IR 05000266/1993011
| ML20046A745 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 07/21/1993 |
| From: | Jackiw I NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20046A743 | List: |
| References | |
| 50-266-93-11, 50-301-93-11, NUDOCS 9307290299 | |
| Download: ML20046A745 (12) | |
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U.S. NUCLEAR REGULATORY COMMISSION
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REGION III
Report Nos. 50-266/93011(DRP); 50-301/930ll(DRP)
Docket Nos. 50-266; 50-301 License No. DPR-24; DPR-27
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Licensee: Wisconsin Electric Company 231 West Michigan Milwaukee, WI 53201
Facility Name:
Point Beach Units 1 and 2 Inspection At: Two Rivers, Wisconsin l
i Dates: May 25 through July 14, 1993 i
Inspectors:
K. R. Jury i
J. Gadzala J. Gavula I
G. O'Dwyer t
Approved By:
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[' 2/ ~ [l I, N. Jackiw, Chief Date Reactor rbjects Section 3A Inspection Summary
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i Inspection from May 25 throuah July 14. 1993 (Reports No. 50-266/93011(DRP): No. 50-301/93011(DRP)
Areas Inspected:
Routine., unannounced inspection by resident ir.spectors of corrective actions on previous findings; plant operations; radiological
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controls; maintenance and surveillance; emergency preparedness; security;
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engineering and technical support; and safety assessment / quality verification.
Results: No violations of NRC requirements and one inspector follow up item
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was identified. An Executive Summary follows.
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Plant Operations
i Unit 1 Power was reduced to 53% on June 12 to remove the IP-25B condensate pump from service for inspection and repair of a flange leak.
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9307290299 930712 E
PDR ADOCK 05000266
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Maintenance / Surveillance
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Only one of two isolation valves were red tagged shut during replacement of an
auxiliary feedwater steam trap.
Simultaneous implementation of a newly written procedure and.two supporting procedures for pump seal replacement appeared to be cumbersome and confusing During control rod exercise test s o., May 28 and June 25, control rod banks did
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not move upon demand.
Operators verified that the cause of the fault for both l
events was electrical and that rods were not physically bound.
The two faults were caused by unrelated failures of different ci cuit cards in the' respective
units' rod control system.
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Emeroency Preparedness
i Point Beach performed an emergency plan drill on June 24 involving declaration of an emergency classification level, activation of the technical support
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center and operations support center, and plant accountability exercising.
l Overall drill performance was good. Minor weaknesses were noted in
communications and coordination among the control room, technical support center and operations support center.
Engineering and Technical Support
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Ground breaking and concrete pours for the new emergency diesel generator i
building began the weeks of June 7 and 30, respectively.
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DETAILS-
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1.
Persons' Contacted '(71707) I(30702)J
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- G, J. Maxfield, Plant Manager l
T. J. Koehler, Site Engineering Manager
R. D. Seizert, Training Manager
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- J. F. Becka, Regulatory Services Manager J. G. Schweitzer,. Maintenance Manager J. C. Reisenbuechler, Manager - Operations N. L. Hoefert, Manager - Production Planning i
J. J. Bevelacqua, Manager - Health Physics
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F. P. Hennessy, Manager - Chemistry a
J. A. Palmer, Manager - Maintenance
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G. R. Sherwood, Manager.- Instrument & Controls l
W. B. Fromm, Sr Project Engineer - Plant Engineering i
T. G. Staskal, Sr. Project Engineer - Performance Engineering
W. J. Herrman, Sr. Project Engineer - Construction Engineering
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- F. A. Flentje, Administrative Specialist l
- R. C. Hetue, Quality Specialist a
t Other company employees were also contacted including members of the
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technical and engineering staffs, and reactor and auxiliary operators.
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- Denotes the personnel attending the management exit interview for i
j summation of preliminary findings.
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2.
Corrective Action on Previous Inspection Findinos (92701) f(92702)1 a.
