ML20056F429
| ML20056F429 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 08/11/1993 |
| From: | Rogge J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20056F425 | List: |
| References | |
| 50-289-93-14, NUDOCS 9308270218 | |
| Download: ML20056F429 (47) | |
See also: IR 05000289/1993014
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U. S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No. 93-14
Docket No. 50-289
License No. DPR-50
Licensee: GPU Nuclear Corporation
P.O. Box 480
Middletown, PA 17057
Facility: Three Mile Island Station, Unit 1
Location: Middletown, Pennsyh'ania
Inspection Period: June 22,1993 - July 31,1993
Inspectors: Michele G. Evans, Senior Resident Inspector -
David P. Beaulieu, Resident Inspector
Paul Kaufman, Project Engineer
Approved by: w >L ([ewse///, /993
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f[Jr(1 F. Rogge, Chief// 9 (/ Date'
eactor Projects Section No. 4B, DRP
Inspection Summary: The NRC Staff conducted safety inspections of Unit 1 power
operatious. The inspectors reviewed plant operations, maintenance, engineering, radiological
controls, and security activities as they related to plant safety.
Results: An overview of inspection results is in the executive summary.
9308270218 930818
ADDCK 05000289
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EXECUTIVE SUMMARY
Three Mile Island Nuclear Power Station
Report No. 50-289/93-14
Onerations
Overall, the licensee conducted plant operations in a safe and conservative manner. The
inspector found that shift turnovers were comprehensive and accurate, and adequately
reflected plant activities and status. Control room operators effectively monitored plant
operating conditions and made necessary adjustments. There was extensive management
involvement in daily activities.
Following identification of c lubricating oil leak on the 'B' emergency diesel generator
(EDG) which could have potentially made the EDG inoperable, the licensee did not consider
the event reportable per 10 CFR 50.73. However, following discussions with the inspector,
the licensee appropriately reported the event within the required time frame.
Maintenance
Due to an inadequate surveillance procedure, the bolts for the 'B' (EDG) lubricating oil cover
were not properly torqued, resulting in a lubricating oil leak. The licensee's immediate
corrective actions in response to this event were appropriate. However, the surveillance
procedure was inadequate because it failed to specify a torque value for the lubricating oil
filter cover. As a result, during the period June 24,1993 through July 1,1993, the ability of
EDG 'B' to continue to " perform with reasonable assurance or reliability" was uncertain, A
repair under very difficult conditions would have been required to recover the EDG, while
plant operators contended with an emergency. If licensee personnel could not promptly
identify and correct the oil leak, the EDG would have been challenged to perform its safety
function. The failure to establish and maintain an adequate written procedure is a violation of
Technical Specification 6.8.1 (Violation 50-289/93-14-01).
The licensee continues to idendfy instances where Instrument and Controls (I&C) technicians
have not properly returned equipment to service following maintenance or surveillance.
During this report period, due to an Instrumentation and Controls technician error, a post
accident hydrogen monitor analyzer panel function selector switch was left in the wrong
position following conduct of a surveillance test. The safety significance of this incident was
minimal. However, the licensee appropriately included the event in an ongoing root cause
analysis involving I&C technicians not properly returning equipment to service. NRC review
of the results of that evaluation is being trued as an unresolved item. Therefore, the
inspector will evaluate the adequacy of the licensee's corrective actions addressing this event
during review of the unresolved item (Update URI 50-289/93-13-01).
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Engineering
The licensee's corrective action regarding reactor protection system (RPS) channel 'B'
bistables found out-of-tolerance during surveillance testing was adequate to prevent
recurrence. The inspector concluded that the licensee's evaluation of the cause of the out-of-
tolerance condition was thorough and testing conducted to determine if any other RPS
channels were affected was good.
Plant Suonon
On a sampling basis, the inspector verified that radiological surveys were current and that
radiological area postings were consistent with these surveys. The inspectors noted no
discrepancies and concluded that overall radiological controls were good.
