ML20149E209
| ML20149E209 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 01/14/1988 |
| From: | Burnett P, Jape F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20149E192 | List: |
| References | |
| 50-369-87-42, 50-370-87-42, NUDOCS 8802100403 | |
| Download: ML20149E209 (13) | |
See also: IR 05000369/1987042
Text
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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REGION ll
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101 MARtETTA STREET, N.W.
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ATLANTA, G EORGI A 30323
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Report Nos.: 50-369/87-42 and 50-370/87-42
Licensee: Duke Power Company
-422 South' Church Street
Charlotte, NC 28242
Docket Nos.: 50-369 and 50-370
License Nos.: NPF-9 and NPF-17
Facility Name: McGuire 1 and 2
Inspection Conducted:
December 15 - 18, 1987
Inspector: h d , E /v
/- / v-gg
[,v- P. T. Burne/t'
Date Signed
Approved by:
M
d
F. Jape, Section Chief
V
r
Date Signed
Engineering Branch
Division of Reactor Safety
SUMMARY
Scope: This routine, announced inspection addressed the areas of therrnal power
monitoring and nuclear instrument calibration and operability.
Results: One violation was identified - failure to make a timely report -
paragraph 6.
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REPORT DETAILS
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1.
Persons Contacted
Licensee Employees
N. Atherton, Compliance
- R. Banner, Compliance
R. Browder, Engineer, Operations
- M.
Case
D. Ethington, Engineer, Ccmpliance
- G. Gilbert, Test Engineer
- B. Hamilton, Technical Services Superintendent
M. Kitlan,* Reactor Engineer
- W. Kronenwetter, General Office, Design Engineerir.g
- L. Kunka, Engineer, Reactor Group
- S. LeRoy, General Office, Licensing
- T. McConnell, Station Manager
- E. McCraw, Compliance Engineer
M. Mustian, Engineer, Instrumentation and Electrical
- M. Nazar, Engineer, Performance
- M. Sample, Superintendent of Integrated Scheduling
- D. Smith, Performance Test Engineer
- J. Snyder, Performance Engineer
W. Suslick, Engineer, Test Group
- R. Travis, Operations Superintendent
Other licensee employees contacted included engineers, operators, and
c.ffice personnel .
NRC Resident Inspectors
- W. Orders, Senior Resident Inspector
- D. Nelson, Resident Inspector
- Attended exit interview
2.
Exit Interview
The inspection scope and findings were summarized on December 18, 1987,
with those persons indicated in paragraph 1 above. The inspector de-
scribed the areas inspected and discussed in detail the inspection find-
ings.
Dissenting coments were not received from the licensee.
Proprietary information is not contained in this report. The inspection
findings included:
a.
Violaticn 369/87-42-01: Failure to make a required report within the
allotted time - paragraph 6.
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b.
Licensee identified violation 369/87-42-02: Thennal power averaged
over eight hours exceeded 100% RTP on severa1' occasions during the
period May 20 to June 11, 1987 - paragraph 6.
3.
Licensee Action on Previous Enforcement Matters
This subject was not addressed in the inspection.
4.
Unresolved Items
No unresolved items were identified during this inspection.
5.
' Calibration of Nuclear Instrument Systems (61705)
When Unit I returned to power on November 15, 1987 following a refueling
outage, the bottom chamber of PRNI N43 was found to be producing about 50%
more current than predicted; while the other seven chambers had outputs
close to predicted currents.
Channel N43 was then placed in trip pending
review of the situation by the licensea and Westinghouse, the manufacturer
of both the chamber and the NSSS.
The reviewers concluded there were no known failure modes of the chamber,
which would lead to an increased current.
The most likely cause of the
high current, then, was the addition of moderator, water, to the instru-
ment thimble surrounding the hermetically sealed canister containing the
neutron-detecting chambers. Since virtually all of the neutrons leaking
from the reactor ate epi-thermal, ar.d the chambers are most sensitive to
thermal neutrons; thermalization of the neutrons would lead to increased
current.
That the upper chamber did not show an increased current indi-
cated that the thimble was only partially filled with water.
The channel was returned to service and calibrated to eliminate any
quadrant power tilt to permit raising power above 50% RTP.
