ML20056A289
| ML20056A289 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 07/30/1990 |
| From: | Joel Wiebe NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20056A286 | List: |
| References | |
| 50-482-90-26, NUDOCS 9008060301 | |
| Download: ML20056A289 (14) | |
See also: IR 05000482/1990026
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APPENDIX B
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U.S. NUCLEAR REGULATORY COMMISSION
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REGION IV
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NRC Inspection Report:
50-482/90-26
Operating License: NPF-42
Docket: 50-482
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Licensee: Wolf Creek Nuclear Operating Corporation (WCN00)
P.O. Box 411
Burlington, Kansas 66839
Facility Name: Wolf Creek Generating Station (WCGS)
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Inspection At: WCGS, Coffey County, Burlington, Kansas
Inspection Conducted: June 1-30, 1990
Inspectors:
M. E. Skow. Senior Resident Inspector
Project Section D, Division of Reactor Projects
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L. L. Gundrum, Resident Inspector,
Project Section D Division of Reactor Projects
D. V, Pickett, Project Manager
Project Directorate IV Division of Reactor Projects III, IV, and V
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and Special Projects, NRR
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W. B. Jones, Senior Project Engineer
Project Section D Division of Reactor Projects
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Approved:
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T S. Wiebe, Chief, Project Section D
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ivision of Reactor Projects
Inspection Summary
Inspection Conducted June 1-30, 1990 (Report 50-482/90-26}
Areas Inspected:
Routine, unannounced inspection including plant status -
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operational safety verification, monthly surveillance observation, monthly
maintenance observation, plant startup from refueling, onsite followup of
events at operating power reactors, and review of licensee event reports.
Results: The licensee's corrective action program failed to assure that
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previous corrective actions precluded recurrence of similar events.
This
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was illustrated in the corrective action response to licensee event
reports (LERs)90-004 and 90-010. These LERs failed to address adequate
corrective action _for scheduling of surveillance tests (paragraph 8.a) and the
control' of troubleshooting activities (paragraph 8.b).
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0008060301 900731
ADOCK 05000482
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Plant personnel were cognizant of surveillance and corrective maintenance
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requirements. Good radiation protection support was observed during
observation of surveillance maintenance activities.
The licensee's control of
troubleshooting activities continued to be a weak area, requiring additional
licensee management oversight.
Operations personnel performed their duties in accordance with the facility
license-requirements. The inspector reviewed the licensee's reactor posttrip
evaluations from the previous inspection period.
It was identified that the-
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evaluations were not performed in accordance with the posttrip review
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procedure.
The licensee's documentation of these posttrip reviews did not
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clearly identify the root cause, nor prescribe and verify that the necessary
corrective actions have been completed (paragraph 9).
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DETAILS
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1.
Persons Contacted
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Principal Licensee Personnel
B. D. Withers, President and CEO
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- J. A. Bailey, Vice President, Nuclear Operations
F. T. Rhodes, Vice President, Engineering and Technical Services
- G. D. Boyer, Plant Manager
- R. S. Benedict, Manager, Quality Control (QC)
- H. K. Chernoff, Supervisor, Licensing
- M. E. Dingler, Manager, Nuclear Plant Engineering (NPE) Systems
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- T. J. Garrett, Manager, Nuclear Safety Analysis
- W. Goshorn, Wolf Creek Coordinator, Kansas Electric Power Company
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- R. L. Gourley, Supervisor, Mechanical Maintenance
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- C, W. Fowler. Manager, Instrumentation and Control (l&C)
R. W. Holloway Manager, Maintenance and Modifications
- R. L. Logsden, Manager, Chemistry
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J. MacTaygart, NPE, Wichita
- A. S. Mah, duelear Safety Engineering
- T. S. Morrill, Manager, Radiation Protection
- D. G. Mosety, Supervisor, Operations
- W. B. Norton, Manager, Technical Support
- C. E. Parry, Director, Quality
- J. M. Pippin, Manager, NPE
- C, M. Sprout, Section Manager, NPE, WCGS
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- J. D. Weeks, Manager, Operations
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- S. G. Wideman, Senior Licensing Specialist
- M. G. Williams, Manager, Plant Support
- R. T. Wright, Supervisor, QA Audit
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The inspectors also contacted other members cf the licensee's staff during
the inspection period to discuss identified issues.
