ML20245D419
ML20245D419 | |
Person / Time | |
---|---|
Site: | Sequoyah ![]() |
Issue date: | 06/14/1989 |
From: | Belisle G, Burnett P, John Zeiler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
To: | |
Shared Package | |
ML20245D384 | List: |
References | |
50-327-89-16, 50-328-89-16, NUDOCS 8906270105 | |
Download: ML20245D419 (19) | |
See also: IR 05000327/1989016
Text
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UNITED STATES '
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NUCLEAR REGULATORY COMMISslON
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REGION Il
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101 M ARIETTA STREET, N.W.
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ATLANTA, GEORGI A 30323
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Report Nos.: 50-327/89-16, 50-328/89-16
. Licensee: Tennessee Valley Authority
6N 38A Lookout Place
,
1101 Market Square
Chattanooga, TN 37402-2801
Docket Nos.: 50-3?7 and 50-328
License Nos.: DPR-77 and DPR-79
Facility Name: Sequoyah Units 1 and 2
Inspection Conducted: May 8 - 12, 1989
Inspector:ha.
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Date Signed
(-/V-87
p P. T. Burne'tt
Inspector:
6.T,7[r
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/" J. Zeiler /
Date Signed
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Approved by:
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fv G. A. B61isic, Chief
Date Signed
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Test Programs Section
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Engineering Branch
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Division of Reactor Safety
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SUMMARY
Scope:
This routine, announced inspection addressed the areas of review of
completed Unit 2, cycle 4 initial criticality and post-refueling
startup tests; evaluation of thermal power measurements for Unit 1;
evaluation of reactor coolant system leakage measurements for both
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units; and followup of previously identified items.
Results: Unit 2, cycle 4 precritical, initial criticality, and los power
physics testing was conducted in accordance with approved procedures.
Test results met acceptance criterie and documentation associated -
with the tests was complete and well-maintained. The new location of
the SRMs and IRMs' relative to the core were found to be acceptable,
provided IRM trip setpoints are appropriately adjusted (paragraph 2).
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The licensee's procedures and methods for calculating and monitoring
thermal power were found to be acceptable. Monitoring core power by
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use of computer point U1118 was found to be more reliable than
monitoring nuclear instruments to assure conformance to the license
limit (paragraph 3).
A weakness was identified in the licensee's reactor coolant system
leakrate testing which resulted in a violation for accepting negative
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unidentified leakage results (paragraph 4).
8906270105 890620
ADOCK 05000327
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REPORT DETAILS
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1.
Persons Contacted
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Licensee Employees
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R. Beecken, Maintenance Manager
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- W. Condon, Reactor Engineer
- M. Cooper, Compliance Licensing Supervisor
- T. Flippo, Quality Assurance Manager
- R. Fortenberry, Technical Support Manager
- G. Gault, Reactor Engineering Supervisor
- G. Hipp, Licensing Engineer
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- W. Lagergren, Operations Manager
- J. Patrick, Operations Superintendent
- S. Smith, Plant Manager
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Other licensee employees contacted included engineers, technicians,
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operators, and office personnel.
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NRC Resident Inspector
- P. E. Harmon, Senior Resident Inspector
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- Attended exit interview May 12, 1989
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Acronyms and initialisms used throughout this report are listed in the
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last paragraph.
2.
Unit 2, Cycle 4 Post-Refueling Tests (72700, 61705, 61708, 61710)
a.
Introduction
Unit 2 had reached 80% RTP by the start of this inspection, which was
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necessarily limited to review of test documents and discussion of
the documented results with test personnel.
Prior to the review of
the individual test procedures, the two docements listed below were
reviewed to cet an overview of the management of the test program and
of the expr: -ad test results:
(1) Test Logbook for RTI-1 and AI-47 and
(2) Nuclear Parameters and Operations Package for Sequoyah Unit 2,
cycle 4, which was prepared by Westinghouse, the fuel and NSSS
vendor.
Review of the NUPOP disclosed that the fuel loading for cycle 4 is
not optimized for low leakage, hence relatively large fluxes should
be incident upon the excore neutron detectors.
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On pages 2-7 of the NUP0P, shutdovin margin is described as based upon
compensating for the HFP to HZP reactivity defects, which in turn
means that cooldown below 547 F is not considered in the analytical
determination that the shutdown margin is adequate throughout core
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life.
Excessive post-trip cooldown of Sequoyah reactors and the
concomitant reduction in shutdown margin was discussed in Inspection
Report No. 50-327,328/88-35.
Clearly, there is no cooldown margin
provided in the cycle 4 design, and shutdown margin must continue to
be maintained by limiting the cooldown and providing prompt
compensatory action when the average RCS temperature does fall below
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547 F.
