IR 05000413/1987010
| ML20214K890 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 05/14/1987 |
| From: | Lesser M, Peebles T, Skinner P, Van Doorn P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20214K867 | List: |
| References | |
| 50-413-87-10, 50-414-87-10, IEB-84-03, IEB-84-3, NUDOCS 8705290083 | |
| Download: ML20214K890 (13) | |
Text
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UNITED STATES
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jo NUCLEAR REGULATORY COMMISSION
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j 101 MARIETTA STREET, N.W.
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ATL ANTA, GEORGI A 30323
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Report Nos.
50-413/87-10 and 50-414/87-10 Licensee: Duke Power Company 422 South Church Street Charlotte, N.C.
28242 Docket Nos.:
50-413 and 50-414 License Nos.: NPF-35 and NPF-52 Facility Name: Catawba 1 and 2 Inspection Conducted: March 26 - April 25, 1987 Inspectors:
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P. K. Van Doorn
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Date ' Signed p T. A.~Peebles, Secit on Chief Projects Branch 2 Division of Reactor Projects SUMMARY Scope: This routine, unannounced inspection was conducted on site inspecting in the areas of review of plant operations; surveillance observation; maintenance observation; review of licensee nonroutine event reports; followup of previously identified items; and followup of IE Bulletins.
Results: Of the six (6) areas inspected, no apparent violations or deviations were identified.
8705290083 870514 PDR ADOCK 05000413 G
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REPORT DETAILS 1.
Persons Contacted Licensee Employees J. W. Hampton, Station Manager H. B. Barron, Operations Superintendent W. F. Beaver, Performance Engineer W. H. Bradley, QA Surveillance S. Brown, Reactor Engineer B. F. Caldwell, Station Services Superintendent R. N. Casler, Operating Engineer R. H. Charest, Station Chemistry Supervisor
- M. A. Cote, Licensing Specialist
- T. E. Crawford, Integrated Scheduling Superintendent W. P. Deal, Health Physics Supervisor C. S. Gregory, I. & E. Support Engineer
- C L. Hartzell, Compliance Engineer
- J. Knuti, Operating Engineer F. N. Mack, Project Services Engineer
- W. W. McCollough, Mechanical Maintenance Supervisor C. E. Muse, Operating Engineer F. P. Schiffley, II, Licensing Engineer G. T. Smith, Maintenance Superintendent J. Stackley, I. & E. Engineer D. Tower, Shift Operating Engineer
- R. F. Wardell, Technical Services Superintendent J. W. Willis, Senior QA Engineer, Operations Other licensee employees contacted included technicians, operators, mechanics, security force members, and office personnel.
- Attended exit interview.
2.
Exit Interview The inspection scope and findings were summarized on April 24, 1987, with those persons indicated in paragraph 1 above. The inspector described the areas inspected and discussed in detail the inspection findings.
No dissenting comments were received from the licensee. The licensee did not identify as proprietary any of the materials provided to or reviewed by the inspectors during this inspection.
The following new items were described at the exit interview:
Unresolved Item 413/87-10-01, 414/87-10-01: Single Failure Vulnerability of the Nuclear Service Water Syste _=
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Unresolved Item 413/87-10-02: Potential Failure to Implement Procedures for a Degraded Standby Shutdown Facility (SSF)
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Unresolved Item 414/87-10-02: Review of Possible Safety Significant Problem Involving Safety Injection System J
Inspector Followup Item 414/87-10-03: Review of ND Hot Leg Suction Valves Maintenance
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Licensee Identified Violation 413/87-10-03: Operability of Available Power i
Sources Not Verified Due to Management Deficiency l
Licensee Identified Violation 414/87-10-04: Technical Specifications Violation Oue to a programming error in the Operator Aid Computer Reactor Coolant Leakage Program.
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Inspector Followup Item 413/87-10-04:
Safety Tag Verification by Work
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Supervisor for Block Tagouts i
3.
Licensee Action on Previous Enforcement Matters (92702)
I a.
(CLOSED) Unresolved Item 413/85-26-03: Mechanism to Assure Support j
Systems are Functional to Support Major System Operation. Operations Management Procedure 2-29, Revision 12.,
Technical Specification
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Logbook, contains in Attachment 3 a listing of systems / components that are required as support systems.
