ML17342A964
| ML17342A964 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 10/08/1987 |
| From: | Burnett P, Jape F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML17342A961 | List: |
| References | |
| 50-250-87-41, 50-251-87-41, NUDOCS 8710160225 | |
| Download: ML17342A964 (17) | |
See also: IR 05000250/1987041
Text
UNITED STATES
NUCLEAR REGULATORY'COMMISSION
REGION II
ATLANTA,GEORGIA 30323
Report Nos.:
50-250/87-41
and 50-251/87-41
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:
September
21 - 25,
1987
Inspector:
. T.
8 rn tt
Approved by:
F. Jape,
Sect>on Chief
Engineering
Branch
Division of Reactor Safety
lo
ate
>gne
ate
cygne
SUMMARY
Scope:
This routine,
unannounced
inspection
addressed
the areas
of review of
post-refueling
startup tests
(Unit 3), evaluation of the thermal
power mea-
surements
(Unit 4),
and followup on outstanding
items.
Results:
One violation was identified:
Failure to perform adequate
post-
modification testing - paragraph 7..
E
87 )0f60225: 87~0~PgP
ADOCK 0
pD
Q
REPORT DETAILS
Per sons
Contacted
Licensee
Employees
- C. J. Baker, Plant Manager
J. Arias, Jr., Regulatory Compliance Supervisor
- W. Bladow, guality Assurance
Superintendent
- J.
W. Brown, Technical
Department
- R. J. Earl, guality Control Supervisor
- S.
D. Ferrell, Licensing Engineer
- D. D. Grandage,
Operations
Superintendent
D.
W. Hasse,
Safety Engineering
Group Supervisor
- V. A. Kaminskas,
Reactor Engineering Supervisor
C. Lenhart,
DDPS Coordinator
- G. L. Marsh,
Reactor
Engineer
- G. Salamon,
Compliance
Engineer
B. Shimkus, Plant Supervisor -
Nuclear
- J. C. Strong,
Maintenance
Superintendent
Other
licensee
employees
contacted
included
engineers,
operators,
and
office personnel.
NRC Resident
Inspectors
D. R; Brewer, Senior Resident
Inspector
- J.
B. Macdonald,
Resident
Inspector
- T. F. McElhinney, Resident
Inspector
"Attended exit interview
Exit Interview
The inspection
scope
and findings were
summarized
on September
25,
1987,
with those
persons
indicated
in paragraph
1 above.
The inspector
de-
scribed
the areas
inspected
and discussed
in detail
the inspection find-
ings.
No
dissenting
cooments
were
received
from the
licensee.
Proprietary
material
information
was
reviewed
in the
course
of this
inspection,
but is" not
included
in this report.
One violation
was
identified:
VI0.250/251/87-41-01:
Failure to perform adequate
post modification
testing - paragraph
7.
3.
Licensee Action on Previous
Enforcement Matters
(Closed)
VIO 251/87-16-03:
Inadequate
procedure
for surveil lance
of
reactor
coolant
system
leakage.
Revised
procedure
4-0SP-041.1
was
approved
on
May 29,
1987.
The inspector
reviewed three recently complet-
ed examples
of the procedure
and compared
the results with those obtained
.
using microcomputer
program
RCSLK9.
Acceptable
agreement
was obtained in
all cases.
A similar revision
has
been
made to 3-OSP-041. 1, the procedure
for Unit 3.
This item is closed.
'Closed)
UNR 250/251/87-34-01:
Absent
records
for post modification
testing required to
be performed following modifications
made in response
to
IEB 80-06. It has
been
concluded that the required testing
was
not
performed prior to placing the modified systems
in operation.
A violation
will be issued
responding
to that failure, see
paragraph
7.
This item is
closed.
4.
Unresolved
Items
No unresolved
items were identified.
5.
Post-Refueling
Startup Tests
- Unit 3 (72700,
61708,
61710)
a.
OP 0204.3
(7/31/86), Initial Criticality after Refueling,
was
begun
on 9/4/87
and completed
on 9/5/87.
Prior to pulling rods,
both source
range
channels
were tested for
operability using the chi-squared
test.
The inspector
independently
verified the
analyses
from the
raw data.
Both systems
were well
behaved.
The use of the chi-squared test confirms the detectors
are
responding'proportionally
to neutrons,
an assurance
not obtained
from
simply satisfying
the operability surveillance
required
in the
technical specifications.