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Section 8.4 of the NUPOP addresses at-power measurement of MTC at E0C
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using a reactivity computer. That method of measurement was found to
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be unacceptable during Inspection Report No. 50-307,328/88-16 and was the
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subject of UNR 50-327,328/88-16-02. Adequate corrective action using
a different measurement technique was reported in Inspection Report
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No. 50-327,328/88-53, and the item was closed.
When questioned, the
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licensee stated it was their intention to continue using the accepted
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method for measuring MTC at E0C, and that section 8.4 was an arti-
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fact, which would be removed form future issues of the NUPOP.
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b.
Precritical Tests
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SI-43 (Revision 12), Rod Drop Time Measurement, was performed on
April 8,1989.
The inspectors independently analyzed the Visicorder
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traces and confirmed that the drop times were as recorded by the
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licensee.
Measured drop times to dash pot entry ranged from 1.17 to
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1.33 seconds.
Hence all rods satisfied the TS 3.1.3.4 acceptance
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criterion that drop time to dash pot entry be less than 2.2 seconds,
c.
Initial Criticality for Cycle 4
Initial criticality of Unit 2, cycle 4 was achieved in a generally
well-controlled and conservative manner under the guidance of RTI-3
(Revision 3), Initial Criticality.
Particular care was taken to
assure true OPERABILITY of the SRMs by frequent and successful use of
the chi-squared test.
The inspector independently verified and
confirmed the resulte, of that definitive statistical test of accept-
able SRM performance.
Procedure RTI-3 contained a good statement on
the application of the test as guidance for test personnel.
During the data collection for the ICRR measurements, only single
observations were recorded for each SRM at each statepoint and the
counting interval was too small to assure at least 1000 counts per
observation.
The better statistical precision available from either
multiple observations and/or longer counting intervals should reduce
the scatter observed in the ICRR plots for this startup.
During
dilution, ICRR was plotted against time, gallons of dilution water,
and boron concentration.
Dilution rate was reduced as ICRR dropped
in magnitude.
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The initial critical conditions for cycle 4 were:
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D bank at 196/197 steps withdrawn with all other rods out,
RCS T-average at 549.4 F (minimum temperature for criticality
at 1670 ppmB, and IRM currepps (N35/N36) at
is540F),R_gyC n
1.3/1.1 x 10
a@s (minimum indication is 10
amps).
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Following criticality, the reactivity computer, an analog device, was
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calibrated for use in the low power physics tests in accordance with
TI-25 (Revision 16), Setup and Operation of the Reactivity Computer.
Appendix G, Installed Reactivity Dynamic Response Check, compared
computer solutions of reactivity for various positive and negative
periods with those obtained from the inhour equation.
In all cases,
the comparisons satisfied the performance acceptance criterion of
4%.
d.
Boron Endpoints and Isothermal Temperature Coefficient Measurements
RTI-4 (Revision 4), Boron Endpoint Determination and Isothermal
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Temperature Coefficient Measurement, was performed on April 13, 1989.
The measured AR0 C 1688 ppmB was in good agreement with prediction
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in the NUPOP of 16M ppmB.
The ITC measurements at AR0 differed during heatup and cooldown by
nearly 1 pcm/ F.
It was noted that RTI-4 did not have an acceptance
criterion for agreement between the two measurements.
At most
Westinghouse plants the acceptance criterion is a 1pcm/ F span of
acceptable measurements as well as agreement of 3 pcm/ F between the
average test result and the predicted value.
At the exit interview,
management made a commitment to add- the acceptance criterion to the
procedure on agreement among measurements used in determining the
average ITC of 1 pcm/ F span
(Inspector Followup Item
50-327,328/89-16-01).
When corrected to 547 F, the average ITC at
ARO was -3.02 pcm/ F and the corresponding MTC was - 1.22 pcm/ F,
which was in acceptable agreement with the NUP0P MTC value of -0.1
pcm/ F.
e.
Control Rod Worth Measurements
RTI-5 (Revision 3), Rod Bank Worth Using Dilution /Boration Method,
was performed on control bank D on April 13, 1989.
The inspector's
independent evaluation of the reactivity computer traces obtained
during the measurement produced a differential worth curve in good
agreement with that obtained by the licensee, as shown in the plot in
Attachment 1 to this report.
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Control bank D was then used as the reference bank for measurements
of worth of the other control and shutdown banks, under the control
of RTI-7 (Revision 4), Rod Worth Measurement Using Rod Swap.
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Shutdown bank A had a measured worth 2.5% greater thac predicted in
the N9 POP.
All other banks were less than predicted by 2.8 to 5.9%.
The sum of the bank worths of 5086 pcm was 3.34% less than the
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predicted value of 5265 pcm.
RTI-1 set the minimum acceptable sum of
the measured values at 4894 pcm, which was consistent with the assumed
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7% difference between measured and predicted values used in the
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shutdown margin analysis in the NUPOP.
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f.
SRM/IRM Performance during Startup and Low Power Testing
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At the start of procedure RTI-3, the two SRMs had countratss of 15 to
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18 cps, well above the acceptable minimum.
