IR 05000348/1989022
| ML19332E475 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 11/09/1989 |
| From: | Cantrell F, Maxwell G, Miller W, Palmain P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19332E470 | List: |
| References | |
| 50-348-89-22, 50-364-89-22, NUDOCS 8912070271 | |
| Download: ML19332E475 (16) | |
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'a UNITED STATES
,g NUCLEAR REGULATORY COMMISSION
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101 MARIETTA STREET.N.W.
- jr ATLANTA, oEoRGI A 30323
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Report Nos.: 50-348/89-22.and 50-364/89-22
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Licensee:
Alabama Power Company
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600 North 18th Street Birmingham, AL 36291
' Docket Nos.:
50-348 and 50-364 License Nos.:
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i-Facility name:
Farley 1 and 2
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-Inspection Conducted: September 11 - October 10, 1989
Inspection at Farley site near Dothan, Alabama Inspectors:
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L g. F. Maxwell, Senior Resident Inspector Date Signed
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@. H.
iller, Jr., Resident Inspector Date Signed
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I l ~ 9 - 9'Cr P. A. Balmain Project Engineer Date Signe'd
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Approved by:
n/?,/79 Floyd5.Cantrell,yfef Dhte Sign 6d Projects Section 1B Reactor Projects Branch 1 Division of Reactor Projects SUMMARY Scope:
This routine onsite inspection involved a review of operational safety verification, monthly surveillance observation, monthly maintenance observa-tion, quality verification program, action on previous inspection findings, and 10 CFR 50.59 Safety Evaluations.
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Results:
Within the areas inspected a violation was identified:
Unapproved overtime worked by licensed operators - paragraph 7.
One apparent violation is being considered for escalated enforcement:
Breach of containment integrity -
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paragraph 3.b.(2).
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8912070271 891106
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PDR-ADOCK 05000348 Q
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. Two unresolved items ~ were identified involving temperature liniit requirements
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equipmenti areas paragraph 3.b (5). and operability of personnel containment.
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airlock paragraph 3.b.(3)
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Certain tours were conducted on.' deep backshift or' weekends; these tours were
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conducted on September 23. October 3 and 9 -(deep backshift inspections-occur between 10 p.m. and 5.a.m.).
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REPORT DETAILS
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Licensee Employees Contacted l
. R. G. Berryhill, Systems Performance and Planning Manager
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C. L. Buck, Plant Modification Manager -
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L L. W. Enfinger, Administrative Manager
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t R. D. Hill, Assistant General Manager - Plant Operations
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- D. N. Morey, General Manager - Farley Nuclear Plant
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C. D. Nesbitt, Technical Manager l
J. K. Osterholtz, Operations. Manager t
L. M. Stinson, Assistant General Manager - Plant Support
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J. J. Thomas, Maintenance Manager L. S. Williams, Training Manager
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Other licensee employees contacted included, technicians, operations personnel, maintenance and I&C personnel, security force members, and office personnel.
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Acronyms and abbreviations used throughout this report are listed in the
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last paragraph.
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Plant Status Unit'l
.l Unit 1 operated at approximately 100 percent reactor power from the l
beginning of the reporting period until September 22.
On that date power
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was reduced and the unit was shutdown for a refueling outage.
The outage
'is scheduled for 45 days.
Unit 1 is scheduled to return to power on November 7, 1989.
Unit 21
Unit 2 operated at approximately 100 percent reactor power from the beginning of this reporting period until September 20.
On that date power
was reduced approximately 60 percent to-repair an electro-hydraulic fluid
1eak on steam generator feedwater pump 28.
At 7:22 a.m. on September 20,
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.the reactor was manually tripped following a loss of the remaining feedwater pump 2A, This pump was tripped due to a low auto-stop oil pressure signal which was generated when a worker inadvertently bumped the i
overspeed trip test valve lever and the valve-opened.
The Unit returned to power on September 26 and operated at approximately 100 percent reactor power-throughout the remainder of the reporting period.