Between 50 and
80% RTP, all PRNI channels were calibrated against the incore flux maps to
obtain consistent axial offset. As operation continued, current from the
lower N43 chanber decreased relative to the other chambers, increasing the
N43 A0 relative to that of the other channels. This was judged to be the
result of water draining or evaporating from the thimble.
'
Technical Specifications require an incore-excore calibration every
.
90 days. Technical Specification Table 4.3-1 requires agreement between
incore and excore A0s of 3%. The licensee established a criterion to
perform the incore-excore calibration whenever the N43 A0 changed by 3%
relative to the other three channels. Three recalibration have been
ree ired since the initial one to satisfy the criterion. The licensee's
v
trending of the current indicates it is decreasing linearly at the rate of
<
0.33%/ day.
Other PRNI parameters being trended include excore tilt from
the lower detectors, excore axial offset, and lower and upper excore
powers.
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With one exception, the contribution of a PRNI channel to control and
safety systems is based on the total channel, sum of both chambers, out
put.
For any channel, adjusting the output to agree with calorimetric
power is easily and routinely done whenever the absolute difference
approaches 2%. The over temperature delta temperature circuit trip
setpoint, however, does include a tenn dependent on A0. A sensitivity
study performed by Westinghouse showed that the lower chamber current
could decrease by 50% without requiring an adjustment in the circuit
constant.
The inspector concluded that channel N43 was operable and that the licens-
ee's trending and recalibration activities were adequate to maintain
operability or to promptly detect failure.
In the course of reviewing the incore-excore recalibration activities, the
inspector learned that there is an incore quadrant power tilt of cpproxi-
mately two percent. That tilt was not reflected by the excore PRNIs when
they were recalibrated. They showed no tilt.
It was not clear that this
action is permitted by the Technical Specification requirement that power
be reduced when a quadrant power tilt measured from the PRNIs exceeds two
percent. This procedure was discussed with members of the Reactor Systems
Branch in NRR. They confirmed that when an incore power measurement shows
the hot channel factors to be within limits in the presence of a tilt the
PRNIs may be recalibrated to show no tilt. The purpose of the tilt limit
determined by the PRNIs is to monitor core changes between incore power
maps.
Unit I tripped on December 28, 1987. During that brief outage between
five and six gallons of water were drained from the N43 thimble, and the
entire instrument canister was replaced. Subsequent operation of N43
appears to be normal.
No violations or deviations were identified.
6.
Statistical Analysis of Power History (61706)
References:
(1) NRC Memorandum, "DISCUSSION OF LICENSED POWER LEVEL," dated
August 22,1980, from E. L. Jordan, Assistant Director for
Technical Programs, Division of Reactor Operations Inspection,
Office of Inspection and Enforcement, to multiple addressees.
(2) McGuire Nuclear Station Memorandum to File, "Statistical Analy-
sis Showing Unit 1 Did Not Violate 100% Rate (sic) Thennal Power
Between May 20 and June 11, 1987, Revision 1."
(3) Standard Mathematical Tables, Tenth Edition, Chemical Rubber
l
Publishing Company, Cleveland, Ohio (1954), pp 239-244.
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(4) Acheson J. Duncan, Quality Control and Industrial Statistics,
Richard D. Irwin, Inc, Homewood, Illinois (1974), pp 580-584 and
709-711.
The criteria for determining significant overpower operation in excess of
the license limit were established in reference 1.
One criterion is that
thermal power averaged over an eight-hour shift exceeds 100% of RTP.
In a
review initiated on June 11, 1987, the licensee identified occurrences of
eight-hour averages exceeding 100% RTP during the period May 20 to that
date. The largest average identified was 100.18%.
The thermal power monitoring program on the OAC performs a new calculation
every minute and updates the power level displayed on the operators screen
with that frequency. However, at the time of this event, the results were
saved to computer storage only once per hour, with the hourly printout of
operating parameters. The stored value was the most recent vne-minute
calculation, not the average over the hour. The licensee's initial
evaluation was based on the assumption that the recorded observation
represented the average power of the preceding hour.
(This is connon
practice in most surveillance-related sampling activities.)
The subsequent corrective action was excellent.
(That much of it had been
planned earlier as a routine operating improvement is also to the licens-
ee's credit.) The programming of the OAC was changed to save the
minute-by-minute thermal power calculations to calculate and display the
running average power for the preceding one and eight-hour periods. The
program modifications also provided computer-generated alarms when any one
of the operational limits discussed in reference 1 is approached.