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1990.
T. P. Gwynn, Deputy Director Division of Reactor Projects,. also
attended the exit.
2.
Plant Status
The plant operated at or near full power throughout the month. On June 2
9, and 23, 1990, power was reduced to 95 percent reactor thermal power
while maintenance work was performed on low pressure feedwater heater
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trains.
3.
~0yerational Safety Verification (71707)
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'.The purpose of this inspection was to ensure that the facility was being
operated safely and in conformance with license and regulatory
requirements.
It also was to ensure that the licensee's management
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control system was effectively discharging its responsibilities for
continued safe operation.
The methods used to perform this inspection
included direct observation of activities and equipment, tours of the
facility, interviews and discussions with licensee personnel, independent
verification of safety system status and limiting conditions for
operation (LCO), corrective actions, and review of facility records.
Areas reviewed during this inspection included, but were not limited to,
control room activities, routine surveillances, engineered safety feature-
operability, radiation protection controls, fire protection, security,
plant cleanliness, instrumentation and alarms, deficiency reports, and
corrective actions.
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Plant operations were generally performed in an acceptable manner.
The
inspectors noted that, in the plant cleanliness and health physics (HP)
areas, the licensee was continuing to clean up and perform decontamination
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of areas and equipment used during the refueling outage that ended in May
1990.
Equipment was being prepared for storage onsite until iequired for
the next outage. A discussion was held with the licensee concerning the
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general state of plant preservation.
The licensee stated that there'is an
ongoing long-term program in place to paint equipment and piping
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throughout the plant. This effort has been observed by the_ inspectors.
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In addition, the licensee has begun to paint some walls in the aux 1111ary
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building which would make decontamination easier in the event of a small
4.
Monthly Surveillance Observation (61726)
The purpose of this inspection was to ascertain whether surveillance of
safetj-significant systems and components was being conducted in
accordance with Technical Specification (TS). Methods used to perform this
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inspection included direct observation of licensee activities and review
of records.
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Items inspected in this area included, but were not limited to,
verification that:
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Testing was accomplished by qualified personnel in accordance with an
approved procedure;
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The surveillance procedure was in conformance with TS requirements;
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The operating system and test instrumentation was within its current
calibration cycle;
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Required administrative approvals and clearances were obtained prior
to initiating the test;
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LCOs were met and the system was properly returned to service; and
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The test data were accurate and complete and the test results met TS
requirements.
Surve111ances witnessed and/or reviewed by the inspectors are listed
below:
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STS10-913, Revision 7 " Containment Hydrogen Analyzer G5065B
Calibration Test"; and
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CHM 01-060, Revision 5, " Sampling of the Reactor Coolant System."
Selected inspector observations are discussed below:
a.
On June 7,1990, the inspector observed the performance of STS10-913
for the Containment Hydrogen Analyzer "B" train. This surveillance
is performed to verify operability of this accident monitoring
instrument as required by TS 3.3.3.6.
The I&C technicians were noted to be cognizant of the precautions and
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limitations specified in the surveillance test procedure. The
equipment was energized and maintained in standby for at least 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
prior to beginning the test.
The redundant channel was operable as
required with the reactor in Mode 1.
Prior to beginning the
surveillance test, an onshift licensed operator was n..ified of
the intent of the procedure and permission was granted by the operator
before the test was initiated.
The inspector noted that the technicians utilized test equipment
which was within its calibration test frequency.
Each required lead
was lifted in accordance with the test procedure.
In each case, the
"as found" data was within its specified tolerance range. The data
was properly recorded within the procedure,
b.