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Other Inspections / Visits September 25 - 29, 1989, Report 89-23, followup on environmental equipment L
_ inspection.
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September 27
29, 1989, Visit by NRR Senior Project Manager -
10 CFR 50.59 and overtime issues discussed.
October 2 - 6,1989,. Report 89-25 ISI inspection and followup on maintenance open items.
October 2 - 6,~ 1989 Report 89-26 IST inspection.
October 2 -6, 1989, Report 89-27, Radiological Controls Inspection.
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' Operational Safety Verification (71707, 92700)
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Plant Tours L
The inspectors conducted routine plant tours during this inspection
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period to verify that the' licensee's requirements and commitments
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were being implemented. Inspections were conducted at various times including week-days, nights, weekends and holidays.
These tours were L
- performed to verify that: systems, valves, and breakers required for safe plant operations were in their correct position; fire protection equipment, spare equipment and materials were being maintained and
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r stored properly; plant operators were aware of the current plant'
status; plant operations personnel were documenting the status of
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out-of-service equipment;'there were no undocumented cases of unusual L
fluid leaks. piping vibration, abnormal hanger or seismic restraint i-movements; all reviewed equipment requiring calibration was current; I
and general housekeeping was satisfactory.
Tours of the plant included review of site documentation and interviews with plant personnel.
The inspectors reviewed the control i
room operators' logs, tag out logs, chemistry and health physics
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logs, and control boards and panels.
During these tours the inspectors noted that the operators appeared to be alert, aware of changing plant conditions and manipulated plant controls properly.
The inspectors evaluated operations shift turnovers and attended shift briefings.
They observed that the briefings and turnover provided sufficient detail for the next shift crew and verified that the staffing met the TS requirements.
Site security was evaluated by observing personnel in the protected and vital areas to ensure that these persons had the proper authorization to' be in the respective areas.
The inspectors also verified that vital area portals were kept locked and alarmed. > The
. security personnel appeared to be alert and attentive to their duties and those officers performing personnel and vehicular searches were
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thorough and systematic.
Responses to security alarm conditions l
appeared to be prompt and adequate.
Selected activities of the licensee's Radiological Protection Program were reviewed by the inspectors to verify conformance with plant procedures and NRC regulatory requirements.
The areas reviewed
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included: operation and management of the plant's health physics
'g, staff, ALARA implementation, Radiation Work Permits (RWPs) for compliance to plant. procedures, personnel exposure records, observa-N tion of work.and personnel in radiation areas to, verify compliance to (
- radiation protection
- procedures, and control of radioactive N
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Plant Events and Observations E-(1) Loss of Main Feedwater - Unit 2
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On September 20, the NRC duty officer was informed by the
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licensee that Unit 2 had been manually tripped. When the trip occurred, Unit 2 was operating at about 60 percent power to
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p allow ~ maintenance activities to be completed on the 2B turbine F
driven main. feedwater pump E-H system.
The 2A main feedwater t
pump - was running and the -2B feedwater pump was secured.
At about 7:22 a.m. the control room operator observed that the 2A
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pump speed and discharge pressure were dropping. -Realizing that the pump-had tripped, he promptly tripped the reactor to minimize
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while operating at 60 percent power.
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'The licensee. investigated the circumstances and conditions associated with' the 2A main feedwater pump trip.
The licensec
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' determined that the turbine overspeed trip test valve for the main feedwater pump turbine, had been accidentally opened by a plant. worker who was conducting work adjacent to the turbine overspeed trip test valve.
When this valve is placed in the open position, a flow path is provided to allow E-H low pressure oil fluid to equalize at a s1'ghtly reduced pressure throughout the > lines associated with the turbine overspeed trip system.
The E-H low pressure oil system normally operates at about 70 psi.
When the trip test valve was/is placed in the open position, this pressure equalized at about 56 psi.
I&C personne1' tested various pressure switches associated with the 2A, main feedwater pump turbine E-H system and found that the controlling pressure switch (PS) 63TT for the high pressure
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fluid trip solenoid valve and for the turbine E-H lube oil trip solenoid valve was incorrectly set.