The SRMs staved on scale
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until criticality was achieved, and the IRMs displayed t
.r a' decade
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of overlap before the SRMs were taken out of service by procedure.
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At about 80% RTP, the licensee discovered that the 25% RTP trip on
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the IRMs had not actuated, although the trip had been appropriately
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bypassed once the PRNIs came on scale.
Investigation by the licensee
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revealed that two instrument carts, each containing an SRM and IRM
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detector, had been moved during maintenance and trouble shooting and
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had been left about 21 inches laterally displaced from the reactor
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vessel.
Consequently, the detectors were exposed to a significantly
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lower flux than had been assumed in setting the flux level trip
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setpoints for the two IRMs.
At full power, the Unit 2 IRM currents
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were approximately one-fourth those of Unit 1.
Based upon the actual
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performance of the detectors during the initial startup for this
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cycle, the inspector concluded that both the SRMs and IRMs are
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capable of operating in the current reduced flux environment as well
as the one that might exist with lower leakage cores in future
cycles.
It is axiomatic that the reduced flux at the detectors must
be considered in setting trip setpoints to correspond to specific
power levels.
The failure to adjust the trip setpoints in the
current instance and the control of the process by which the instru-
ment carts were moved are being addressed by the Resident
Inspectors in NRC Inspection Report 50-327,328/89-15.
No violations or deviations were identified in the review of initial
criticality and low power testing.
3.
Thermal Power Measurement (61706)
The microcomputer program TPDWR2, which is described in NUREG-1167,
TPDWR2: Thermal Power Determination for Westinghouse Reactors, Version 2,
was used to evaluate plant data to make an independent assessment of the
licensee's adherence to the rated thermal power limit.
This program was
written as part of the NRC Independent Measurements Program.
To customize
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a version for use at Sequoyah, data on steam generator design features,
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and pressurizer and steam generator volumes were obtained from the FSAR,
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and plant drawings.
The resultant heat balance data list is given in
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Attachment 2.
Since the units are identical, only the list for Unit 1 is
provided.
Considerable manipulation of the raw data obtained from the licensee was
required before they could be input to TPDWR2.
All pressures had to be
changed from gauge to absolute; steam generator and pressurizer levels had
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to be converted from narrow range level in percent to relative level in
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inches; letdown and charging flows had to be converted from volts to gpm;
RCS narrow range average temperature had to be averaged to a single value;
and RCS cold-leg temperatures had to be calculated from RCS hot-leg and -
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average temperatures.
The inspectors also adjusted the total heat losses
through component thermal insulation term used in TRPDWRZ in order to
match the equivalent value determined in the licensee's calculation of
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thermal power.
All of the manipulations of data were performed using the
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SUPERCALC3 spread;heet program.
The spread sheets for the final data for
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Unit 1 is given in Attachment 3.
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Using Unit 1 online data from the licensee's plant computer and manipulat-
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ing the data using the inspector's spreadsheet program, TPDWR2 was used to
calcuiate thermal power at ten different time periods between 1002 and
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1442 for May 10, 1989.
The following table shows the results between the
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licensee's and inspector's calculation of the thermal power for these time
periods.
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SEQUOYAH 1: HEAT BALANCE SUMMARY (5-10-89)
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A
B
TIME LICENSEE TPDWR2 A-B
MEAN Percent
(Mwth) (Mwth) (Mwth)
(Mwth) Error
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1002
3379.9 3373.5
6.4
3376.7
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1022
3384.0 3363.7 20.3
3373.8
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1042
3385.8 5357.7 28.1
3371.7
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1102
3379.8 3378.1
1.7
3378.9
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1142
3397.0 3383.8 13.2
3390.4
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1222
3386.7 3374.8 11.9
3380.7
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1302
3383.6 3353.8 29.8
3368.7
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1322
3370.7 3356.1
14.6
3363.4
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1402
3385.5 3361.7 23.8
3373.6
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1442
3387.0 3380.8
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3383.9
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Averages
15.6
3376.2
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As shown in the table, the results from TPDWR2 were consistently lower
than those calculated by the licensee's online computer calculation
displayed at computer point U1118.
The differences ranged from 1.7 to
28.1 megawatts thermal .
However, in perspective, the largest difference
was still within one percent error.
A review of the outputs did not
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reveal any obvious source of 'the disagreements. The results of a typical
set of calculations by TPDWR2 are given in Attachment 4.
The inspectors also independently verified the licensee's Unit 2 precision
heat balance calculation. using raw data ' taken by the licensee on
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May 11,1989, at five minute' intervals between time 1100 to 1200. . Thei
spread sheets for the final data used for Unit 2 thermal calculation is
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given in Attachment.3.
After first averaging the value of each variable
used in the calculation, the licensee calculated the= average thermal power
over this time period to be 3344.5 me'gawatts thermal.
