ML20245D867
| ML20245D867 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 06/14/1989 |
| From: | Belisle G, Burnett P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20245D852 | List: |
| References | |
| 50-425-89-20, NUDOCS 8906270255 | |
| Download: ML20245D867 (6) | |
See also: IR 05000425/1989020
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION '
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101 MARIETTA STREET,N.W.
ATLANTA, GEORGI A 30323
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Report Nos.: 50-425/89-20.
' Licensee: Georgia Power' Company
P. 0.'. Box l1295
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Birmingham, AL.35201
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Docket Nos.: 50-425'
License No.: NPF-81
Facility. Name:
Vogtle 2.-
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Inspection Conducted- May.16'- 20, 1989
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Inspector:bd. _M
4-/4 ~8 7
p. T. Bifrpett
Date Signed
Accompanying Personnel: B
R. Eaton, PWR Instructor
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NRC Technical Training Center
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Approved y:
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4 -/ t'-99 -
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G. A. BeAigle, Chief
Date Signed
Test Programs Section
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Engineering Branch
Division of Reactor Safety
SUMMARY
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Scope:
This special, announced inspection addressed the areas of Unit 2
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response to loss of turbine load transient of May 14,1989, review of
completed startup tests, and witnessing of 100 percent loss of ^1oad
test.
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Results:
It was determined that Unit 2 did not reach a condition during the
loss of turbine load transient that should have. initiated an automat-
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ic trip or demanded a manual trip from.the. operators..-These conclu-
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sions were supported by approximate simulation .at the plant
simulator.
The licensee's procedures for review of plant events that do not-lead
to a plant trip were found to have two weaknesses: There is.insuffi-
cient guidance to personnel-on the collection of data' for detailed
review of such events ~ and- early involvement of operations personnel
in the review of all plant events is not assured.
No violations or deviations were identified.
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8906270255 890621
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REPORT DETAILS
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Persons Contacted
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Licensee Employees
- M. J. Aj1 uni, Operations Superintendent
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- G. B. Bockhold, Jr., General Manager, Vogtle Nuclear Operations -
- W. L. Burmeister, Operations Superintendent-
J. A. Dobbs, Onshift Operations Superv.isor
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S.- Dodds, Simulator Instructor, Licensed Operator Training
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R. E. Dorman, Licensed Operator Training Supervisor .
- W. :C. Gabbard,1 Senior Regulatory Specialist
- C, G. Garrett, Operations Engineer
D. E. Gustafson, Plant Engineering Supervisor.
J. B.'Joyner,. Plant Engineer
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- W. F. Kitchens, Assistant. General Manager, Operations and Maintenance-
- A..L. Mosbaugh, Plant Support Manager
R. O. Odom, Plant Engineering Superintendent
Other licensee employees contacted included engineers, shift supervisors,
operators, and office personnel.
Other Organizations
R. J. Florian, Southern Company Services '
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C. B. Holland, Westinghouse
W. C. Phoenix, Consul Tec
0. D. Hayes, Consul Tec
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NRC Resident Inspectors
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- R. F. Aiello, Resident Inspector
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J. F. Rogge, Senior Resident Inspector
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- Attended exit interview on May 19, 1989.
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Acronyms and initialisms used throughout this ' report .are listed in the
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last paragraph.
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2.
Turbine Load Transient on May 14,1989(92705)
During the morning of May- 14, 1989, Vogtle Unit 2 was- operating at a
nominal 90%. RTP ' with a generator output of approximately 980 Mwe. 'All-
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control systems were in automatic 'except for rod control, which was in
manual.
Two of three banks.of backup pressurizerLheaters were turned on
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manually.
The RCS C had recently been. diluted .to compensate for xenon
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buildup, and the heaters were on to increase pressurizer spray to bring
the pressurizer C in equilibrium with that of the RCS.
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At approximately 8:02 am CDT, the turbine control and combined intermedi-
ate valves closed.
Generator output dropped to zero. 'The valves then
opened and generator output increased to 920 Mwe. The total elapsed time
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for this event was less than two minutes, and the generator unloading and
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loading rates were approximately 200%/ min, each way.
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The following observations of plant performance during that two minute
period were made from review of plant recorder traces and output of the
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alarm typer.