IClosed) Inspection Follow Up item (266/92007-02):
Technical Specification Upgrade Project-
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i An FSAR review performed by the licensee identified that several equipment items required by the plant's accident analysis, did not contain appropriate limiting conditions for operation nor surveillance requirements in the plant technical specifications.
As a result, the licensee committed to upgrade their technical specifications. During the period that changes to TS were being proposed, the licensee committed to. inform NRC Region 111 of.any safety related equipment that was removed from service, for which there was no associated limiting condition for operation.
Technical specification change requests 154, 156 and 157 have since been submitted to address this weakness.
These change requests were being reviewed by the NRC and amendments are to be issued as appropriate. Additionally, a new procedure, QP 6-9,
" Technical Specification and Bases Change Preparation, Review and Approval", was developed to control the process by which changes to the' technical specifications are identified and evaluated.
This item'is closed.
Closure of this item also concludes the
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L licensee's special regional reporting requirements.
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(Closed) Inspection Follow Up Item (266/92027-0111 Potential (
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Inoperability of Safety Injection Pumps During Inservice Testing
i Safety injection pumps were found to be susceptible to damage during inservice testing if a single failure of the isolation valves in their minimum flow line occurred under certain
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conditions.
Although performance of this inservice test created i
the potential for a single failure to render a safety system
inoperable, the situation is_ analogous to that experienced during testing of any safety system when the single failure criteria is
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temporarily relaxed to permit removing a component from service.
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However, to reduce the risk of pump damage, the licensee submitted a justification to change the test frequency of these valves'from
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quarterly to cold shutdown.
This allows testing when the safety I
injection pumps may be taken out of service.
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Licensee Event Report (LER)92-010 was submitted to document this situation and propose corrective action (see paragraph B.a.).
j Inservice test procedures were subsequently revised to remove the I
i two minimum flow isolation valves from the quarterly inservice test and add them to the cold shutdown inservice test procedure.
The inspector reviewed the event report and the revised procedures and had no further concerns.
This item is closed.
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(Closed) Inspection follow Up item (301/93009-03):
Containment j
Hatch Leak Testing Methodology i
The inspector observed leak rate testing of the Unit 2 containment i
upper personnel airlock on May 13, 1993, and questioned certain aspect s of the test methodology.
The licensee's subsequent review of the test results determined that leakage was unsatisfactory.
The hatch seal was subsequently refurbished and a retest yielded satisfactory results.
Additionally, the leakage rate recorded on j
the test data sheet as 5.870 ccm was changed to the correct value
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of 5,870 ccm.
This item is closed.
3.
Plant Operations (71707)
The inspectors evaluated licensee activities to confirm that the facility was being operated safely and in conformance with regulatory i
requirements.
These activities were confirmed by direct observation,
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facility tours, interviews and discussions with licensee personnel and j
m nagement, verification of safety system status, and review of facility records.
To verify equipment operability and compliance with technical specifications (TS), the inspectors reviewed shift logs, Operations'
records, data sheets, instrument traces, and records of equipment mal f unc t ions.
Through work observat ions and disassions with Operations staff members, the inspectors verified the staff was knowledgeable of plant conditions, responded promptly and properly to alarms, adhered to procedures and applicable administrative controls, was cognizant of in
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progress surveillance and' maintenance-activities, and 'was aware of
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inoperable equipment. status. The inspectors performed channel
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verifications and' reviewed component status and safety related
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parameters to verify conformance'with TS.
Shift. changes were observed, i
verifying that system status continuity was maintained and that' proper control room staffing existed.
Access to the control room was
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restricted and operations personnel carried out their assigned duties lin l
an effective manner.
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Plant tours and perimeter walkdowns were conducted to verify' equipment operability, assess the general condition of plant equipment, and to
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verify that radiological controls, fire protection controls, physical protection controls, and equipment tag out procedures were properly-implemented.
