ML18004B850
| ML18004B850 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 05/11/1987 |
| From: | Burnett P, Jape F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18004B849 | List: |
| References | |
| 50-400-87-16, IEIN-86-014, IEIN-86-14, IEIN-87-005, IEIN-87-5, NUDOCS 8706150233 | |
| Download: ML18004B850 (9) | |
See also: IR 05000400/1987016
Text
PS AEVI,
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+a*a+
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.W.
Report No.:
50-400/87-16
Licensee:
Carolina
Power
and Light Company
P. 0.
Box 1551
Raleigh,
NC
27602
Docket No.:
50-400
Facility Name:
Harris
1
License No.:
Inspection
Conducted:
April 6 - 10,
1987
Inspector:
P.
T.
Bu ne t
Date Signed
Approved by
F. Jape,
Chief
Engineering
Branch
Division of Reactor Safety
Date Signed
SUMMARY
Scope:
This routine,
unannounced
inspection
addressed
the review of startup
tests
completed at 50 and
75 percent of rated
power and licensee
response
to IE
Information Notices.
Results:
No violations or deviations
were identified.
870eisoaaa
87asco
ADOCK 05000400
Q
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REPORT DETAILS
Persons
Contacted
Licensee
Employees
- R. A.
J.
L.
"H.
W.
J.
M.
R. J.
"G. L.
- J
L
- A. J.
- C
- C. E.
- J.
R.
- J.
H.
R.
B.
- W. R.
R.
R.
Watson,
Vice President,
Harris Nuclear Project
Wills, Plant General
Manager
Bowles, Director, Onsite Nuclear Safety
Collins, Manager,
Operations
Duncan,
Test Program
Development
Engineer,
Technical
Support
Forehand,
Director Quality Assurance/Quality
Control
Harness,
Assistant Plant General
Manager,
Operations
Howe, Regulatory
Compliance
McKensie, Principal Quality Assurance
Engineer
Rose, Jr., Quality Assurance
Supervisor
Sipp,
Manager
EERC
Smith, Operations
Support Supervisor
Van Metre,
Manager,
Technical
Support
Wilson, Principal Engineer,
Technical
Support
Wojonarowski,
Reactor
Engineering
Leader, Technical'upport
Other
licensee
employees
contacted
included
shift
foremen,
startup
engineers,
control
room
operators,
and office personnel.
NRC Resident
Inspectors
G.
F. Maxwell, Senior Resident
Inspector
S.
P. Burris, Resident
Inspector
- Attended exit interview
Exit Interview
The inspection
scope
and findings were
summarized
on April 10, 1987, with
those
persons
indicated in paragraph
1 above.
The inspector described
the
areas
inspected
and
discussed
in detai
1
the
inspection
findings.
No
dissenting
comments
were received
from the licensee.
Proprietary material
was not reviewed in the course of the inspection.
Licensee Action on Previous
Enforcement Matters (92702)
(Closed)
Violation 86-96-01:
Inadequate
procedure
for measuring
reactor
coolant
system
leakrate.
The
inspector
reviewed
the
revised
procedure
OST-1026,
and
analysed
survei llances
performed
under it using
micro
computer
program
RCSLK9,
from the
NRC Independent
Measurements
Program.
Six
completed
copies
of
OST-1026,
Reactor
Coolant
System
Leakage
Evaluation,
which were performed in early March,
1987,
were reviewed,
and
0
0
the results
compared with calculations
using
RCLSK9.
(System temperatures
ranged
from 345 to 571
F, all at 2235 psig.)
Agreement for both gross
and
was within 0.2
gpm in all cases.
In those
cases
in
which there
was
no net change
in reactor
coolant
system
average
temperature
there
were
no differences
in the gross
leakage
calculations.
The
small
differences
that
arose
when
there
were
changes
in
average
temperature
derived
at least
in part
from the licensee
correcting for both
system
volume
and coolant density
changes,
while
RCSLK9 accounts
only for the
latter.
Whether
the
vessel
and
piping actually
respond
to changes
in
coolant temperature
in the time frame of the test is open to question.
4.
Unresolved
Items
No unresolved
items were identified during this inspection.
5.
Fifty Percent
Power Tests
(72608)
a.
9105-S-01,
Calibration of Steam
and Feedwater
Flow Instrumentation
at
Power -
50% (Retest)
was
performed
March
13,
1987
and accepted
on
April 7,
1987.