Using TPDWR2 and
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unaveraged (snapshot) values of the. variables, the inspectors calculated-
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an. average thermal power 'using five sets of data in this time period..
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This average resulted in 3331.4 megawatts thermal which compared favorably
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with the licensee's calculation. ' Since blowdown flow was isolated during
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the precision heat balance, .it was not a factor in. either calculation.
Differences in the methods of calculating blecdown enthalpy had been
thought to be part of. the cause of differences between the TPDWR2 and
U1118 r'.sults for Unit 1.
In any case, the licensee's calculations of
thermal power were consistently more conservative (higher) than those-
produced by TPDWR2.
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It was concluded that the licensee's methods of calculating thermal power
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to assure conformance to the license limit are acceptable.
Furthermore,
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it is preferable for the operators to monitor' U1118 rather than NI power
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to assure conformance to the license limit of 3411 megawatts thermal.
Experience at other facilities has shown that xenon driven changes in core
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power distribution can cause the NI calibrations to become non-conserva-
tive in a matter of hours after being calibrated against thermal power.
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No violations or deviations were identified in the licensee's surveillance
C thermal power.
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4.
R n Leakrate Measurement (61728)
The microcomputer program RCSLK9 for measurement of reactor _ coolant system
leakage is ' described in NUREG-1107.
This program was written as part of
the NRC Independent Measurements Program.
To customize a_ version for use
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at Sequoyah, data on system and tank volumes were obtained from the'FSAR,
the plant curve book, and plant drawings.
The resultant parameter list _is
given in Attachment 5.
Since tFe units are _ identical, only the ' list for
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Unit 2 is provided.
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RCS leak' age calculations were performed on both units by. the licensee on
May 8,
1989, using SI-137.2, Reactor ' Coolant System Inventory.
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calculations satisfy Technical Specification surveillance requirement
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The reported total leakage for Hqit 2 was 0.51 gpm anc the
total identified leakage was .0.69 gpm. This resulted in a total unidenti-
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fied leakage of . negative 0.18 gpm, which was accepted by the licensee
without further review or documented justification regarding the negative
result.
Negative unidentified leakage is a physical impossibility. When
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responsible licensee personnel were' questioned concerning this matter, .
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they stated that negative unidentified leakage'results indicated that the
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unidentified leakage is relatively small and was, therefore, considered to
be zero.
The inspectors concluded that negative unidentified. leakage
results could be indicative of problems -such as ; improperly calibrated
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level instrumentation or inleakage of fluid from non-reactor coolant
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system sources to the PRT or'the RCDT.
The inspectors expressed concern
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for the potential for masking of the true unidentified: leakage should '
- problems such as-these occ'ur.
The licensee could unknowingly exceed the.
1.0 gpm technical specification limit in these' circumstances.
To verify the Unit 2 leakage results, the .oe . data recorded by the
licensee were analyzed using RCSLK9.
The results of the. RCSLK9 calcula--
tion (Attachment 6) were in good agreement'with that calculated by' the
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licensee.
The total leakage was calculated as 0.47 gpm' and the unidenti-
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fied leakage was negative 0.33 gpm. The inspectors reviewed all completed '
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tests of SI-137.2 for Unit 2 from April 1 through May 8,1989, to. deter-
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mine if there were other instances where negative leakages were accepted
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by the licensee without justifications included. Within this period, four--
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other examples were noted on dates April 14 and 27, and May I and 2,1989.
However, the majority of these-involved small negative values, and in no
case did the values exceed the negative 0.-18 gpm found on May 8, 1989.-
The licensee's acceptance of negative . unidentified leakage results is
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identified as violation 50-327,328/89-16-02.
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5.
Action on Previous Inspection Findings (92701)
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(Closed)
Violation
50-327,328/88-53-01,
Failure' to
Follow
Administrative Procedures During Low Power Physics Testing
NRC Inspection Report No. 50-327,328/88-53 identified four examples
during Unit 1 initial criticality and. low power physics testing in
which the licensee failed to follow precaution steps -in procedures
AI-47, Conduct of -Testing, RTI-3, Initial Criticality, .and RTI-4,
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Boron Endpoint Determination and Isothermal Temperature Coefficient
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Measurement.
During that inspection, it was determined that no
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discernible adverse effects on-the- test results occurred as a result
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of the infractions and adequate corrective action was taken or
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planned by the licensee.
The violation met the NRC enforcement
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policy criteria for discretionary enforcement and was not cited.
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During this inspection, 'he inspectors reviewed licensee documents
which described the corr. ;tive action taken.
The licensee counseled
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personnel involved on the importance of following procedures and
proper procedural documentation.
Also, the inspectors reviewed
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licensee memorandum dated April 6,1989, addressed to each test
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member involved, emphasizing requirements and management expectations
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of following procedures.
In addition, review of completed Unit 2,
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cycle 4 initial criticality and low power physics test results.during
this inspection did not identify further instances where test
personnel failed to follow test procedures.
