IR 05000315/1990013
| ML17334B371 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 08/01/1990 |
| From: | Burgess B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17334B369 | List: |
| References | |
| 50-315-90-13, 50-316-90-13, IEB-88-011, IEB-88-11, NUDOCS 9008140476 | |
| Download: ML17334B371 (21) | |
Text
U.
S.
NUCLEAR REGULATORY COMMISSION
REGION III
Repor t Nos.
50-315/90013(DRP);
50-316/90013(DRP)
Docket Nos.
50-315; 50-316 License Nos.
American Electric Power Service Corporation Indiana Michigan Power Company 1 Riverside Plaza Columbus, OH 43216 Facility Name:
Donald C.
Cook Nuclear Power Plant, Units 1 and
'I Inspection At:
Donald C.
Cook Site, Bridgman, Michigan Inspection Conducted:
June 6 thr'ough July 17, 1990 Inspectors:
B. L. Jorgensen D.
G.
Passehl E.
R. Schweibinz Approved By:
B.
L. Burge
, Chief Projects Section 2A E/
Da e
Ins ection Summar Ins ection on June 6 throu h Jul
1990 Re ort Nos.
50-315/90013 DRP; 50-316/90013 DRP of:
plant operations; one reactor trip; radiological controls; maintenance; surveillance; engineering and technical support; emergency preparedness; security; safety assessment and quality verification; and Bulletins.
Also, a
routine periodic Management Meeting was held in the NRC Region III offices on June 7, 1990.
No Safety Issues Management System (SIMS) items were closed.
Results:
Of the ten areas inspected, no violations or deviations were identified in any areas.
The 'inspection disclosed no notable new strengths nor weaknesses in the licensee's programs and activities inspecte DETAILS 1.
Persons Contacted a.
Ins ection June 6 throu h Jul
1990 A. Blind, Plant Manager
- J. Rutkowski, Assistant Plant Manager Technical Support L. Gibson, Assistant Plant Manager Projects
"K. Baker, Assistant Plant Manager - Production
'B. Svensson, Executive Staff Assistant
"J.
Sampson, Operations Superintendent P. Carteaux, Safety and Assessment Superintendent T. Bei lman, Maintenance Superintendent
- J. Droste, Technical Superintendent
- Engineering
"T. Postlewait, Design Changes Superintendent L. Matthias, Administrative Superintendent
- J. Wojcik, Technical Superintendent
- Physical Sciences M. Horvath, guality Assurance Supervisor D. Loope, Radiation Protection Supervisor The inspector also contacted a
number of other licensee and contract employees and informally interviewed operations, maintenance, and technical personnel.
"Denotes some of the personnel attending the Management Interview on July 18, 1990.
b.
Mana ement Meetin June
1990 Licensee Personnel D. Williams, Senior Executive Vice President, AEP M. Alexich, Vice President - Nuclear, AEP A. Blind, Plant Manager J.
Rutkowski, Assistant Plant Manager Technical Support J. Droste, Technical Superintendent
- Engineering T. Bei lman, Maintenance Superintendent M. Evarts, Chief, Nuclear Maintenance Support Section, AEP T. Roxey, Nuclear Operations Support, AEP NRC Personnel C. Paperiello, Deputy Administrator, Region III J. Zwolinski, Assistant Director, NRR Division of Projects III, IV and V.
H. Miller, Director, Division of Reactor-Safety, Region III W. Forney, Deputy Director, Division of Reactor Projects, Region III B. Clayton, Chief, Division of Reactor Projects, Branch
B. Burgess, Chief, Reactor Projects, Section 2A B. Jorgensen, Senior Resident Inspector E. Schweibinz, Senior Project Engineer
2.
0 erational Safet Verification 71707 71710 42700 Routine facility operating activities were observed as conducted in the plant and from the main control rooms.
Plant startup, steady power operation, plant shutdown, and system(s)
lineup and operation were observed as applicable.