On June 8, 1990, the inspector observed the drawing of a reactor
coolant sample and subsequent sample analysis. The control room was
notified of the intent to obtain the sample. The required valve
lineup was established by the operator to allow the chemistry
technician to draw the sample. The proper radiological controls were
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in place prior to obtaining the sample.
The inspector verified that
the chemistry technician had signed the appropriate procedure.
The system was purged for greater than 30 minutes prior to drawing
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the sample. The sample hood was noted to be properly positioned and
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the ventilation system taking a suction from inside the hood.
The sample was subsequently drawn and analyzed in accordance with
the surveillance procedure. The technicians were noted to be
cognizant of the test requirements and the tests were performed in
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accordance with the procedure.
Each test result was within the
established acceptance criteria.
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The inspector noted tnt both surveillances were conducted in accordance
with the surveillance p edure by qualified personnel.
In each case, the
surveillance procedure accomplished its intended verification. A
potential violation has been identified in paragraph 8.a of this report
for tne failure to resolve surveillance test scheduling errors.
5.
Monthly Maintenance Observation (62703)
The purpose of inspections in this area was.to ascertain that maintenance
activities on safety-related systems and components were conducted in
accordance with approved procedures and TS.
Methods used in this
inspection included direct observation, personnel interviews, and records
review.
Items verified in this inspection included:
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Activities did not violate limiting conuttions for operations and
that redundant components were operable;
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Required administrative approvals and clearances were obtained before
initiating work;
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Radiological controls were properly implemented;
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Fire prevention controls were implemented;
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Required alignments and surveillances to verify postmaintenance
operability were performed;
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Replacement parts and materials used were properly certified;
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Craftsmen were qualified to accomplish the designated task and
additional technical expertise was made available when needed;
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QC hold points and/or checklists were used and QC personnel observed
designated work activities; and
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Procedures used were adequate, approved, and up to date.
Portions of selected maintenance activities regarding the work
requests (WRs) were observed. The following WRs and related documents
were reviewed by the inspectors:
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No.
Activity
WR 50882-90
Oil sample from the "B" emergency diesel generator
WR 50273-90
Perform preventive maintenance on essential service
water Valve EF HV0052
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WR 50275-90
Perform preventive maintenance on essential service
water Valve EF HV0060
WR 02895-90
Perform corrective maintenance on centrifugal charging
pump flow control Valve BG FCV121.
WR 03534-90
DRPI Data "A" failure on Rod K-10
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Selected inspector observations are discussed below:
a.
On June 1,1990, the digital rod position indication (DRPI) system
simultaneously generated a number of alarms including the rod nonurgent
alarm, rod urgent alarm, rod bottom alarm, and rod position
indication (RPI) rod deviation alarm. Upon control room
acknowledgement of the alarms, only the rod nonurgent alarm for the
K-10 shutdown rod locked in. Panel lights indicated that only the
"B" train was providing input to the DRPI system with the "A" train
input faulted.
Review of the core parameters and the demand position
indication system by the control room operators verified that rod
movement had not occurred and that the K-10 shutdown rod remained
fully withdrawn from the core. A condition referred to as " half
accuracy" existed for the DRPI system monitoring the K-10 shutdown
rod. The licensee stated that a " half accuracy" condition does not
degrade DRPI system operability and that entry into TS 3.1.3,2 is not
appropriate.
Troubleshooting by I&C personnel on June 1, 1990, resulted in
replacing the K-10 circuit card in the DRPI cabinet in containment.
Upon replacement of the circuit card, the DRPI alarm cleared on the
control room panel.
However, it was noted by the inspector that
subsequent testing of the circuit card by I&C personnel did not
identify any apparent pr +' ems.
The DRPI alarms of June I surrounding the K-10 shutdown rod were
again repeated when troubleshooting was being performed on the DRPI
cabinet in containment on June 18, 1990.