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a.
Unit 1 Operatioral Status
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Power was reduced to 53% on June 12 to remove the IP-25B
- i condensate pump from service for inspection and repair.
The pump
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had developed a leak in the upper flange and was temporarily.
repaired by welding a patch.over the defect. Temporary repairs
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were successfully completed and full power restored the same day.
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On July 11, power was reduced several megawatts to_ compensate for l
a multiplexing unit's failure which was' apparently caused by high -
i room area temperature.
Following troubleshooting, the' unit.was returned to full power the next day. ~The unit operated at full'
power during the remainder of this period with only requested; load-
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following power reductions.
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Unit 2 Operational Status The unit continued to operate at full power during_this period
with only requested load following power reductions.
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4.
Maintenance / Surveillance Observation (62703) (61726)
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Maintenance
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The inspectors observed safety related maintenance activities on systems.and components to ascertain that these activities were conducted in accordance with 1S, approved procedures, and apprcpriate industry codes and standards.
The inspectors determined that these activities did not violate limiting conditions for operation (LCOs) and that required redundant components were operable, The inspectors verified that required administrative, material, testing, and radiological and fire prevention controls were adhered to.
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i Specifically, the inspectors observed / reviewed the following.
i maintenance activities:
i Replacement of auxiliary feedwater steam trap 2MS-327B
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The inspector noted that only one of two available isolation
valves between the steam generator and the valve being
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worked were red tagged shut. Although both valves were-l shut, the steam trap root valve was not specified on the
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equipment isolation sheet as being required to be shut.
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Operations personnel corrected this situation by promptly.
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tagging the valve and revising the' equipment isolation
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sheet. The licensee has no formal requirement for two valve protection in high energy systems but encourages the-i practice informally.
Licensee management stated their j
intention to incorporate such guidance into their equipment'
i isolation procedure'in an upcoming revision, Interim i
guidance was communicated to the operating shifts via a
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Night Order.
2P-29 auxiliary feedwater pump oil chaage j
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r IP-llB component cooling water pump seal repair l
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A newly written procedure, RMO-Pil-2, (Revision 0),
" Component Cooling Water Pump Seal Replacement", was used to i
perform this maintenance action.
Although the main' body of
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the procedure provided good direction for the evolution, two
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supporting procedures were also required, which made the administrative portion of the assignment cumbersome and
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All three procedures were intended to simultaneously support each other; however, there was
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duplication of effort such as multiple requirements to
obtain shift supervisor permission to perform the maintenance. Maintenance supervision stated that this difficulty would be addressed for procedures generated in
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the future.
Control rod drive system troubleshooting and repair (see
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paragraph 4.c.)
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Removal of service water header isolation valve SW-7 l
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Surveillance t
The inspectors observed certain safety related surveillance
activities on systems and components to ascertain.that these
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activities were conducted in accordance with license requirements.
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For the surveillance test procedures listed below, the inspectors determined that precautions and LCOs were adhered to, the required i
administrative approvals and tag-outs were obtained prior to test i
initiation, testing was accomplished by qualified personnel in
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i accordance with an approved test procedure, test instrumentation was properly calibrated, the tests were completed at the required
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frequency, and that the tests conformed to TS requirements.
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test completion, the inspectors verified the recorded test data was complete, accurate, and met TS requirements; test
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discrepancies were properly documented and rectified; and that-the systems were properly returned to service.
Specifically, the inspectors witnessed / reviewed selected portions of the following test activities:
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TS-6 (Revision 15), Control Rod Exercise, Biweekly l
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ICP 13.8 Appendix A (Revision 3), Auxiliary Feedwater System j
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I RESP 6.1 (Revision 12), Core Power Distribution and Nuclear
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Power Range Detector Calibration
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Control Rod Drive syste-Malfunctions l
l During performance of the Unit 2 biweekly control rod exercise l
test (15-6) on May 28, group 1 control rods in shutdown bank-A did not insert upon demand.