The normal setting for PS
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63TT is 35 psi, however, the setpoint had drifted to 59 psi.
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The ~ licensee concluded that when the test valve was in the open position it allowed the E-H low pressure oil system pressure to equalize at 56 psi which was below the PS 63TT setpoint, This caused PS 63TT to activate, causing the previously noted solenoid valves to.open which resulted in dumping the turbine E-H low pressure oil system pressure.
This resulted in loss of steam to the turbine and ultimately loss of main feedwater flow.
I&C replaced PS 63TT and adjusted the new pressure switch to its correct setting (35 psi).
The feedwater pump turbine E-H
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-system was successfully.. tested and returned to service.
On
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September.26, the unit was returned to power operation.
(2) Loss of' Unit 1 Containment Integrity-Unit:1 core alterations were performed 'on October 1, without maintaining containment integrity.
On September 30, the operations crew established containment integrity in preparation for remcving the fuel from the reactor.
The on-duty crew
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believed - that' the atmospheric relief valve from each steam generator was closed; however, the relief valves had been opened to drain the, steam generator for outage work activities by ETP-1013, Steam Generator Cooldown Using AFW Flow Through Main
Steam Lines.
These valves had 'not been closed after the generators had been drained.
Procedure ETP-1013 did not require these valves to be closed.
The manways on the secondary side of
the steam generators were removed on September 28 and 29, for i
steam generator-inspection activities.
Thus,.an open flow path existed from the containment atmosphere to the outside environ-ment through the open manways and open atmospheric relief
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valves.
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l Procedure FHP-1.0, Controlling Refueling Procedure, requires i
containment integrity to be established prior to removal of the i
reactor vessel: head.
However, ' the TS does not - require containment-_ integrity for core alterations-until after the
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reactor head has been removed from the vessel.
states that during refueling operations each - containment j
building penetration providing direct access from the contain-ment atmosphere to-the outside ' atmosphere will be closed by an isolation valve, blind flange or manual valve.
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On. September 30, from 10:00 to 10:45 p.m., the reactor head was
removed from the reactor vessel.
The manway for steam generator
.1A was reinstalled on October 1, at 7:00 a.m., but the manways for generators 18 and IC remained open.
A core alteration i
occurred-on October 1, between 12:45 and 7:45 p.m., when the
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control rod drive mechanisms were unlatched from the control
rods.
During this time containment integrit> was not maintained.
An open flow path existed from the containment-l atmosphere to the outside atmosphere through the open manways of steam generators 18 and 10 and the open 6 inch atmospheric relief valves on B and C steam lines.
At 2:00 p.m., on October 1, the manway to steam generator 1B was replaced.
The manway to generator 10 remained open.
At 9:00 p.m., on October 1, the
shift supervisor noted that the atmospheric relief valves were l
open.
He directed the crew to close these valves to establish containment integrity.
L The licensee advised the resident inspectors of the potential breach of containment integrity on the afternoon of October 2, and upon verification officially advised the NRC via the i
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emergency notification (red phone) system at 6:55 p.m.
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October 2.
A followup LER report is to be issued.
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The licensee promptly discontinued all fuel alterations until all containment penetrations were inspected and verified to be
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Procedure 1-STP-18.4, Containment Refueling Integrity
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Verification, was revised to include inspection and verification h' t that all containment penetrations are closed prior to any core b
alterations.
This should prevent recurrence of this violation.
L No ; additional Unit 1 open containment penetrations pere P
identified during the performance of 1-STP-18.4.
'This event is similar to another loss of containment integrity l_
that occurred on April 19, 1989, during the previous Unit 2
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refueling outage.
For details refer to inspection report
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- No M3,369/89-11.
Based on the above, an open flow path existed to the outside
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atmosphere.through steam generator 1B for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 45 minutes and through steam' generator 1C for 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> 3 minutes.
This containment i_ntegrity issue is an apparent violation and will be
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discussed at an Enforcement Conference at the Region II Office on November 15, 1989.