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TABLE OF CONTENTS
EXECUTIVE SUMM A RY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . il
1.0 SUMM ARY OF FACILITY ACTIVITIES . . . . . . . . . . . . . . . . . . . . . . . . I
1.1 Licensee Activities ................................. I
1.2 NRC Staff Activities ................................ I
2.0 PLANT OPERATIONS (71707) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
3.0 M AINTENANCE (61726, 62703, 71707) . . . . . . . . . . . . . . . . . . . . . . . . 2
3.1 Maintenance Observations ............................. 2
3.2 Surveillance Observations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
3.3 Emergency Diesel Generator Fuel Oil Ink Due to an Inadequate
Procedure (Violation 50-289/93-14-01) ..................... 3
3.4 Post-Accident Hydrogen Monitor Switch Out of Position (Update URI
50-289/93- 13-01 ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
4.0 ENGINEERING (71707) .................................. 7
4.1 Reactor Protection System Bistables Out-of-Tolerance . . . . . . . . . . . . . 7
5.0 PLANT SUPPORT (71707) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
5.1 Radiological Controls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
5.2 Securi ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
6.0 NRC MANAGEMENT MEETINGS AND OTHER ACTIVITIES ......... 9
6.1 Routine Meetings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
6.2 TMI-l 10R Outage Briefing . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
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DETAILS
1.0 SUMMARY OF FACILITY ACTIVITIES
1.1 Licensee Activities
Unit I remained at 100% power throughout the inspection period.
1.2 NRC Staff Activities
The inspectors assessed the adequacy of licensee activities for reactor safety, safeguards, and
radiation protection, by reviewing information on a sampling basis. The inspectors obtained 2
information through actual observation of licensee activities, interviews with licensee
personnel, and documentation reviews.
The inspectors observed licensee activities during both normal and backshift hours: 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br />
of direct inspection were conducted on backshift. The times of backshift inspection were
adjusted weekly to assure randomness.
2.0 PLANT OPERATIONS (71707)
The inspectors observed overall plant operation and verified that the licensee operated the
plant safely and in accordance with procedures and regulatory requirements. The inspectors
conducted regular tours of the following plant areas:
- Control Room - Auxiliary Building
- Switch Gear Areas - Turbine Building
-- Access Control Points - Intake Structure
- Protected Area Fence Line -- Intermediate Building
-- Fuel Handling Building - Diesel Generator Building
The inspectors observed plant conditions through control room tours to verify proper
alignment of engineered safety features and compliance with Technical Specifications. The
inspectors reviewed facility records and logs to determine if entries were accurate and
identified equipment status or deficiencies. The inspectors conducted detailed walkdowms of
accessible areas to inspect major components and systems for leakage, proper alignment, and
any general condition that might prevent fulfillment of their safety function.
The inspector found that shift turnovers were comprehensive and accurate, and adequately
reflected plant activities and status. Control room operators effectively monitored plant
operating conditions and made necessary adjustments. There was extensive management
involvement in daily activities. The inspector concluded that the licensee conducted overall
plant operations in a safe and conservative manner.
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3.0 MAINTENANCE (61726, 62703, 71707)
3.1 Maintenance Observations
The inspector reviewed selected maintenance activities to assum that: the activity did not
violate Technical Specification Limiting Conditions for Operation and that redundant
components were operable; required approvals and releases had been obtained prior to
commencing work; procedures used for the task were adequate and work was within the skills
of the trade; maintenance technicians were properly qualified; radiological and fire prevention
controls were adequate; and, equipment was properly tested and returned to service.
Maintenance activities reviewed included:
- General Maintenance Procedure 1405-3.2, " Diesel Engine Maintenance."
Job Order Number 070268, "EDG-T-1B Air Start Check Valve Inspection."
Corrective Maintenance Procedure 1410-Y-II, " Threaded Piping and Fitting
Maintenance."
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Overall, the inspectors found that individuals involved in maintenance activities were
knowledgeable and work was conducted using appropriate procedures. Review of the
licensee's activities associated with the diesel generator engine maintenance are described
further in section 3.3.
3.2 Surveillance Observations
The inspectors observed conduct of surveillance tests to verify that approved procedures were
being used, test instrumentation was calibrated, qualified personnel were performing the tests,
and test acceptance criteria were met. The inspectors verified that the surveillance tests had
been properly scheduled and approved by shift supervision prior to performance, control
room operators were knowledgeable about testing in progress, and redundant systems or
components were available for service as required. The inspectors routinely verified adequate
performance of daily surveillance tests including instrument channel checks and reactor
coolant system leakage measurement.
Surveillance activities reviewed included:
Surveillance Procedure 1303-4.1, " Reactor Protection System."
