ML18029A613
| ML18029A613 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 05/21/1985 |
| From: | Brooks C, Cantrell F, Patterson C, Paulk G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18029A611 | List: |
| References | |
| 50-259-85-25, 50-260-85-25, 50-296-85-25, NUDOCS 8507010476 | |
| Download: ML18029A613 (12) | |
See also: IR 05000259/1985025
Text
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UNITEDSTATES
NUCLEAR REGULATORY COMMISSION
'EGION
II
10t MARIETTASTREET, N.W.
ATLANTA,GEORGIA 30323
Report Nos.:
50"259/85-25,
50-260/85-25,
and 50-296/85"25
Licensee:
Val'ley Authority
500A Chestnut Street
Chattanooga,
TN
37401
Docket Nos.:
50-259,
50-260
and 50-296
License Nos.:
DPR"52,
and
Facility Name:
Browns Ferry Nuclear Plant
inspection
Conducted:
March 26 - Apri1 25,
1985
Inspectors:
G.
L. Paul k
C. A. Patterson
Dat
S gned
Da
e
igned
C.
R, Brooks
Approved by:
F.
. Cantrell, Sectio
Division of Reactor Proj
s
Date
gned
~ ~,)
g'g
ate Signed
SUMMARY
Scope:
This routine,
unannounced
inspection entailed
290 inspector-hours
in the
areas
of operational
safety,
maintenance
observation,
surveillance,
reportable
occurrences
and reactor trips.
Results:
One violation with four examples
of technical specification 6.3.A for
failure to
follow procedures
related
to battery
surveillance
and
clearance
procedures.
0
8507010476
850522
ADOCK 05000259
8
REPORT DETAILS
Persons
Contacted
Licensee
Employees
J. A.
G. T.
J.
E.
J.
R.
J.
H.
J,
0.
0.
C.
Ray
H
C.
G.
T. 0.
R.
E.
A.
W.
R.
E.
T. L.
T.
F.
J.
R.
B. C.
A. L.
R.
R.
T.
W.
S.
R.
G ~
R.
W.
C.
A. L.
R. L.
Coffey, Site Director
Jones,
Plant Manager
Swindell, Superintendent
- Operations/Engineering
Pfttman, Superintendent - Maintenance
Rfnne, Modifications Manager
Carlson, Quality Engineering
Supervisor
Mfms, Engineering
Group Supervisor
unkapillar, Operations
Group Supervisor
Wages,
Mechanical
Maintenance
Super visor
Cosby, Electrical Maintenance
Supervisor
Burns,
Instrument Maintnenace
Supervisor
Sot rell, Health Physics
Supervisor
Jackson,
Chief Public Safety
Chinn,
Senior Shift Manager
Ziegler, Site Services
Manager
Clark, Chemical Unit Supervisor
Morris, Plant Compliance Supervisor
Burnette, Assistant Operations
Group Supervisor
Smallwood, Assistant Operations
Group Supervisor
Jordan,
Assistant Operations
Group Supervisor
Maehr, Planning/Scheduling
Supervisor
Hall, Design Services
Manager
Thomfson,
Engineering
Section Supervisor
Clement,
Radwaste
Group Controller
Lewis, Senior Shift Manager
Other
licensee
employees
contacted
included
licensed
reactor
operators,
auxiliary operators,
craftsmen,
technicians,
public safety officers, Quality
Assurance;
Design
and engineering
personnel.
Exit Interview
The inspection
scope
and findings were
summarized
on April 26 and 29,
19S5,
with the Plant Manager and/or Assistant Plant Managers
and other members of
his staff.
The licensee
acknowledged
the findings and took no exceptions.
The licensee
dfd not identify as proprietary
any of the .materials
provfded to or reviewed
by the inspectors
during this inspection.
This subject
was not add~essed
in the inspection.
Unresolved
Items~
There
were
three
new unresolved
items
as identified in paragraphs
5',
7,
and 9.
5.
Operation Safety (71707,
71710)
The
inspecto'rs
were kept informed
on
a daily basis of the overall plant
status
and
any significant safety
matters
related
to plant operations.
Oaily discussions
were held
each
morning with plant management
and various
members of the plant operating staff.
The inspectors
made frequent visits to the control
rooms
such that each
was
visited at least daily when
an inspector
was
on site.
