ML20138B669
| ML20138B669 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 10/02/1985 |
| From: | Bruce Bartlett, Cummins J, Martin L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20138B646 | List: |
| References | |
| 50-482-85-30, NUDOCS 8510210298 | |
| Download: ML20138B669 (16) | |
See also: IR 05000482/1985030
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APPENDIX B
US NUCLEAR REGULATORY COMMISSION
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NRC Inspection Report:
50-482/85-30
LP: NPF-42
Docket:
50-482
Licensee:
Kansas Gas and Electric Company (KG&E)
Post Office Box 208
Wichita, Kansas 67201
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Facility Name:
Wolf Creek Generating Station (WCGS)
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Inspection At: Wolf Creek Site, Coffey County, Burlington, Kansas
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Inspection Conducted: August 1 to 31, 1985
Inspectors:
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J. E. Cummins, Senior Reactor Inspector,
Date
Operations, (pars. 2,3,4,5,7,8,
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9, 10, 11, 12, 14, and 15)
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B. L. Bartlett, Resident Reactor Inspector,
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Operations, (pars. 2, 3,
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5, 6, 7, 8,
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Approved:
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L.Y Martin, CKef, Project Section B
Da(e/
Reactor Projycts Branch
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Inspection Summary
Inspection Conducted August I to 31,-1985 (Report 50-482/85-30)
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Areas Inspected:
Routine, unannounced inspection including plant status;
followup on previously identified items; operational safety verification;
engineered safety features system walkdown; startup test witnessing; startup
test data review; onsite followup of events; security; plant tours; monthly
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maintenance observation; surveillance witnessing; independent inspection; and
Ruskin fire dampers. The inspection involved 210 inspector-hours onsite by two
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NRC inspectors including 32 inspector-hours onsite during offshifts.
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8510210298 851009
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ADOCK 05000482
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Results: Within the 14 areas inspected, three violations were identified (test
port not reclosed in accordance with procedure, para. 4; inadequate procedure,
para. 8; and flow orifices installed backwards, para. 10).
One open item was
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identified (Cutler-Hammer Type E-30 switches, para.13).
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DETAILS
1.
Persons Contacted
Principal Licensee Personnel
- G.
L. Koester, Vice President-Nuclear
- C. C. Mason, Director-Nuclear Operations
- F. T. Rhodes, Plant Superintendent
- R. M. Grant, Director-Quality
- J.
A. Zell, Operations Superintendent
- M. G. Williams, Supt. of Regulatory, Quality, and Administrative
Services
0. L. Maynard, Licensing Supervisor
H. K. Chernoff, Licensing
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- K. Peterson, Licensing
W. B. Norton, Reactor Engineering Supervisor
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J. J. Johnson, Chief of Security
- W. M. Lindsay, Quality Systems Supervisor
R. Hoyt, Emergency Plan Supervisor
- R. Flannigan, Site Representative, Kansas City Power & Light
- W. J. Rudolph, Site QA Manager
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- L. E. Borders, Shift Supervisor
- A. A. Freitag, Site NPE Manager
- C. M. Herbst, Assistant Project Engineer-Bechtel
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- D. L. Reed, Lead Engineer-ISEG
The NRC inspectors also contacted other members of the licensee's staff
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during the inspection period to discuss identified issues.
- Denotes those personnel in attendance at the exit meeting held on
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September 4,1985.
2.
Plant Status
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On August 8, 1985, the plant initially reached 100 percent of full power
and commenced the 250-hour warranty run.
On August 12, 1985, the 250-hour
run was interrupted fc:' approxiaiately 7 days when main feed pump (MFP) 'B'
was removed frcm service for maintenance which was completed on
August 20, 1985.
The 250-hour wseranty run at 100 per'.ent power was
resumed on August 21, 1985, and completed on August e/, 1965.
During this
inspection period, power ascension testing at the 90 and 100 percent
levels was completed.
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3.