The performance of licensed Reactor, Operators and Senior Reactor Operators, of Shift Technical Advisors, and of auxiliary equipment operators was observed and evaluated including procedure use and adherence, records and logs, communications, shift/duty turnover, and the degree of professionalism of control room activities.
The Plant Manager, Assistant Plant Manager-Production, and the Operations Superintendent.
were well-informed on the overall status of the plant, made frequent visits to the control rooms, and regularly toured the plant.
Evaluation, corrective action, and response to off-normal conditions or events, if any, were examined.
This included compliance with any reporting requirements.
Observations of the control room monitors, indicators, and recorders were made to verify the operability of emergency systems, radiation monitoring systems and nuclear reactor protection systems, as applicable.
Reviews of surveillance, equipment condition, and tagout logs were conducted.
Proper return to service of selected components was verified.
a.
Unit 1 operated routinely throughout the inspection period.
Unit 2 operated at normal routine power except for a period on June 11-14, 1990 following a reactor trip (See Paragraph 3, below), until shutdown commenced on June 29,
)990 for cycle 8 refueling.
The outage is expected to last until about the middle of, September.
b; On June 18, 1990, a leak detection alarm was received in the Unit 2 control room while a monthly incore flux map surveillance was in progress.
Just prior to the alarm, heavy resistance was encountered while trying to insert a flux thimble into core location C-7.
Upon withdrawal, indications of thimble tube leakage became evident in the form of rising particulate radiation (about one-half decade per minute)
and a slight containment dewpoint increase.
The leak alarm was also received.
Reactor coolant leakrate showed a step increase of approximately one gallon per minute.
The licensee entered a four-hour Technical Specification Action Statement (No. 3.4.6.2)
based on unidentified leakage greater than one gallon per minute.
Shortly thereafter, a team of operators and radiation protection personnel entered containment at the location of the suspected leak (where radiation levels measured about ten-millirem per hour) and successfully identified and isolated the leak by closing the associated isolation valv The licensee continued to monitor containment conditions, and parameters gradually returned to normal.
A consultation with the vendor is planned in addition to an inspection of all thimble tubes during the Unit's refueling outage (See Paragraph 6.a).
c.
On July 3, 1990, a non-licensed auxiliary equipment operator was found asleep at his station, in the 2AB emergency diesel room, by the Operations Superintendent.
The operator was relieved of duty and suspended pending a review of the situation.
The sleeping individual had been assigned to the diesel room to take hourly performance readings on the diesel, which was in the midst of a 24-hour surveillance run.
No data significant to the test was omitted.
A roving firewatch was also assigned to the area, and indicated the subject individual had been awake a few minutes earlier.
d.
Routine inspector review of licensee corrective action program documents disclosed several which appeared to involve errors or weaknesses in equipment status control:
(1)
Problem Report PR 90-0577:
the power supply breaker (No.
1-IMO-120) for a Unit 1 accumulator outlet valve was found energized, contrary to requirements, with reactor'oolant system pressure above 2,000 psi; (2)
Problem Report PR 90-0631:
the boric acid inlet manual valve (No. 2-CS-425S) to the blender was found closed, contrary to the valve lineup sheet requirements; (3)
Problem Report PR 90-0660:
the "A" Train SSPS was placed in
"Test,"
and thus not fully operable, while the "B" Train motor-driven auxiliary feedwater pump was concurrently inoperable (the test line valve was open) contrary to licensee administrative control requirements; and, (4)
Problem Report PR-90-0759:
the East turbine auxiliary cooling water pump breaker was racked to the "test" versus the
"disconnect" position, contrary to the requirements of the effective testing lineup.
Although each condition was found and corrected safely within established time limits (where applicable),
the occurrence of several such problems in a relatively short time was discussed at the Management Interview.
No violations, deviations, unresolved or open items were identified.
The Unit 2 reactor tripped from about 85"percent power on June 11, 1990, with the "first out" indicating a negative rate trip.
Ten rods apparently dropped on a momentary loss of power supply 2AC; post-trip
testing found one phase of 2AC electrically "open."