Observations
included
instrument readings,
setpoints
and recordings;
status of operating
systems;
status
and
alignments
of emergency
standby
systems;
onsite
and offsite
emergency
power
sources
available
for automatic
operation;
purposes
of
temporary tags
on equipment controls
and switches;
alarm status;
adherence
to procedures;
adherence
to limiting conditions for operations;
nuclear
instruments
temporary
alterations
in effect;
daily
journals
and logs; stack monitor recorder traces;
and control
room manning.
This inspection activity also
included
numerous
informal discussions
with
operators
and their supervisors.
General
plant tours were conducted
on at least
a weekly basis.
portions of
the turbine building, each reactor building and outside areas
were visited.
Observations
included
valve positions
and
system
alignment;
and
hanger
conditions;
containment
isolatiotl alignments;
instrument
readings;
housekeeping;
proper
power supply
and
breaker
alignments;
radiation
area
controls;
tag controls
on equipment;
work activities in progress;
radiation
protection
controls
adequate;
vital
area
controls;
personnel
search
and
escort;
and vehicle search
and escort.
Informal discussions
were held with
selected
plant
personnel
in their functional
areas
during
these
tours.
Weekly verifications 'of system
status
which included major flow path valve
alignment,
instrument
alignment,
and
switch
position
alignments
were
performed
on the high pressure
coolant injection systems.
I'
complete
walkdown of the accessible
portions of the
D.C. battery supply
system
was
conducted
to verify system operability.
Typical of the
items
checked
during the
walkdown were:
lineup procedures
match plant drawings
and the as-built configuration,
hangers
and supports
housekeeping
adequate,
electrical
panel
interior
conditions,
calibration
dates
appropriate,
system
instrumentation
on-line,
valve
position
alignment
correct,
valves
locked
as
appropriate
and
system
indicators
functioning
properly.
"An Unresolved
Item is
a matter
about which,more
information is required
to
determine whether it is acceptable
or may involve a violation or deviation.
Ouring
a routine tour of the Unit 3 Reactor Building, the inspector noted
several
discrepancies
associated
with the
control
rod drive hydraulic
control units
(HCU).
Directional control
valves
on twelve
HCUs were found
to
be missing
the valve
cap which encases
the
needle
valve
used for rod
timing adjustment.
One directional
control valve was
found to
be missing
its solenoid
enclosure
cover thus
exposing
the coil
and terminals
to the
environment.
Channel
nuts
which are
used to fasten
the
HCU frame to the
.channel
embedded
in the concrete
pad were
found to
be rotated
90-degrees
such that they performed
no useful
function on several
HCUs.
Channel
nuts
on all
HCUs showed
signs of excessive
deterioration
from rust.
Still other
were
found with visibly loose
mounting
hardware
(several
threads
visible on the bolt beneath
the bolt head).
Although the majority of the
are
mounted
back-to-back,
several
on the
end of a string are
free-standing
with no additional restraints
to substitute
for the
support
otherwise
provided
by the
mating
HCU frame..
The inspector
informed the
licensee of these
problems.
The inspector
found similar problems with the
Unit 1
HCUs.
Also, large fIat washers
were
used with the Unit 3
HCU frame
bolts but smaller
lockwashers
were
used
on the Unit
1 frame bolts.
These
concerns
are identified
as
an unresolved
item pending further analysis
by
the licensee
{259/260/296/85-25-01).
The licensee
reported
on Parch
28,
1985, that
a design error was discovered
in the electrical circuit for two handswitches
(63-24 and 63-25) which al.)ow
bypassing
the interlock for drywell purging in the
RUN mode of operation.
With these
switches
in the bypass
positions
and the'mode
switch in RUN, the
standby
gas
treatment
system,
the control
room
emergency
pressurization
system
and
some
group six i solations
would be inoperable.
This item will be
inspected
further
and will be carried
as
an inspector
followup item (IFl
259/85"25-02).
Haintenance
Observation
(62703)
Plant
maintenance
activities
of
selected
safety-related
systems
and
components
were observed/reviewed
to ascertain
that they were
conducted
in
accordance
with requirements.