Discus-
sions with operators in the control room confirmed they had become more
sensitive to the need to monitor and respond to average power.
Prior to issuing a LER, the licensee performed additional analysis of
their power history and determined, to their satisfaction, that they had
not averaged over 100% RTP over the entire period in question. This
conclusion was based on a statistical analysis performed by one of the
operations engineers.
His work is summarized in raference 2.
The essence the analysis was that in an eight-hour period 480 calculations
of power would be perfonned, but on'y eight retained, and those eight did
not give a good estimate of the average of the 480. Therefore, an accu-
rate eight-hour-average power could not be obtained fram eight data, but
could be represented by a larger number of data collected over a longer
period of time. To that end, the engineer reviewed the 576 power-level
data retained over the period of interest. He discarded all observations
less than 98.5% RTP as not being representative of full power operation
and created sequential eight-hour averages of the remaining 546 readings
as well as an overall mean. He then calculated the standard deviation of
the eight-hour averages and a standard deviation of of the eight-hour
mean. His results for the mean of eight-hour averaus and standard
deviation of that mean were 99.91 and 0.0215% RTP, respective?y. Since
the mean plus three standard deviations of that mean did not exceed 100%
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RTP; he concluded that there was 99% confidence that in the average
eight-hour period power did not exceed the limit.
Although accepted by licensee management as valid, the argument is weak
from several aspects.
First, the regulatory issue is not what the average
power was over the period of interest or the confidence that can be placed
in that average.
The issue is whether in any eight-hour period power
averaged above 100% RTP, and that issue is not addressed in the licensee's
analysis.
Second, from a statistical stand point, there is an assumption
that power, except for periods of reduced power operation, was held at a
constant value by the automatic rod control system, and that any varia-
tions observed about that constant value are from random variations of
process variables. However, no analysis of variance or test of the mean
was performed.
The inspector obtained the licensee's power data file for this period on a
micro floppy disk, and converted the file to a format compatible with the
micro computer program SUPERCALC3. He then used that program for inde-
pendent statistical analyses of the licensee's power history data. Review
of the data in sequence shr,wed there were five periods of apparent full
power operation separated by periods of reduced-power operation. The five
periods contained in order 32, 201, 178, 92, and 33 power records. The
lowest recorded power in any set exceeded 98.9% RTP. Assuming the data
were truly random, a mean and variance were calculated for each set of
power data. The means of the sets were compared in pairs using the
Students' T test.
In tnree of the seven pairings considered, the means
were significantly different at the 95% confidence level and two pairings
were different at the 99% confidence. level.
These results challenge the
assumption of constant power over all subperiods of the total period of
interest. The variances were analyzed in pairs using the F test.
In one
of four pairings considered, there was a significant difference in vari-
i
ances. This challenges the assumption of constant variability among the
,
data. The T test and F test and their tables are given in reference 3.
Bartlett's test of homogeneity of variances is discussed in reference 4.
,
It provides a means of testing the variances of the five sets as a group.
The results were that there was less than 1% probability that all sets
have a common variability.
The text of the reference notes that the test
is based upon an assumption of random variation of the data within each
!
group of data.
If this is not true, a finding of significant difference
in variability may reflect a departure from normality rather than hetero-
geneity of variance.
If the data are not normally distributed, then no statistical test is
valid for the analysis of the data, and no valid conclusions may be drawn
if the tests are performed.
Qualitatively, this lack of normality is
demonstrated from calculating the running eight-hour-average power level.
This calculation showed that there were 146 occurrences of power averaging
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over 100% RTP over the inmediately preceding eight-hour period. Many of
these periods were sequential with cluster sizes of 22, 20, 17, 16, 13, 9,
9, 7, and 6.
A cluster size of 22 means that the average of 29 sequential
power records obtained with hourly frequency exceeded 100% RTP,
Finally, a chi-squared test was performed on the 536 data as a whole as a
test of nonnality. The result was there was less than five percent
probability that a normal distribution would have produced as large a
statistic. The test results are summarized in Table 1 below. This test
was also performed for a sequence of 96 one-minute-interval data obtained
form Unit 2 with a similar result. This result does not impugn the
validity of the average of such data being representative of a one or
eight-hour period of reactor operation.