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b.
(Closed) IFI 327,328/88-53-02, Followup on Issues Identified During
Initial Criticality
During NRC inspection of Unit 1, cycle 4 initial criticality, two
issues were identified for inspector followup. One issue involved a
review of the licensee's investigation of the failure to meet the 50
ppm acceptance criterion on the agreement between measured and
predicted AR0 boron endpoint measurement.
The second issue involved
a procedural inadequacy for the absence of a plot of ICRR versus
measured boron concentration in procedure RTI-3, Initial Criticality.
NRC review and resolution of the critical boron concentration issue
was performed in NRC Inspection Report 50-327, 328/88-60.
One
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violation and two unresolved items resulted from this
eview and is
documented in that report.
During this inspection, the inspectors verified that adequate
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licensee investigation and analysis was conducted to resolve the
critical boron issue and this part of the IFI is considered closed.
The inspectors also reviewed procedure RTI-3, Revision 6, dated
February 15, 1989, and verified the incorporation of a plot of ICRR
versus measured boron concentration.
6.
Exit Interview (30703)
The inspection scope and findings were summarized on May 12, 1989 with
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those persons indicated in paragrrph I above.
The inspectors described
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the areas inspected and discussed in detail the inspection findings listed
below.
Dissenting comments were not received from the licensee. Proprie-
tary information was reviewed in the course of the inspection, but is not
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contained in this report.
Item Number
Description and Reference
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50-327,328/89-16-01
Inspector followup item - commitment to add
the acceptance criterion of I pcm/ F span
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for agreement among
measurements used in
determining the average ITC , paragraph 2.d.
50-327, 328/89-16-02
Violation - Inadequate surveillance
procedure to determine RCS leakage, para-
graph 4.
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7.
Acronyms and Initialisms Used in This Report
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administrative instruction
AI
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all rods out
ARO
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CAQR -
condition adverse to quality report
counts per second
cps
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E0C -
end of cycle
FSAR -
Final Safety Analysis Report
gpm -
gallon per minute
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HFP -
hot full power
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HZP -
hot zero power
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ICRR -
inverse count rate ratio
inspector followup item
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ITC -
isothermal temperature coefficient
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moderator temperature coefficient
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MTC
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Mwth -
megawatts thermal
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nuclear instruments
NI
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NSSS -
nuclear steam supply system
NUPOP-
Nuclear Plant Operations Package
percent millirho, a unit of reactivity
pcm
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ppmB -
parts per million boron
PRNI -
power range nuclear instruments
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pressurizer relief tank
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RCDT -
reactor coolant-drain tank
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RCS -
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refueling test instruction
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rated thermal power
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surveillance instruction
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source range monitor
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technical instruction
TI
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technical specification
TS
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Attachments:
1.
Sequoyah 2, Cycle 4, Control Bank D
2.
Heat Balance Data
3.
Sequoyah Unit 1 Power Data
4.
Heat Balance
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5.
RCS Parameter List
6.
'!CS Leak Rates
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Attachment 1
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Differential Worth (pem/ step)
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Attachment 2, page 1 of 2
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HEAT BALANCE DATA
SEQUOYAH 1
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05-09-89
PLANT PARAMETERS:
REFLECTIVE INSULATION
Pusp Power (MW eath)
4.0
InsideSurfaceArea(sqft)
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15,958
Pump Efficiency (1)
90.0
HeatLossLoefficient(BTus/hrsqft). 660.00
PressuriterinsideDiaseter(inthes)
83.6
NONREFLECTIVE INSULATION-
,
STEAM 6ENEHATDRS
InsideSurfaceArea(sgit)
11,575-
Dose Inside Diameter (inches)
168.50
Thickness (inches)
4.0
RiserOutsideDiaseter(inches)
56.50
TherealCondectivity(BTUs/hrftF)
0.420
Number of Risers
3
MoistureCarry-over(%)inA
0.250
LICENSEDTHERMALPCsER(MWt)
3411
MoistureCarry-over(1)inB
0.250
Moisture Carry-over (1) in C
0.250
Moisture Carry-over (1) in D
0.250
DATA:
SET 1
SET 2
SET 1
SET 2
. TIME
M02
1022
TIME
1002
1022
STEAM SENERATOR A
STEAM SENERATOR B
StrasPressure(psia)
876.6
874.9
SteasPressure(psia)
880.2
882.9
FeedwaterFlow(E6lb/hr.)