The following items
were considered
during
this review:
the limiting conditions for operations
were
met; activities
were
accomplished
using
appr'oved
procedures;
functional
testing
and/or
calibrations
were
performed prior to returning
components
or
system, to
service;
quality
control
records
were
maintained;
activities
were
accomplished- by qualified personnel;
parts
and materials
used
were properly
certified;
proper
tagout
clearance
procedures
were
adhered
to; Technical
Specification
adherence;
and
radiological
controls
were
implemented
as
required.
Maintenance
requests
were
reviewed to determine'.status, of outstanding
jobs
and
to
assure
that priority was .assigned 'o
safety-related
equipment
maintenance
which might affect plant safety'.
The inspectors
observed
the
below listed maintenance activities during this report period:
a.
Battery corrective maintenance
for main and diesel batteries
b.
Limitorque valve pinion gear inspection
c.
Unit 2 refueling operations
d.
"C" fire pump maintenance
e.
LPCI W set maintenance - Unit 2
During
a
routine
tour of the reactor
building
on April 23,
198S,
the
inspectors
noted
that
the
2DA
low pressure
coolant
injection
(LPCI)
motor-generator
set
was
tagged
out for maintenance
and the motor
removed.
However,
several
alarm lights were illuminated at the local control station.
~At junction
box
5991,
a white light for TRIP HI-HI
MOTOR
TEMP
and at
junction box
5952
a white light for WARNING MOTOR TEMP HI and
a red light
for TRIP
MOTOR
TEMP HI-HI were illuminated.
The local
switches
at
the
junction boxes for the
LPCI motor-generator
were tagged
under hold order
85-150A.
The shift engineer
and electrical
maintenance
supervisor
were
notified of the inspector's
concern
that voltage still might be applied to
some of the lifted motor leads
on April 23,
1985.
On April 29,
1985, the inspector'as
notified by the electrical
maintenance
section that the thermistor
were lifted when the motor was.removed.
These
were not included in the hold order
and
upon followup inspection
were found to be "hot" ( 18 volts).. The tagout
was to be revised to include
an additional
boundary to secure
power to the thermistor leads.
Plant
Standard
Practice
BF 14.25,
Clearance
Procedure,
requires all sources
of electrical
power
be
removed
from
equipment
for work to
be
safely
performed.
This item is included
as the fourth example of the violation for
failure to follow procedure
(260/85-2S-03).-
Further review of hold order 85-1SOA indicated that the hold order tag (83)
placed
on the
main supply circuit breaker
to the
2DA LPCI
MG set
was
an
incorrect tag.
The
2EN LPCI motor-generator
set that was tagged out on hold
order 85-118 also
had
an incorrect tag placed
on fts main
supply circuit
breaker.
The tag for the
2DA LPCI
MG set
had
been inadvertently placed
on
the main breaker for the
2EN LPCI
MG set
and vice versa.
This violation of
clearance
procedures
is included in the fourth example
noted
above.
Surveillance
Testing Observation
(61726)
The
inspectors
observed
and/or
reviewed
the
below listed
surveillance
procedures.
The inspection
consisted
of
a
review of the
procedures
for
technical
adequacy,
conformance
to technical specifications,
verification of
test instrument calibration, observation
on the conduct of the test,
removal
from service
and return to service of the
system,
a review of test data,
limiting condition for operation
met,
testing
accomplished
by qualified
personnel,
and
that
the
surveillance
was
completed
at
the
required
frequency.
S.I.
S.I.
S.I.
S.I.
4.5.E.l.c
3.2
4.9.A.2.b
4.9.A.2.a
MOV Operability'est,
Inserv fce Section
XI Valve Testing
Auxiliary Electrical
Equipment - Battery Analysis
Auxiliary Electiical Equipment - Battery Check
The
inspectors
reviewed
survei 1 lances
established
to satisfy
Technical
Specification
Surveillance
Requirement
4.9.A.2, Unit Batter ies (250-volt).
This closep inspector followup item (259/260/296/ORP
85-01), Station Battery
Operation,
Maintenance
and Inspection.
Ouring
the
station
battery
inspection
phase,
the
inspectors
identified
several
discrepancies
related to the seismic qualification of the batteries
and their racks.
These
shall
be tracked
as
an unresolved
item (259/260/
296/85-25-04)
and .are
as follows:
a 0
b.
C.
The Unit 250-volt battery
racks
are not fastened
to the floor-mounted
pedestals
as
depicted
in the
as-constructed
drawings.