The demand counter for these control rods also did not move, indicating an electrical fault in the control circuitry.
The cause was not determined because the fault cleared
before diagnosis could be completed. Attempts to recreate the l
condition were unsuccessful.
Operators verified that.the
l condition was not caused by mechanical binding and that the rods l
l could f all into the core if required.
On June 10, during the next scheduled performance of this test,
additional personnel were stationed to monitor the evolution.
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f aulty condition reocc urred, but ; gain cleared before diagnosis-
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'fechnicians rea,cced the slave cycler counter l
could be completed.
circuit card in the rod control.jnen whose intermittent failure
would have been most likely to ha e psnerated the symptoms noted.
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Testing f requency was increased fr.n Siweekly to semiweekly.
Bench testing of the circuit card tM t had been removed from the control system revealed that it was in fact defective.
Subsequent testing of Unit 2 control rod drive c.ystem revealed no further system abnormalities and the test interval was returned to normal on June 25.
On June 25. during performance of biweekly control rod exercise test on Unit 1 (TS-5), rods in control banks B, L and D did not move upon demand.
As they had previously done for the Unit 2 event, operators verified that the cause of the f ault was electrical and that rods were not physically bound.
The Unit i fault was determined to be different than that on Unit 2 because whereas only one group on Unit 2 did not respond, three entire banks on Unit 1 did not respond. An attempt to recreate the
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Unit 1 condition was unsuccessful as all banks subsequently' moved as demanded.
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During the next Unit 1 test performed July 6, technicians observed l
that two banks were simultaneously selected for movement following l
positioning of the bank selector switch from one bank to the nex1.
This symptom indicated a fault with the supervisory buffer memory circuit card, which was consequently replaced.
Frequency of Unit.
I testing was also increased to semiweekly to verify no further abnormalities.
Bench testing of the removed card indicated a faulty gate which caused the input from the bank selector switch to be incorrectly output.
Continued testing of Unit I rod control
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l has not indicated additional abnormal operation.
l The inspectors reviewed the licensee's maintenance work plans and.
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schedule for addressing this issue and closely monitored the resultant corrective actions and testing.
The inspectors will continue to follow this issue's resolution and document their findings in a future report (266/93011-01).
l No other discrepancies were noted during the observance of any of l
the above tests.
5.
Emergency Preparedness (71707)
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An inspection of emergency preparedness activities was performed to l
assess the plant's implementation of the site emergency plan and I
implementing procedures.
The inspection included a monthly review and l
tour of emergency facilities and equipment, discussions with licensee staff, and a review of selected procedures.
An emergency plan drill was performed on June 24 involving declaration of an emergency classification level, activ6 ion of the technical support center and operations support center, and plant accountability exercising.
The drill scenario consisted of a steam generator tube rupture on Unit I leading to a minor offsite release, a short duration
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loss of AC power to the affected unit, and two missing personnel.
The control room simulator was used to provide realism and enhance drill etfectiveness.
Overall drill performance was good.
Emergency classification levels were appropriately declared and notifications were made within specified time requirements.
Unit I was shut down as directed by emergency procedures and a shutdown of Unit 2 was commenced as required due to both emergency diesels being inoperable.
Communications among personnel were inconsistent, ranging from informal exchanges to very effective interactions with " repeat backs" However, operators were prompt in requesting clarification of any ambiguous guidance and no erroneous actions occurred due to misunderstanding of directions.
Control room personnel responded well to the scenario even though a second cont rol l
operator was assigned to assist the Unit 1 operator.
Communications were initially weak among the control room, technical support center, i
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and operations support center.
This resulted in degraded coordination
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of activities and some confusion regarding response facilities'
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activation status and responsibility for specific corrective actions.
Although overall corrective action prioritization was good, attention
was not fully focussed on reducing or eliminating the offsite release.