The control room operators
were not told by the-troubleshooting technicians that the alarms
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could come in, nor did the operators expect the indication and alarms
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to occur. Again the "A" train faulted and a " half accuracy" situation
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resulted in the indication for K-10 shutdown rod position, This
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failure to adequately control troubleshooting activities is a repeat
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of a similar problem identified in LER 90-010. " Technical
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Specification Violation - Simultaneous Inoperability of Two Auxiliary
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Feedwater Pumps Because of Support System Inoperability." The
corrective actions specified in the LER (described in detail in
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Section 8.b of this report) did not preclude repetition of the
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problems.
Even though the troubleshooting activity performed on
June 18, 1990, did not result in equipment inoperability, the event
illustrates that the licensee has not established the appropriate
management control of troubleshooting activities.
This event is used
as additional information which further demonstrates that the
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corrective action specified in LER 90-010 was not adequate. See
Section 8.b of this report for a description of the violation.
Following the second June 18, 1990, event, I&C personnel made a
containment entry on June 21, 1990. By disconnecting the power cable
from the K-10 circuit card at the DRPI cabinet and connecting it to a
specially designed rod position simulator, I&C personnel could
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identify whether the fault existed in cable connections between the
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control room and the DRPI cabinet or between the DRPI cabinet and the
actual K-10 rod.
The inspector obscrved the testing from the control
room. The inspector observed that operations and I&C personnel
discussed the need to maintain RPI in the control room throughout the
testing process.
In order to test the "A" train input, the control
room operators initially arijusted the DRPI mode selection switch from
monitoring both "A" and "B" train input to monitoring "B" train
only, multiple spurious ORPI alarms were generated (i.e., rod urgent
alarm, rod bottom alarm, and RPI rod deviation alarm).
However, when
the alarms were acknowledged, only the rod nonurgent alarm locked in.
General warning DRP1 alarm lights existed from all control and
shutdown rods during the testing due to the " half accuracy" condition
that existed due to isolation of the "A"
train input.
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Troubleshooting by I&C personnel using the rod position simulator
concluded that a fault existed in the power cable somewhere between
the DRPI cabinet in containment and the K-10 shutdown rod. The areas
of concern are physically located near to and behind the biological
shield wall and leads down to the reactor vessel upper internal area.
Due to as low as reasonably achievable (ALARA) considerations, the
licensee decided that further troubleshooting and entry behind the
biological shield wall would not be appropriate with the plant at
100 percent power.
Repairs to this system were deferred to the
forced outage list.
The DRPI system was returned to the " half
accuracy" condition for monitoring the K-10 shutdown rod position,
b.
WRs 50273-90 and 50275-90 were performed on June 6, 1990, to verify
proper positioning of the limit switches and to lubricate the valve
actuator including the geared limit switch.
The work was performed in accordance with approv9d WRs.
Prior to
beginning the work, the appropriate clearance w.s implemented, and
the operators entered into the required LCO for one train of
essential service water out of service.
No une;9ected conditions
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were noted by the craf tpersons or the inspector curing the
maintenance activity.
Following completion of the maintenance activity, a QC inspector
observed the setting of the limit switches. A postnaintenance test
was performed on the two valves prior to returning the system to
service.
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WR 02895-90 was performed on June 7, 1990, to repair a body-to-bonnet
1eak on centrifugal charging pump flow control Valve BG FCV121.
This valve had been worked during the previous refueling outage to
repair a similar type leak.
Prior to beginning the work, the required clearance boundaries were
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established and independently verified. A specific radiation work
permit (RWP) was initiated to identify the radiological hazards
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within the contaminated area around the valve and to specify the
necessary personnel controls. The operators followed the appropriate
LCO when the system was t en out of service.
The inspector observed that the maintenance personnel were adhering
to the RWP requirements.
HP assistance was requested.to control the
potentially contaminated water that would be released when the body-
to-bonnet was separated. The system configuration did not allow
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draining of the 1tne prior to performance of the maintenance
activity.
The potentially contaminated water was well controlled
within the designated ares.