Although TVA
drawing
48N949RA
shows
the Unit 3
Main Battery
rack bolted to the
pedestal
with 5/8-inch bolts,
the
rack is actually
welded to the
pedestal.
TVA drawing
48N958RA
shows
the Units
1
and
2 Main Battery
racks
welded to the pedestal;
however,
the
racks
are actually welded
and
bolted to the
pedestal.
The
licensee
has initiated
a
safety
evaluation
and discrepancy
reports
on the drawings.
None of the Oiesel
Generator
Battery racks are fastened
to the
embedded
plates
as
depicted
in
the
as-constructed
drawings.
TVA drawing
shows
field supplied
or finish concrete
should
be
installed for level
rack installation.
The
racks. were
found to .be
elevated
about
2-inches
above
the
embedded
plate with
no
or
concrete.
The licensee
performed
an analysis
which indicated that the
racks
were
not seismically qualified in their present
condition
and
initiated the installation of shims per the as-constructed
drawing.
Many battery
rack fasteners
were
not installed
per the
vendor
manual
(C&0 Installation
and Operating Instructions for Stationary Batteries,
Section
12-600-1).
Channel
nuts
are
used
to fasten
the rail to the
frame
as
depicted
in Figure
3 of the
vendor manual.
The nuts
were
found rotated 90-degrees
such that they performed
no fastening function
on several
locations of the Shutdown
Board B,
C and
0 Battery racks
and
the Hain Unit 1
and
3 Battery
racks.
Several
loose tie
rods
were
additionally found on Unit
1 and
3 Hain battery racks.
d.
Although Section 3.2.3 of the Vendor Hanual requires "furnished plastic
spacers"
be placed
between
each cell, plywood spacers
were found on the
Hain Unit 1,
2
and
3 batteries
and either
styrofoam
or
foam rubber
spacers
were found in the remaining locations.
e.
The
end cells
on
the
3EB
Shutdown
Board battery
were
found
about
3-inches
away
from the
battery
rack
end rail.. The
licensee
has
initiated action to move the
end rai l
such that it butts against
the
end cells.
A detailed
review of Surveillance
Instruction (SI) 4.9.A.2.b, Auxiliary
Electrical
Equi pment - Battery
Ana lysi s,
and
SI 4.9.A.2. a,
Auxi 1 iar y
Electrical Equipment - Battery Check,
found other
items in conflict with the
Vendor Manual.
These
items will be tracked
as
an Inspector
Followup .Item
(IFI 259/85-25-05)
and are
as follows:
a.
SI 4.9.A.Z.b Step 3.7 requires that individual cell voltages
be within
a0. 1 volt of the
average
battery cell
voltage.
Section
7.Z of the
Vendor
Manual,
however,
indicates
that cell voltages
should
be within
10.04 volt and provides
a list of potential
problems
which may cause
cell voltages outside this range.
The licensee
has indicated that they
have historically had difficulty meeting
a .04 volt acceptance criteria
and initial contact with the vendor resulted
in concurrence
with an 0. 1
volt criteria.
b.
SI 4.9.A.2.a
contains
an
acceptance
criteria of 267 +3.0 volts for
Shutdown
Board battery overall float voltage.
Since
Shutdown
Board
B
Battery currently has three cells jumped out due to a broken .intercell
terminal
post,
the float voltage
per cell is 2.28 a .025 volts.
This
is outside
the
Vendor Manual
requirements
of 2.20 to 2.25 volts per
cell for float voltage.
The
licensee
is evaluating
the effect of
maintaining excessive
on the battery.
C.
The
battery
cell
temperature
recorded
for the
Shutdown
Board
3EB
battery during the performance of SI 4.9.A.2.a
on February 5,
1985,
was
45'.
According to the Vendor Manual (Section 7.3), battery capacity
decreases
to about
81% of rated capacity
at this temperature.
Since
the acceptance
criteria for the batt:ery capacity test required
by T.S.
4.9.A.2.c is 80~, temperatures
less
than
45
F.
have the potential for
making the battery inoperable.
The licensee
is evaluating this event.
A review of recently
completed
Surveillance
Instruction
Data
Sheets
was
performed.
Three examples of failure to follow procedures
were identified:
a.
