
significant safety concern
noted in Inspection
and
Enforcement
Information Notice 84-06, which discussed
a problem of steam binding of the
AFW pumps at
the
H.
B. Robinson Plant.
Other more detailed notification of this problem
16
was received
by the licensee
from the Institute of Nuclear
Power Operations
(INPO).
The inspector
noted that
no
immediate
problem existed
as neither
the suction
nor discharge
piping of any of the three
AFW pumps
was hot to
the
touch during
several
walkdowns
during
the
inspection
period.
The
licensee
has trained the operators
on this problem and procedural
guidelines
are to be provided by June
1,
1985.
The monitoring of pipe temperatures
by
touch will begin
on June
1,
on
a once per shift basis.
It is noted that the
three
AFW pumps for the
two units
are
self-venting
due
to
a
continuous
recirculation
path
from the
pump discharge
through the
lube oil cooler to
the
condensate
storage
tank.
The outlet
valves
downstream
of the
check
valves are normally-closed
The testing of the check valves
as
part of the
ASME Code,
Section XI; program is to assure
that the
pump
can
put out full flow through
the
check
valves.
The
adequacy
of this test
method to support
that
the
check
valve will perform its
intended
check
function (restriction
of backflow) is
an
Unresolved
Item pending further
discussions
with the Region
and
No violations or deviations
were identified.