Surveillance Procedure (SP) 1303-4.23, " Channel Test of Reactor Building Post-LOCA
Hydrogen Monitor."
Surveillance Procedure 1301-8.2, " Diesel Generator Annual Inspection."
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Overall, the inspectors found that surveillance activities were performed in a controlled
a manner using appropriate procedures. However, specific concerns regarding testing of the
reactor building post-LOCA hydrogen monitor and testing of the emergency diesel generator
are described in Sections 3.3 and 3.4.
3.3 Emergency Diesel Generator Fuel Oil Leak Due to an Inadequate Procedure
(Violation 50 289/93-14-01)
On July 1,1993, the licensee operated emergency diesel generator (EDG) 'B' for 35 minutes
to warm the diesel to support testing in accordance with Surveillance Procedure " Emergency
Imading Sequence and High Pressure Injection logic Channel / Component Test." After the
licensee secured the diesel, they observed that oil had been deposited throughout the inside of
the fan compartment. The EDGs use air to cool the diesel jacket cooling water and
lubricating oil systems. There is an engine driven fan located at the top of the fan
compartment which draws air past radiators that make up two sides of the fan compartment.
The licensee found that the lubricating oil filter, which is located inside the fan compartment
room was the source of the leaking oil. The licensee found that all bolts on the filter cover
were under-torqued and two bolts were only finger tight. After all ten bolts were torqued to
100 ft-lbs, the licensee operated EDG 'B' again and there was no evidence of oil leakage.
The licensee checked EDG 'A' and found that the bolts were adequately torqued.
The licensee and inspector reviewed Surveillance Procedure 1301-8.2, " Diesel Generator
Annual Inspection," Step 8.2.20.3.f, and General Maintenance Procedure 1405-3.2, " Diesel
Engine Maintenance," Step 8.4.7.1.f, which involve tightening of the filter cover bolts. The
procedures say to tighten the bolts evenly, but do not specify a torque value. The Fairbanks-
Morse vendor manual specifies a torque value of 150 ft-lbs for an 8 bolt filter. Plant
Engineering calculated a torque of 120 ft-lbs for the 10 bolt filter that experienced the
leakage. Past licensee experience demonstrated that 100 ft-lbs was sufficient torque. Plant
Engineering plans to determine what the proper torque is and chcnge SP 1301-8.2 and
GMP 1405-3.2, accordingly.
The Plant Review Group (PRG) met to perform an operability determination for EDG 'B'.
The PRG noted that EDG 'B' was operated for two hours on June 24,1993, for post-
maintenance testing following its annual inspection which began on June 21,1993.
Following the post-maintenance testing, Plant Operations took daily readings inside the
radiator housing and did not notice oil leakage. Based on the reduction in oil level that
occurred on July 1,1993, the licensee estimated that 20 gallons of oil leaked during the
diesel run. The PRG estimated that a low lubricating oil level alarm would be received in
approximately 1/2 hour and that the oil level would have fallen to a critical level (lubricating
oil pump suction) 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> later. The Plant Review Group determined that 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> would
be sufficient time to find and remedy the oil leak and therefore, concluded that the diesel
remained operable and capable of fulfilling its safety function. Since the PRG determined
that the diesel was operable, they did not consider if this event was reportable under
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The inspector reviewed the licensee's operability determination by first evaluating the
licensee's ability to detect the leak while the diesel is operating. The licensee contended that
they would have been alerted to the leakage upon receipt of a low lubricating oil level alarm.
The inspector reviewed the alarm records for July 1,1993, and found that the licensee
rxeived the low level alarm the same minute the diesel was first started and it remained in an
alarm condition throughout the diesel operation. The inspector, as well as plant operators,
have observed that it is not uncommon to receive a low lubricating oil level alarm after the
diesel is first staned. Some engine cavities get refilled with lubricating oil once the diesel is
staned, thereby lowering the lubricating oil sump level. The inspector interviewed five
operators ranging from Shift Supervisor to Auxiliary Operator. All operators indicated that
they do not enter the fan compartment to look for leaks after receipt of a low oil level alarm,
because the alarm is common and they know the reason for the alarm. The operators
indicated that they periodically check the lubricating oil sump dip stick while the alarm is in.