Licensee Actions on Previously identified Identified items
(Closed) Violation (50-482/EA05-27-IIC):
failure to Operate the Plant to
Prpcodural Requirements.
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This' item involved the failure to tag the removal of the blind flanges to
four flow elements in accordance with Administrative Procedure (ADM)02-101.
The NRC inspector reviewed the licensee's response to the Notice
of Violation, verified the change to ADM 02-101 for clarification had been
incorporated, reviewed the letter from the plant manager to all section
superintendents involved in test procedure implementation, and verified
through. field inspections of selected temporary modifications that the
requirements of ADM 02-101 are being met.
This item is closed.
(Closed) Violation (50-482/8511-03):
Failure to Operate the Plant to
Procedural Requirements.
This is a duplicate documentation of enforcement action violation
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(50-482/EA8527-IIC) which is discussed and closed above.
This item is
closed.
(Closed) Open Item (50-482/8508-05):
RD #28 Spent Fuel Pool Leak
Monitoring System.
This open item involved preoperational test deficiency deferral RD #28
concerning testing on the spent fuel pool leak monitoring system.
The NRC
inspector reviewed Test Discrepancy Report (TDR)-004.to test SU3-EC02,
Revision 0, and verified the required leak rate readings were obtained.
This item is closed.
(Closed) Open Item (50-482/8511-10):
RD #42 Rod Control Cluster Assembly
Change Fixture Engage Light.
The NRC inspector reviewed completed copies of Wolf Creek Work Requests
(WR) 2985-85 and 7606-85 which documented, respectively, the installation
and the functional testing of the engaged light for the rod control
cluster assembly change fixture.
This item is closed.
(Closed) Violation (50-482/8519-02):
Failure To Respond To Control Room
This item involved a failure to respond to Annunciator F-79.
The NRC
inspector has questioned on-duty personnel on their awareness of
annunciator panels and observed personnel follow the alarm procedures when
required.
This item is closed.
4.
Operational Safety Verification
The NRC inspectors verified that the facility is being operated safely and
in conformance with regulatory requirements by direct observation of
Itcensee facilities, tours of the facility, interviews and discussions
with licensee personnel, independent verification of safety system
status and limiting conditions for operations, and reviewing facility
records.
The NRC inspectors, by observation and direct interview,
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verified the physical security plan was being. implemented in accordance
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with the security plan.
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NRC inspector observations noted'during this inspection period are
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discussed below:
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During a tour of the plant on August 5, 1985, the NRC inspector observed
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the plug missing to a test port located in the ductwork on the discharge
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of containment purge filter Adsorber Unit FGT01.
At the time of the
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observation, the containment purge system was not in operation and there
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was no testing involving this system in progress.
This failure to control
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activity affecting quality in accordance with documented procedures is a
violation.
(50-482/8530-01)
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. Engineered Safety Features (ESF) System Walkdown
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The NRC inspectors verified the operability of ESF systems by walking down
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selected accessible portions of the systems.
The NRC inspectors verified
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valses and electrical circuit breakers were in the required position,
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power was available, and valves were locked where required.
The NRC
inspectors also inspected system components for damage or other conditions
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that might degrade system performance.
The ESF systems listed below were
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walked down during this inspection report period:
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Safety injection system
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Residual heat removai system
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Reactor coolant charging system
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Fire protection sprinkler system
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Essential service water system
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Control building heating, ventilating, and air conditioning
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Containment purge system
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No violations or deviations were identified.
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6.
Startup Test Data Review
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The NRC inspector reviewed completed power ascension test' procedures.
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Verification that all test changes, including deletions,-were
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approved, reviewed, and incorporated properly.
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Verification that all test deficiencies were resolved in accordance
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with the appropriate procedures.
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Verification that deficiencies which constitute a reportable
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occurrence as defined by Technical Specifications (TS) have been
properly recorded.
Verification that the as run copy of the completed test data package
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was properly completed.