In addition this attachment
provides a list of possible compensatory measures that can be implemented to maintain a piece of equipment in an operable status if
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the normal support equipment is not available.
This list was
developed by station personnel with assistance from design j
engineering.
Based on this review this item is closed.
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(CLOSED)
Violation 413/86-51-01, 414/86-54-01: Failure to Follow
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Procedures and TS 4.11.2.1.1 Involving Radioactive Gas Release. The
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response to this item was provided in correspondence dated March 6, j
1987.
The inspector reviewed the corrective action taken and
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considers this item closed.
It is noted that Station Directive i
3.10.1 is being changed as part of the corrective action. Although this is not yet complete, followup inspection will be conducted
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through followup of the LER issued regarding this event (LER 413/
l 87-01).
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(CLOSED)
Violation 413/86-51-02, 414/86-54-02: Failure to Follow i
j Procedure for Annunciator Response.
The licensee response to this item was provided in correspondence dated March 6, 1987.
The inspector reviewed the corrective actions taken and considers this
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i item closed.
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(CLOSED) Violation 414/87-05-03: Failure to Follow Procedures
Resulting in a Violation of Technical Specification 4.0.2 and 4.0.4.
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This violation was addressed in licensee correspondence dated i
April 8, 1987.
The inspector has reviewed the corrective actions described in the licensee's correspondence and considers this item
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closed.
l No violations or deviations were identified.
4.
Unresolved Items
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t Unresolved items are matters about which more information is required to determine whether they are acceptable or may involve violations or deviations. Three new unresolved items are identified in paragraphs Se,
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5.
PlantOperationsfeview(Units 1 & 2) (71707 and 71710)
I a.
The inspectors reviewed plant operations throughout the reporting period to verify conformance with regulatory requirements, Technical Specifications (TS), and administrative controls. Control room logs, danger tag logs. Technical Specification Action Item Log, and the
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removal and restoration log were routinely reviewed. Shift turnovers were observed to verify that they were conducted in accordance with approved procedures.
The inspectors verified by observation and interviews, the measures taken to assure physical protection of the facility met current requirements.
Areas inspected included the security organization,
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the establishment and maintenances of gates, doors, and isolation zones in the proper condition, that access control and badging were proper and procedures followed.
In addition to the areas discussed above, the areas toured were
observed for fire prevention and protection activities.
These included such things as combustible material control, fire protection systems and materials, and fire protection associated with maintenance activities. The inspectors reviewed Problem Investigation Reports to i
determine if the licensee was appropriately documenting problems and implementing appropriate corrective actions.
b.
The inspectors conducted a detailed walkdown of portions of the Nuclear Service Water System on both Units and portions of the
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Chemical and Volume Control System.
IAn Unresolved Item is a matter about which more information is required to determine whether it is acceptable or may involve a violation or deviation.
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c.
Unit 1 Summary On April 6,1987, Unit 1 entered Technical Specification (TS) 3.0.3 and commenced a power reduction from 100% due to discovery that IWV testing on Residual Heat Removal (ND) pump suction check valves FW-28 and FW-56 had not been performed. (Unit 2 also entered TS 3.0.3 and i
was in Mode 3 at the time). Required tests were performed on both units and Unit 1 halted the power reduction at 93% and returned to 100% power.
On April 9, Unit I tripped from 100% power on Steam Generator Low Low Level, when valve ICF-33 (Feedwater Containment Isolation Valve) failed shut, due to a blown fuse in the actuator control circuit.
Later that same afternoon while attempting to restart, the reactor was manually tripped prior to criticality when rod position indication failed for one rod in Shutdown Bank A.
Unit I returned on-line on April 10, 1987.
d.
Unit 2 Summary On March 26,1987, Unit 2 was in Mode 3 repairing stem leakage on 2NC-33, a pressurizer PORV block valve, when the actuator shaft was inadvertently sheared by maintenance technicians.
The unit entered into a mini-outage and returned on-line on April 10. On April 13 the unit shutdown from 100% power in accordance with Technical Specification 3.4.6.2 due to excessive unidentified leakage measured at 1.8 gpm. The leak was isolated to leakage through a relief valve on the sample line to the Reactor Coolant Radiation Level Monitor (EMF-48).
e.