It does suggest that the
minute-to-minute variations in power are systematic rather than rindom.
The inspector concluded there was no statistically valid means of refuting
the straight-forward interpretation of the reactor power data and the
conclusion that power had averaged over 100% RTP for an eight-hour period.
The analyses discussed above were performed by the inspector subsequent to
the on-site portion of the inspection and the exit interview. At the exit
interview, the licensee responded to the inspector's qualitative concerns
about the validity of their statistical analysis by volunteering to issue
an LER. The failure to 'ssue a timely LER has been identified as a
violation of 10 CFR 50.73.a(2) (VIO 369/87-42-01). The extended overpower
operation has been identified as a licensee identified violation (LIV
369/87-42-02).
No requirement for additional corrective action to pre-
clude further overpower operation has been identified.
Table 1: Chi-Squared Test of Normality
Class
Actual
Theoretical
Upper
Frecuency
Frequency (F-f)2
Limit
(F)
(f)
f
..........................................
99.548
30
24.9
1.034
99.663
28
39.9
3.565
99.778
60
71.6
1.882
99.893
101
99.6
.020
100.008
128
107.3
3.990
100.121
98
88.2
1.095
100.236
50
58.7
1.302
100.351
29
29.6
.014
infinity
12
16.1
1.035
...............
3:..........
3'.............
SUMS =
536
536.0
13.938
For 6 degrees of freedom, the probability, P(6,13.9), that
a random sample gives no better fit is 5%. Hence, the
hypothesis of nornality is re.jected.
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7.
Independent Analysis of Thermal Power (61706)
a.
References
(1) NUREG-1167. TPDWR2: Thermal Power Determination for Westinghouse
Reactors, Version 2,
(2) -McGuire Nuclear Station FSAR, Chapter 5,
(3) Westinghouse Technical Manual 1440-C247, Pressurizer Instruc-
tions ... ,
(4) Westinohouse Te:hnical Manual 1440-C250, Vertical Steam Genera-
tor ?.n ruction:
... ,
(5) 0AC Manual:
(a) Thermal Outputs Calculation, Section 3.2.10, and
(b) Thermal Outputs Calculation Dump, Section 3.2.14.
b.
Parameter and Cata Acquisition
The micro computer program TPDWR2, developed by the NRC's Indepen-
dent Measurements Program for analysis of licensee thennal power data
is described in Reference (1).
In order to customize the program for
use at FcGuire, plant specific physical and performance parameters
were obtained from references (2) to (5).
Those parameters are given
on page 1 of attachment 3 along with typical input data for the
calculations described below.
Using the plant computer to log the input data for TPDWR2 assures
better numerical resolution and contemporaneousness of the data than
manual collection from MCB indicators can provide.
All of the
necessary data were obtained using edits frem computer point identi-
fication tables 1, 4, 10, and 11. The points were printed out with
one-minute frequency for over an hour. The inspector selected six
sets of data for analysis in pairs separated by about 30 minutes.
Most of the data were not in a form that could be input directly to
TPDWR2.
Data sources, with the loop A computer points used in the
examples of loop-specific parameters, and the required manipulations
are described below:
!
S/G pressure = pressure (psig, A1107) + atmospheric pressure (P0117)
!
FW flow (Mlb/hr) + average (P1412, P1413)
FW temperature = A0454 (no manipulation necessary)
i
BD ficw (gpm, S/G conditions) = A0652(lb/hr)*0.002514
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- _ _ _ _ _ - _ _ _
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'S/G level (inches) = A1059(%)*2.33-+ 394
LD flow e A0764 (no manipulation necessary)
LD temperature = A1088 (loop C cold leg)
CHG flow (gpm) = A0758 - 32gpm (flow to the seals does not return
enthalpy from the regenerative heat
exchanger)
CHG temperature = A0758 (regenerative heat exchanger outlet
temperature)
PZR pressure (psia) = P1389
PZR level (inches) = A0976(%) * 5.205 + 25.75
NC average temperature = P1461 (no manipulation required)
NC average cold leg temperature = average (A1064, A1076, A1088)
A SUPERCALC3 spreadsheet was used to perform all of the necessary
calculations and to organize the results in an order best suited for
input to TPDWR2.
c.