The level
2 acceptance
criterion for the
flow transmitter signals
was that they agree within
5% of full scale
d/p of the special test instruments.
This is
a large
number, of the
order of
61
inchs-water,
and
about
one-third
the
reading
at
50%
power.
However,
the
acceptance
criterion
in 9107-S-03
(90%
power
data)
is
no discrepancy
in excess
of 0.5% full scale d/p (approxi-
mately
6. 1
inches
water)
special
test
instrument
accuracy
(1.4 inches - water).
This will translate
to
about
1% instrument
error at full power, which, if attained will be acceptable.
b.
9105-S-05,
Core
Performance
at
50%
Power,
was
completed
on
February
21,
1987,
and
the resul ts were
approved
on March 3,
1987.
The
heat
flux hot
channel
factor
and
nuclear
enthal py ri se
hot
channel
factor each satisfied its technical
specification limit at
50
and
75% power, thus justifying escalation
of power to the
75% testing
plateau.
The maximum quadrant
power tilt ratio was 1.007, well below
the limit of 1.02.
The INCORE-calculated
average
reaction rate error
was 4.7%, again, well below the level
2 acceptance
criterion of 10%.
The reaction
rate error is the difference
between
predicted
power
production
in an
assembly
and the
measured
value.
Determination of
control
rod position
by use of the
incore
nuclear
instruments
was
also
demonstrated
to agree within
12 steps
with main control
board
indications.
The
review included
the
following procedures,
which were
performed
in
support of this test:
(1)
FMP-101 (Revision 2), In-Core Thermocouple
and Flux Mapping,
(2)
EST-710 (Revision 3), Hot Channel
Factor Tests,
(3)
EST-722
(Revision
0),
Control
Rod Position
Determination
Via
Incore Instrumentation,
and
(4)
EPT-052
(Revision
0),
Power
Range
Heat
Balance
Via Precision
Calorimetric.
9105-S-06,
Thermal
Power Measurement
and Statepoint
Data Acquisition
at
50%
Power
(Retest
2)
was
performed
on
March 13,
1987
and
the
results
accepted
on April 7,
1987.
The
test
had
no
level
1
acceptance
criteria
and
was judged
successful
when acceptable
data
were obtained,
submitted for use
in other tests,
and
thermal
power
calculated.
9105-S-07,
NIS Overlap Verification,
Power
Range
Calibration
and
Setpoint
Adjustment -
50%
was
completed
on
February
23,
1987
and
approved
on
March 3,
1987.
Adjustment of all
four
power
range
nuclear
instruments
was within two percent
of the indicated thermal
power
as determined
by test
procedure
9105-S-06.
Plots of chamber
current versus
power at zero,
30,
and
50% thermal
power all appeared
to be linear.
9105-S-09,
Power
Coefficient
at
50%
Power,
was
performed
on
February
22,
1987,
and the results
were
accepted
on
March 3,
1987.
Xenon stability during the test
was confirmed by use of the computer
code
EXSPACK, which indicated that prior to starting the test
was changing
by about 6pcm/hr.
The total reactivity worth of xenon
was about
2450
pcm.
The initial thermal
power
was determined
using
procedure
OST-1004,
Power
Range
Heat Balance-Daily Interval.
9105-S-11,
Loss of Feedwater
Heaters
Test at
50% Power,
was performed
on
March 29,
1987
and
the results
accepted
on April 7,
1987.
There
were
no level
1 acceptance
criteria.
The level
2 criterion was that
no
measurement
of feedwater
temperature
drop
more
than
44
F.
The
three
recorded
measurements
of temperature
drop were
11, 5,
and
11F;
satisfying the criterion.
9105-S-14,
System
Flow Measurement
at
50%
Power,
was
performed
on
February
22,
1987
and
the results
were
accepted
on
February
25,
1987.
The measured
flow was
309420
gpm, which satisfied
the acceptance
criterion that flow be greater
than
298948
gpm, which
includes
allowances
for error.
Data
sheet
10.3
and
acceptance
criterion 7.2. 1 indicate the flow rate is 309970
gpm, in contrast to
the value
above
from section
1 of the procedure.
The data
used were
obtained
from test procedure
6105-S-06.