(3) Unit 2 Containment Air Lock F
With the reactor in mode 3, on~ September 25 at 3:00 a.m., the operating crew found that the personnel containment airlock had
_ not been demonstrated operable as required by TS 4.6.1.3.a.
O When an airlock is being used for multiple entries, TS 4.6.1.3.a requires the air lock.be demonstrated operable at least once per
'72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Beginning September 20, at 1:46 p.m. and continuing for several days, multiple entries were made into the Unit 2
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containment following the 7:22 a.m.
However,
operability of the air lock was not demonstrated-at 1:46 p.m. on September 23, as required.
When the licensee identified this
discrepancy, the air lock was promptly verified to be operable by
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conducting procedure test 1-STP-15.0, Containment Air Lock Seal Operability Test.
-An incident report was written.
The licensee is evaluating the circumstances associated with this event to determine the cause and appropriate actions required to prevent recurrence.
This item is identified, as unresolved item 364/89-22-04, Operability of Unit 2 personnel containment air lock.
(4)
Loss of Control Room Pressurization System The control room is provided with two full capacity redundant air pressurization systems to maintain the control room at a positive pressure following a loss of coolant accident.
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system operates automatically upon receipt of a containment L
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isolation signal from either Unit 1 or Unit 2.
Control room pressure 'is maintained above. 0.25 inches of water by a -
i modulating pressure bleed off damper (HV-27688) located in the
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discharge flow path from each Jan unit.
On October 3, the'"A" train control room pressurization system was removed from service to perform some-design changes.
TS 3.7.7 allows one' of the two systems to be removed from service,
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but requires an inoperable control room emergency ventilation system to be restored to service within 7 days or be in hot
standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
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At 10:15 p.m. on October 3, during investigation of abnormal high control room pressure, the Unit 1 shift supervisor and Unit 2 shift foreman noted that the MOV 27698, "B" train control room ventilation exhaust damper indicator light on the B0P panel was not on.
Further investigation found breaker FG-84, which supplies power to MOV 2769B to be open.
Valve ~2769B is down stream. of damper 2768B.
With valve 27698 closed the full pressure from the pressure fan is discharged into the control
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room.
Pressure above 0.25 inches of water is maintained in the control room which makes the doors into the control room difficult to open.
However, technically both control room pressurization systems were out of service.
This situation placed the plant outside the provisions of the TS.
The breaker supplying the power to valve M0V 2769B was returned to service at 10:18-p.m. (3 minutes after the problem was identified).
The licensee's investigation found that the power.to MOV 2769B was verified during the shif t supervisor turnover at 6:30 p.m.
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and reactor operators turnover at 7:30 p.m. on October 3.
-The open bre iker was found at 10:15 p.m.
Therefore, the maximum time that power was not available to this valve was less-than 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
Although, the plant was operating outside the-TS, the safety = significance of this event is considered small since full pressurization was available to the control room and the licensee demonstrated the ability to quickly restore the system to normal condition once the problem was identified.
The breaker was apparently accidentally opened by contractors working adjacent in the corridor to the motor control panel.
The licensee is conducting an investigation to identify the person responsible.
This event was promptly reported to the NRC over the emergency notification (red phone) system at 11:31 p.m. on October 3, and subsequently retracted on October 23, based on the licensee
L determination that the control room ventilation system wao.iot inoperable.
Since the problem was identified by the licensee, promptly reported to the NRC and has only minor safety significance, no violation is being issued.
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I (5) High Temperature in Unit 2 Electrical Penetration Room
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On October 5, the inspectors found the temperature in electrical
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penetration room '2347 to be very warm..This room contains
- safety related "A" train motor control center MCC-20.
By using L
a-calibrated' temperature indicator, the inspectors identified L
the ambient _ temperature in ' the room to be approximately 108 F
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and the temperature of the motor control center cabinets varied b
from 110"F to 132 F.
FSAR section 3.11.2.2 states that active safety related equipment located outside of containment normally
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Upon further
investigation the inspectors found that the fan cooling unit for o
this space was not fully operable.