A momentary interruption on a
second phase, or any rod motion with one phase lost, would have the same effect.
The associated SCRs (silicon-controlled rectifiers) could not "hold" under those conditions.
Both these possibilities, and other questions, were investigated.
All safety-related equipment responded per design following the trip, One rod bottom light failed to indicate, but other indicators, confirmed by direct coil stack voltage readings, verified the rod had fully inserted.
A failed rod bottom bistable module was identified and replaced.
The. precise root cause of the negative flux rate reactor trip could not be absolutely determined.
False signals originating in either the excore nuclear instruments or in the SSPS logics, however, were essentially ruled out.
This left a genuine negative rate signal, probably the result of dropped control rods, as the presumed cause.
The single phase of the power supply to one group of rods that was found electrically "open" (in electrical continuity/resistance checks)
shortly after the event was confirmed as the result of bad contact at a fuse clip and knife switch in the circuit.
This could have contributed to the dropping of ten rods, but could not have been sole contributor.
The fuse clip was repaired and the switch exercised.
Detailed electrical checks were performed in various configurations on all the rod controls and power supplies and no discrepancies were found.
The most likely scenario involved a second phases in the power supply to the "suspect" ten rods being momentarily
.interrupted as a consequence of being bumped by a worker who was on top of the power supply cabinet performing a ceiling fire detector test.
The reactor was again taken critical on June 14, 1990.
No violations, deviations, unresolved or open items were identified.
Radiolo ical Controls 71707 During routine tours of radiologically controlled plant facilities or areas, the inspector observed occupational radiation safety practices by the radiation protection staff and other workers.
Effluent releases were routinely checked, including examination of on-line recorder traces and proper operation of automatic monitoring equipment.
Independent surveys were performed in various radiologically controlled areas.
No violations, deviations, unresolved or open items were identifie.
Maintenance 62700 62703 42700 Maintenance activities in the plant were routinely inspected, including both corrective maintenance (repairs)
and preventive maintenance.
Mechanical, electrical, and instrument and control group maintenance activities were included as available.
The focus of the inspection was to assure the maintenance activities reviewed were conducted in accordance with approved procedures, regulatory guides and industry codes or standards and in conformance with Technical Specifications.
The following items were considered during this review: the Limiting Conditions for Operation were met while components or systems were removed from service; approvals were obtained prior to initiating the work; activities were accomplished using approved procedures; and post maintenance testing was performed as applicable.
The following activities were inspected:
Job Order A050964:
"Replace existing Ingersol-Rand pump with new Worthington pump."
The replacement applied to the Unit 1 North miscellaneous sealing and cooling water pump.
Upon review of the job order work package the licensee found that a weld was performed without proper welder qualifications.
The weld in question was on a
one-inch branch connection off the three-inch suction pipe, and the welder was not qualified to one-inch pipe.
The welder has since passed a performance test to qualify down to one-inch pipe.
The cause of the problem (documented in Problem Report 90-0687)
was attributed to the Supervisor's failure to ensure that his welder held proper qualifications for the weldment which was to be performed.
It was also recognized that it is in each welders interest to know his/her own current qualification status.
b.
Job Order B018055:
"Investigate and Repair Damaged Fan 2-HV-SGRS-1A (Unit 2 Control Room instrumentation distribution supply).
The fan was damaged in January 1990, when pieces of a stationary turning vane were dislodged, due to metal fatigue, and impacted the operating fan blades (ref.
NRC Inspection Report 50-315/90009(DRP);
59-316/90009(DRP)
Paragraph 9.b).
The repair-plan included fabrication of a new elbow section with heavier metal (16 vs.
and removal of loose parts thrown downstream into the duct after the failure occurred.
C.
- "12 MHP 5021.082.001 (Rev. 7): "Inspection and Repair of 4KV Circuit Breakers."
Inspection and lubrication of prescribed parts of circuit breaker 2D5 (Unit 2 normal incoming feed from transformer TR2CD) was performed in conjunction with Job Order B000628 and the above procedure.
