ML20150C128
| ML20150C128 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 06/23/1988 |
| From: | Burnett P, Jape F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20150C120 | List: |
| References | |
| 50-348-88-20, 50-364-88-20, NUDOCS 8807120321 | |
| Download: ML20150C128 (13) | |
See also: IR 05000348/1988020
Text
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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101 MARIETTA STREET,N.W.
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ATLANTA, GEORGI A 30323
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Report-Nos.: 50-348/88-20 and SC-364/88-20
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Licensee:
Alabama Power Company
600 North 18th Street
Birmingham, AL 36291
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Docket Nos.:
50-348 and 50-364
License Nos.:
Facility name:
Farley 1 and 2
Inspection Conducted: May 16 - 20 and June 6 - 10, 1988
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Inspector:
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b - 2 b fB
g4. P. T. Burnett
Date Signed
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Approved by:
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dedsr
- Date Signed
F. Jape, Section Chief
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Test Programs Section
Engineering Branch
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Division of Reactor Safety
SUMMARf
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Scope:
This routine unannounced inspection addressed the areas of witnessing
post-refueling startup tests, review of completed core surveillance
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procedures, independent measurements of reactor thermal power and
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reactor coolant system leakage, and review of the licensee's related
procedures,
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Results: One violation was identified.
The procedure used to calculate
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reactor coolant system inventory was inadequate in that the constant
used to make corrections for changes in pressurizer level was neither
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correct nor conservative (Violation 348,364/88-20-03) - Paragraph 5.
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Management made a commitment to evaluate the feasibility of moving
the source range detectors to a region of lower flux so that
criticality would occur below P-6 (Inspector Followup Item 348,364/
88-20-01) " Paragraph 2.
The licensee n.ade a commitment to upgrade the U1118 plant computer
calculation of thermal power (Inspector Followup Item 348,364/88-
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20-02) - Paragraph 4.
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8807120321 880628
ADOCK 05000348
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REPORT DETAILS
1.
Persons Contacted
Licensee Employ'<es
- R. G. Berryhill, Systems Performance and Planning Manager
J. A. Collier, Junior Engineer
- R. D. Hill, Assistant General Plant Manager - Plant Services
W. S. MacDonald, Reactor Engineer
- R. H. Marlow, Technical Supervisor
- C. D. Nesbitt, Technical Manager
- J. K. Osterholtz, Operations Manager
- W. D. Shipman, Assistant General Plant Manager - Operations
- J. J. Thomas, Maintenance Manager
- J. D. Woodard, Vice President
Other licensee employees contacted included, operations personnel,
security force members, and office personnel.
Other Organization
L. Grobmeyer, Westinghouse
NRC Resident Inspectors
- W. H. Bradford, Senior Resident Inspector
W. H. Miller, Resident Inspector
- Attended exit interview on May 20, 1988.
- Attended exit interview on June 6, 1988.
Acronyms and initialisms used throughout this report are listed in the
final paragraph.
2.
Post-Refueling Startup Tests (72700, 61708, 61710)
a.
Test Witnessing - Unit 1
Jhe inspector witnessed the initial criticality of Unit 1 to start
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Cycle 9.
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Prior to pulling the shutdown banks, a statistical reliability check
(Chi-squared test) of the two SRNIs, N31 and N32, was performed
successfully using procedure FNp-0-ETP-3635 (Revision 1), Reliability
Check of Source Range Instrumentation.
The initial countrates were
in excess of 500 cps.
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After the two shutdown rod banks were withdrawn, countrate data were
i obtained for reference countrates for ICRR calculations. 'These
calculations were performed and the results plotted every 50 steps as
the control banks were withdrawn in overlap to a final configuration
of 0 bank at 190 steps.
After the first 50 steps increment, no
. additional rod withdrawal was performed until the ICRR' plot for both
SRNIs confirmed that criticality was not likely in the next incre-
ment. At the end of rod withdrawal, SRNI countrates were in excess 1
of 3000 cps.
While observing these activities, the inspector noted that the two
data takers were not well prepared to understand and perform their
duties.
Both needed training on data collection and use of the data
sheets.
It was necessary for the test engineer on shift to provide
instruction that more properly should have been performed as part
of a pretest briefing. Before the next phase of the approach to
criticality was performed, the data takers were briefed on the test
and their responsibilities.