A work request,'MWR-194636A, I
to repair' the system was issued on August 26.
Maintenance was performed on the system; however, the system was found to be not functionally acceptable by operations on August 29.
A revised work order _was issued-by operations on August 29, but apparently nocadditional maintenance was performed prior to October 5.
On
October 6, operations listed this work order as a work request requiring ' attention.
After the inspectors informed operations of the hot temperature in Room 2347, compensatory measures were
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promptly initiated to provide interim cooling for this t com.
The non-fire-rated door into an adjacent room was blocked opened to provide additional cooling.from another cooling unit.
This arrangement provided some improvement; however, 'the room remained warm due to heat being transferred through an open vent-shaf t into the room from the' mechanical penetration room on the 121' elevation beneath room 2347.
t TS 3.7.13 ' and Table 3.7-8 do not include MCC-2U (room-2347) as
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an area which requires the temperature to be maintained within
established limits.
Nevertheless, it appears that the temperature in this room needs to be maintained below 105 F.
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Also, it appears that the temperature limits established by the TS for electrical equipment areas may not be consistent with the a
FSAR and the environmental design criteria.
The licensee was requested to review the following concerns of the inspectors:
determine if equipment cooling is required to be maintained in safety 'related electrical equipment areas to meet the operability requirements of TS 1.18 and, determine if the conflict between the temperature limits of TS Table 3.7-8 and i
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FSAR Section 3.11.2.2 is acceptable.
This item is identified as Unresolved Item 348,364/89-22-03, Temperature limit requirements
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for electrical equipment areas.
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.4.
Monthly Surveillance Observation (61726)
i The inspectors witnessed the licensee conducting maintenance surveillance h
test activities on safety-related systems and components to verify that the licensee performed the activities in accordance with TS and licensee requirements.
These observations included witnessing selected portions of
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'each surveillance, review of the surveillance procedures to ensure that
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administrative controls and tagging procedures were in force, determining
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i that approval was obtained prior to conducting the surveillance test, and the individuals conducting the test were qualified in-accordance with-plant-approved procedures.
Other observations included ascertaining that test instrumentation used was calibrated, data collected was within the
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specified. requirements of.TS, any identified discrepancies were properly i
' noted, and;the systems were correctly returned to service.
The following
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specific activities were observed:
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0-STP-80.1 Diesel Generator 1-2A Operability Test
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p 1-STP-158 RCS Pressure Isolation Valve Leak Test
1-STP-608.1 Main Steam Safety Valve Tests
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2-STP-24.1 SW Pumps 2A, 2B and 2C Quarterly IST 2-STP-29.2 Shutdown Margin Calculations l
.2-STP-35.1 Unit 2 Startup TS verification 2-STP-152.0 Steam Generator Feedwater Pump 2A and 28 Stop Valve Stem Movement Verification.
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0-ETP-3635 Reliability Check of source Range Instrumentation l
1-ETP-204 Containment Atmospheric Sample System Leakage Assessment (R-11 and R-12).
1-ETP-4289 Steam Jet Air Ejector Performance Test
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The inspectors witnessed the test of CCW pump 2C using procedure 2-STP-23.3.
The flow rate for the pump was found tu be in the alert range.
Operations promptly initiated an investigation to determine the cause of the flow deficiency.
The discharge gauge was found out of calibration.
The gauge was replaced and pump was satisfactorily retested.
On September 23, the inspectors witnessed a portion of the tests performed on the main steam safety relief valves which were tested using procedure 1-STP-608.1.
Of the 15 valves tasted 9 were found to operate at a pressure slightly out of the required acceptable ranges.
These valves were adjusted and satisfactorily retested.
No violations or deviations were identified.
The results of the inspections in this area indicate that the program was effective with respect to meeting the safety objectives.
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Monthly Maintenance Observation (62703)
LThe inspectors ' reviewed the licensee's maintenance activities to verify
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.the following: maintenance personnel were obtaining the appropriate tag i.