' Verification that the test summary and evaluation were completed in
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accordance with procedure.
Verification that the test results were properly approved.
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The following test data packages were reviewed by the NRC inspectors for:
SU7-008.4
" Power Coefficient Determination-90 Percent Power,"
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Revision 1, dated July 22, 1985
SU7-0010.1
"Large Load Reduction at 75 Percent Power," Revi-
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sion 1, dated July 27, 1985
5U7-0020.6
" Turbine generator tests (90' Percent Reactor Power),"
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Revision 0, dated July 29, 1985
SU7-5014
" Test Sequence at 75 Percent Power," Revision 3, dated
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July 16, 1985
SU7-5015
" Test Sequence at 90 Percent Power," Revision 3, dated
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July 16, 1985
SU7-SC03.6
" Thermal Power and Statepoint Data Collection at
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90 Percent Power," Revision 1, dated July 30, 1985
SU7-SE02.8
" Operational Alignment of Nuclear Instrumentation,"
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Revision 1, dated July 8, 1985
SU7-SE03.2
" Axial Flux Difference (AFD) Instrumentation
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Calibration at 75 Percent Power," Revision 1, dated
July 25, 1985
SU7-SF06.5
" Operational Alignment of Process Temperature
Instrumentation," Revision 1, dated July 25, 1985
SU7-SR02
"Incore Moveable Detector and Thermocouple Mapping at
Power," Revision 1, dated January 23, 1985
STS RE-013
"Incore-Excore Detector Calibration," Revision 0,
dated July 3, 1985
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" Power Range Adjustment to Calorimetric," Revision 3,
dated July 27, 1985
SYS SR-200
" Moveable Incore Detector Operation," Revision 1,
dated November 1, 1984
No violations or deviations were identified.
7.
Startup Tcst Witnessing
Selected portions of the startup. tests listed below were witnessed to
ascertain conformance of the licensee to license and procedural
requirements, to observe the performance of the staff, and to ascertain
the adequacy of test program records, including preliminary evaluations of
test results.
SU7-SE03.3
" Axial Flux Difference Instrumentation Calibration,"
Revision 1
SU7-009.3
" Load Swing Tests-100 Percent Power," Revision 1
SU7-0010.2
"Large Load Reduction-100 Percent Power," Revision 1
SU7-0011
" Plant Trip From 100 Percent Power," Revision 0
SU7-0013
"NSSS Acceptance Test," Revision 1
Selected observations made by the NRC inspectors are discussed below:
SU7-SE03.3 - The difference between the indicated axial flux
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difference (AFD) and the measured flux difference exceeded the
1.5 percent acceptance criteria specified in Step 9.3 of the test
procedure.
The indicated axial flux differences were approximately
2 percent greater than the measured flux difference.
To correct this
situation, the nuclear instruments were realigned using an AFD
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instrumentation fine tuning procedure which was furnished by the
nuclear instrument vendor, Westinghouse, and after this realignment,
the axial flux difference indications were verified to be within
specifications.
A followup incore flux map was run and the new axial
flux difference indications were verified to meet the acceptance
criteria when compared to the data obtained from this incore flux
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map.
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SU7-009.3 - Step 6.11 of the test procedure specified a power increase
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of 10' percent power, but the maximum power attained was approximately
7 percent when the step was performed.
The controlling rod bank
reached its upper limit restricting further powerfincrease. .The
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licensee's evaluation of this test deficiency determined that the
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Final Safety Analysis Report (FSAR) requirements (Section 14.2.3.9.3)
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was to achieve an approximate 10 percent load decrease / increase and
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the 7 percent increase met the intent of the FSAR requirements.
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response of changes in plant parameters during this test was compared
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to test data taken for the same plant parameter changes at the
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75 percent power plateau and the licensee determined from this
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comparison that the response to a 10 percent load increase would have
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No violations or deviations were identified.
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8.