With an initial boron concentration
of 2080
ppm,
shutdown
and
then
control
banks
were withdrawn in 50 step
increments until
D bank was
at
160 steps.
The inverse count rate ratio (ICRR) was calculated
and
plotted for each
increment.
The final
ICRR for the source
range
was
0.59.
The
ICRR was renormalized to 1.0 and dilution was initiated at
the rate of 100
gpm until the
ICRR reached
0.53
on the source
range.
At that point, dilution was reduced to a rate of 50
gpm
and contin-
ued until criticality was achieved.
The logged critical configura-
tion was
1710
ppmB in the reactor coolant system
(RCS) with
D bank
inserted- to 117 steps
to level off the flux.
The 43 steps
on bank
D
had
an approximate reactivity worth of 107 pcm, the equivalent of 14
ppmB.
A similar reactivity overshoot
had
been
observed .in the review
of the last post-refueling startup of Unit 4 (see
inspection report
251/87-31).
It would appear
prudent to stop dilution earlier,
such
as at
an
ICRR of 0.2,
and allow criticality to occur during mixing
with additional
rod withdrawal
as
needed.
This approach
would also
reduce
the
amount of water processing
required for the tests.
This
consideration
was discussed
at the exit interview.
The nuclear
heating flux level
was determined,
and. an
upper flux
level for zero power tests established
below the heating
range.
The reactivity computer
was
checked
out..
Reactivities
determined
from each
of four stop-watch
periods
agreed
with the reactivity
computer within 2X.
From these,
the calibrated
range of the
reactivity computer
was from -36 pcm to +38.5 pcm.
b.
OP 0204.5 (10/29/86),
Nuclear Design
Check Tests
During
Startup
after Refueling,
was
begun
on 9/5/87.
The measured
all-rods-out
(ARO) boron concentration
was
1711
ppmB and
.agreed with
the predicted
value (from WCAP-11454) of 1744 within 50
ppmB, thus, satisfying the acceptance
criterion.
The
isothermal
temperature
coefficient for ARO was
measured
to
be
+0.96 pcm/F, the average of two heatup
and two cooldown measurements.
The average
was corrected for
a doppler coefficient of -1.9
pcm/F
yielding
a moderator
temperature
coefficient
(MTC) of +2.86 pcm/F,
which
was
in
good
agreement
with the prediction of -2.76 for the
prevailing
temperature
and'oron
concentration.
The
Technical
Specification
3. 1.2. 1 limit is +5.0 pcm/F.
The inspector
indepen-
dently verified the test results
from analysis of the
raw data.
No
measurement
with one rod bank inserted
was performed.
However, the
fuel
vendor,
provided
a moderator
temperature
coef-
ficient control correlation to extrapolate
the
measured
zero-power
. MTC to other .core conditions
described
by the boron concentration
(C ), core
average
moderator
temperature,
and control
rod position.
Th) constants
of the algorithm
are
cycle-dependent.
Using the
algorithm, the licensee
generated
curves for the 'plant curve book to
assist
the operators
in maintaining core
parameters
so that the
MTC
would trend to zero at full power.
.The inspector applied the algo-
rithm to specific core conditions
and concluded that maintaining the
MTC less
than zero at
power would not
be difficult throughout the
cycle.
Control
rod worths were determined
by rod swap.
The reference
bank
was bank
C, which had
a measured
worth against
boron dilution of 1325
pcm.
The remaining, bank worths
were
determined
by rod
swap
and
ranged
from -5X less to
7C more than prediction.
The total rod worth
. was 0. 15% more than predicted.
The reactivity computer stayed within
the calibrated
range throughout the test.
~ t
Using the integral
worth of control
bank
C and the change
in boron
concentration
durin'g that measurement,
a differential boron worth of
7.57
pcm/ppmB was calculated.
Using data obtained during the measurement
of bank C, the boron worth
'was calculated
to be 7.57 pcm/ppmB;
11.5X less
than predicted.
The inspector
reviewed
the strip chart records
from the reactivity
computer for the
C bank measurement
and independently
confirmed the
differential
and integral worth curves.
Attachment I is
a graphical
display of the differential worth curves
determined
by the inspector
and the licensee.
The unusual
shape of the differential worth curve
is the result of part-length
burnable
poison rods being installed at
the core vertical center line.
The reference
document
for predicted
test results
was
The
Nuclear Design
and
Core Management of the Turkey Point
3 Power Plant .Cycle
ll, (PROPRIETARY).
No violations or deviations
were identified.
6.