Discussions with licensee personnel and review of Problem Investigation Report 2-C87-0072 indicated that Unit 2 may have been in an unreviewed safety situation regarding train B of the Safety Injection (NI) System. Operations personnel had opened valves 2NI968 and 2NI120B to bleed off pressure in the piping on the discharge side
of the B NI pump to the Refueling Water Storage Tank (RWST).
This
was done since Reactor Coolant (NC) System leakage into the line was causing pressure to increase to the relief valve setpoint of 1750
psig. A single B train power supply failure could have resulted in NI flow diversion to the RWST resulting in low NI flow or a possible
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unmonitored radiation release via the RWST.
This problem was discovered by a licensee Test Review Committee formed in response to an NRC violation (see RII Report 413,414/87-08).
Licensee Design Engineering is analyzing the situation.
Since further analysis is
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required and this could have been a significant problem this is Unresolved Item 414/87-10-02: Review of Possible Safety Significant i
Problem Involving 3afety Injection System.
f.
On April 8, 1987, Unit 2 experienced a control rod bank Demand Position Indicator failure for control bank 0 while in Mode 1.
Technical Specification (TS) 3.1.3.2 requires the systerr to be I
operable and with the Digital Rod Position Indication System, to be l
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also give indication of bank Demand Position and were verified to be
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i operating correctly. The licensee was able to localize the fault to the mechanical counter. Since an alternate means of determining bank
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Demand Position was available the licensee did not enter the action
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statement of TS 3.1.3.2 on the basis that the system was determined
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After discussions with responsible engineers, this was j
determined acceptable by the inspector.
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On April 10,-1987, the inspector noted that the Standby Shutdown
Facility (SSF) Standby Makeup pump for Unit I was logged as
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inoperable in the Technical Specifications Action Item Log (TSAIL).
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In discussions with the Shift Supervisor it was determined that
the Security Duty Team Supervisor had not been informed that the
SSF was in a degraded condition as required by enclosure 4.7 of
OP/0/8/6100/13, Standby Shutdown Facility Operations.
This resulted
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in a failure to implement applicable portions of the station security plan.
Discussions with licensee personnel indicate that the TSAIL i
entry may have been made conservatively when in fact the standby i
makeup pump was operable and the SSF was not degraded.
This is j
identified as Unresolved item 413/87-10-02: Potential Failure to
Implement Procedures for a Degraded SSF, pending conclusion of j
licensee review of this problem.
f h.
The inspectors participated in a special management meeting held in j
Charlotte, N.C. licensee corporate office on April 23, 1987.
The meeting attendees included Mr. M. L. Ernst, NRC:RII Deputy Admini-l strator; K. Jabbour, NRC:NRR Project Manager; L. A. Reyes, NRC:RII
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Director of Division of Reactor Projects; T. A. Peebles, NRC:RII
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Section Chief Division of Reactor Projects; Mr. H. B. Tucker, Duke
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Power Company Vice President Nuclear Production Department; l
J. W. Hampton, Catawba Station Manager and selected members of the licensee's corporate and site staff.
The Itcensee decision to conduct the 100*.' loss of electrical load test for Unit 2 on March 31,
1 1987, while the unit was experiencing unidentified leakage between 1
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and 2 gpm and performance indicators for Catawba were discussed at l
this meeting.
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1.
The inspector observed a practice emergency drill on April 24, 1987, j
and attended the licensee critique.
The licensee appropriately recognized areas needing improvement.
The overall drill was much
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i improved relative to the drill held on February 19, 1987 (See IE Inspection Report No. 50-413/87-07 and 50-414/87-07).
No violations or deviations were identified.
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6.
Surveillance Observation (Units 1 & 2) (61726)
a.
During the inspection period, the inspector verified plant operations l
were in compliance with various TS requirements.
Typical of these
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requirements were confirmation of compliance with the TS for reactor
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coolant chemistry, refueling water tank, emergency power systems, safety injection, emergency safeguards systems, control room ventilation, and direct current electrical power sources.
The inspector verified that surveillance testing was performed in accordance with the approved written procedures, test ~ instrumentation was calibrated, limiting conditions for operation were met, appropriate removal and restoration of the affected equipment was accomplished, test results met requirements and were reviewed by personnel other than the individual directing the test, and that any deficiencies identified during the testing were properly reviewed and resolved by appropriate management personnel, b.