.Onsite Followup of Events
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The NRC inspector performed onsite followup of the nonemergency events
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listed below. -The NRC inspector observed control room personnel response,
observed instrumentation indicators of reactor plant parameters, reviewed
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logs and computer printouts, and discussed the event with cognizant
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personnel.
The NRC inspector verified the licensee had responded to the
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event in accordance with procedures and had notified the NRC and other
agencies as required in a timely fashion.
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Engineered safety feature actuations that occurred during the report
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period are listed in the table below.
The NRC inspector will review the
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license event report (LER) for each of these events and will report any
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findings in future NRC inspection reports.
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Date
Event
Plant Status
Cause
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CRVIS
Mode 1
Lightning strike on
transmission line
8/7/85
Rx Trip
Mode 1
MSR Hi-Hi level
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Specifics of-the reactor trip and related events that occurred on
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August 7, 1985, are discussed below:
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At 06:26 CDT on August 7 1985, with reactor plant power at-
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approximately 92 percent, a turbine trip occurred due to moisture
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separator reheater (MSR) Hi-Hi level.
Immediately after the turbine
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trip, the reactor tripped on P-9 (turbine trip with reactor power
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greater than 50 percent). During the trip an auxiliary feedwater
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actuation signal (AFAS), main feedwater isolation signal (FWIS) and
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steam generator blowdown isolation signal were received as expected
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on low steam generator level.
All equipment actuated properly except
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for two blowdown valves which did not indicate closed as requir'ed.
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Blowdown isolation valve BM HV-1 indicated closed at the radwaste
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control panel but did not indicate closed on the ESF status panel but
the closed indicator light on its control room handswitch failed to
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filuminate. ', Limit switches for the two valves were adjusted and.the
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valves were tested satisfactorily.
The cause for the MSR high level
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that initiated the trip could not be identified.
Minor adjustments
to approximately 24 secondary system controllers were performed and
the plant was restarted.
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In addition to the reactor trip and CRVIS listed above, the NRC inspectors
also performed followup on the defect / deficiency reports listed below.
The NRC inspectors reviewed the reports and verified:
The reportability determination was accurate and complete.
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The plant safety review committee reviewed the report when required.
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The proper corrective action was taken when required.
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The following defect / deficiency reports were reviewed:
85-105
" Air Flow From Containment to Environment Without a
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Release Permit"85-106
" Reactor Trip"
" Noncompliance with Technical Specifications-Missed Hourly
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Fire Watch"85-109 " Reactor Trip From Turbine Trip-MSR Hi-Hi Level"
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NRC inspector comments are noted below:
The review of Defect / Deficiency Report 85-105 " Air Flow From
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Containment to Environment Without a Release Permit," identified
deficiencies in Licensee Procedure STS IC-275B which allowed the
release of containment air to the outside atmosphere without first
obtaining and analyzing a grab sample as required by Technical
Specification surveillance requirement 4.11.2.1.2.
When the STS was
performed, Step 5.1.8 required that' isolation dampers GT HZ-4, 5, 6,
7, 8, 9, 11, and 12 be opened or checked open, with these isolation
dampers open there were only closed exhaust fan discharge dampers to
prevent the release of conttinment air to atmosphere. These
discharge dampers leaked allowing a release of containment air to
occur 'for 40 minutes before an operator observed the drop in
containment pressure and terminated the discharge.
When STS IC-275B
was written and reviewed, there was a failure to consider that all of
the isolation dampers did not need to open simultaneously and that a
grab sample should be taken in the event of leakage.
This failure to
prescribe a procedure appropriate to the circumstances is a
violation.
(50-482/8530-02)
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9.
Security
The NRC inspectors verified the physical security plan was being
implemented by selected observation of the following items:
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The security organization is properly manned and the security
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personnel are capable of performing their assigned functions.
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Persons,within the protected area (PA) display their identification
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badges, when in vital areas are properly authorized and, when-
required, are properly escorted.