Thermal
Power Monitoring - Unit 4 (61706)
The
microcomputer
program
TPDWR2,
which is described 'n
NUREG-li67,
TPDWR2:Thermal
Power Determination for Westinghouse
Reactors,
Version 2,
was
used to evaluate
plant raw data to make
an independent
assessment
of
the licensee's
adherence
to the rated thermal
power
(RTP) limit.
In order to customize the program for use
on Turkey Point 4, the following
documents
were reviewed to obtain plant specific data:
a.
System Description 9: Pressurizer
and Relief System,
b
System Description ll: Steam Generators,
c.
Technical
Manual
1440-C92, Pressurizer,
d.
Technical
Manual
1440-C302,
e.
Plant curve Book,Section VI, and
f.
Revised Final Safety Analysis Report.
This search
did not provide all of the parameters
required.
The steam
generator
dome inner diameter
and riser outer diameter were estimated
from
small scale
drawings.
The reactor coolant
pump efficiency, steam
genera-
tor moisture carryover,
and insulation heat losses
were obtained
from the
default values built into the
program. 'he
steam
generator
manual
had
been
updated to describe
the the tube bundle replacement.
However, later
discussion
with licensee
personnel
revealed
that the risers
had
been
modified internally
by replacing
the swirl vane separators
with tubular
steam
separator s.
The total effect of the lack of these
plant specific
parameters
is estimated
to
b'e less
than
3 Mwth in a measurement
of the
order of 2200
Mwth.
Nevertheless,
to
improve future measurements,
a
member of the licensee staff has
agreed
to obtain better
values for the
estimated
and default parameters.
The
licensee's
procedure
for reactor
heat
balance
is
contained
in
4-0SP-059.5,
Power
Range
Nuclear
Instrumentation Shift Checks
and Daily
Calibrations.
Although the calorimetric portion is usually performed
by a
program in the digital data
processing
system
(DDPS), Attachment
4 is the
manual
equivalent
using
the
same
or alternate
data
sources,
and is in
effect
a description of the
DDPS calorimetric.
The procedure
does
not
perform
a separate
heat balance
on each
as
does
TPDWR2 and
all other
licensee
procedures
reviewed to date.
Instead
the performance
of an average
is calculated
based
upon average
temperature
and pressure,
average
steam
generator
pressure,
and average
feedwater flow.
To obtain the latter, the
DDPS across
the individual flow
venturis,
in units of inches
of water,
are
averaged,
square-rooted,
multiplied by the thermal
expansion factor, available
from the plant. curve
book in the manual calculation, multiplied by the squareroot
of feedwater
density,
and, finally, multiplied by
a flow constant.
The flow constant
is the average
calibration factor of the three calibrated
feedwater flow
venturis.
The
DDPS calorimetric requires
four minutes
to perform.
Three calori-
metrics were performed in a
25 minute span.
At the start
and near the end
of each,
raw data for use in
TPDWR2 were obtained
from the
DDPS and the
ERDAS computer.
These
data pairs were averaged for input in the compari-
son calculation
by TPDWR2.
Considerable
manipulation of the
raw data
was required before they could
be input to
TPDWR2.
All pressures
had to be
changed
from gauge to abso-
lute;
steam
generator
and pressurizer
levels
had to
be converted
from
narrow range
level in percent to absolute
level in inches;
and cold-leg
temperatures
had to be averaged
to a single value.
The raw data indicated that feedwater
temperature
was constant
throughout
the tests
and that the variations in feedwater
pressure
were insignifi-
cant.
Therefor e, constant
values of thermal
expansion factor and square-
root of feedwater density were used in all calculations of feedwater flow.
The individual venturi flow constants
were obtained
from a system engi-
neer's
notebook.
The average of the three values
was less
than the single
flow constant
used
in the licensee's
calculation.
The individual flow
constants
were multiplied by the ratio of the
averages
to provide
an
adjusted
average
equal,to that used in the procedure.
Finally,
a single
constant
was .derived for each
steam
generator
to convert the variable
dP(in-H 0) to feedwater flow in millions of pounds
per hour.
All of the
manipulftions of data
were
performed
using the
SUPERCALC
3 spread
sheet
program;
The
spread
sheets
for the
raw
and .final data
are
given in
Attachment 2.
Typical results
from
TPDWR2 are
given in Attachment
3.
The agreement
between
TPDWR2
and
the
DDPS .calorimetric
was excellent,
with
TPDWR2
averaging
0.17K
RTP less
than the 'calorimetric for the three measurements.