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Operators carried out the actions of the emergency operating procedure j
directing an attempt to equalize pressure between the primary and the-l faulted steam generator to reduce the leak rate, but these actions were
not aggressively pursued by the technical support center when problems l
were initially encountered in equalizing pressure.
However, the minimal release that was simulated under the drill scenario likely contributed to the reduced attention paid to this factor.
Good engineering support was available in the technical support center.- However, it was under-
utilized due to managers becoming overly involved in the details of j
certain activities.
l A critique was held shortly after the drill to provide feedback to the participants.
Drill observers provided objective and critical comments indicative of accurate and impartial observation of drill activities.
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Many of the comments corresponded to the inspector's own observations.
i However, the critique was often sidetracked by detailed discussions of
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possible corrective actions which caused it to lose effectiveness.in
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relaying weaknesses to participants.
Portions of the critique were not widely attended, which further reduced its training value.
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Security (71707)
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Tne inspectors, by direct observation and interview, verified that
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portions of the physical security program were being implemented in
accordance with the station security plan.
This included checks that
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identification badges were properly displayed, vital areas were locked and alarmed, and personnel and packages entering the protected area were
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appropriately searched.
The inspectors also monitored any compensatory measures that may have been enacted by the plant.
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The inspector observed testing of sections of the plant's perimeter i
intrusion detection system.
Appropriate measures were taken to
compensate for any weaknesses found.
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Engineerina and Technical Support (71707) (450531
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I The inspectors evaluated engineering and technical support activities to determine their involvement and support of facility operations.
This was accomplished during the course of routine evaluation of facility i
events and concerns, through direct observation of activities. and
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discussions with engineering personnel.
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Construction of New Emeroency Diesel Generator Buildinq j
Ground breaking'for the building to house two new emergency diesel generators and the new diesel fuel oil system began the week of L
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June 7.
The concrete pours for the main building's foundation and walls commenced June 30.
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The inspectors monitored excavation and grading activities, preparations for foundation setting, concrete batching and j
testing, placement of steel reinforcing bar, and pouring of
foundation and structural concrete.
Concrete testing was
witnessed at both the discharge from the trucks and.at the pour point.
Point Beach used an independent testing laboratory for
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concrete testing and analysis.
Discussions were held with craft
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workers and supervisors to evaluate their ' knowledge' of the job
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requirements.
Various personnel involved with construction -
management were interviewed to discuss soil and concrete testing
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and results, quality assurance oversight, and control of contractors.
No concerns were noted. The inspectors will j
continue to monitor progress of this construction.
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8.
Safety Assessment /0uality Verification (40500) (90712]_L92700]
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a.
Licensee Event Report (LER) Review The inspectors reviewed LERs submitted to the NRC to verify that the details were clearly reported, including accuracy of the description and corrective action taken.
The inspector determined-whether further information was required, whether generic implications were indicated, and whether the event warranted onsite follow up.
The inspector also verified that. appropriate ~
corrective action was taken or assigned and-that continued operation of the facility was conducted in accordance with Technical Specifications and did not constitute an unreviewed safety question as defined in 10 CFR 50.59.
The following LERs were reviewed and closed:
266/301/91-010 Four Hydrogen Monitors found with Default Calibration Parameters During Instrumentation and Control i
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procedure 13.2 l
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This report describes an event wherein all four containment hydrogen monitors were found with their default calibration parameters inserted.
This condition, which in effect renders the monitors inoperable, was caused by their microprocessor being reinitialized following a momentary power interruption when instrument bus power supplies were shifted.
Details appear in inspection report 266/91022; 301/91022.
Corrective action i
included installation of a modification to lock-in the hydrogen monitor trouble alarm upon a loss of power.
This prevents the i
operator from inadvertently resetting the alarm in the control l
room until the loss of power alarm is physically reset at the local instrument, rack. The alarm response sheet then alerts operators of the requirement to have the monitors recalibrated.