The licensee has increased his emphasis on use of procedures for maintenance, focusing both on workers obtaining and using the correct procedure and on literal compliance with procedure detail Evidence of this emphasis was noted in review of corrective action documents relating to the maintenance area:
(1)
Problem Report PR 90-0758:
valve 2-TBP-105-30 (steam dump valve URV-130 bypass strainer inlet shutoff valve) was disassembled and repaired without an approved procedure; (2)
Problem Report PR 90-0760:
procedure
"*2 IHP 4030 STP.168 was being performed with a data sheet tolerance of 25, when the correct tolerance was 0.25 (the erroneous instruction was not complied with, nor was it corrected);
(3)
Problem Report PR 90-0780:
valves 2-CS-130 (a deborating demineralizer outlet check valve)
and 2-B-175 (a blowdown steam shutoff valve) were repacked using a superseded (not current revision)'rocedure.
There were numerous lesser examples.
Also, there were more numerous instances of procedure revisions for clarity, to correct typos, etc., to ensure they could be complied with literally.
The general concepts, management monitoring, and status of maintenance procedure
"quality" were discussed at the Management Interview.
No violations, deviations, unresolved or open items were identified.
6.
Surveillance 61726 42700 The inspector reviewed Technical Specifications required surveillance testing as described below and verified that testing was performed in accordance with adequate procedures, that test instrumentation was calibrated, that Limiting Conditions for Operation were met, that removal and restoration of the affected components were properly accomplished, that test results conformed with Technical Specifications and procedure requirements and were reviewed by personnel other than the individual directing'the test, and that'eficiencies identified during the testing were properly reviewed and resolved by appropriate management personnel.
The following activities were inspected:
C a.
THP 6040 PER.323,
"Flux Mapping System Operation and Supportive Data Collection."
The licensee was performing this procedure at the time the leak in the flux thimble tube occurred (Paragraph 2.b).
Performance of the procedure was terminated at the onset of the leak, when data was collected on 37 of the 44 thimbles required to be OPERABLE per Technical Specifications (No. 3.3.3.2).
At issue was whether the data retrieved satisfied the surveillance requirements-in the Technical Specifications'
0,
Discussions with Cook Nuclear Plant Reactor Engineering confirmed an incore flux map was obtained on May 19, 1990, and that this flux map satisfied the criterion contained in the Technical Specification
"Moveable Incore Detectors (No. 3.3.3.2)."
The results wele expected to be-valid for most surveillances through the beginning of the Unit 2 refueling outage.
However, the results of a surveillance requirement to compare incore to excore axial offset appeared to "expire" a few days prior to the outage.
This would have required the licensee to investigate other options to fulfillthe requirement.
The licensee prepared and supplied a technical justification to show the partial core flux map was sufficient to meet the Technical Specification requirement to measure.,and compare axial offset.
The information associated with the justification was forwarded through the NRC for evaluation.
- "12 MHP 4030 STP.048,
"Main Steam Valve Setpoint Verification Using Trevitest."
Trevi-testing of the main steam safety valves on Unit 2 was performed by Furmanite representatives on June 25-28, 1990 and supervised by site engineering personnel.
Twelve of the twenty valves tested lifted within the specified pressure range on the first test and repeated the lift successfully.
The remaining eight valves all lifted at higher than specified pressures on their initial test.
Seven of these eight returned to within specification without adjustment on subsequent tests.
Only one valve required adjustment.
After adjustment it returned to within specification when tested and exhibited the required repeatability.
Deviations for the eight "failures" ranged from 2 psig to 24 psig.
The licensee's corporate and site engineering groups are working towards determining a root cause for the setpoint drift (ref.
Problem Report 89-1359).
Howev'er, based upon review of previous test data it appears the cause of the setpoint drift may be inherent in the valve design for the licensee's applications A change to the Technical Specification from plus/minus 1-percent tolerance to plus/minus 3-percent tolerance has been recommended by the licensee's corporate engineering division.
This is consistent with ASME Section XI, 1986 and later editions.
At present D.
C.