SI 4.9.A.2.a, Auxiliary Electrical
Equipment Battery Check,
is intended
to satisfy
the
weekly pilot cell
checks
requi red
by
Technical
Specification
4.'9.A.2.a.
Battery 'pilot cells are rotated
every four
months
and are designated
in Electrical
Maintenance
Instruction
No.
4
(EMI-4), Batteries.
SI 4.9.A.2.a performed
on February
25 through 27,
1985, =checked cell
no.
60 for the Unit batteries
and cell
no.
30 for
the Diesel batteries
rather
than
the pilot cells specified
in EMI-4
(cell
no.
68 for the Unit batteries
and cell
no.
38 for the
Diesel
batteries).
Thus,
the
weekly pilot
. cell
checks
required
by T.S.
4.9.A.2.a were not completed during the
week, of February
24,
1985.
b.
SI 4.9.A.2.a is also
intended
to satisfy the weekly check of overall
battery voltage
required
by T.S. 4.9.A.2.a.
The acceptance
criteria
specified
in SI 4.9.A.2.a
fs 133.5 s 1.5 volts for Diesel
Generator
Battery
overall
battery
voltage.
4.0.A.2.a,
performed
on
C.
February
11 - 12,
1985,
has
130.4 volts recorded
for, Diesel Generator
A
overall voltage.
This is outside
the acceptance criteria; however,
the
surveillance
data
sheet indicates that overall float voltage acceptance
criteiia was satisfied
and
no corrective action
was initiated.
Surveillance
Instruction
4.9.A.2.b,
Auxiliary Electrical
Equipment
Battery Analysis,
provides
a check that individual cell voltages
are
within
a
O.l volts
of the
average
cell
voltage.
SI 4.9.A.2.b,
performed
on
February
20,
1985,
contains
an error that essentially
negated
the
check which was
performed.
The
SI 4.9.A.2.b data
sheet
originally recorded
262.4 volts
as
the overall
battery
voltage
and
2. 186 volts
as
the average
cell voltage
(262.4 volts divided by
120
cells).
Individual cell voltages
were
then
compared with this value.
Subsequent
to this comparison,
the overall battery voltage
was revised
to 268.4 volts (a single line was drawn through the original value with
initials of the individual who made the revision);
however,
the change
was not carried
through to the calculation
of average
cell
voltage
which should
have
been
revised to 2.237 volts.
The verification that
individual cell voltages
were within 0. 1 volt of average
cell voltage
was not repeated
using the revised figures.
The
above
violation
(259/260/296/85-25-06)
of
Technical Specification 6.3.A., Failure to Adhere to Procedures,
was disc'ussed
with the licensee
during an exit interview on April 26,
1985.
During
a review of Surveillance
Instruction (SI) 4.5.E. l.c., High Pressure
Coolant Injection {HPCI) System
Motor Operated
Valves Operability Test,
an
unexplained
difference
between
units in the
maximum allowed valve stroke
timing was
found,
The
following times
in
seconds
were
given
in
the
instruction for the two series
HPCI torus suction valves
73-26
and 73-27:
Valve
.
Time (sec.)
1-73-26
1-73-27
2-73-26
2-73-27
3-73-26
3-73-27
73.5
78
87
81
86.3
44.3
The
value
given for Unit 3
73-27
was
about half the value of the other
valves.
The inspector questioned
the reason
for the difference
and if the
direct current
motor operated
valve 73-27 could
have
an
open
shunt motor
field causing
the valve to operate at
a faster
speed.
Data reviewed for the
past
several
years
indicated
the difference
has 'existed
over
a period of
years.
The licensee
is evaluating the. timing- difference.
This will remain
an unresolved
item pending resolution of the time difference
by the licensee
(296/85-25"07).
S.
Reportable
Occurrences
(90712,
92700)
The below listed licensee
event reports
(LERs) were reviewed to determine if
the information provided
met
NRC requirements.
The determination
included
adequacy
of event description,
ver1fication of compliance
with technical
specifications
and
regulatory
requ1rements,
correct1ve
action
taken,
existence
of potential
generic
problems,
report1ng
requirements
satisfied,
and
the relative
safety
sign1ficance
of each
event.
Additional in-plant
reviews
and discussion with plant personnel,
as appropriate,
were conducted
for those
reports
indicated
by
an asterisk.
The following licensee
event
reports
are closed:
. LER No.