3.690
3.673
FeedwaterFlow(E6lb/hr)
3.669
3.641
FeedwaterTraperature(F)
433.5
433.2
Feedwater Temperature (F)
432.8
432.6
Surface Blendown (gps)
0.0
0.0
SurfaceBlowdown(gps)
0.0
0.0
BottosBlowdown(gps)
34.5
34.5
BottonBlowdown(gps)
58.0
50.0
WaterLevel(inches)
61.8
63.6
Water Level (inches)
65.7
65.8
STEAM SENERATOR C
STEAM BENERATOR D
Steas Pressure (psia)
875.6
878.3
SteasPressure(psia)
876.4
878.7
Feedwater Flow (E6 lb/hr)
3.786
3.776
FeedwaterFlow(E6lb/hr)
3.681
3.690
FeedwaterTemperature(F)
433.5
433.4
FeedwaterTempetiture(F)
433.1
432.9
SurfaceBlowdown(gps)
0.0
0.0
Surface Blowdown (gps)
0.0
0.0
BottonBlowdown(gps)
4B.0
40.0
Botton Blowdown (gps)
54.5
54.5
WaterLevel(inches)
63.9
62.6
WaterLevel(inches)
65.1-
65.5-
LETDOWNLINE
CHAR 61NGLINE
l
Flow (gps)
89.4
90.8
Flow (gp)
88.5
87.7
leaperature(F)
547.4
547.4
Temperature LF)
475.2
475.5
PRESSURl!ER
REACTOR
Pressure (psia)
2246.6 2245.2
I ave (F)
577.3
577.3
WaterLevel(inches)
304.4
304.9
i cold (F)
547.4
547.4
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Attachment 3, page 1 of 2
SEQUOYAH UNIT 1 POWER DATA
================================================================
STEAM GENERATOR A(1)
STEAM GENERATOR B(2)
FW-T Bottom
FW-T Bottom
Time Press. FW-F1
( F) BD(gpm) Level
Press. FW-F1
( F) BD(gpm) Level
(psia) (Mbh) T0418A U0512 (in.)
(psia) (Mbh) T0438A UO532 (in.)
-_____.._....__.._____ _____.._________._.__..._. ..._ ._...__ __... .......
1002 876.6 3.690 433.5
34.5 61.8
880.2 3.669 432.8
58.0 65.7
1022 874.9 3.673 433.2
34.5 63.6
882.9 3.641 432.6
58.0 65.8
1042 877.1 3.691 433.5
34.5 60.2
879.7 3.642 432.8
58.0 66.1
1102 875.6 3.706 433.7
34.5 63.2
879.2 3.658 432.9
58.0 65.7
1122 877.3 3.673 433.3
34.5 64.1
884.1 3.671 432.6
58.0 63.2
1142 871.9 3.711 433.7
34.5 62.6
878.0 3.661. 432.8
58.0 65 7
1202 879.2 3.685 433.1
34.5 62.6
884.4 3.658 432.3
58.0 64.2
1222 875.9 3.693 433.3
34.5 63.2
881.5 3.691 432.7
58.0 64.9
1242 876.1 3.663 433.4
34.5 60.9
880.7 3.646 432.7
58.0 66.5
1302 876.8 3.666 433.3
34.5 62.1
881.9 3.658 432.7
58.0 65.7
1322 878.0 3.671 433.1
34.5 64.9
883.6 3.639 432.3
58.0 65.5
1342 871.7 3.688 433.9
34.5 62.9
877.1 3.680 433.1
58.0 65.4
1402 877.1 3.668 433.4
34.5 62,5
881.2 3.642 432.8
58.0 65.1
1422 875.6 3.721 433.4
34.5 62.4
881.2 3.685 432.7
58.0 66.2
1442 875.8 3.706 433.5
34.5 62.4
880.5 3.683 432.8
58.0 66.0
1452 874.9 3.683 433.3
34.5 61.1
881.2 3.641 432.6
58.0 65.1
= = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = =
STEAM GENERATOR C(3)
STEAM GENERATOR D(4)
FW-T Bottom
FW_T Bottom
Press. FW.F1
( F) BD(gpm) Level
Press. FW-F1
( F) BD(gpm) Level
(psia) (Mbh) T0458A U0552 (in.)
(psia) (Mbh) T0478A UO572 (in.)
-_______..............___.___.....__...__.._ ____.......___..___.__... .
875.6 3.786 433.5
48.0 63.9
876.4 3.681 433.1
54.5 65.1
878.3 3.776 433.4
48.0 62.6
878.7 3.690 432.9
54.5 65.5
876.1 3.756 433.5
48.0 62.5
876.1 3.666 433.1
54.5 65.8
875.1 3.771 433.7
48.0 62.2
876.1 3.713 433.3
54.5 64.9
880.0 3.787 433.3
48.0 62.1
880.2 3.700 432.8
54.5
64.1-
873.7 3.815 433.5
48.0 62.1
874.4 3.683 433.2
54.5 62.8
880.5 3.779 433.2
48.0 60.6
880.5 3.715 432.7
54.5 63.8
877.8 3.716 433.4
48.0 63.4
878.2 3.728 433.1
54.5 63.2
876.3 3.766 433.4
48.0 62.1
877.1 3.691 433.1
54.5 64.9
877.3 3.764 423.4
48.0 61.5
878.0 3.659 433.1
54.5 65.2
880.0 3.766 433.1
48.0 64.5
880.0 3.676 432.7
54.5 64.9
873.2 3.782 433.9
48.0 64.5
873.6 3.705 433.5
54.5 63.8
876.8 3.766 433.5
48.0 62.2
877.5 3.696 433.1
54.5 63.2
876.8 3.787 433.5
48.0 62.9
878.0 3.678 433.1
54.5 64.2
876.1 3.771 433.5
48.0 62.8
876.8 3.696 433.2
54.5 63.6
877.3 3.764 433.3
48.0 62.6
877.7 3.700 432.9
54.5 63.8
1
l
- - - -
_ _ _ _ _ - _ _ _ _____ _ _ __
,
,
,
.