TPDWR2 Calculational Results
The comparison between TPDWR2 and the licensee's calculations of
power was very good, with TPDWR2 consistently the lower by 0.2 to
0.6% RTP. Typical results for TPDWR2 are given on pages 2 and 3 of
attachment 2.
Some of the differences in results may come from the
calculation of blowdown enthalpy. TPDWR2 uses an average of steam
generator saturation conditionr. and feedwater conditions to calculate
enthalpy for bottom BD flow.
The licensee's value of 75% net efficiency for the reactor coolant
pumps was used in the TPDWR2 calculations, hence that quantity does
not contribute to the differences in results. However, 75% is the
lowest value of pump efficiency the inspector has seen clained by any
licensee, and, if in error, would slightly penalize power generation.
The licensee's calculational method is acceptable as prograrned. The
differences noted, if real, would lead to over estimating thermal
power rather than ur. der estimating it.
No violations or deviacions were identified.
Attachments:
1.
List of Acronyms and Initialisms Used in This Report
2.
TPDWR2 Parameters, Data, and Results
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ATTACHMENT 1
LIST 0F ACRONYMS AND INITIALISMS
- AO - axial offset, the difference in percent'pewer between the upper and
Iower chambers of a PRNI channel
BD - blowdown (water removeii from S/Gs as part of a continuous cleaning
process)
CHG - charging or makeup water
FSAR- Final Safety Analysis Report
LD - ietdown (water removed from the primary system its part of a purification
process)
LER - licensee event report (required by 10 CFR 50.73)
MCB - main control board
NC - nuclear coolant (system)
NSSS- nuclear steam supply system
0AC - operator aid computer ( .he plant computer)
PRNI- power range nuclear instrument
PZR - pressurizer
RTP - rated thermal power
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HEAT BALANCE CATA
Mc6UIRE 2
12-15-87
PLANT PARAMETERS:-
FEACTOR COOLANT SYSTEM
FEFLECilVE INSULAi!0N
Pump Power (MW each) .
5.2
Inside Surface Area (sq ft)
.
15,958
-
>
"
.Puep Efficiency (I)
74.7
Heat Loss Coefficient (BTUs/hr sq f t)
.55.00
Pressurizer Inside Diaseter. finches)
84.0
N0hREFLECTIVE INSULATION
. Inside Suriace Area (sq f t)
!!,575
Dose inside Diaseter (inches)
168.50
Thickness (inches)
- 4.0
-
Riser Outside Diaseter (inches)
21.00
Thereal Conductivity (BTUs/hr it F)
0.035
Number of Risers
12
4
Moisture Carry-over (1) in A
0.100
LICENSED THERMAL F0WER (MWt)
3411-
'
Moisture Carry-over (Il in 9
0.100
Moisture Carry-over (U in C
0.100
Moisture Carry-over (I) in D '
O.100
DATA:
SET I
SET 2
SET 1
SET 2
e
TIME
1742
1811
TIME
1742
1811
'
STEAM SENERATOR A
1
i
Steae Prassure (psial
1000.2
999.6
Steas Fresnre (psia)
996.3
996.3
I
Feedvater Flow (E6 lb/hr)
3.889
3.906
Feedsater Flou (E6 lb/hr)
3.722
3.779
Feedsater Temperature (F)
438.7
438.8
Feedsater Temperature (H
438.0
438.0
1
Surface Blewdown (gps)
0.0
0.0
Suriace Bloudown (gps)
0. 0 '
O.0
Bettca Bloedown (gpal
151.0
150.1
Botten Blondenn (gpal
148.6
142.1
Water Level (inches)
558.0
556.4
Water Level (inches)
548.7
548.7
[
4
STEAM SENERATOR C
STEAh EENERATOR D
Steae Pressure (psia)
1001.5 1001.2
Steas Fressure (psia)
998.6
998.6
'
Feedmater Flcu (E6 lb/hr)
3.911
3.902
Feedsater Flce (E6 lb/hr)
3.708
3.716
)
Fredeater Temperature (F)
438.1
438.3
Feedsater feeperature (F)
439.0
4*8.7
L
Suriace Blcadoun (gpa)
0.0
0.0
Suriace Blendcun (gps)
0.0
0.0
l-
Botton Bhudenn (gps)
162.5
165.0
Bettes 81cedean (g;s)
138.4
144.0
Water Level (inthes)
545.7
545.5
Water Level (inches)
545.9
546.1
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LETDC W LINE
CHAR 81N6 LIXE
e
Fica (gps)
82.8
83.1
Fles (gse)
60.5
60.8
iesperature (F)
559.5
559.4
Teaperature (F)
510.1
510. 'e
i
PEES!JRl!ER
F.EACTOR
Pressure (psia)
?267.0 2266.0
i ave (F)
559.0
587.9
kater Level (inches)
357.3
3 7.8
i cold (F)
559.3
55'.3
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_ _ _ _ _ _ _ - _ _ _ - _ .
