ML20035G936
| ML20035G936 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 04/23/1993 |
| From: | Burnett P, Crlenjak R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20035G930 | List: |
| References | |
| 50-327-93-11, 50-328-93-11, NUDOCS 9304300199 | |
| Download: ML20035G936 (6) | |
See also: IR 05000327/1993011
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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101 MARtETTA STREET. N.W.
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ATLANTA, GEORGt A 30323
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Report Nos.: 50-327/93-11 and 50-328/93-11
Licensee: Tennessee Valley Authority
3B Lookout Place
1101 Market Street
,
Chattanooga, TN 37402-2801
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Docket Nos.:
50-327 and 50-328
License Nos.:
Facility Name: Sequoyah 1 and 2
Inspection Conducted: March 22-25, 1993
Inspector: f'T. buy
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Approved by:
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R. V. Crienjak,Jasief
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Operational Pr6 grams Section
Operations Branch
Division of Reactor Safety
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SUMMARY
Scope:
This routine, unannounced inspection was conducted in the areas of close out
of open items and witnessing of training for post-refueling start-up tests.
Results:
Two inspector follow-up items were closed. One unresolved item was determined
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to be a violation because of inadequate procedures for evaluating setpoints in
excess of the allowable values of LIMITING SAFETY SYSTEM SETTINGS
(paragraph 2.c).
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Questions regarding calibration of Eagle-21 and the response of that system to
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a potential over-voltage condition were resolved satisfactorily (paragraph 3).
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The reactor engineering group was judged to be performing well (paragraph 4).
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9304300199 930423
ADOCK 05000327
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REPORT DETAILS
1.
Persons Contacted
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Licensee Employees
- R. J. Beecken, Plant Manager
D. R. Branham, Instrument Engineer
- M. A. Cooper, Site Licensing Manager
- J. B. Espy, Reactor Engineering
- R.
A. Fenech, Site Vice President
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- T. A. Flippo, Site Quality Manager
R. G. Gladney, Instrumentation and Control Supervisor
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- J. F. Lemons, Reactor Engineering
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- Z. I. Martin, Nuclear Fuel Project Manager
- H. R. Rogers, Technical Support Manager
- M. A. Skarzinski, Reactor Engineering Supervisor
- J. D. Smith, Regulatory Licensing Manager
- R. R. Thompson, Compliance Licensing Manager
- G. T. Tiner, Reactor Engineering
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P. G. Trudel, Engineering Manager
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- J. N. Ward, Manager, Engineering and Modifications
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- N. A. Welch, Operations Superintendent
C. H. Wittemore, Compliance Engineer
Other licensee employees contacted during this inspection included
engineers, training department staff, and administrative personnel.
NRC Resident Inspectors
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- W. E. Holland, Senior Resident Inspector
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S. M. Shaeffer, Resident Inspector
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A. R. Long, Resident Inspector
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- Attended exit interview
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2.
Action on Previous Inspection Findings (92702, 61705)
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a.
(Closed)
IFI 50-327 and 50-328/92-01-01: The licensee is
considering changes in the use of the alternate dilute mode to
prevent over-dilution of the VCT during extended use of that mode of
operation.
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Flow to the VCT spray was not shutoff by the procedure then in use,
and it did not include a specific requirement to make charging flow
greater than dilution flow. Without such controls on the dilution
process, experience has shown that a reactivity overshoot may occur
during the subsequent mixing process. The overshoot can challenge
the PDIL.
Furthermore, it had been postulated, in the PRA analysis
for another facility, that restart from a station blackout with a -
diluted VCT may lead to a severe reactivity transient.
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Report Details
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Precautions in procedure 0-RT-NUC-000-0003.0 (Revision 1), Initial
Criticality, required only that the charging rate exceed the
dilution rate.
In the course of this inspection, the licensee
revised this procedure (Revision 2) to require that FCV-62-128 be
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closed when the alternate dilute mode of operation is used. The
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action prevents dilution water from entering the VCT through the
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spray line.
In addition, 0-RT-NUC-000-0003.0, Rod Bank Worth
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Measurement Using Dilution /Boration Method, was revised (Revision 2)
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to return FCV-62-128 to P-AUTO once dilution was terminated.
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b.
(Closed)
IFI 50-327 and 50-328/92-01-02: The acceptance criteria
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for control rod worth measurements will be reviewed for consistency
with assumptions made in SDM analysis.