'
.
,
.
Attcchment 3, page 2 of 2
====================================================
LETDOWN LINE CHARGING LINE PRESSURIZER
REACTOR
Power
Flow Temp.
Flow Temp (F)
Press.
Level T-avg T-cold (Mwt)
(gpm)
(F)
(gpm) T0126A
(psia)
(in.)
(F)
(F)
U1118
89.4 547.5
88.5 475.2
2246.6 304.4 577.3 547.4 3379.9
90.8 547.5
87.7 475.5
2245.2 304.9 577.3 547.4 3384.0
86.6 547.4
88.3 475.8
2247.1 304.9 577.3 547.4 3385.8
88.7 547.3
89.1 475.8
2243.2 303.9 577.3 547.3 3379.5
86.2 547.6
88.7 476.4
2244.7 304.4 577.4 547.6 3383.4
85.4 547.0
90.0 474.6
2241.3 300.3 576.9 546.9 3397.0
90.3 547.8
88.7 474.6
2247.1 304.9 577.5 547.7 3377.7
88.6 547.4
89.2 474.6
2243.2 303.9 577.3 547.4 3386.7
85.6 547.5
89.0 474.6
2248.1
304.4 577.3 547.4 3397.9
89.0 547.5
89.8 474.9
2245.7 304.9 577.3 547.5 3383.6
85.9 547.8
88.7 475.5
2246.2 306.5 577.5 547.7 3370.7
89.8 546.9
89.0 475.2
2241.7 302.3 577.0 547.0 3395.8
,
89.4 547.5
88.5 475.2
2246.2 303.9 577.3 547.5 3383.5
87.5 547.5
89.1 475.2
2246.6 303.9 577.3 547.4 3385.5
90.3 547.3
89.5 474.9
2245.7 30?.4 577.2 547.2 3387.0
87.9 547.6
89.4 475.2
2247.1 304.9 577.3 547.6 3384.1
Thermal power, U1118, is the plant computer online calculation.
l
Column headings including computer point labels such as U0512, T0418A, etc
were taken directly from the trend computer output. All other column
headings reflect manipulation of the trend computer output.
,.
, , ' '
Attachment'4,.page 1 of 2
,
'
HEAT BALANCE
I
.
SEQUOYAH 1
05-09-89
t
DATA SET 1 OF 2
ENTHALPY
FLOW
POWER
POWER
1002 hours0.0116 days <br />0.278 hours <br />0.00166 weeks <br />3.81261e-4 months <br />
(BTUs/lb)
(E6'1b/hr)
(E9 BTUs/hr)
(MWt)
.
Steam
1195.5
3.673
4.391
o
412.2
-3.690
-1.521
1
Surface Blowdown
522.9
0.00000
0.00000
Bottom Blowdown
465.8
0.01386
0.00646
__
_
Power Dissipated
2.8770
842.6
Steam
1195.3
3.646
4.358
411.4
-3.669
-1.510
Surface Blowdown
523.4
0.00000
0.00000
j
Bottor Blowdown
465.7
0.02330
0.01085
1
3
___
Power Dissipated
2.8593
837.4
i
l
Steam
1195.5
3.768
4.504
412.2
-3.786
-1.560
Surface Blowdown
522.7
0.00000
0.00000
Bottom Blowdown
465.8
0.01928
0.00898
q
_______
'
Power Dissipated
2.9531
864.9
_
Steam
1195.5
3.659
4.374
411.8
-3.681
-1.516
i'
Surface Blowdown
522.8
0.00000
0.00000
Bottom Blowdown
465.6
0.02190
0.01020
Power Dissipated
2.8684
840.1
OTHER COMPONENTS
!
(
Letdown Line
543.9
0.03372
0.01834
i
Charging Line
459.3
-0.03622
-0.01664
Pressurizer
623.7
0.00020
0.00012
Pumps
-0.05839
Irasulation Losses
0.01749
i
Power Dissipated
-0.03906
-11.4
I
______
1
REACTOR POWER
3373.5
,
.-
>
!
- - _ _ _ _ .
..
..