Vehicles are properly authorized, searched, and escorted or
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controlled within the PA.
Persons and packages are properly cleared and checked before entry
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into the PA is permitted.
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The effectiveness of the security program is maintained when security
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equipment failure or impairment requires compensatory measures to be
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employed.
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Response to threats or alarms, or discovery of a condition that
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appears to require additional precautions is consistent with
procedures and the physical security plan.
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No violations or deviations were identified.
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10.
Plant Tours
-At various times during the course of the inspection period the NRC
inspectors conducted general tours of the reactor building, auxiliary
building, radwaste building, fuel handling building, control building,
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turbine building, essential service water building, and the secured area
surrounding the buildings.. During the tours, the NRC inspector observed
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housekeeping practices, fire protection barriers and equipment,
maintenance on equipment, and discussed various subjects with licensee
personnel.
NRC inspector findings are discussed below:
As a part of a plant tour,.the NRC inspectors inspected the installation
of flow orifice plates installed in safety-related systems ~ to verify the
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plates were installed in accordance with design drawings.
The orifice
plates are stamped with the work " Inlet"'which identifies the ' side of the
plate that should be installed facing upstream to the flow.
Drawing 556-32170, Revision 4, " Outline Drawing Concentric Orifice
. Plates," shows the correct orientation for installing the orifice plates.
The NRC inspector determined that Orifice Plates-EG-FE-83, " Centrifugal
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Charging Pump 'A' 011 Cooler," and EG-FE-89, " Residual Heat Removal Pump
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Seal' Coolers," were installed backwards from the orientation shown on
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Drawing 556-32170.
Failure to install these orifice plates in accordance
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with the drawing is a violation.
(50-482/8530-03)
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11.
Monthly Maintenance Observation
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The NRC inspectors observed maintenance activities performed on
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safety-related system, and components to verify that these activities were
conducted in accordance with approved procedures, Technical
Specificatfor.s, and applicable industry codes and standards.
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folicwing elcments were considered by the NRC inspectors during the
observation and/or review of the maintenance activities:
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Limiting conditions for operation (LCOs) were met and, where
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applicable, redundant components were operable.
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Activities complied with adequate administrative controls.
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Where required, adequate, approved, and up-to-date procedures were
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used.
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Craftsmen were qualified to accomplish the designated task and
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technical expertise (i.e., engineering, health physics, operati is)
were made available when appropriate.
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Replacement parts and materials being used were properly certified.
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Required radiological controls were implemented.
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Fire prevention controls were implemented where appropriate.
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Required alignments and surveillances to verify post maintenance
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operability were performed.
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Quality control hold points and/or checklists were used when
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appropriate and quality control personnel observed designated work
activities.
Selected portions of the maintenance activities listed below were observed
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and/or related documentation reviewed:
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WR No. 3292-85, Revision 1, Centrifugal Charging Pump 'A'
(CCP-A)
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rotating assembly replacement.
This WR documented the replacement of
the rotating assembly for CCP-A.
The pump shaft was gouged during an
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earlier removal of the shaft sleeve.
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WR Nos. 09603-85, 09601-85, 090602-85:
These work requests
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documented the replacement of terminal blocks in the terminal box
mounted.
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No violations or deviations were identified.
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Surveiltance Witnessing
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The NRC inspector observed selected portions of the performance of
surveillance procedures and verified the following items:
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Testing was being accomplished by qualified personnel in accordance
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with an approved procedure.
The surveillance procedure conforms to TS requirements.
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Required test instrumentation was calibrated.
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Technical Specification limiting conditions for operation were
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satisfied.
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Test data was accurate and complete.
The NRC inspectors performed
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independent calculations of selected test data to verify their
accuracy.
The performance of the surveillance procedure conformed to applicable
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administrative procedures.
The surveillance was performed within the . required f requency and the
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test results met the required limits.