0-RT-NUC-000-008.0 (Revision 0), Low Power Physics Testing
Acceptance Criteria, was the source of the acceptance criteria for
control rod reactivity worth measurements. The acceptance criterion
for the sum of the measured control rod worths was that the sum
exceed 93% of the sum of predicted worths.
That criterion was
consistent with the assumptions in SDM calculations. A criterion of
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17% agreement between predicted and measured reference bank worth
appeared to be equally appropriate and necessary, because all other
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measurements of control rod worth are dependent upon it. The
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licensee did not agree to the additional acceptance criterion.
As discussed in inspection report 50-413 and 50-414/92-26, this
issue is generic to most PWRs, and will be referred to NRR for
resolution.
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c.
(Closed) URI 50-328/92-21-01, Reportability of IR N-36 Miscalibra-
tion described in Incident Investigation S-92-049.
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During inspection 92-21, the inspectors questioned the licensee
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about the reportability of IR N-36 miscalibration.
The licensee
stated the miscalibration was not reportable because the rack error
was within the TS allowable limits and the miscalibration was not
safety significant. The licensee cited the Instrumentation Society
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of America standard for setpoint methodology and the Eagle-21 SER as
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the reason for considering only rack error.
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Following Unit 2 Cycle 6 refueling, the NI trip setpoints were
adjusted to correct for expected differences in flux to the NIs from
the time of their last calibration to that expected with the new
core. The licensee used 0-PI-NUC-092-081.0 (Revision 2), Prestart-
up NIS Calibration Following Core Load, to derive the prestart-up
calibration data. After initial start-up and low power physics
testing was complete, 0-PI-NUC-092-082.0 (Revision 1),.Poststart-up
NIS Calibration Following Core Load, was performed at a nominal 4%
thermal power. This procedure checks the effective IR reactor trip
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setpoint.
Using this procedure, IR N-36 reactor trip setpoint was
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Report Details
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investigated by personnel from operations, instrument maintenance,
and engineering departments. The licensee's investigation
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determined that incorrect weighting factors were the most probable
cause.
Inaccurate power determination and IR N-36 instrument drift
in the nonconservative direction were contributing causes.
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The licensee found, during the investigation, that IR N-36 trip
setpoint, in the instrument rack, had drifted 0.034 Vdc in the non-
conservative direction. Once the unit was at higher power, and a
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more precise heat balance had been performed, the licensee found
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that IR N-36 would have tripped at 32.3% RTP, during start-up. This
was 12.3% RTP higher than the intended, conservative, trip setpoint
of 20% RTP.
The licensee's assessment was that two-thirds of this
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difference was a result of the flux weighting error, which they
classified as process error. The remaining one-third of the
difference was from the 0.034 Vdc non-conservative drift in the
setpoint, which the licensee classified as rack error.
The licensee was proactive in the testing and adjustment of IR N-36.
The flux-weighting adjustment to trip setpoints had yielded
conservative results in the past. The miscalibration was found at
the earliest possible opportunity, before reactor power was
increased to a significant level.
Reactor. power was not increased
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until IR N-36 was recalibrated and the reactor trip function
verified within TS limits. The licensee has developed an improved
flux-weighting scheme for pre-calibration of the irs. When applied
to historical data from both units, the new scheme yields
consistently conservative results for the trip setpoint.
Unfortunately, the review of the incident by the site compliance /
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licensing group was incorrect. The ACTION statement for TS 2.2.1
requires that, for any setpoint less conservative than the allowable
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value given in Table 2.2-1, the channel be declared inoperable. An
inoperable trip channel should be reported pursuant to 10 CFR 50.73.
The site licensing group concluded that the difference between trip
setpoint and allowable values of Table 2.2-1 was limited to rack
error. Since the rack error (licensee classification) did not by
itself lead to violation of the allowable value, and process error,
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in their judgement, did not need to be considered;
the site
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licensing group concluded that the channel was not inoperable;
hence, no report was required.
Their basic premise is incorrect.
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The allowable value includes all sources of error, including process
error and rack drift. Anytime the allowable value is exceeded, the
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specification is violated or the function is inoperable.