_out.and clearance approvals prior to commencing work activities, correct F
documentation was1available for all requested parts and material prior to use, procedures were available for all requested parts and material prior to; use, procedures were : available and - adequate for the work being conducted, maintenance personnel performing work activities were qualified
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L to accomplish these ' tasks, no maintenance activities reviewed were
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violating any limiting conditions for operation.during the specific evolution,'the required QA/QC reviews and QC hold points were implemented, post-maintenance testing activities were completed, and equipment was properly returned Lto service after the completion of work activities.
Activities reviewed included:
MWR 183524-Replace existing Unit 2 fire damper No. 2-121-116-05 with Pullman type fire damper.
MWR 210141 Repair seal leak on boric acid pump 18.
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MWR 193383 Install new EQ cabling to Unit 1 solenoid valve Q1N23SV3227AC-B to motor driven auxiliary feedwater pump.
MWR 210510 Repair leak on Unit 2 reactor coolant RTD valve Q2013V005A.
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WA-WOO 302456 SW pump 1E supply breaker Q1R158RKDLO4 maintenance (procedure EMP-1313.03).
0-50P-0.6 Limitorque valve lubrication and valve stem inspection (several valves in Unit 1 121' mechanical penetration room).
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No violations or deviations were identified.
The results of the inspections in this area indicate that the program was effective with
respect to meeting the safety objectives.
6.
Inspection of Quality Verification Function (35702)
The inspectors; reviewed the quality control (QC) organization established to provide independent review and verification of work performed by the site contractor (" Fluor") who will perform most of the PCN modifications during the current Unit I refueling outage.
The group includes 21 inspectors.
Four are permanently assigned at Farley.
The remainder are permanent Alabama or Georgia Power Co. employees who have been temporarily assigned to Farley to provide coverage during this refueling outage.
The number of inspectors assigned to each discipline is as follows:
eleven electrical, six mechanical, three civil and one nondestructive examination.
All are certified to meet the Level II inspector j
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L qualifications of ANSI N45.2.6.
The NRC' inspectors reviewed -the qualifications of a sample of-the QC inspectors and determined that they were qualified to perform their assigned duties.
The licensee's QC. program to verify - compliance with construction requirements by the principle site contractor, Fluor, appears to be effective.
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.7.
' Action on Previous Inspection Findings (92702)
(Closed) Unresolved item 50-348/364-89-14-01.
Apparent Excessive Work-hours for Licensed Operators.
- A management meeting was apparent held at the Region II office on July 31, 1989, to discuss the licensee's apparent excessive use of overtime for licensed operators.. Licensee representatives discussed their method of scheduling licensed operator crews for shift work and explained that
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'during periods when one unit was in an outage that they maintained i
essentially the'same. schedule, but extended shifts from eight to. twelve i
hours such that crews' worked 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> in a seven consecutive day period.
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They opposed any changes to their scheduling method because it would be disruptive to crew morale and have possible ALARA considerations.
After the meeting, NRC representatives acknowledged the. licensee's
. position but expressed concern about several cases in which operators used l
~ overtime' excessively, without first obtaining required management
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approval. - This was evidenced by one reactor operator who worked eight consecutive = days (he -worked at least 92 hours0.00106 days <br />0.0256 hours <br />1.521164e-4 weeks <br />3.5006e-5 months <br />)on the operating unit;
another reactor operator worked eight consecutive days (96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />) on the
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operating unit; and a SR0 who worked 16 consecutive days'(184 hours0.00213 days <br />0.0511 hours <br />3.042328e-4 weeks <br />7.0012e-5 months <br />) on the shutdown unit.
The NRC was also concerned that the 1989 Unit 2 a
refueling outage schedule for shif t supervisors (SR0s) and reactor operators was not approved by the required level of management (General
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Manager - Nuclear Plant, his designee, i.e. Emergency Director or higher
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authority).
The schedule contained numerous instances where employees I
were allowed / required to work in excess of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in seven consecutive days.
Allowing employees to ' work in excess of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during any seven consecutive day period without the required management approval is a violation of TS Section 6.2.2.f.