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Surveillances witnessed by the NRC inspectors are listed below:
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STS IC-5048 - Channel Calibration of Reactor Coolant System Flow
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Transmitters
STS BG 100A - Centrifugal Charging System 'A' Train Inservice Pump
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Test, Revision 0
STS RE-004 - Shutdown Margin Determination, Revision 2
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No violations or deviations were identified.
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13.
Independent Inspection
a.
Unqualified Terminal Blocks Installed in Valve Terminal Boxes
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On August 8, 1985, the licensee informed the NRC resident inspector
and Region IV that unqualified (lower temperature rating) terminal
blocks had been used to replace damaged terminal blocks in electrical
terminal boxes on the main steam isolation valves (MSIV) and the main
feedwater isolation valves (MFIV).
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Fourteen of the unqualified terminal blocks were installed during the
period from December 1984 through July 1985.
The following table
shows the number of unqualified terminal blocks installed in each
affected valve and the approximate date they were installed.
Number of Unqualified
Approximate
Valve No.
Blocks Installed
Date Installed
MFIV AE-FV-42
6
December 7,1984
MFIV AE-FV-39
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December 8,1984
MSIV AB-HV-14
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July 31,1985
The unqualified terminal blocks are Type NU-2 blocks manufactured by
Connectron Incorporated and supplied to WCGS by Anchor-Darling Valve
Company.
These unqualified blocks were identical in design
configueation to the higher temperature qualified blocks but were
made out of polyamide (nylon) material whereas the qualified blocks
were made out of a polysulfone material.
Anchor-Darling supplies
both the low temperature polyamide terminal blocks (Part No. W 30164)
and the nign temperaidre polysulfone terminal blocks (Part
No. W 32676). On July 9,1984, WCGS received a lot of 16 of the low
temperature blocks which were incorrectly identified by
Anchor-Darling on all~ the accompanying documentation as Part
No. W 32676, the part number for the high temperature qualified
terminal block.
The licensee has determined that the 14 unqualified
terminal blocks that were installed in the MSIV and the MFIV terminal
boxes came from this lot of 16.
The licensee has replaced the unqualified terminal blocks with
qualified terminal blocks.
b.
Cutler-Hammer Type E-30 Switches
On August 9, 1985, the. licensee informed the.NRC' resident inspector
of a potential problem with Cutler-Hammer Type E-30 push button
switches.
The effects of the potential problem with the Type E-30
switches were first observed when three fire dampers failed to
automatically close as required when a control room ventilation
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isolation signal actuation occurred on July 19, 1985 (LER 85-057-00).
Followup investigation by the licensee determined that the dampers
had failed to close because a Type E-30 Cutler-Hammer handswitch
operating mechanism had hung up (failed to return to its normal
position).
This was a spring operated return-to-normal momentary
contact switch.
This type failure cannot be detected by observation
of the portion of the switch that is visible to the operator on the
front of the control panel but can be detected by observing the
switch plunger position at the back of the switch behind the control
panel.
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Actions taken by the licensee to evaluate / correct the potential
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problem included the following.
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Fifteen switches have been removed and sent to Cutler-Hammer for
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evaluation.
Five of these switches had exhibited some form of
failure whereas the.other ten were randomly selected to provide
additional data for evaluation.
A total of 457 Cutler-Hammer Type E-30 switches were identified
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to be installed in safety-related circuits.
These switches were
inspected and five of them were found in a stuck position during
this initial inspection.
Each of the 457 switches was evaluated
for its impact on safety and it was determined tnat 294 of the
switches could defeat an automatic safeguards actuation if it
failed.
The licensee has identified each of the 294 switches
located on the main control board in the control room with a
small label, and implemented administrative controls which
require verification that each switch operated has functioned as
required.
A weekly visual inspection of all identified switches
is being performed to verify that all switch plungers are in
their normal position.
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The licensee conducted a review of WCGS emergency operating
procedures (EMGs) and the activities related to each reference
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in the EMGs to operation of one of the E-30 type switches were
evaluated, and it was determined that there were no situations
where there would not be time to verify normal switch operation
prior to a required subsequent operation.