Increased
confidence
in TPDWR2 results
could be obtained
by resolving the
question
of proper
flow coefficients for the
three
flow
venturis,
consideration
of measured
heat
loss
and reactor
coolant
pump
efficiency, and by a'better
physical description
of the steam generators.
Based
upon the agreement
obtained,
the licensee's
method of performing the
reactor
heat
balance
(calorimetric) is acceptable.
However, it does
not
provide the ability to trend the performance of individual
steam
genera-
tors inherent in other methods.
No violations or deviations
were identified.
Followup of Unresolved
Item (92701)
(Closed)
UNR 250/251/87-34-01:
Demonstrate
that
systems
modified in
response
to IE Bulletin 80-06 were tested following the modifications.
The licensee
was unable to find records of appropriate
post-modification
tests
following modifications
PC/M 80-51,
PC/M 80-52,.PC/M
80-167,
and
PC/M 80-168,
which were completed during refueling outages
that started
in
1981.
The modifications affected
the post-safeguards-initiation
reset of
control to the normal containment coolers,
containment isolation valves
2911,
2912,
and
2913,
and containment
pumps.
A licensee
review of
the Engineered
Safeguards
Integrated
Test
(procedures
3/4-0SP-203),
which
is
performed
following each
refueling
outage,
revealed
that not all
modified functions
were
tested
in that procedure.
In 3/4-OSP-203
the
safeguards
trains are activated
simultaneously,
and
a train-by-train test
is required for those
components
powered from telemand
swap motor control
centers
(MCC).
The licensee
then wrote special
test
procedures
to accomplish the neces-
sary testing.
TP-374 for Unit 3 and TP-375 for Unit 4 were both completed
on
September
2,
1987.
Review
by quality control
was
completed
on
September
16,
1987.
The Unit 4 test
revealed
that the
B and
C normal
containment
coolers
did not trip off because
of disconnected
This
is
more evidence
that
adequate
testing
had not been
performed following
the modification; since
there
are
no maintenance
work orders
on record
that could have led to the disconnections.
The system, functioned properly
after the leads
were connected.
The initial test failure led the licensee
to test other safeguards
compo-
nents
drawing power from telemand
swap
MCCs.
These tests
were performed
under
and TP-378 for Units
3 and 4, respectively.
The components
tested
were
blowdown isolation valves,
component cooling
water
system inlet, valves,
steam
generator
water
sampling
valves
and
control building isolation dampers.
All tests
were satisfactory.-
Discussions
with licensee
personnel
revealed
that since the time of the
modifications addressed
here there
have
been major improvements
in assur-
ing and controlling post-modification testing.
These include:
'a ~
Revision of guality Instruction 3.1, Control of Design
Performed
by
JPE,
to require
engineering
design
packages
to contain
a startup
testing
section.
That section
provides
guidelines
for startup
testing
procedures
as
recommended
by manufacturers,
required
by
technical
specifications,
or desired
by engineering.
Testing
requirements
must
be specified
in the design
package
when necessary
to assure
the modification does not alter the plant's
design basis.
b.
A Startup
Department
was organized
as
part of the performance
en-
hancement
program
(PEP).
That department
is responsible for system
acceptance
walkdowns,
system 'acceptance
testing
and other duties to
assure
a system will work as designed
before being turned over to the
plant for operation.
c.
Administrative
procedure
0190. 15,
Plant
Changes
and Modifications,
requires
the plant technical
staff perform
an engineering
review
following a
PC/M implementation
and that quality control perform
a
post-implementation
review.
d.
Revsion of 3/4-OSP-203 to require train-by-train safeguards
tests
to
assure
proper function of components
powered
by telemand
MCCs is in
process.
Failure to perform adequate
post-modification testing prior to placing the
affected
'components
in service
has
been identified as
a violation against
10CFR50 'Appendix B, Criterion
11 (VIO 250/251/87-41-01).
.ATTACHMENTS:
1.
Control
Bank C, Cycle XI
Differential Worth
2.
Raw Data for Thermal
Power
Calculation
3.
Plant Parameters
ATTAC
1
ROL BANK C, CY
XI, TURKEY POI
Eiffelenf~al I'mph
M~ Inspector
~ Licenses
10
0
ATTACHM
TP4. Rm Data Ar Than+I pawr Calculahion
generator
A
Beneratcs
8
~ Bai~r, C
Faeclsater
Presses
kza.