On the secondary side, all steam dumps and atmospheric
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relief valves opened as steam pressure rose from 980 psig to a peak of
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1180 psig.
As load picked up and stabilized, these valves and dumps
reseated without any difficulty.
Steam generator levels fell in response
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to the secondary pressure increase, but did not fall to the trip setpoint.
Once steam flow was reestablished, the levels swelled, but not to the trip
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setpoint.
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On the H mary side, pressurizer pressure rose from the normal 2235 psig
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to 2306 psig. One pressure-rate sensitive PORV opened briefly. The alarm
typer records do not indicate the degree of PORY opening, but do show that
the valve was not fully shut for 13 seconds.
Subsequently, pressure
dropped to 2120 psig in response both to the PORV opening and the cooldown
of the RCS as the heat sink was reestablished.
With the loss of heat
sink, average RCS temperature rose from 586 F to 601 F with a concomitant
increase in pressurizer level from 60 to 80%.
Also, with the rapid
increase in pressurizer level, the third stage of backup pressurizer
heaters came on automatically.
The inspector calculated the loss of RCS
mass by thermal dilution to be 12,512 lbs of water.
The corresponding
increase in pressurizer inventory from the 20% level change was calculat-
ed, by the inspector, to be 12,350 lbs of water. This close agreement in
inventory values indicates the instrumentation used was performing accept-
ably.
The 162 lbs of water difference in the two masses is equivalent to
3 seconds of - full-flow PORV operation.
Since the PRT 1evel was not
recorded, no further correlation could be made.
The difference between
the two mass calculations is probably as indicative of the precision
possible in the calculations as it is of PORV operation.
The records show that, during the transient, both the OPdT and 0 tdt trip
setpoints lowered as expected, but with the loss of heat sink the actual
dT also dropped and maintained a margin to the changing trip setpoints.
The reactor power, as indicated by the recorder for the selected PRNI,
dropped from 91% to 87% and returned to 91% during the transient.
This
response is consistent with nominal values of MTC and DPC, but can not be
used to evaluate either. .The response of the recorder is much slower than
the response of the reactor to changes in temperature and power level.
Furthermore, the increased temperature of the water in the downcomer, at
the time of minimum power, would have increased the leakage flux at the
detectors partially offsetting the flux reduction from power reduction.
Hence, the actual reduction in ieactor power may have been greater than
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indicated on the PRNI recorder, but precise determination of the' drop in
reactor power is not necessary for the evaluation of this transient.
Following. review of the records discussed above and discussions with
members of the event evaluation team, the inspectors interviewed ^ separate-
ly the members of:the control room crew on duty during the transient. 'By
sheer coincidence, ,the 0 SOS wasc in the: Unit 2 control room when the
transient began.- From his description of his activities and recollections
of the details of the event, the inspectors concluded he had performed his
duty of maintaining an. overview of the entire plant. status and the activi -
. ties of control room personnel in a thoroughly professional. and proficient
manner.
His.past experience as a simulator instructor contributed to his
ability to maintain a broad. overview rather than becoming ;too focused on a
particular plant parameter.
He did maintain. cognizance of:those parama-
ters, which would dictate ordering a manual trip of the unit.
The Unit 2 shift supervisor too was in the control room throughout. the
transient, and also appeared to have a good recollection of the entire
status of the unit as it went through its perturbation. 'He-stated that
one of his first actions following the apparent recovery of the unit was
to confirm by reference to the TS that the safety limit on temperature and
pressure had not been exceeded.
It was his judgement also that no plant-
parameter he observed had reached a limit that should initiate an automat-
ic trip or require a manual trip of the reactor.
There were two licensed operators in the control . room.
One was assigned
to B0P and the other to MCB.
Both appeared to the inspectors to have
focused on the significant parameters within their . individual purviews and
to have responded appropriately to those parameters.
In particular, the
B0P operator took feedwater pressure control in hand .immediately after
power recovery to increase feed flow and reduce the chance of a trip'on
steam generator low level.
Although there was no similar event programmed on the plant simulator,
manual simulations were performed with the active cooperation of the
simulator instructor staff.
When the turbine control . valves were closed
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and reopened ' rapidly, the simulator unit did not trip, nor did the
response lead to a condition requiring a trip.