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A dilution rate of 20gpm was begun using the alternate dilute mode,
and with it a new set of ICRR plots was started. Once those plots
showed the trend of reactivity increase, the dilution rate was
increased successively to 30 and then 45gpm.
Throughout the rod withdrawal phase and in the early stages of
dilution, the inspector made independent checks of the reliability of
the SRNI by performing Chi-squared tests based upon three to five
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observations of countrate.
All tests were successful until the
countrates exceeded 10,000 cps, af ter which excessively large values
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of Chi squared were calculated, probably the result of high resolving
time losses in the system.
P6, the source range block permissive,
was obtained well in advance of criticality, and the remainder of the
approach to criticality was monitored solely by the two IRMs.
N36
appeared to be noisy, but responding, while N35 was well-behaved.
At one point, prior to P6, instrument technicians took N36 out of
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service to adjust the compensating voltage in an attempt to reduce
its irratic reading.
Dilution continued; since no mode change was
imminent. Criticality was announced after about a four-fold increase
in flux above P6.
It appears the SRNIs are too sensitive, or located in a region of too
high flux, to satisfactorily monitor the entire approach to criti-
cality.
Thus, as in this instance, the final stage was monitored by
the IRMs, which had yet to demonstrate acceptable performance.
An
initial countrate of 0.5 cps has been found acceptable (Regulatory
Guide 1.68), although a rate one or two orders of magnitude higher
does facilitate data collection.
If the Unit 1 source range
detectors could be moved to a location where the flux is a factor of
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four 'to ten lower, then both the initial and at-critical countrate
requirements could be satisfied.
The current core is' designed for -
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low leakage; so it does not seem 'likely that future cores- will
provide a significantly reduced source.
At the. exit interview,Emanagement made a commitment to' evaluate the
feasibility of moving the source range detectors to a region of lower
flux so that criticality would ' occur below P-6 (Inspector Followup
Item 348,364/88-20-01).
b.
Review of Completed Procedures - Unit 1
The following completed startup test procedures were reviewed:
(1) FNP-1-ETP-3601 (Revision 5), Zero Power Physics Test Procedure,
was used to perform boron endpoint measurements at ARO and
reference bank-in configurations, determine the ITC and MTC, and
measure control rod reactivity worth. For the ITC measurement,
both the procedure and.the vendor recommend a temperature change
of at least 4 F.
In practice, the heatup was 2.8 F and the
cooldown was 3.3 F.
For this core the MTC is far from the
limiting value, and the potentially. reduced precision of the
measurement is not of concern. However, the procedure _ guidance
should be followed more closely in the future to assure the MTC
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is better resolved when it approaches the limiting value. The
reactivity computer trace from the reference bank worth meas-
urement was spot checked.
In all cases, the inspector's
determination of the reactivity increment was slightly greater
than that reported by the licensee.
Thus it appears the
licensee's reported control rod worths are conservatively low,
but in satisfactory agreement with design values.
(2) FNP-1-STP-3605
(Revision 6),
Startup and Power Ascension
Procedures.
(3) FNP-1-STP-29.3 (Revision 0), Shutdown Margin Verification with
Control Rods at the Rod Insertion Limit.
(4) FNP-1-STP-114 (Revision 7), Determination of Moderator Tempera-
ture Coefficient at All Rods Out Zero Power and at 70% Power,
was performed on May 19, 1988.
Using the ITC measured in the
zero power tests, the procedure confirmed by appropriate
adjustment of design parameters that the MTC limits of TS 3.1.1.3 were satisfied at ARO and at 70% RTP. This revision of
the procedure reflects a significant improvement over the last
time this area was inspected.
(5) FNP-1-STP-111 (Revision 6), Overall Reactivity Balance.
No violations or deviations were identified.
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3.
Core Power Distribution Monitoring (61702)
. Although the procedures are essentially the same, those for. Unit 2 were
chosen for review; since that unit is currently in its' fif th month of
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the current cycle, and review of data for an ongoing cycle was of more
interest than for a completed cycle,' The documents reviewed were:
a.
FNP-2-STP-110 (Revision 12), Determination of Limiting Hot Channel
Factors, FQ(Z) and FdHN;
b.
FNP-2-STP-121 (Revision 16), Power Range Axial Offset Calibration,
and;
c.
Unit 2 Flux Map Log, which contains copies of the data sheets
generated in performing the above surveillances during the current
cycle.