,.
After further review, the NRC agrees that the refueling outage schedule i
may be used as a mechanism for overtime approval.
However, the outage schedule - should state clearly that it is also serving as overtime o
approval.
Proper management approval signatures would be needed to comply with Technical Specifications 6.2.2. f.
While the licensee presented a generally acceptable reason for their present work schedule (e.g., five, i
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six, seven and seven consecutive work days,12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per day during each five week period), the NRC staff expressed the opinion that use of this schedule is not prudent during extended periods because of the tendancy of fatigue to degrade performance.
The NRC considers that the "five, six, l
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seven, and seven days" rotational shif t schedules during trefueling at
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for tracking purposes. unresolved item 50-348/364-14-01 is closed and will
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now be identified as a violation 50-348/364-89-22-02, Unapproved Overtime L
fforfLicensed Operators.
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Bi. ~10 CFR 50.59 Safety Evaluations'(92700)
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During September 27 and 28, 1989,- the NRC Senior Project Manager conducted an on-site audit of certain Farley administrative procedures and safety evaluations done per 110 CFR 50.59.
The purpose.of' the audit was to
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. determineithe extent of licensee actions taken as a result of a prior headquarters' audit -(Inspection Report 88-19' dated July 7,1988, as corrected' on August 17,-~1988.)
Another purpose of the audit was to determine the extent of licensee actions taken as c result of U U issuance of industry guidelines for 10 CFR Safety Evaluation (NSAC-125 r
dated June:1989).
The-following is a summary of. the audit results:
a.
-Persons contacted (Licensee personnel):
+
- D. Morey, General Manager
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- C. Nesbitt, Technical Manager
- S. Fulmer,. Supervisor Safety Audit and Engineering
- Attended exit interview
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b.
The following documents were discussed with licensee personnel:
(1)
FNP-0-AP-1,- Development, Review and Approval of Plant
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' Procedures, Revision 25, dated July 13, 1989.
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(2) Memorandum'from Dave Morey to Holders of FNP AP 1, Designation of
'
. Qualified Reviewers, dated July 25, 1989.
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'(3)- Nuclear Safety Analysis Center, NSAC-125," Guidelines for 10 CFR s
50.59 Safety Evaluations, dated June 1989.
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(4) NRC Letter from C. E. Rossi to Director 0 MSS Davision, NUMARC, dated May 10, 1989.
(5) -PCN No. B 88-1-5244, Revision 2, dated September 13, 1989 Farley
~ Nuclear Plant Nuclear Safety Evaluation Checklist for reactor vessel lead vent line extension.
c.
Discussion:
The NRC staff completed its review of the joint NUMARC/NSAC working
. group's final draft of NSAC-125, and has commented by letter dated May 10, 1989.
Enclosure 1 to the May 10 letter includes a discussion of three key issues:
1) Margin of safety, 2) Consequences of an
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accident' or malfunction of equipment, and 3) Probability of an accident or malfunction of equipment.
A copy of this letter was g
provided to:the licensee for information.
p In addition, the NUMARC response (letter Tipton to Rossi dated June
.,
9,1989) to the NRC letter of May 10, 1089 was provided to the licensee' for information.
With these two letters-and the June 1989
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NSAC-125'. the licensee has all available 'information needed to
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upgrade the administrative procedures and training program at the Farley site.
f
.In view of the information provided above, the following questions F
were asked of the licensee:
'
1)1.What is the current status of the Farley site procedures for 10 CFR 50.59 reviews as compared to NSAC-125 guidance?
s Answer:
Plant staff is still trying to clearly understand L
NSAC-125 guidance.
As understanding becomes clear, f
modifications will be made.
2)
What is the current status of the Farley site training for change initiators and reviewers of 10 CFR 50.59 evaluations?
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Answer:
No specific 10 CFR 50.59 training exists at Farley
,
site.
3)
.What is the current schedule for implementing NSAC-125 and for initiating 1) and 2) above?
Answer:
Once the questions on NSAC-125 are resolved, an attempt would be made to make necessary changes by the end of 1989.