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Attachinent'2 '
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HEAT BALANCE
.McGUIRE 2
4
12-15-87-
'
DATA SET 1 OF 2
ENTilALPY.
, FLOW
POWER
POWER
1742 hours0.0202 days <br />0.484 hours <br />0.00288 weeks <br />6.62831e-4 months <br />
(BTUs/lb)
(E6 lb/hr)
(E9 BTUs/hr)
(MWt)
>
Y
Steam
1192.3
3.831
4.567
418.0
-3.889
-1.626
Surface Blowdown
542.6
0.00000
0.00000
Bottom Blowdown
478.0
0.06005
0.02870
_______
Power Dissipated
2.9703
869.9
Steam
1192.4
3.663
4.368
417.2
-3.722
-1.553
Surface Blowdown
.542.0
0.00000
0.00000
Bottom Blowdown
477.3
0.05913
0.02822
,
_______
Power Dissipated
2.8429
832.6
Steam
1192.2
3.847
4.586
Feudwater
417.4
-3.911
-1.632
Surface Blowdown
542.8
0.00000
0.00000
Bottom Blowdown
477.7
0.06464
0.03088
_______
Power Dissipated
2.9847
874.1
,
Steam
1192.3
3.653
4.355
418.3
-3.708
-1.551
.
!
Surface Blowdown
542.3
0.00000
0.00000
Bottom Blowdown
478.1
0.05504
0.02631
___
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Power Dissipated
2.8303
828.9
OTHER COMPONENTS
Letdowa Line
559.0
0.03072
0.,,717
Charging Line
499.2
-0.02389
-0.01193
<
Pressurizer
637.8
0.00010
0.00012
Pumps
-0.05296
,
Insulation Losses
0.00147
_______
Power Dissipated
-0.04613
-13.5
__=_ _
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REACTOR POWER
3392.1
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Attachment 2'
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HEAT BALANCE-
McGUIRE 2
1:
12-15-87
,
DATA SET 2 OF 2
ENTHALPY
FLOW.
POWER
POWER
1811 hours0.021 days <br />0.503 hours <br />0.00299 weeks <br />6.890855e-4 months <br />
(BTUs/lb)
(E6 lb/hr)
(E9 BTus/hr)-
(MWt ) -
Steam
1192.3
3.848
4.588
418.1
-3.906
.t.633
Surfa'ee Blowdown
542.5-
0.00000
0.00000
Bottom Blowdown
478.0
0.05969
0.0285i
. . _ _ _ _ _ . .
Power Dissipated
2.9833
873.7.
i
. STEAM GENERATOR B
l
'
Steam
1192.4
3.722
4.439
i
417.2
-3.779
-1.577
I
Surface Blowdown
542.0
0.00000
0.00000
Bottom Blowdown
477.3
0.05655
0.02699
>
Power Dissip0ted
2.8890
846.1-
Steam
1192.2
3.837
4.574
417.6
-3.902
-1.629
,
Surface Blowdown
542.7
0.00000
0.00000
Doktom Slowdown
477.8
0.06563
0.03136
..______
Power Diusipated
2.9762
871.6
Steam
1192.3
3.658
4.362
41L 0
-3.716
-1.553
Surface Blewdown
542.3
0.00000
0.,00000
Bottom Blowdown
477.9
0.05727
0.02737
_______
Pows. Dissipated
2.8361
830.6
OTHER COMPONENTS
Letdown Line
558.9
0.03083
0.01723
Charging Line
499.3
-0.02401
-0.01199
Pressurizer
637.5
0.00018
0.00012
Pumps
-0.05296
Inse.lation Losses
0.00147
_______
Power Dissipated
-0.04613
-13.5
______
REACTOR FDWER
3408.6
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