The licensee also performed an engineering analysis and.
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determined that the primary use of the switches is for valve or
ventilation damper control.
Each switch circuit was analyzed
and all situations were identified where switch binding could
inhibit an engineered safety features operation, where a bound
switch would not be obvious (i.e., an associated alarm
energized), and where subsequent operation of the affected
component could be inhibited.
The licensee, Cutler-Hammer, and Bechtel are analyzing the
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potential switch problem to determine if the observed switch
failures were isolated cases or if there is a generic problem.
This is an open item pending.the results of this analysis.
(50-482/8530-04)
No violations or deviations were identified.
14. -Ruskin Fire Dampers
- Three concerns related to the' ability of louvered fire dampers installed
at WCGS and manufactured by Ruskin Manufacturing Company to function per
design'under required conditions have been addressed by the licensee.
Each of these concerns and the corrective action is discussed below:
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a.
Failure of Dampers To Fully Close Against Ventilation System Air
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Flow.
The licensee, in conjunction with Ruskin and Bechtel (architect /
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engineer), made modifications to the damper closure spring and
associated hardware and then performed an operational test of all
dampers (vertical and horizontal).
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Mullion Welds on Multisection Horizontal-Dampers.
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By letter dated April 22, 1985, Ruskin Manufacturing Company, in
accordance with the requirements of 10 CFR Part 21, notified the NRC
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that welds used to join damper sections and mullion plates were not
in accordance with design drawings, Ruskin Drawing 5415.
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Underwriters Laboratory's evaluation determined that the existing
welds 1-inch on 9-inch centers in lieu of 1-inch on 6-inch centers
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were acceptable and would.not affect the 3-hour _ fire rating of the
Design drawings were changed to show the existing weld
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conditions at WCGS.
The license inspected all affected dampers and
rewelded one damper where the weld spacing exceeded the 9-inch
criteria.
c.
-Inadequate Clearance Between Damper' Frames and Embedded Penetration
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By letter dated May 1, 1985, KG&E, in accordance with the-
requirements of 10 CFR Part 21, notified the NRC that some of the
Ruskin fire dampers installed at WCGS had less than the designed
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clearance between the fire damper frame and the penetration sleeve in
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which the damper was installed.
This created a condition that might
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prevent the affected dampers from funtioning as designed due to
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buckling caused by-thermal expansion during a fire.
The licensee has
implemented a program to inspect all affected fire dampers and to
replace or modify the dampers so that each damper installation has
the required gap between the damper and its associated penetration
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The above concerns with Ruskin fire dampers were reported and discussed in
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WCGS LER 85-017-00. The modifications,and retesting have been
- satisfactorily completed on all the installed dampers-except for eight
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dampers which will require cold shutdown plant conditions due to their
location in safety-related ventilation systems.
NRC Open Item
50-482/8515-01 was initiated to track concerns with the Ruskin fire
dampers. This open item was discussed in NRC Inspection Report
50-482/85-19. This item will remain open pending completion of all work
and testing of Ruskin' fire dampers related to the concerns discussed
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above.
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15.
Open Items.
Open items are matters which have been discussed with the licensee, which
will be reviewed further by the inspector, and which involve some action
on the part of the NRC or licensee or both.
An open item disclosed
during the inspection is discussed in paragraph 13.
16.
Exit Meetings
The NRC inspectors met with licensee personnel to discuss the scope and.
findings of this inspection on September 4, 1985.
The NRC inspectors also
attended entrance / exit meetings of other NRC region-based inspectors
identified below:
Inspection
Lead.
Area
Inspection
Period
Inspector
Inspected
Report No.
8/12-16/85
R. Baer
Radwaste
85-32
8/19-23/85
C. Hackney.
Emergency
Preparedness
85-33
8/26-30/85
J. Kelly
-Security'
85-34
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