259/84-38
"259/85-01
"Oate'2-19-84
01"21-85 /
Event
Ramp Generator
failure
Leakage in drywell
due to core spray test
hoses
being left on after
surveillance.
"259/85-02
Ol"24--85
Inadvertent start of all
diesels
and core spray
pumps during surveillance
"259/85-03
02-05-85
Inadvertent start of
diesel
generators
"C"
and "0" during funct1onal
testing of protective
relays.
Trip of 161kv
offs1te lines occurred
during this event.
9.
No violations or deviations
were 1dentified.
Reactor Trip (93702)
The inspectors
reviewed activities associated
with the below listed reactor
trips during
th1s
report period.
This
rev1ew
included
determination
of
cause,
safety
significance,
performance
of
personnel
and
systems,
and
corrective action.
The inspectors
examined
instrument recordings,
computer
printouts,
operat1ons
journal entries',
reports
and
had
discussions
with
operations,
maintenance
and
engineering
support
personnel
as
appropriate.
Unit I was
manually
scrammed
from 30.9X
power. as
part of
a controlled
shutdown
on August 21,
1984.
The shutdown
was forced by a .7-day
LCO entered
on August 14,
1984 per T.S.
3.5.A.2 following an inadvertent
overpressuri-
zat1on
of core
spray
system
loop 1.
Failure of
a
Rod Worth Hinimizer
System
(RWM)
surveillance
during
the
controlled
shutdown
forced
the insertion of the
manual scram at 30.9% power to complete
the shutdown.
There were
no safety
system challenges
during the shutdown.
On January
16,
1985,
a low reactor water level
scram occurred
from 99% power
on Unit
1
due
to
a failed level controller (LIC 46-5).
As water level
continued to decrease,
the recirculation
pumps tripped,
main steam isolation
valves
closed,
high
pressure
coolant
injection
{HPCI) and
reactor
core
.isolation
cooling
(RCIC)
systems
initiated
automatically
as
designed;
however,
several
problems
developed.
HPCI recovered
reactor
water level,
but RCIC did not inject since it immediately tripped
on overspeed
and high
exhaust
pressure.
Reactor pressure
was manually controlled by opening Main
Steam relief valves
(MSRV).
False position indication developed
in the
acoustic monitoring system which led the operators
to believe that
a relief
valve was sticking open.
All safety
systems
performed
as designed
except
as
noted.
On
March 19,
1985,
Unit 1
was
scrammed
from
44K
power
as
part of
a
controlled
shutdown
required
by Technical Specification 4.7.A.2:b.
Two
isolation valves
(HCV 71-14
and
HCV 73-23) 'had failed
local leak rate tests
placing the Unit in an
LCO.
Unit 3
was
scrammed
from
47%
power
as part of
a controlled
shutdown.to
investigate
the
source
of
excessi ve
unidentified
drywel 1
1 eakage
on
February 9,
1983.
The unit
was
shutdown
by
rather
than
controlled
rod insertions
since
two
IRMs {B&F) were inoperable.
The leak
was identified on
a 3/4-inch test connection
near the inboard reactor water
cleanup
(RWCU) isolation valve and was determined
to be
a vibration induced
fatigue crack.
There were, no safety
system challenges
during the trip.
Unit 3 was
scrammed
from less than
3% power on March 9,
1985,
as part of a
controlled
shutdown
to
investigate
reactor
water
level
discrepancies
observed
during
the
previous
startup.
There
were
no
safety
system
challenges
during the trip.
The
was
forced
by
(RWM) problems which were eventually traced to omissions
in the
RWM program.
On
June
16,
1984,
the Unit 2 reactor
scrammed
from 60.9%
power
due to
a
false main turbine oil tank low level signal
which initiated a turbine stop
valve closure.
The cause
of the oil tank
low level
signal
coul'd not
be
determined Out
was
believed
to
be
due
to
an
operator
who inadvertently
bumped the switch while performing
a weekly check of the level
No
other safety
systems
were challenged during the event.
Various recurring administrative
errors
were
noted. in the licensee'
post
trip review packages
for these events.
These
items were discussed
with the
licensee
during an exit meeting on"April 26; 1985;
Examples
are
as follows:
Trip Report
Problem
U-l, Trip 176
(1)
STA completed Preliminary Scram Evaluation at
'840
on August 21,
1984.