Reactor
Thaa
Sha P
Flo dP Level
Sta P
Flo dp Lovel
St@ P
Flo dP
Level
Tasp
pres
Cpsig> Cin H20
CX)
Cpeig> Cin H20
CX>
Cps'> Cin H20>
CX>
Cdeg R
Cpsig>
Fras
Level
T Av
Cpsig>
<X)
<deg F>
TC A
TC B
LD Flo
CHRB Flo
QRS T
deg F)
Cdeg F>
<<pe)
Cpm>
Cdag F>
13.49
81F 49
180 '?2
55o69
832 SF
IIS 83
59o81
830e99
189.14
13 54
816 74
182e 15
55e51
832 12
1M 43
59o75
830 06
189e 00
14 00
815 99
182e95
55.lH)
8%499
189 44
59e62
829 8?
18L13
14:06
816 18
182e22
55e3?'30e43
189e44
59o99
829 31
IM 34
14&6
815.06
1?9e93
55 49
830 62
Qlge21
59eBO
829 31
185 67
14 19
815e99
182o43
55o55
8X481
186e65
60e3F
828e?%
189o 14
60 00
59o81
60o 12
60.24
60e18
59 SF
430.00
430 00
430.00
430o00
430.00
429o9F
1073.9
1D?5 5
1071 5
10F3o9
1073o4
10?6o4
2255.7
51.88
2235o5
51.95
2235 I
51.99
2254e6
51e92
2254o6
51 86
2235e2
51 ~ 86
573.50
5?3.86
5?3.?2
5?3o84
546.83
546.81
546,98
546 76
573 59 546.88
5?3o65
546 79
547 48
54?.53
547 53
547o50
54F.62
547 47
52.42
52e?5
52.81
52o74
52.61
52o48
32 48
32 26
31 93
32 91
511 23
510e91
512 48
512o79
512 46
511e53
TP4e Input Data For Powr Calculahion
Staaa generator
A
Stean Operator B
Staaa Banerator
C
Letdovn Line
Charging Line
P
FM Flo
FM T LGKL
P
FM Flo
FM T LEVB.
P
'M Flo
FM T LELEL
Flou
Tesp.
Flou
Teap,
Cpsia) CHIbhw.) Cf)
Cins)
Cpsia) <Hite) CF)
Cine)
(psia>
CIGbhw ) CR
Cine)
Cgpa)
CR
Cpm)
CF)
(psi
Level
T eva
T cold"
<R
CF>
831 8
3 1380
4%)
531 4
84?o2
3 2025
430
53?o5 '45 2
3 2D11
430
53? 5
52 6
54F 5
32 9
51io1
2250 3
2LLO
5?3 7
54?e2
850 8
3 14?9
430
531 5
845 4
3 2094
430
53?o5
844 3
3 1941
430
53?o5
52 8
54F 5
32 4
512 6
2249 6
213 2
573 8
54F 2
830,2
5 1359 ~
531 4
845 4
3 i966
430
53?o9
843,7
3 18?0
430
53?o9
52 5
54F 5
32 4
512 0
2249I6
212 8
5?3o6
547 2
ATTACH)IENT 3
PLANT PARANETERSc
HEAT BALANCE DATA
TURKEY POlNT I
9-22-87
REACTOR COOLANT SYSTEN
Puap
Power
(NM each)
Puap Efficiency (l)
Pressurizer
Inside Disaster
(laches)
STEAN BENERATORS
Done Inside Disaster
(inches)
Riser Outside Diaaeter
linches)
Nuaber of Risers
Noistwe Carry-over
(9) in A
Nofsture Carryover
(L) fn B
Moisture Cary-aver (I) fn C
4.1
90.0
84.0
160.00
52,00
0. 125
0.125
0.125
NONREFLECTIVE INSULATION
lnsfde Surface Area (sq ft)
Thickness
(fnches)
Theraal Conductivity (BTUs/hr ft F)
LICENSED THERNAL )ONER (NMt)
7IBS)
4.0
0.035
REFLECTIVE INSULATION
Inside Surface
Area (sq ft)
11,122
Heat Loss Coefffcfent (BTUa/hr sq ft)
SS.OO
DATAc
TINE
STENI SENERATOR A
Stean
Pressure
lpsfa)
Flow (E6 lb/hr)
Teaperature
(F)
Surface
Bfowdown (gpa)
Bottoa Blowdown (gpa)
Mater Level (inches)
SET
1
SET 2
1354
1419
831.8
830.2
3.138
3.136
QO.O
QO 0
0.0
0.0
0.0
0.0
531.4
S31.4
TINE
STENI BENERATOR B
Stean Pressure
(psfa)
Flow lE6 lb/hr)
Teaperature
(F)
Surface
Biowdown Cgpa)
Bottoa Bfowdown (qpa)
Mater Level
Cinches)
SET
1
SET 2
)354
1419
847.2
845.4
3.203
3.197
QO.O
430.0
0.0
0.0
0.0
0.0
537.5
537.9
STEAN GENERATOR C
Steaa
Pressure
(psfa)
Flow (E6 lb/hr)
Teaperature
(F)
Surface
Bfowdown lgpa)
Bottoa Blo>>down (gpa)
Mater Level (inches)
845.2
843.7.