The following surveillance activities were observed:
PT/1/A/4200/05A NI Pump 1A Performance Test PT/1/A/4450/05 Containment Air Return Fans Quarterly Performance Test (partial)
No violations or deviations-were identified.
7.
Maintenance Observations (Units 1 & 2) (62703)
a.
Station maintenance activities of selected systems and components were observed / reviewed to ascertain that they were conducted in accordance with requirements.
The inspector verified licensee conformance to the requirements in the following areas of inspection:
the activities were accomplished using approved procedures,' and functional testing and/or calibrations were performed prior to returning components or systems to service; quality control records were maintained; activities performed were accomplished by qualified personnel; and materials used were properly certified. Work requests were reviewed to determine status of outstanding jobs and to assure that priority is assigned to safety-related equipment maintenance which may effect system performance.
b.
The following maintenance activity was observed:
2996 SWR Diesel Generator Air Start Compressor 182 Annual Preventive Maintenance While observing this activity, the inspector noticed a technician verifying the safety tag on the compressor power supply after maintenance had begun. The circuit breaker prior to authorization to begin work had been properly locked opened and verified by the operations group, however, an additional verification was required by the work supervisor in accoordance with administrative procedure.. - - -
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This particular work request was being performed along with others
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.under a block tagout.
There does not appear to be a definitive
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method for maintenance technicians to know if the safe working -
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J condition verification performed by the work supervisor has been performed -when using block tagouts.
This guidance needs to be
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clarified.
This was discussed with the licensee and will be l
identified as Inspector Followup Item 413/87-10-04:
Safety Tag i
Verification by Work Supervisor for. Block Tagouts, pending licensee clarification of this process.
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c.
On March 30, 1987, the inspectors became aware that' two normal hot
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leg suction valves for the Residual Heat Removal (ND) System were inoperable. Discussions with the licensee indicated that one valve
had failed to open due to an electrical problem with an interlock-at the valve providing suction from the Refueling Water Storage Tank.
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This problem was easily corrected and the valve was opened allowing an operable condition to exist for one train of the ND system. The d
other valve was stuck and later had to be manually opened, then
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cycled to verify operability.
Discussions with the licensee
indicated that these valves had stuck before.
Further review is
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necessary to verify whether adequate maintenance has been performed on these valves to assure operation.
i Although the suction path provided through these valves is not used
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for accident scenarios it is the normal path for ND cooldown and the valves are safety related. The valves.in question are ND18, 2A, 368,
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j and 37A. This is Inspector Followup Item 414/87-10-03: Review of ND J
Hot Leg Suction Valves Maintenance.
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j No violations or deviations were identified.
8.
Review of Licensee Nonroutine Event Reports (Units 1 & 2) (92700)
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The below listed Licensee Event Reports (LER) were reviewed to determine if the information provided met NRC requirements..The determination included: adequacy of description, verification of
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compliance with Technical Specifications and regulatory requirements,
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corrective action taken, existence of potential generic problems,
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i reporting requirements satisfied, and the relative safety significance of each event.
Additional inplant reviews -and
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discussion with plant personnel, as. appropriate, were conducted for i;
those reports indicated by an (*).
The following LERs are closed:
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- LER 413/86-56 Rv.1 Containment Purge System Isolation Due to j
Defective Procedure
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- LER 413/87-06 Rv.1 Reactor Trip Due to Defective Procedure and l
Training Deficiency
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i LER 413/87-07 Swapover of Train B Nuclear Service Water
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System to Assured Source Due to Unknown l
Cause
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LER 413/87-08 Nuclear Service Water System Manway Hatch
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Covers Missing Due to Unknown Cause
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LER 413/87-09 Rv.1 Fire Watch Not Established Due to Personnel Error
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- LER 413/87-10 Operability of Available Power Sources Not-Verified Due to Management Deficiency (LIV
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413/87-10-03)
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- LER 413/87-11 Rv.1 Unit Shutdown Due to Diesel Generator Inoperability Because of Manufacturing Deficiency
LER 413/87-13 Manual Reactor Trip After a Main Feedwater Control Valve Failed Closed Due to Unknown Cause
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- LER 414/86-08 Termination of Containment Air Release Due to a Spurious Radiation Alarm Signal
- LER 414/86-17 Degraded Auxiliary Feedwater Flow Due to
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- LER 414/86-29 Rv.1 Feedwater Isolation and Auxiliary Feedwater Failure - Due to Equipment and Personnel Failures
- LER 414/87-02 Rx Trip Due to Accidental Shorting of Test
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Cable and Auxiliary Feedwater Start Due to
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Personnel Error f
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- LER 414/87-04 Technical Specifications Surveillances j
Missed Due to Personnel Error
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- LER 414/87-07 Rx Trip Due to a Management Deficiency and i
Equipment Failures (Note: Generic corrective
.l actions relative to errors on a specific shift are being addressed via UNR
413,414/87-05-01)
LER 414/87-08 Technical Specification Violation Due to A i
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Design Deficiency In the Operator Aid
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Computer Reactor Coolant Leakage Program
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(LIV 414/87-10-04)
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b.