Cook safety valves are tested in accordance with ASME Section NI 1983 Edition, S-83 Addenda.
"*2 OHP 4030 STP.018,
"Steam Generator Stop Valve Dump Valve Surveillance Test."
During the surveillance, valve 2-MRV-232 (Train B dump valve for No.
3 Steam Generator) failed its monthly inservice test (IST) with an initial stroke time of 8.89 seconds and a
secondary stroke time of 6;49 seconds (IST criteria is 2.00 seconds).
Prior to the second test maintenance electricians checked the solenoid actuator and found no problems.
After the second test, the valve was lubricated and retested without improvemen The stroke seemed to occur in two steps - the first 11/16 inch of stem travel happened smoothly within the required time, but the last 1/16 inch was delayed.
The failure of the valve to stroke properly was peculiar in that the valve was recently rebuilt (May 1990) with post-maintenance testing indicating three stroke times under 1.5 seconds.
Failure of the control valve to stroke properly placed the licensee in a four-hour Limiting Condition for Operation (LCO) for the associated steam generator stop valve (Technical Specification 3.7. 1.5).
The failure was documented in Problem Report 90-0681.
In response to this, the licensee examined this particular control valve's flow characteristic curve, among other things, to evaluate the relationship between percentage of maximum flow capacity and percentage of total travel range.
The contour of the valve plug surface provides for a "quick-opening" flow characteristic, used where significant flow rate must be established quickly as the valve begins to open.
The curve is basically linear through the first 40-percent of valve plug travel, then flattens
. out noticeably to indicate little increase in flow rate as travel approaches the wide open position.
Hence, the licensee concluded the opening function of the valve had been met without the last 1/16 inch of travel.
To correct the discrepancy, the licensee adjusted the limit switch to indicate open at the end of the 11/16 inch of travel based on the foregoing conclusion.
The valve is scheduled to be repaired during the current refueling outage.
- "2 THP 4030 STP.217B,
"Unit 2 Diesel Generator AB Load Shedding and Performance."
The "Safety Injection With Blackout" portion of this multifaceted procedure was observed from the Unit 2 control room.
Applicable requirements of the procedure were met including diesel generator starting and load sequencing.
Plant conditions precluded testing of certain components, which were scheduled for testing later during the refueling outage.
""2 IHP 4030 STP. 129,
"Power Range Nuclear Instrument Protection Set IV (N-43) Surveillance Test."
The surveillance was performed prior to main steam safety valve lift setpoint tests in conjunction with Job Order B50204, which documented adjustment of the power range neutron flux reactor setpoint to less than or equal to 87.2 percent power in accordance with Technical Specifications (No. 3.7. 1. 1).
OHP 4030 STP.024E,
"East Control Room Pressurization/Cleanup Filter System Operability Test."
Data sheets for tests performed on May 12 and June 9,
1990 were reviewed with no problems noted.
"*2 OHP 4030 STP.020W, West Component Cooling Water Loop Surveillance Test."
Data from surveillance tests performed on May 14 and June 12, 1990 were reviewed and no problems were note '
h.
Inspector review of the licensee's corrective action program documentation identified three items involving the apparent failure'o complete specified surveillance.
The items had in common the fact that each was a "special" surveillance whose performance is conditional - they're required only if certain equipment is inoperable:
( 1)
Problem Report PR 90-0704:
with the Unit 2 "CD" emergency diesel generator inoperable, the continued operability of the alternate Train was not verified every eight hours, but (in one case)
was verified some 14 minutes late; (2)
Problem Report PR 90-0733:
with the 12CD reserve feed supply inoperable, the operability of the remaining offsite A.C. power supply was not verified within 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> s, but was verified some
hours, 23 minutes late; and, (3)
Problem Report PR 90-0748:
with the Unit 2 Auto Gas Analyzer inoperable, and the Unit degassing,
"grab" samples were not taken at the specified 8-hour interval on four occasions.
These matters may be reviewed further on receipt of anticipated Licensee Event Reports.