Data sheet
10. 1 demonstrates
that the
average
reactor
coolant
system
(RCS) temperature
and
steam
generator
and pressurizer
levels were stable
over the period of the
test.
The final determination of reactor coolant
system flow will be
performed
using
a precision
heat
balance
once
the reactor
achieves
100% rated thermal
power.
0
h.
9105-S-18,
Preliminary
Incore/Excore
Calibration,
was
performed
on
February
22,
1987 and accepted
on March 3,
1987.
Performance
at this
power
level
precedes
the
Technical
Specification
requirement
to
perform at
75% power.
Hence,
the use of only two quarter-core
flux
maps
at
only
two different axial flux differences
is acceptable.
This test
provided
an opportunity
to
make
practical
use
of the
engineering
and
. maintenance
surveillance
tests
(ESTs
and
MSTs,
respectively) that were to be applied in the required surveillance.
No violations or deviations
were identified.
Seventy-five
Percent
Power Tests
(72616)
Several
70% power tests
were left to
be performed at the
time of
this'nspection.
Most of the
completed
tests
were
in the
review cycle.
However,
data
from the
incore-excore
calibration test at
75% power were
available,
and
the
inspector
performed
independent
calculations
of the
least
squares fit of axial offset to chamber current'or
each of the eight
chambers.
Exact
agreement
on zero-offset,
full-power current,
slope,
and
correlation coefficient was obtained in every case.
Five data pairs
were
used
in
each
calculation
and all correlation coefficients
were greater
than 0.995.
No violations or deviations
were identified.
7.
Licensee
Followup of IE Information Notices (92703)
PUMP
TURBINE
CONTROL
PROBLEMS
and
SUPPLEMENT
1:
OF
AFW,
HPCI,
AND RCIC
TURBINES have
been
considered
by
the
licensee.,
The
review of
the
original
notice
was
completed
on
April 16,
1986
and
the
review of the
supplement
was
completed
on
January
19,
1987.
Based
upon
discussions
with
representatives
of
Woodward
Governor
and utilities experiencing
mechanical
trips
from oil pressure
in the governor,
licensee
engineers
concluded
that
no
similar
mechanism
existed
at
Harris.
They did,
however,
identify
a
potential for condensate-buildup-induced
overspeed trips and
made specific
recommendations
for corrective action.
The modification portion of the
corrective action
had not been
implemented at the time of this inspection.
trips of the turbine driven
AFW pump
were
experienced
during
the remote
shutdown
and loss of offsite power tests.
Initially the
cause
was assigned
to
a loose electrical
speed
probe
connection.
Maintenance
findings and post-maintenace
testing
appeared
to justify that conclusion.
Subsequently,
additional
trips were experienced,
and they were
ascribed
to
condensate
buildup in the
steam
supply lines.
Currently,
the
condensate
is being drained
several
times
per shift in response
to
alarms,
and the recent experience
has sensitized
the operators
to the
need for prompt response
to the alarms.
Corrective
modifications
are
currently
being
planned.
The
need
for
modifications to eliminate
condensation
in the
steam
supply line to the
pump turbine
stems
from the current practice of keeping
the
steam
supply
valve closed.
This results
in unheated
line length of over
100 feet in
which
steam
leakage
past
the
supply
valve
can
condense
without being
drained
by automatic action.
If the supply were controlled at the turbine
governor valve,
there
would be less
than three feet of line to heat
on
pump start,
and condensation
in the remainder of the
steam line would be
forced through
the condensing
pot drains
by
steam pressure.
Apparently,
the decision
to control
steam
at
the
supply
valve
was
made
to avoid
performing high energy line break analysis
to the
100 plus feet of steam
line
between
the
supply
valve
and
the
governor
valve.
Subsequent
discussions
with members, of the Plant Systems
Branch,
confirmed that
classifying the
steam line as
a non-high energy line in the current
mode
of operation
was acceptable.
MISWIRING IN WESTINGHOUSE
ROD
CONTROL
SYSTEM was
reviewed
by the licensee,
who prepared
a test
procedure
to
determine if the
problem
existed
at Harris.
That test,
EPT-041T,
was
performed
on March 30,
1987.
A review of the
completed test
procedure
confirmed that it was
responsive
to the
concern
expressed
in the notice
and that the wiring error did not exist
in the Harris
1
rod control
system.
No violations or deviations
were identified.