On March 2, 1987, Diesel Generator 1B was declared inoperable at 1030 due to loss of shutdown air pressure.
Technical Specification 3.8.1.1 requires that operability of A.C. offsite power sources be demonstrated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter. The Shift Supervisor failed to ensure initial performance of the alternate power source operability check resulting in the check not being performed for a period of approximately eight and one half hours. At 1906 the operability check was oerformed and it was verified that
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A.C. power sources had been available the entire time.
This event was identified by-the licensee and described in Licensee Event Report (LER) 413/87-10. This is identified as Licensee Identified Violation (LIV) 413/87-10-03 Operability of Available Power Sources Not Verified Due to Management Deficiency.
This item is closed since followup inspection has been completed via review of the LER.
c.
On March 3, 1987, an error was discovered in the Operator Aid Computer (OAC) program that calculates Reactor Coolant System leakage for Unit 2.
Technical Specification (TS) 3.4.6.2 limits unidentified leakage to 1 gallon per minute and (TS) 4.4.6.2.1.d requires leakage to be calculated at least once every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
The program error occurred when a revision was implemented into a copy of the leakage program which contained an error.
The error was not related to the revision and was not on the version of the leakage program in use at the time. The copy, which contained the revision and the error, was loaded into the OAC on June 18, 1986, and remained until March 3, 1987, when it was corrected.
The error, resulted in unidentified leakage to be calculated in a non-conservative direction.
Of the calculations performed during the questionable time period, 19 of 79 potentially exceeded the 1.0 gpm limit with the largest leakage rate corrected to 2.9 gpm.
This incident was identified by the licensee and is described in Licensee Event Report (LER) 414/87-08.
This is identified as Licensee Identified Violation (LIV) 414/87-10-04 Technical Specification Violation Due to a Programming Error in the
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Operator Aid Computer Reactor Coolant Leakage Program. This item is
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closed since followup inspection has been completed via review of the LER.
Two (2) Licensee Identified Violations (LIV) are documented in paragraphs 8b and 8c.
9.
Previously Identified Inspector Findings (92701)
a.
(CLOSED)
Inspector Followup Item 413/85-12-05:
Develop Systematic Approach to Correct Numerous Leaking Valves.
The licensee has developed a formal review of catch containment status and has significantly reduced the number of leaking valves. Therefore, this item is closed.
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b.
(CLOSED)
Inspector Followup Item 413/85-30-01: Revise Station Directive 4.4.4 to Include Details of Implementation Plans for NSM's.
The above Station Directive has been changed and, therefore, this item is closed.
c.
(CLOSED) Inspector Followup Item 413/86-05-03, 414/86-07-02: Various Maintenance Procedure Deficiencies.
This followup item identified several examples of maintenance procedures that an inspector considered to have weaknesses and that the licensee agreed that changes would be made to correct the potential problem areas.
A review of each of the procedures that were identified in this item has been conducted and the weaknesses previously found have been corrected. Based on this review, this item is closed.
d.
(CLOSED)
Inspector Followup Item 413/86-30-01, 414/86-33-01:
Followup of Clarification of Nuclear Service Water System Procedure.
The inspector reviewed change 47 to OP/0/A/6400/06C, Nuclear Service Water System.
Initial concerns, regarding the assurance that all operating scenarios had been covered in the procedure, have been adequately addressed, however, additional concerns with the operation of the system have arisen.
The Nuclear Service Water System (RN)
consists of two, train specific pumps per unit for a total of four pumps. Under normal operating conditions only one of the four pumps is required to supply loads on both units and does so through common discharge piping.