The implications of these events as related to possible lapses in equipment status control knowledge (see also Paragraph 2.d above).were discussed at the Management Interview.
No'iolations, deviations, unresolved or open items were identified.
7.
En ineerin and Technical Su ort The inspector monitored engineering and technical support'ctivities at the site and, on occasion, as provided to the site from the corporate office.
The purpose of this monitoring was to assess the adequacy of these functions in contributing properly to other functions such as operations, maintenance, testing, training, fire protection and configuration management.
The following Condition and Problem Reports were noted:
a.
Problem Report PR 90-0766:
"Requirements of PMSO-091 ("Scaffolding Erection" ) were not observed prior to installing scaffolding inside containment in MODE 5 during the Unit 2 'refueling outage."
Clarification was needed as to how extensive scaffolding could be without compromising adequate equipment protection during a seismic event.
The problem was reviewed by the licensee's Nuclear Safety and Licensing section and D.
C.
Cook plant management.
Condition Report CR 2-07-90-1028:
"Unit 2 Reactor Coolant Pump No.
3 area ventilation exhaust ductwork had cracks."
The report was forwarded to the licensee's corporate Problem Assessment Group, along with other similar reports on newly-discovered ductwork cracking, for generic followup review.
c
~
Problem Report PR 90-0782:
"Unit 2 Train N Battery distribution cabinet was found to have the wrong fuses in circuits 7 and 8."
Applicable prints show 35 amp.
and 10 amp.
fuses for circuits 7 and 8,=respectively.
Actual installation shows circuit 7 with 70 amp.
fuses and circuit 8 with 60 amp.
fuses.
The licensee verified correct fuses in Unit 1 and'installed correct qualified fuses in Unit 2.
No violations, deviations, unresolved or open items were identified.
8.
Emer enc Pre aredness 82201 82203 At about 1:00 p.m.
(EDT) on July 13, 1990, an electrical accident occurred in D.
C.
Cook Unit 2.
One worker, a contractor, was killed.
Three others were seriously injured.
At the time of the accident, Unit 2 was shut down in MODE 5 and the residual heat removal system was in service fo'r reactor cooling.
The workers involved (two licensee and two contractor electrical workers)
were scoping electrical modification work in or around a 4160 V. breaker on plant bus T21C.
The breaker was apparently open and tagged, and the bus grounded, but power feeds from 4160 V. bus 2C were "live."
The fatally injured workman apparently contacted the "live" 4160 V.
feeder somehow, faulting the supply.
He was electrocuted, and an electrical flash/explosion ignited clothing on the other affected
~ personnel causing serious burns.
An Emergency Plan "Unusual Event" was declared on the basis of the small explosion and fire.
Offsite medical assistance was summoned and responded.
The "Unusual Event" was secured after the injured were sent by ambulance to area hospitals.
Upon removal of the injured, the area was quarantined pending establishment of safety conditions.
Licensee training of electricians following the event was reviewed and found to be adequate.
I'he licensee conducted a,press briefing for local media, and a press release was issued by the corporate office.
The licensee also made notifications to agencies of the State of Michigan as required.
No violations, deviations, unresolved or open items were identified.
9.
Securi t 71707 Routine facility security measures, including control of access for vehicles, packages and personnel, were observed.
Performance of dedicated physical security equipment was verified during inspections in various plant areas.
The activities of the professional security force in maintaining facility security protection were occasionally examined or reviewed, and interviews were occasionally conducted with security force members.
On July 6, 1990, the licensee notified the NRC of a "for-cause" Fitness For Duty alcohol screen of a Maintenance Planner.
The Planner was relieved from duty after a confirmed breath analyzer level showed him to be over the cut-off level.
The "for-cause" test was administered after the onsite security staff had received an anonymous call that the individual had the smell of alcohol on his breath onsite.
His unescorted access was suspended pending further review.
The individual was not involved in safety-related work.
No violations, deviations, unresolved or open items were identified.
Safet Assessment/
ualit Verification 37701 38702 40704 92720 The effectiveness of management controls, verification and oversight activities, in the conduct of jobs observed during this inspection, was evaluated.