The logged flux maps, numbers 152 to 158, were all full core maps, and
with the exception of the first, which was at 32% RTP, were obtained at
full power.
All maps were obtained by traversing 49 to 50 instrument
thimbles; the minimum allowable, TS 3.3.3.2,.is 38. The maximum exposure
between maps was 30.8 EFPD, and the typical span was about 22 EFPD. TS
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4.2.2.2.d(1)(b) and 4.2.3.1.b require these surveillances on 31 EFPD
intervals.
The results of the surveillances were:
Maximum FQ(Z) = 1.79 (TS 3.2.2 limit = 2.32)
Maximum FdHN = 1.48 (TS 3.2.3 limit = 1.55)
Incore QPTRs were all less than 1%.
STP-110 does not require that flux maps be obtained during power escala-
tion if the measured Fxy does not exceed the limiting Fxy of the first map
taken, No. 152 at 32%. However, not taking intermediate power maps before
reaching 100% RTP is contrary to practice at other similar facilities.
Prudence would argue for at least one intermediate mar at a power level
between 50 and 90% RTP,
-At the start of the cycle, an incore-excore nuclear instrument correlation
was performed at 32% RTP in accordance with section 10 of STP-121. The
correlation was based upon seven flux maps (146 to 152), of which only
the first and last were full core maps and the remainder were quarter
core maps.
Incore measured axial off sets ranged from +27.8% to -27.1%.
Individual chamber currents ranged from 44 to 60 microamperes.
.STP-121 permits additional incore-excore correlation data to be taken
during power escalation to enhance the fit, but no additional maps were
made.
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The inspector independently analyzed the incore-excore correlation data
using a least squares spreadsheet with the microcomputer program
SUPERCALC3 (Release 2.1).
For the eight neutron chambers involved, the
correlation coefficients ranged from 0.973 to 0.994.
The inspector is
accustomed to finding all coefficients in excess of 0.99 when the data for
the correlation were obtained in the range of 50 to 75% RTP. Attachment 1
is a typical plot of correlation data and results for the pair of chambers
comprising Unit 2 PRNI Nd?,
In the second week of the inspection, incore-excore correlation data were
available from the recent startup of Unit 1 (Procedure FNP-1-STP-121). A
similar analysis of those data yielded correlation coefficients in the
range 0.985 to 0.999. Attachment 2 is a typical plot of correlation data
and results for the pair of chambers comprising Unit 1 PRNI N44. As power
was escalated on Unit 1, additional flux maps were obtained at 50 and 80%
RTP.
Those results were used to adjust rather than augment the earlier
correlations.
The adjustments affected only the zero offset currents and
not the slopgs of the offset versus current relationships.
The meth-
odology used and the bases for those adjustments will be reviewed in later
inspection.
The currents produced by the chambers are only 40 to 50% of those observed
for similar instruments at other facilities.
Hence at the relatively
low powers, about 30% RTP, that the licensee uses for the incore-excore
correlation, chamber currents are only 40 to 60 microamperes.
Thus the
observed currents are resolved to only two significant figures instead of
the three available at other facilities. The low currents appear to have
a direct and negative impact on the quality of the correlations.
The
inspector referred the licensee to another facility that has been success-
ful in increasing the currents produced by the PRN!s.
Unit I limiting hot channel factors were satisfactory at 32 and 48% RTP as
analyzed using FNP-1-STP-110 (Revision 17) during startup operations.
No violations or deviations were identified.
4.
Thermal Power Determination (61706)
The NRC independent measurement program for determination of reactor
thermal power is described in NUREG1167, TPOWR2: Thermal Power Determin-
ation for Westinghouse Reactors, Version 2.
To customize the program for
use at Farley 1 and 2, the necessary system parameters were obtained by
review of the FSAR and vendor documents.
Tne parameters for asulation
losses were adjusted to approximate the licensee's measured losses on
Unit 1. To obtain Unit 2 data for use with the microcomputer program
TPDWR2, the operators adjusted one of the standard plant computer edits
to output the required process data every 15 minutes.
Simultaneously,
the inspector collected feedwater flow data manually from the Barton
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dif ferential pressure meters.
(The Bartons are the source of feedwater-
flow measurements for the . licensee's calorimetric procedure).
The data
obtained, although sufficient for use in TPDWR2, were not in the order
or, in all cases, in the units required for input to that_ program. A
SUPERCALC3 spreadsheet was created to facilitate ordering and conversion
of the data for input to TPDWR2.