3.201
3.187
QO.O
QO.O
0.0
0.0
0.0
0.0
537.5
537.9
. LETIHNN LINE
CHN81NS LINE
Flow (gpa)
Teapicrature
(F)
52.6
52.6
547.5
547.6
Flow (gpa)
Teaperature. (F)
32.9
32.4
511.1
512.0
PRESSURIZER
Pressure
(psia)
Mater Level (inches)
2250.3
2249.6
213.1
212.8
REACTOR
T ave
CF)
T cole)
CF)
573.7
573.6
547.2
547.2
ATTACHMENT 3
HEAT BALANCE
TURKEY,POZgT 4
9-22-87
DATA SET
1
OF 2
1354 hours0.0157 days <br />0.376 hours <br />0.00224 weeks <br />5.15197e-4 months <br />
STEAN SENERATOR A
ENTHALPY.
(BTUs/lb)
FLOW
(E6 lb/hr)
POWER
(E9 BTUs/hr)
POWER
(NWt)
Steam
Surf ace
Blowdown
Bottom Blowdown
Power Dissipated
STEAN GENERATOR B
Steam
Surf ace Blowdown
Bottom Blowdown
Power Dissi pated
STEAN GENERATOR C
1197.6
408. 3
515. 3
460. 3
1197.2
408. 3
517. 9
461 5
3.138
-3.138
0 00000
0 00000
3. 202
203
0. 00000
0 00000
3 758
-1 281
0 00000
0 00000
2 4768
h. 833
-1 308
0.00000
0 00000
2.5258
725. 4
739. 7
Steam
Surf ace Blowdown
Bottom Blowdmm
Power Dissipated
TNER
Letdown Line
Charging Line
Pressuri zer
Puaps
Insulation Losses
Power Dissipated
REACTOR POi4ER
1197. 2
408. 3
517 6
461 4
544
1
500. 3
701 3
3 201
-3 201
0 00000
- 0 00000
0 01984
-0 01297
~.00005
3 832
-1 307
0 00000
0 00000
2.5249
0 01079
-0.00649
-0 00004
~ 03750
0 00100
~ 03223
-9 4
2195 2
ATTACHMENT 3
DATA SET 2 OF 2
~1419'our+
STEAN GENERATOR A
HEAT BALANCE
TURKEY POINT 4
9-22-87
ENTHALPY
FLOW
POWER
(BTUa/lb)
(Eb lb/hr)
(E9'TUa/hr)
POWER
(NWt)
Steam
Surface
Blawdawn.
Bottom Blowdawn
Power Dieaipated
Steam
Surf ace
Blawdown
Bottom Blawdawn
Power Dfsarfpated
STEAN GENERATOR C
Steam
Surface Blawdown
Bottom Blowdawn
Power Dieaipated
OTHER CCNPONENTS
Letdawn Line
Charg fng Lgne
Pressur fzer
Pumps
Xnaulatf on Laaaea
1197 7
408. 3
515 0
460 2
1197 2
408. 3
517 6
461. 4
1197 3
408 3
517-3
461.2
544
1
501 .4
701-3
3. 136
-3. 136
0.00000
0.00000
196
-3. 19'7
0 00000
O.OOOOO
3 187
187
'
00000
0 00000
0. 01982
-0 01277
~.00005
3. 756
-1.280
0 00000
0.00000
2.4753
827
-1 305
0 00000
0 00000
2 5213
3 815
-1'01
0. 00000
0 00000
2.5139
0 01078
.=0 00640
~ 00004
-0.03750
0 00100
725. 0
738. 4
736. 3
Pawer Diaaf pated
REACTOR
POINTER
~ 03216
-9 4
219'0 2