!
- This licensee response was considered commendable.
However, during subsequent discussions,- the. licensee advised that a-corporate approach was preferred.
A task force has been established with the Vice President Technical Support of Southern Services as chairman.
This task force will have two members from each SON 0PC0 plant (Farley, Vogtle, and Hatch).
The licensee stated that since the task force was only recently established, a schedule for implementing NSAC-125 guidelines w
will be formulated as soon as the extent of the changes are determined. Although the corporate approach has certain advantages, NRC expressed a concern that this approach (which started the first week of October 1989) could delay training of personnel at Farley site in using NSAC-125 guidance.
It was suggested'that the qualification and training of site personnel that are currently authorized to initiate and review 10 CFR 50.59 mondifications and procedure changes begin at an early
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date.'
Along these lines, the listing of qualified reviewers (dated; July-25,1989) contains over 200 personnel from all groups at the site.
For training and qualification purposes, it
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would-offer that the listings should be reduced.
This action L.
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would minimize training requirements and should improve the 10 4.
CFR 50.59 safety evaluation ef fort.
This.will be an inspector-
followup: item, 348/364-89-22-05, Implementing 10 CFR 50.59 R
y Guidance.
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'9..
Exit. Interview L'
The inspection.. scope and findings were summarized during ' management -
L interviews; throughout -the report period and on October 11, with the plant
?
manager <and selected members of his staff.
The inspectors stated relative O
to violation 89-22-02, that while the licensee's extensive use of overtime-may be technica11yDiegal',. the extent 'of overtime used during outages j
h appears to violate the intent of guidance provided on this subject.
The
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NRC staff believes that Alabama Power Company should develop other methods to-provide part ~ of the additional shif t coverage needed during outages.
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'The 111censee' acknowledged.the inspection findings and did not identify as
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proprietary any material reviewed by the inspection during this
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. inspection.
i Item Number-Description and Reference g
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- .348,364/89-22-01 Apparent violation of - containment integrity
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during refueling, paragraph 3.
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'348,364/89-22-02 Violation - Unapproved overtime for licensed
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t operators, paragraph 7.
- 348,364/89-22-03 URI Temperature limit requirements for
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electrical equipment areas, paragraph 3.
364/89-22-04 URI - Operability of Unit 2 personnel airlock, paragraph 3.
-348,364/89-22-05 IFI Implementing 10 CFR 50.59 guidance, d
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paragraph 8.
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- Licensee was informed that the item discussed in paragraph 7 was closed.
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'cronyms and Abbreviations
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A
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.AFW' -
A0P Abnormal Operating Procedure
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AP Administrative Procedure
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APC0 -
Alabama Power Company CFR Code of Federal Regulations
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CCW Component Cooling Water
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- DC Design Change
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L DR-Deviation' Report
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ECP Emergency Contingency Procedure
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.E-H Electro-Hydraulic Control System
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~EIP Emergency Plant Implementing Procedure
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EQ Environmental Qualifications
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ESF.
Engineered Safety Features
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-EWR Engineering Work Request-
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F Fahrenheit
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GPM Gallons Per Minute
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-ISI
' Inservice Inspection
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.IST
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Inservice Test
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LCO_
Limiting. Condition for Operation
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.MOV Motor-Operated Valve
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M0 VATS. -
Motor-0perated Valve Actuation Testing o
Maintenance Work Request MWR
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NCR Nonconformance. Report
-
NRC Nuclear Regulatory Commission
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NRR NRC Office of Nuclear Reactor Regulation
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Plant Modifications Department
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i PMD
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PSI-Pounds Per_ Square Inch
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' QA Quality Assurance
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QC Quality Control
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RCP Radiation' Control-and Protection Procedure
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SI Safety Injection
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SAER Safety Audit and Engineering Review
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S/G'
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SSPS Solid State' Protection System
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SOV Solenoid 0perated Valve
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STP Surveillance Test Procedure
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SW'-
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Technical Specification i
TS-
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WA
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Work Authorization
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