The customized plant parameters for
Unit 2 (Unit l's are identical) and a-typical set of input data are given
in Attachment 3.
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TPDWR2 was first run using feedwater flow as measured through the plant
computer and then using flow calcu'uted using the Barton dP cells,
In
general, the steam generator power calculations were lower using the
Barton flows than with the installed flowmeters, but only the differences
for Steam Generator B appeared to be significant, about 1.4% gr eater
averaged over four sets of calculations.
Typical results for Unit 2,
corresponding to the input data in Attachment 3, are given in Attach-
ment 4.
This calculation used the same feedwater flow data as used in the
licensee's routine' surveillance (FNP-2-STP-109), which gave a result of
2654.' megawatts thermal. The difference from the TPDWR2 result of 2655.3
is not considered significant.
The licensed power level is 2652 Mwth.
The indicated 0.1% overpower is a typical and tolernble variation from the
setpoint
Reading the Barton dP meters installed for Unit 2 is not as straight-
forward operation as it should be. The installed meters are scaled for 0
to 450 inches of water column, and that two-inch-increment scale is backed
by a mirror so needle position may be read without parallax. On Steam
Generator A, the scale has been extended to 475 inches of water, and the
calibrated 475-inch point is scribed on the solid face of the meter. The
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actual operating point of the meter is about 460 inches, bu* that point
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must be determined by the reader without the aid of scribed subdivisions
between 450 and 475 inches and without the aid of the mirror to avoid
parallax in the reading. For Steam Generator C, two insulated pipes cover
the face of the meter with less than two feet of clearance.
Reading the
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meter accurately is difficult whether the viewer places his head between
the pipes and the meter or attempts to read the meter through the space
between the pipes.
At the exit interview, the licensee stated that three new Barton meters
with 500 inch scales were on order and that it was their intent to install
them where needed as soon as they arrive.
Changing meters does not
require and outage. Management also took notice of the comments on the
difficulty in reading the meters.
The installed flowmeters also provide input to the plant computer calcu-
lation of thermal power at computer point U1118.
That point does not
provide a reliable estimate of power because no attempt has been made to
keep the installed flowmeters calibrated to the precision required for
measurement.
(The licensee contends the flow meter calibration is
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sufficient _to support the safety funt tion and level control). Management
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agreed that having.an unreliable inuicator on or available for display was
poor practice. At the exit : interview, the licensee made a commitment to
upgrade the U1118 calculation (Inspector Followup Item 348,364/88-20-02).
No violations o- deviations were identified.
5.
Reactor Coolant System Leakage Measurements (61728)
The microcomputer program RCSLK9, which was develcped by the NRC Indepen-
dent Measurements Program, is described in NUREG-1107, RCSLK9: Reactor
Coolant System Leakage Determination _ for PWRs.
Farley uses FNP-1/2-STP-
9.0 for surveillance of RCS leakage.
Data obtained by the licensee for
recent surveillances on both units were analyzed using RCSLK9.
The
comparisons revealed that one constant used by the licensee to correct
for changes 'in pressurizer level was in error and was non-conservtive. As
a result, volumetric changes in pressurizer inventory were equated with
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those in the VCT without correcting for the near factor of two difference
in density. A one percent increase in pressurizer le rel would lead to
underestimating RCS -losses by about 25 s'.andard gallons. For a surveil-
lance period of 30 to 60 minutes the leak rate would be underestimated
by 1 to 0.5gpm.
The maximum allowed unidentified leak rate is igpm.
Although no violation of the leakage limit was identified, the inability
of the procedure to perform the required surveillance with acceptable
precision under permitted and anticipated conditions has been identified
as a violation (Violation 348,364/88-20-03).
The licensee initiated
corrective action to revise the procedure as soon as the problem was-
identified, and on June 14, 1988, the licensee reported the corrected
procedures would be implemented the following day.