Trip did not occur until
1440
on August 21,
1984.
10
U-1, Trip 177
U-1, Trip 178
U-3, Trip 116
U"2, Trip 152
(2)
No UO, ASE, or SE signatures.
on GOI-100-11 cover
sheet.
I
(1)
Preliminary Scram Evaluation at
1445
on January
17,
1985.
Thi s
i s
24
hour s after trip
(1440
on
January
16, 1985).
SP 12.8 requires
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
(2)
Independent
technical
review
was
not
independent
(it
was
performed
by
same
man
who
did
the
Preliminary
Evaluation)
and it
was
not
completed within 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />
as
required
by
12.8
(trip at
1440
on
January
16,
1985,
independent
review at January
22,
1985).
(1)
STA did not sign Preliminary Scram Report cover
page.
(2)
Control
rod density listed
as
N/A on Preliminary
Report with no explanation.
(3)
Shift Engineer reviewed Preliminary Scram Report
before
STA completed this ~eport
(SE review at 0155
March 19, .1985,
STA completed
at
0621,
March 19,
1985).
(1)
Independent
technical
review did not meet
32 hour3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />
criteria (Trip at 2315 February 9,
1985,
review at
1200 February
12 1985).
(2)
Shift Engineer decided
not to inform Engineering
Section
Supervisor
to place
the trip on
Immediate
Attention List (IAL) for tracking because
there is
no I.A.L. on the
weekend
and they
were
going
to
restart
immediately.
Procedure
requires placing
on
I.A.L.
During
a routine review
on April 24,
1985, of scram report
number
117 for
the Unit 3 shutdown
conducted
March 9,
1985,
the inspector
noted that
a part
of the
Rod Worth Minimizer (RWM) computer
program was
found to be missing.
During the controlled
shutdown
the
RWM became
When
RWM group
five was completely inserted,
the
RWM latched to group
one rather
than group
four.
Attempts at reinitializing the process
computer
and
RWM to correct
the problem were unsuccessful.
The unit was manually
scrammed.
During troubleshooting
of the
RWM program the alarm message
table portion of
the
program
was
found to be missing;
The
RWM program can
be aborted for a
variety of reasons.
When
an abort occurs,
'a message
appears
on the alarm
typer giving the
reason
for the abort.
All of the
messages
which should
have
been in the message
table were missing.
Below is
a list of the messages:
FAILED
RWM " ROD SELECTED AND DRIVING, ROD NOT SELECTED
11
RWH - INVALID ROD IDENTIFICATION
RWH -
RWM - LOAD SEQUENCES
BEFORE STARTING
RWM " CONTROL
ROD
SCRAM FAILED
RWM - FAILED APPLYING (withdraw or insert) (block or permissive)
LOGIC 0 AND LPAP LOGIC I
RWM - MORE THAN THREE INSERT
ERRORS
RWM - MORE THAN ONE WITHDRAW ERROR
RWM - M.O,D. FAILED
RWM - SEGMENT TRANSFER FAILURE
It was thought that
when the
computer
went to execute this portion of the
program
a message
was sent to the alarm typer which was not understandable.
Each output device
has
a default device
and after attempting
to output to
- all of the available output devices
the computer
locked up.
Discussions with plant computer
personnel
revealed that erratic operation of
Unit 3
process
computer
has
occurred
for the
past
several
years.
The
missing section
of the
program
was discovered
during troubleshooting for a
perceived
hardware
problem believed to be causing erratic operation
of the
computer.
After
loading
the
section
of
missing
program
into
the
write-protected
area of the
computer
memory,
smo'oth operation
of the
RWH
program
was
observed.
The alarm
message
table
was found to be in place in
the Unit I and II process
computer
memory.
This problem occurred
due to
a lack of control in the past for the process
computer.
The plant computer
personnel
have initiated procedures
to make
a
record of the write-protected
area of computer
memory on magnetic tape
each
month.
The previous
month's
tape
would be
used
to verify no changes
had
occurred in the protected
arya of memory.
The
RWM program problems will remain unresolved
pending resolution that the
program
is
functioning
properly
and
review of
computer
software
maintenance
procedures.
This item was discussed
in an exit meeting
with
plant management
on April 26,
1985 (259/85-25-08),