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After removing the bonnet, the licensee identified that the gasket
was off center, resulting in the body-to-bonnet leakage. A new
gasket was installed, and the valve reassembled.
New packing was
also installed. The torquing of the bonnet to the valve body was
observed by a QC inspector. During the maintenance activity, HP
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t :hnicians surveyed the area outside the controlled contamination
area to ensure that the contaminated area was not increasing.
The
valve was subsequently tested in accordance with the established
postmaintenance test.
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The licensee's performance in the area of troubleshooting continued to be
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a weakness.
The corrective maintenance activities were correctly performed
in accordance with the maintenance procedures by qualified maintenance
personnel.
6.
. Plant Startup from Refueling (71711)
In addition, the following additional procedures were. incorporated into
this inspection in accordance with Inspection Procedure 71711:
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Surveillance of Core Power Distribution Limits (61702)
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Calibration of Nuclear Instrumentation Systems (61705)
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Core Thermal Power Evaluation (61706)
The following tests for heatup, approach to critica1'ty, and core phys 4s
were witnessed and/or reviewed and found to have be9n conducted in
accordance with approved procedures:
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STS SF-003, Revision 2, " Digital Rod Position Indication Test";
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STS RE-007, Revision 4, " Rod Drop Time Measurement";
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RXE 01-002, Revision 2, " Reload Low Power Physics Testing";
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STS RE-Oll, Revision 4 and Temporary Procedure Change MI 90-491, "RCS
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- ts1 Flow Rate Measurement";
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F a nE-013, Revision 3, "Incore-Excore Detector Calibration"; and
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STS10-201, Revision 8 and Temporary Procedure Change MI 90-516,
" Analog Channel Operational Test 7300 Process Instrumentation
Protection Set
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Primary personnel responsible for the major steps in obtaining the results
of the computer analysis code calculations from the initial incore flux
map data were interviewed. A flux map printout at greater than 50 percent
power was reviewed with the reactor engineering supervisor.
It was
verified that control rod insertions, core power level, and burnup at the
time of the flux map were part of the input to the code. The review
confirmed that all detectors independently traversed a reference
calibration instrument tube. The printout included the relative set of
measured fission rates as seen by the detector for each thimble location
following normalization.
The axial flux differences and quadrant power tilt limits were maintained
during startup.
The source, intermediate, and power range calibrations
and the incore-excore detector calibrations were performed as required.
The site reactor engineering staff were knowledgeable regarding the
computer codes being used and ADM 01-077, Revision 4, " Computer Software
Administrative Controls." The reactor engineering staff organization is
authorized by the licensee to have four people.
The staff has two
experienced engineers and one engineer being trained for the position.
One position is vacant.
The Cycle 5 reload safety evaluation was reviewed.
The current Updated
Safety Analysis Report (USAR), Revision 3, dated March 12, 1990, was
reviewed to ascertain the parameters for the previous Cycle 4,
Chapter 4,
" Reactor," still reflected the description-for Cycle 3.
Cycle 4 started
in-January 1989. This failure to update the USAR is an apparent violation
of the requirements of 10 CFR 50.71(e)(4), which requires revisions to
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the USAR to be filed annually and to reflect all changes up to a maximum
of 6 months prior to the date of filing (482/9026-03).
The licensee
revised their procedure for updating the USAR prior to the end of the
. inspection period. Because the criteria specified in Section V.A of the
Enforcement Policy was satisfied, a cited violation will not be issued.
The licensee controlled the heatup, approach to criticality, and core
physics test in e.::cordance with the approved procedures. Good
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communication between the operators and site reactor engineering staff
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was evident. The licensee's implementation of the requirements specified
in 10 CFR 50,71(e)(4) needs to be improved.
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Onsite Followup of Events at Operating Pnwer Reactors (93702)
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The purpose of this inspection activity was to provide onsite inspection
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of events at operating power reactors.