Technical Specification (TS) 3.7.4 requires two independent RN loops per unit to be operable while the applicable unit is in Modes 1 through 4.
Licensee Event Report 414/86-31 Rev.1 of August 29, 1986, identified that the RN system could not be considered two independent loops when a diesel generator or a Nuclear Service Water pump from either unit became inoperable.
This is because assuming a LOCA and a station blackout, three RN pumps would theoretically be available to effectively supply 4 RN trains, however a single failure would leave only two RN pumps.
Duke design engineering has shown this condition would result in inadequate flow to heat exchangers. OP/0/A/6400/06C, Nuclear Service Water System, was revised to include special RN system lineups when one or more RN pumps or diesel generators were inoperable for greater tFan 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
These special alignments would ensure adequate RN flow to heat exchangers on the operating unit.
The inspector identified areas where clarification was required and documented concerns under IFI 413/86-30-01, 414/86-33-01. Upon review of various procedure changes up to change 47, it was noted that under certain conditions RN would be isolated to a Component Cooling Heat Exchanger on the unit shutting down, to ensure adequate RN flow for the opposite unit, thus allowing continued operation. Although not specifically addressed by Technical Specifications, isolation of a Component Cooling Heat Exchanger for the sole purpose of keeping an opposite unit operating does not appear to be prudent as it removes a redundant safety system.
Additional concerns on the use of human valve operators under various conditions remain open.
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On January 21, 1987, Station Problem Report #CNPR-2530 was generated i
identifying a postulated different single failure of the RN system and requested evaluation from design engineering.
Part 1 of Design Study (CNDS-080/00) of March 27, 1987, concluded that an unanalyzed condition exists and needs to be addressed. The scenario involves a diesel generator or RN pump on either unit inoperable for greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, requiring shutdown of that unit.
Under LOCA conditions of the operating unit and station blackout RN suction and discharge would automatically shift from Lake Wylie to the Standby Nuclear Service Water Pond (SNSWP), the seismically qualified ultimate heat sink. A single failure of a SNSWP supply valve would block cooling water to two RN pumps. Since one RN pump was already inoperable, the fourth RN pump would alone be left to supply cooling to both units.
With one unit in a LOCA cond', tion and one unit i
shutting down, one RN pump is inadequate to cool all heat exchangers This ' ould result including remaining operating diesel generators.
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in eventual loss of all diesel generators, AC power and cooling water for both units.
Various solutions have been proposed including: installation of redundant parallel SNSWP supply valves or elimination of the requirement to swap RN Suetions to the SNSWP on a LOCA. The licensee has stated that should a diesel generator or RN pump become inoperable on one unit for greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, manually shifting
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RN suction and discharge to the SNSWP would eliminate the single failure vulnerability and justify continued operation of the other unit.
It was noted that there are no provisions to monitor SNSWP
temperature during all months of the year.
The licensee is i
continuing part 2 of the Design Study which includes a Probabilistic Risk Assessment of the problem.
The licensee has generated Problem j
Investigation Report (PIR) 0-C87-0097 to determine if this unanalyzed
condition is reportable.
This is identified as Unresolved Item a
i 413/87-10-01, 414/87-10-01: Single Failure Vulnerability of the Nuclear Service Water System pending licensee resolution and further
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j review by the NRC.
e.
(CLOSED) Inspector Followup Item 413/86-31-01, 414/86-34-01: Lack of
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Formal Notification of Removal of Licensed Personnel From Shift That Fail Annual Requalification Examination. The training department now I
sends written notice to the Operations Engineer of any licensed operator that fails any portion of an annual requalification examination. Operations maintains a file of this notification along
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with information as to the action taken as a result of the failure, i
Based on this review, this item is closed, i
No violations or deviations were identified, i
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10.
FollowupofIEBulletins(92701)
(CLOSED)
IEB 84-03 (Unit 2): Refueling Cavity Water Seal. The licensee responded to this bulletin in correspondence dated November 21, 1984; December 31, 1984; January 18, 1985; January 29, 1985; February 18, 1985; March 22,1985; and November 12, 1985.
Regional and resident inspector review showed the licensee response and actions to be adequate and, therefore, this bulletin is closed.
No violations or deviations were identified.
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