The licensee has vigorously defended their position using language
contained in SAFETY EVALUATION BY THE OFFICE OF' NUCLEAR REACTOR
REGULATION SUPPORTING AMENDMENT NO. 141 TO FACILITY OPERATING
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LICENSE NO. DPR-77. TENNESSEE VALLEY AUTHORITY, SE000YAH NUCLEAR
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. PLANT.-UNIT 1. DOCKET NO. 50-327.
That reference does include the
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Report Details
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statement, "The allowable values of Tables 2.2-1 and 3.3-4 are be ng
revised to reflect rack drift allowances associated with the
Eagle-21 digital process protection system." The term reflect does-
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not limit the error tolerance to rack drift effects only.
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In summary, the inspector concluded that the licensee's procedures
and practices for evaluation of operability of trip system setpoints
are inadequate.
Furthermore, the site compliance / licensing group
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was insensitive to the safety significance of the term LIMITING
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SAFETY SYSTEM SETTING, for this occurrence.
Failure to have
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adequate procedures to evaluate the import and consequence of
violating the allowable value of a LIMITING SAFETY SYSTEM SETTING
has been identified as a violation, which will be tracked as item
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50-327 and 50-328/93-11-01.
3.
Follow-up of Eagle-21 Issues (61705)
Inspection report 92-01 discussed problems with use of a data logger to
obtain cross-calibration data for the Unit 1 RTDs with inputs to
Eagle-21. During the inspection, the licensee stated an intention to
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fully check out the data logger system and_ demonstrate its suitability
for the task at the first possible opportunity; when Unit 2 entered MODE
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3 during the scheduled shutdown of March 1992. The inspector reviewed
the data obtained during the shutdown and during the subsequent heatup to
operating temperature and discussed the results with the responsible
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engineer. All results were acceptable, proving the utility of the data
logger in this application.
In the opinion of the engineer, the vendor-
recommended DMM would not have yielded better results than the data
logger.
On March 1,1993, Unit 2 was shutdown by a feedwater heater line break.
Steam from the break entered the electrical generator and led to an over-
voltage condition on plant electrical circuits powered by the generator.
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The 480 Vac boards supplying the power inverter to Eagle-'21 reached
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potentials as high as 580 Vac during the transient.
The licensee
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contacted the manufacturer of the inverter and received written assurance
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that the inverter output should not have exceeded its 120 Vac output
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during the transient. _Furthermore, the switching power supplies for
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Eagle-21 are designed to fail to protect the circuits they power. None
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of the power supplies were found tripped or failed. The licensee's
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conclusion that Eagle-21 was not adversely affected by the over-voltage
transient appears to be well founded.
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4.
Reactor Engineering Unit (72700)
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The inspector witnessed two half-days of simulator training for the
reactor engineering staff in preparation for start-up testing of Unit 1
at the end of the current refueling outage.
The testing appeared to be
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well-simulated, and the trainees were attentive to the program of
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instruction. This is a productive initiative on the part of the
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Report Details
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licensee.
In inspection report 90-29, concern was expressed for the
number and experience level of the reactor engineering staff.
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performance of the reactor engineering group has removed that concern.
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The current unit supervisor is being promoted to another part of the
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licensee organization. However, the licensee has been successful in
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recruiting an experienced supervising reactor engineer from another
utility.
No diminution of the effectiveness of the reactor engineering
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group is anticipated,
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5.
Exit Interview
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The inspection scope and results were summarized on March 25, 1993, with
those persons indicated in paragraph I, above. The inspector described
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the areas inspected and discussed in detail the inspection results listed
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below. Although reviewed during this inspection, proprietary information
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is not contained in this report.
Dissenting comments were not received
from the licensee at the time of this interview, but they did indicate
that they would like to review further the proposed violation. A
telephone conference on the subject of the characterization of instrument
errors was conducted on April 15, 1993.
Participants included TVA,
Westinghouse, NRR, and Region II personnel. At the conclusion, Region II
management concluded that the violation listed below should be issued.
VIO 50-327 and 50-328/93-11-01:
Failure to have adequate procedures to
evaluate the importance and consequence of violating the allowable value
of a LIMITING SAFETY SYSTEM SETTING (paragraph 2.c).
6.
Acronyms and Initialisms
DMM
digital multimeter
flow control valve
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IFI
inspector follow-up item
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IR
intermediate range nuclear instrument
NI
nuclear instrument
PDIL
power dependent insertion limits
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pressurized water reactor
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resistance temperature detector
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safety evaluation report
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unresolved item
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volume control tank
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violation
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