When the valve positions
were changed slowly, the -simulated unit tripped on low steam generator
level.
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The inspectors concluded that, unusual as the transient was, no plant
parameter reached a state that should have required either an automatic or
manual trip.
An event review team was formed shortly after the occurrence. . By virtue
of its makeup and its charter, the team focused on the cause. of the
turbine valve closures and the performance of the secondary system.
Little early effort was spent by the team in evaluating primary side
performance, and an operations department member was not appointed to the
team until three days after the event.
Team membership aside, it did
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appear to the inspectors that the operations staff did perform a prompt
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and thorough evaluation of primary plant performance.
It was only the
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integration of evaluation efforts that was delayed.
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Because the event did not lead to a reactor trip,much of the data collect-
ed automatically following a trip or manually, in accordance' with proce-
dure 1006-C, Reactor Trip Review, were not captured. More data would have
facilitated the event review.
The cause of the turbine valve closure had not been determined with
certainty by the end of the inspection and will be inspected in the
future.
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No violations or deviations were identified.
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3.
Review of Completed Startup Tests (72578)
The following completed startup test procedures were reviewed:
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a.
2-BB-01 (Revision 0), Reactor Coolant System Flow Measurement at
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Power, was performed during the period April 20 to May 16,1989, at
nominal power levels of
30% 50%, 78%, 90%, and 100% RTP.
The
minimum value of 403,578 gpm, at 100% RTP, satisfied the TS 3.2.5.c
minimum limit of 396,198 gpm.
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2-6SE-02 (Revision 0), Incore Movable Detector and Thermocouple
Mapping, was performed repeatedly during the period April 19 to May
18, 1989.
Proper functioning of the equipment under test was
confirmed.
No violations or deviations were identified.
4.
Witnessing of the Plant Trip from 100 Percent Power (72580)
Startup test procedure 2-700-02 (Revision 0) . Plant Trip from 100% Power,
was performed on Saturday morning, May 20, 1989.
Prior to performance of
the test, the inspector reviewed the procedure and discussed it with test
and operations personnel.
The inspector confirmed the test met the
description of FSAR 14.2.8.2.53 and satisfied the requirements of RG 1.67,
Appendix A, paragraph 1.1.
The inspector attended the pretest briefing of shift and test personnel
and witnessed control room activities during the test.
All systems
appeared to perform satisfactorily.
One minor casualty occurred as a
result of the test: a safety-relief vrive blew off the feedwater side of
the 6B feedwater heater.
The inspectur viewed the area; noted that the
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steam from a line about two inches in diameter was blowing straight up
through a grating floor above and that no safety related equipment ap-
peared to be in the path of the steam.
After conferring with the shift
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supervisor and receiving his assurance that the plant response to the t' rip
had been normal and that a normal shutdown procedure was nearly completed
without identifying, any problems, :the inspector 11 eft the _ site.-
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completed test procedure will be inspected ' during a future- routine
inspection.
No violations or deviations were. identified.
5.
Exit Interview '(30703)
The inspection scope and findings were summarized on May 19,- 1989, with,
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those persons indicated in paragraph _IL above. . The inspectors described'
the areas inspected and discussed in detail the inspection findings.
In
particular, the inspectors' discussed the need for more guidance. to person-
nel on the collection of data for. detailed review of. events that do- not
lead to a plant trip.
The inspectors emphasized .the need for early
involvement of operations personnel in the review of all plant events.
Dissenting comments were not received . from the licensee.
Proprietary
information is not contained in this report.
6.
Acronyms and Initialisms Used in This' Report
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doppler power coefficient
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delta temperature:
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dT
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FSAR -
final Safety Analysis Report
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gallons per minute
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gpm
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pounds
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lbs
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MCB -
main control board
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moderator temperature coefficient
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MTC
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Mwe -
megawatts electrical
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OPdT -
overpower delta temperature
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OSOS -
onshift operations supervisor
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OTdT -
overpower delta temperature'
PORV -
power operated _ relief. valve
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PRNI -
power range' nuclear instrument-
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PRT .-
pressurizer relief tank
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psig -
pounds per square inch - gauge
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reactor. coolant system-
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- Regulatory Guide
rated thermal power
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Technical Specification ~
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TS
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