The inspector verified installation of the modification and i
reviewed the alarm response sheets.
No concerns were noted.
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266/301/92-010 Isolation'of SI Pump Flow Path During inservice'
Testing of Minimum Flow Recirculation Line Isolation Valves
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This' report describes the possible isolation'of all available flow
paths for the safety injection pumps during performance of quarterly Inservice Tests IT-40 (Unit 1) and IT-45 (Unit 2).
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isolation of all available flow paths could result in operating
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the pumps at shutoff head, ultimately damaging the pumps and
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rendering them inoperable.
Corrective' action for this issue is-i detailed in the closecut of the associated inspection follow up l
item in paragraph 2.b above.
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301/93-002 Reactor-Trip During Turbine Trip Testing This report describes a Unit 2 reactor trip from full power at l
1:45 a.m. on March 28.
The trip occurred during main turbine trip testing and was caused by a combination of operators not following.
procedures and a loose test valve handwheel.
Details are
contained in Inspection Report 266/93006; 301/93006, in which the
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procedural violation was identified. The inspectors
reviewed / evaluated the respective corrective actions which.
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included: a human performance evaluation; counseling of the
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individuals involved in the event; clarification of task
separation between the task supervisor and workers-performing the j
test: and, repair of the faulty test valve handwheel.
266/301/93-003 Nonconservative Setpoints for the Low
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Temperature Overpressure System l
This report describes a condition where the setpoints for the low
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temperature. overpressure (LTOP) system were nonconservative.
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location of the reactor coolant system pressure transmitters was
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not adequately considered in the LTOP setpoint development-analysis.
With reactor coolant pumps running and developing a
differential pressure across the core, the reactor vessel pressure
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is greater than that seen by pressure transmitters.
As a result,
.i during the worst case mass input transient, the LTOP setpoint-is
'l approximately 34 psig too high to operate both reactor coolant
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pumps throughout the entire temperature range.
As corrective action, changes were made to procedures OP-1A, " Cold Shutdown to Low Power Operation" and OP-3C, " Hot Shutdown to Cold l
Shutdown" to ensure that one reactor coolant pump is secured when j
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cold leg temperature is below'160 F and to require that the control switch for the idle pump be red tagged out.
Restricting j
operation to one reactor coolant pump reduces the differential
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pressure across the core sufficiently that the maximum allowable vessel pressure is not exceeded even with the existent LTOP
setpoints and pressure transmitter locations.
At temperatures.
above 152 F the maximum allowable pressure is sufficiently high
to permit unrestricted coolant pump operations under the existent ~
configuration.
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. The plant had submitted an exemption request with ' respect to this
'i issue to allow use of recently approved ~ASME code case.N-514.
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-This specific exemption request.was subsequently withdrawn -
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-following'a decision to instead await conclusion of the routine r
ASME code approval process before-proceeding.
The inspectors will
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follow resolution of this issue through unresolved item 93-006.
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M nager's' Supervisory Staff Meetina l
The inspectors observed session 92-13 of the Manager's Supervisory Staff.
Issues discussed included proposed control room ventilation system modifications, Individual _ Plant ' Evaluation
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results,-and a TS change request.
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Outstandina items (92701)
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Inspection follow Up items
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Inspection follow up items are matters which have been discussed with Wisconsin Electric management, will be reviewed further by the inspector, and involve some action on the part of the NRC, company or both. A follow up item disclosed during the inspection is discussed in paragraph 4.c.
10.
Exit Interview (71707)
A verbal summary of preliminary findings was provided to the Wi-sconsin
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Electric representatives denoted in Section 1 on July 15, at the
conclusion of the inspection.
No written inspection material was provided to company personnel during the inspection.
Tha likely informational content of the inspection report with _ regard to documents or processes reviewed during the' inspection was also
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discussed.
Wisconsin Electric management did not identify any doc'ments-
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u or processes that were reported on as proprietary, i
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