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6,
Acronyms and Initialism Used in This Report
AFD - Axial flux difference
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A0
- Axial offset
ARO - all rods out
cps - counts per second
dP
- differential pressure
ETo - Engineering technical procedure
EFPD - Effective full power. day
FdHN - Nuclear enthalpy rise hot che.nnel factor
FQ(Z)- Heat flux bot channel factor
Fxy - Radial (planar) peaking factor
gpm
gallon per minute
ICRR - inverse count rate ratio
IRM - intermediate range monitor
ITC - Isothermal temperature coefficient
MTC - Moderator temperature coefficient
pcm - Percent millirho (unit of reactivity)
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ppmB - Parts per million boron
QPTR - Quadrant power tilt ratio
RTP - Rated thermal power
SRNI - source range nuclear instrument
STP - Surveillance test procedure
TS
- Technical Specification
VCT - volume control tank
7.
Exit Interview
The inspection scope and findings were summarized on May 20, and June 10,
1988 with those persons indicated in paragraph 1 above. The inspector
described the areas inspected and discussed in detail the inspection
findings. Proprietary material was reviewed by the inspector during this
inspection, but is not incorporated in this report.
Dissenting comments
were not received from the licensee.
Attachments
a.
Unit 2, N42, Incore-Excore Correlation
b.
Unit 1, N44, Incore-Excore Correlation
c.
Heat Balance Parameters and Data
d.
Heat Balance Results
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ATTACHMENT 1
Rpt. 348, 364/88-20
Data and Least Squares
Tit
FNP 2, N42
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ATTACHMENT 5 Rpt. 348, 364/88-20
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LEAT IALANCE IATA
FAELEY 2
6/7/S3
PL h i FARANETERS:
FE:CiOR CCOLANT SYSTEM
EEFLECI!VE Ik!JLATICN
Fut; P:.er (N each)
4.5
Ir, side Erf a:e Area (s; it)
34,000
Fas; Efficiency (1)
93.0
Heat L:ss Ccef ficient (BiJs/tr sq f t)
- 2).00
Pressari:er Irside Diateter (tnches)
34.0
!.0\\fEFLECTItE !hSU ATICN
-
STEAP EENE:ATCOS
Ins:de Sarf a:e Area (sq f t)
24,0C0
hse Inside ha: Ster (in:tas)
!!3.5)
15ttiress (:n:hes)
4.0
P.iser Catside Dia eter (ir.thes)
56.75
i?.arsal C ndactivity (STUs/hr it F)
0.14)
N.ater of Risers
3
Maisttre Carrreur (1) in A
0.040
LICEME IFEF*Ai, F:4ER (Nt)
2m
M:itters Carry-:or (1) in 9
0.043
5:isture Ca ry-uer (U in C
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TIME
17X
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STEAM SEhE U T:R A
SiE M SEh!F A13 I
Eteas fressare (;sta)
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Stea Pressa'e (psla)
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Feed.ater flew (E6 It/5r;
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feed.ater Tes;erature (El
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h tt:s !!cwd:n (;;s)
15.2
htt; Pleades, (g;s)
16.0
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Water Level (ir:res)
42.5
Water Le,el (in:res)
47.3
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Eteas fressore (;sia)
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Feefeattr f!:n IR lt/tr)
3.312
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ATTACHMENT 4 Rpt. 348, 364/88-20
HEAT DALANCE
FARLEY 2
6/7/88
DATA SET 1 OF 1
ENTHALPY
FLOW
POWER
POWER
1700 hours0.0197 days <br />0.472 hours <br />0.00281 weeks <br />6.4685e-4 months <br />
(DTUs/lb)
(E6 lb/hr)
(E9 DTils/hr)
(MWt)
GTEAM GENERATOR A
Steam
1199.0
3.966
4.756
418.1
-3.972
-1.661
Surface Dlowdown
510.4
0.00000
0.00000
Bottom Blowdown
463.1
0.00612
0.00283
Power Dissipated
3.0976
907.2
'
Steam
1198.9
3.862
4.630
417.4
-3.869
-1.615
Surface Blowdown
510.8
0.00000
0.00000
Dottom Blowdown
463.0
0.00644
0.00298
Power Dissipated
3.0186
884.1
Steam
1198.9
3.207
4.564
418.4
-3.812
-1.595
Surface Blowdown
511.0
0.00000
0.00000
Bottom Blowdown
463.5
0.00515
0.00239
Power Dissipated
2.9714
870.2
OTHER COMPONENTS
Letdown Line
534.7
0.04956
0.02650
Charging Line
435.7
-0.03939
-0.01716
Pumps
-0.04287
Insulation Losses
0.01226
Power Dissipated
-0.02126
-6.2
--_---
REACTOR POWER
2655.3
.
-
--