Specific inspection activities
included:
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Observing plant status;
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Evaluating the significance of the events, performance of safety
systems, and actions taken by the licensee;
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Confirming that the licensee had made proper notification of the
events and of any new developments or significant changes in plant
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conditions; and
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Evaluating the need for further or continued NRC response to the
events.
The following items were considered during the followup:
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Details regarding the cause of the event;
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n .ctioning of safety systems as required by plant conditions;
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Radiole
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.scensee actions to correct the cause of the event; and
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ctions taken or planned prior to resumption of facility
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Selected events rec Jiring licensee event reports (LERs) that occurred
during this report seriod are listed in the table below:
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Event *
Plant Status
Cause
6/13/90
Mode 1
Loss of No. 7 transformer
(100% power)
caused loss of power to
XNB01 Class 1E transformer
6/19/90
TS violation
Mode 1
Chemistry TS action
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(100% power)
statement time limits
exceeded
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6/19/90
TS violation
Mode 1
GT RE21A went into
(100% power)
accident / isolate mode
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TS V W ica - Technical Specification Violation
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T4 mspracts verified that the licensee's initial response to these
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avens ves oppropriate.
Plant response to the initiative event was also
appropriate when applicale
The inspectors will review the LERs for these events and will report any
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findings in a subsequent inspection report.
8.
Review of Licensee Event Reports (92700) and Evaluation of Licensee
Self-Assessment Capability (40500)
a.
LER 90-004, " Technical Specification Surveillance Requirement Not
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Satisfied Because of Personnel Error in Scheduled Surveillance Test."
The corrective action was to proceduralize the requirement to check
the past due surveillance procedures inquire report (PDSIR) on a
weekly basis as a minimum. A similar commitment was made by the
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licensee in 1987.
This similar committment was documented in
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LER 87-014 to require the frequency of checking the PDSIR weekly,
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More recently, LER 89-016, issued September 8, 1989, addressed a
failure to meet a TS surveillance requirement because of an error in
scheduling. A noncited violation (482/8923-01) was issued in
1989 for-the-licensee's failure to perform the surveillance test on
schedule and enter TS 3.0.3 after discovery.
The failure to perform
the surveillance test in 1989 was due to a schedulin0 error.
The inspectors reviewed the past due surveillance report dated
June 22,'1990,.and found that it had 87 entries. The entries did not-
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accurately reflect actual cases of missed surveillance tests. The
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inspectors concluded that the POSIR has not been effective in
identifying missed surveillance tests. This appears 'o be partially
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because of the number of invalid entries in the PDSIR.
The
licensee's surveillance test scheduling computer preg' Jam, which
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developes the PDSIR, does not minimize the report output sufficiently
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to allow for a simple review of missed surveillance test. The
licensee's failure to identify corrective action which would reduce
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the probability of similar events occuring in the future in
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accordance with 10 CFR 50.73 is an apparent violation (482/9026-01),
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b.
LER 90-010, " Technical Specification Violation - Simultaneous
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Inoperability of Two Auxiliary Feedwater Pumps Because of Support
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System Inoperability." At 7:32 a.m. (CDT), Emergency Diesel
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Generator "A" was removed from service for planned maintenance. This
rendered the "A" auxiliary feedwater pump (AFP) inoperable according
to TS and the appropriate LC0 was logged.
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At 3:45 p.m., it was discovered that the inspection access door
for the AFP "B" room cooler was removed, thereby rendering the "B"
AFP inoperable.
Investigation revealed that the hatch had been
removed at approximately 8 a.m.
Therefore, two AFPs had been
inoperable for a period of approximately 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
Per TS 3.7.1.2,
the plant should have been in at least hot standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
with two AFPs inoperable.
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There were two corrective actions stated in the LER. The first
action was to convey to operations and maintenance personnel that
inspection access doors must be installed in order to ensure
operability of room coolers.
The second action was to make
maintenance personnel aware of the necessity to thoroughly discuss
troubleshooting activities in detail with the shift supervisor prior
to commencement of the activity.