The inspector frequently attended management and supervisory meetings involving plant status and those focusing on proper co-ordination among departments.
\\
The results of licensee auditing and corrective action programs were routinely monitored by attendance at Problem Assessment Group (PAG)
meetings and by review of Condition Reports, Problem Reports, Radiological Deficiency Reports, and security incident reports.
As applicable, corrective action program documents were forwarded to NRC Region III technical specialists for information and possible followup evaluation.
Some reviews, involving Condition Reports and Problem Reports, are discussed elsewhere in this inspection report.
Problem Report 90-0601 involved two "certified" bolts which failed during reuse pre-torquing.
The bolts were two of eight retaining bolts for a steel frame supporting a block wall closing a pipe tunnel.
The Quality Control (QC) group tested the failed bolts, which proved to be of the correct alloy, but found the shanks were not of the specified hardness.
A manufacturing problem was suspected.
The remaining six bolts from the same batch, which had been reinstalled, were removed, tested (they met criteria) and disposed of.
No additional bolts were in stock from the suspect batch.
The inspector determined there is apparently no procedure for controlling or dispositioning materials which fail QC hardness or spectrographic criteria; however, the QC group investigated the technical aspects of the problem in a thorough and timely manner and showed an inquisitive attitude about assuring all potentially suspect bolts were accounted for and controlled.
The inspector was kept well informed.
Quality implications, including Part 21 reportability, remain under licensee investigation.
No violations, deviations, unresolved or open items were identified.
j ll.
NRC Bulletins 92703 The inspector reviewed the NRC communications listed below and verified that: the licensee has received the correspondence; the correspondence was reviewed by appropriate management representatives; a written response was submitted if required; and, plant-specific actions were taken as described in the licensee's response.
(Open)Bulletin 88-11, "Pressurizer Surge Line Thermal Stratification."
The licensee responded to the subject Bulletin with letters dated March 6, 1989 (AEP:NRC: 1086)
and May 31, 1989 (AEP:NRC: 1036A).
The 'first letter established an alternative schedule for evaluation of the issue based on licensee participation in the generic Westinghouse Owner's Group (WOG) review.
The second letter provided a justification for Continued Operation (JCO) for D.
C.
Cook Units 1 and 2 which was based on the generic WOG review.
This was subsequently supported by Westinghouse topical report WCAP-12277 submitted in June, 1989.
On May 31, 1990, the licensee was informed by Westinghouse that plant-specific analyses had determined the WCAP-12277 allowable usage factor of
~ 0 was exceeded; the continued applicability of the JCO was therefore in question.
The licensee initiated Problem Report PR 90-617 to pursue the issue.
Westinghouse was immediately retained and provided with comparative plant specific data in some areas 'of conservatism in the generic analyses.
A plant-specific JCO was thereby derived and provided to the licensee via a Westinghouse letter dated June 28, 1990.
The JCO concludes that operation of both units remains acceptable pending bulletin-required detailed plant specific analyses by January 1991.
No violations, deviations, unresolved or open items were identified.
12.
Mana ement Meetin A management meeting, attended as indicated in Paragraph conducted in the NRC Region III offices on June 7,
1990.
the meeting was to discuss various licensee initiatives, maintenance support area.
1.b, was The purpose of primarily in the Among the topics upon which the licensee representatives made specific presentations were:
b.
C.
d.
e.f.
Formation and activities of a new Corporate Nuclear Section; General status of Maintenance Improvement Plan; System Engineer Program; Computerized Information Management System; Auxiliary Building upgrade status; and, overview of recent operations and trends.
Maintenance
'
The licensee was responsive to questions in all topic areas during and following the specific presentations.
13.
Mana ement Interview 30703 The inspectors met with licensee representatives (denoted in Paragraph l.a)
on July 18, 1990, to discuss the scope and findings of the inspection as described in these details.
In addition, the inspector also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the inspector during the inspection.
The licensee did not identify any such document/processes as proprietary.
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