Attachment 4c page 2 of 2
<
,
.
i
'
.'
HEAT BALANCE
l
SEQUOYAH 1
.
05-09-89
.
DATA SET 2 OF 2
ENTHALPY
' FLOW
POWER
POWER-
i
1022 hours0.0118 days <br />0.284 hours <br />0.00169 weeks <br />3.88871e-4 months <br />
(BTUs/lb)
(E6 lb/hr)
(E9 BTUs/hr)
(MWt)
\\
Steam
1195.5
3.657
4.371
l
1
411.9
-3.673
-1.513
Surface Blowdown
522.6
0.00000-
0.00000
]
1
Bottom Blowdown
465.5
0.013G6
0.00645
_______
Power Dissipated
2.8652
839.2
Steam
1195.3
3.617
4.324
q
411.2
-3.641
-1.497
i
Surface Blowdown
523.9
0.00000
0.00000
Bottom Blowdown
465.8
0.02330
0.01085
__
Power Dissipated
2.8373
831.0
$
1
Steam
1195.4
3.768
4.492
412.1
-3.776
-1.556
"
Surface Blowdown
523.1
0.00000
0.00000
Bottom Blowdown
465.9
0.01928
0.00898
__
_
]
Power Dissipated
2.9455
862.7
Steam
1195.4
3.667
4.384
411.5
-3.690
-1.518
Surface Blowdown
523.2
0.00000
0.00000
]
Bottom Blowdown
465.7
0.02190
0.01020
)
Power Dissipated
2.8755
842.2
'
OTHER COMPONENTS
'!
Letdown Line
543.9
0.03425
0.01863
l
Charging Line
459.7
-0.03588
-0.01649
l
,
L
Pressurizer
623.7
0.00020
0.00012
l
l
Pumps
-0.05839
,
l
Insulation Losses
0.01749
_
-
Power Dissipated
-0.03863.
-11.3
REACTOR POWER
3363.7
.
'
.
!
_ _ _ _ _ _ _ _ _ - - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ .
_ _ _ _ _ _ _ _ -
-
Attachment 5
,
.
.
-
.
,
,,
.
PARAMETER LIST
Unit Identification:
.
i
Plant Name
SEQUOYAH
Unit Number
2
Docket Number
50-328
Nuclear Steam System Supplier
Vessel and Piping:
Volume
10812 cubic feet
,
Pressurizer:
Level Units
%
Temperature Compensated
No
Calibration Curve
S. lope
517.13 pounds per %
Cpper Level Limit
100 %
Lower level Limit
0%
Relief
Relief Tank
.
I
Volume Control Tank:
Level Units
%
-
Calibration Curve
l
Slope
160.54 pounds per %
j
l
Upper Level Limit
100 %
!
Lower level limit
0%
i
'
Geometric Method Available
No
Drain Tank:
!
1
Level Units
%
,
!
Calibration Curve
Slope
26.7 pounds per %
,
'
Upper Level Limit
76.4 %
Lower level limit
26.6 %
Geometric Method Available
No
Belief Tank:
Level Units
%
Calibration Curve
Slope
1100 pounds per %
Upper Level Limit
90 %
l
Lower level limit
60 %
l
Geometric Method Available
No
.
r
- .
_ _ _ - - . _ . _ - _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _
- _ _ _ _ -
_ , , _
l
Attachment 6
l
^
f
' ,,
'
j
\\
.
.
-
.
j
g
,-
NRC
INDEPENDENT MEASUREMENTS PROGRAM
j
i
I
REACTOR COOLING SYSTEM LEAK RATES
STATION: SEQUOYAH
TEST DATE : 05-08-89
UNIT
- 2
START TIME: 0100
{
DOCKET : 50-328
DURATION
2.383 hours0.00443 days <br />0.106 hours <br />6.332672e-4 weeks <br />1.457315e-4 months <br />
{
l
TEST DATA
Initial
Final
System Parameters
Pressure, psia
2262
2265.7
T Ave, degrees F
570.3
570
Water Levels
Pressurizer, %
51.8
51.7
Relief Tank, %
81.8
82.5
Volume Control Tank, %
6771
62.3
Drain Tank, %
18
22
Water Charged
0 gal
Water Drained = 0 gal
-
=
TEST RESULTS
L
l
Change in Water Inventory in pounds:
l
i
Vessel & Piping
263
Relief Tank (1)
840
!
Pressurizer
-52
Drain Tank (1)
107
i
Volume Control Tank (1)
-771
l
Less: Water Charged
0
Collected Leakage
947
!
Plus: Water Drained
O
I
______
Cooling System
-559
!
Leak Rates in gpm (3):
Gross
O.47
Identified
0.80 -
Unidentified
-0.33
,
(1)
Determined from tank calibration curve.
(2)
Determined from tank dimensions.
(3)
The density used for converting inventory change to leak
rate was 62.31 pounds / cubic foot based on standard
conditions.