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TS 6.8 requires that written procedures shall be established,
implemented, and maintained covering the applicable procedures
rcrommended in Appendix A of Regulatory Guide (RG) 1.33, Revision 2,
February 1978.
Section 9, Item a, states, " Maintenance that can
affect the performance of safety-related equipment should be properly
preplanned and performed in accordance with written procedures,
documented instructions, or drawings appropriate to the circumstance."
The next statement in Section 9. Item a, states " Skills normally
possessed by qualified maintenance personnel may not require
step-by-step delineation in a procedure."
The corrective actions stated in the LER are considered inadequate
because the administrative centrols do not require that activities
that can affect safety-related equipment or the control room
operators' ability to assess plant status be controlled with a
written procedure. A further demonstration of the inadequacy of the
corrective action specified in LER 90-010 was the inadequate control
of maintenance troubleshooting which occurred on June 18, 1990, and
is further described in Section 5.a of this report for work on the
ORPI system,
,.
If adequate corrective action'had been-taken, the potential for
io
'
operator confusion during the troubleshooting of the DRPI monitor
might have been avoided.
This is a second example of a potential
violation (482/9026-01).
The licensee's corrective action for this
'
LER and the potential violation will be reviewed by the NRC staff
'
(482/9026-01).
LER 90-10 is closed.
.
9.
Followup on a Previously Identified NRC ltem (92701)
(Closed) Unresolved Item (482/9022-02):
Posttrip Reviews - The posttrip
reviews for May 14, 17, and 19, 1990, were reviewed. Attachment A of
3
ADM 02-400, Revision 7, "Posttrip Review," was used as the basis for
'
documentation.
Section 6.5.1 states that strip-chart recordings must
accurately reflect real time to have meaningful information.
The
s
j
_ _ _ _ _ _ .
.
{e
.
.
i
14
strip-charts included in all three V p reports did not indicate real time.
Another item noted was the failure to complete Section 6 of Attachment A
for the posttrip review for the May 14, 1990, trip.
Section 6.12.1
requires that the actual or suspected cause of the trip and any abnormal
or degraded indication identified during the transient shall be documented
in Part 6 of the posttrip review report.
However, that information was
contained in other portions of the report.
In addition, the posttrip
review for the May 19, 1990, trip did not address all the actions taken to
correct the known cause of the trip as well as the actions taken to check
other similar equipment for possible generic problems.
Efforts taken by
the licensee to look for additional contributing causes were not listed
in the posttrip review. The purpose for the posttrip review, as stated in
ADM 02-200, Sections 1.1 and 2.1, is to diagnose the cause of the trip,
ascertain the proper function of safety-related and other important
equipment prior to restart, and provide a documented review of the results
that will permit a determination to be made as to the readiness of the
plant to safely return to operation.
Part of ensurir.g the plant is ready
for restart is to look for less obvious contributing causes and to ensure
that all necessary corrective actions have been completed.
Failure to
have thorough and complete posttrip reviews is a violation for failure to
follow procedure (482/9026-02).
During the review of the procedure, the following concerns were noted:
o
Section 6.13.1 states, "The maximum and minimum values of parameters
on Attachment A shall be compared with their design specification."
Part 3 of the posttrip review report is to be used to document this
safety assessment.
Part 3 of Attachment A lists the parameters and
provides columns for the minimum and maximum values.
No value for
the design specification is listed nor is there a column to document
the design value,
o
Part 3 of Attachment A lists specified parameters. There are no
additional lines to allow the documentation of any additional
parameters that the reviewer may feel are significant, such as the
. rate of increase or decrease of Tavg.
Unresolved Item 482/9022-02 is closed.
10.
Exit Meeting (30703)
The inspector met with licensee personnel (denoted in paragraph 1) on
June 29, 1990. The inspector summarized the scope and findings of the
inspection. The licensee did not identify as proprietary any of the
information provided to, or reviewed by, the inspectors. M r. T . P . Gwyn n ,
Deputy Director, Division of Reactor Projects, Region IV, also attended
the exit meeting.