In this incident, the licensee became aware of the lubricating oil leak after securing the
diesel, not by dip stick measurements. Therefore, the inspector determined that the low *
lubricating oil level alarm would not have alerted the licensee to a funher lowering of oil
level caused by the leak. The inspector determined that it is likely that the oil leak would
have been detected through dip stick readings, making the diagnosis of a leak more insightful,
which may have delayed tN pursuit of the leak. The licensee also indicated that if the diesel
was called upon in an emergency, the low lubricating oil level alarm would have been
pursued aggressively, thereby revealing a leak existed. The inspector discussed with Plant
Engineering that it is not uncommon to receive a low lubricating oil level alarm when starting
the EDGs. The inspector was concerned that the operators would become desensitized to the ;
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alarm and not respond to the alarm aggressively, in an emergency. Plant Engineering was
not aware of the problem and agreed to research the history of the lubricating oil level alarm
activity and evaluate changing the setpoint.
The inspector also evaluated the licensee's operability determination by determining what
actions would be necessary to repair the leak. The licensee would have to open the fan i
companment door, which is very difficult due to the suction created by the fan, and they
would then have to find the source of the leak. The maintenance technician would then have
to enter the fan companment to evaluate the leak and conduct the repair. Due to the
operation of the fan, the air velocity inside the housing is very high. Since the lubricating oil
is approximately 180 F and was dispersed in all directions, the maintenance technician would
have to don pro'ective clothing that must remain in place with the high velocity air. The ,
inspector agre.:d that it could be possible to make the repair even under the difficult I
conditions. However, the inspector questioned whether the actions could be considered
reasonable.
The inspector reviewed the NRC guidance regarding operability determinations and
reportability. NUREG 1022 Supplement 1, " Licensee Event Report System," states that a
safety system must operate long enough to complete its intended safety funct on. Reasonable
operator actions to correct minor problems may be considered, however, heroic actions and
unreasonably insightful diagnoses, particularly during stressful situations, should not be
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assumed. Generic Letter 91-18, " Operable / Operability: Ensuring the Functional Capability
of a System or Component," paragraph 3.3, states that in addition to providing the specified
safety function, a system is expected to perform as designed, tested and maintained. When a
system is degraded to a point where it cannot perform with reasonable assurance or
reliability, the system should be judged inoperable, even if at this instantaneous point in time,
the system could provide the specified safety function.
The inspector discussed with the licensee the NRC guidance regarding operability l
determinations and the reasonableness of the required actions the licensee was assuming could l
be accomplished. The licensee told the inspector that they still believed that the repair was
reasonable. On July 27,1993, the PRG held another meeting to discuss the diesel lubricating i
oil leak. The PRG again concluded that since the diesel had performed successfully dunng
the two hour run on June 24,1993, and since there was evidence that no significant ]
lubricating oil leakage had occurred prior to the run on July 1,1993, the diesel was operable ;
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during the period of time, because it continued to demonstrate the ability to perform its
specified function. However, the licensee determined that since uncertainty existed as to f
whether EDG 'B' would have continued to " perform with reasonable assurance or reliability"
based on the presence of the oil leak, and its potential effect on engine lubrication and '
cooling, the event was reponable under 10 CFR 50.73. The licensee submitted Licensee
Event Report 93-006-00 on August 2,1993. j
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The inspector concluded that the licensee's immediate corrective actions in response to this I
event were appropriate. In addition, following discussion with the inspector, the licensee l
appropriately reported the event per 10 CFR 50.73. However, the PRG review of the diesel
low lubricating level alarm as the means of leak detection was weak. The inspector found
that Surveillance Procedure 1301-8.2 was inadequate because it failed to specify a torque
value for the lubricating oil filter cover. As a result, during the period June 24,1993
through July 1,1993, the ability of EDG 'B' to continue to " perform with reasonable
assurance or reliability" was uncertain. A repair under very difficult conditions would have
been required to recover the EDG, while plant operators contended with an emergency. If
licensee personnel could not promptly identify and correct the oil leak, the EDG would have j
been challenged to perform its safety function. The failure to establish and maintain an ;
adequate written procedure is a violation of Technical Specification 6.8.1 (50-289/93-14-01).
3.4 Post-Accident Hydrogen Monitor Switch Out of Position (Update URI
50-289/93-13-01)
During this inspection period, the licensee identified several instances where Instrument and
Controls (I&C) technicians had not properly returned equipment to service following
maintenance or surveillance. There were four instances where I&C technicians did not return
instrument air valves for balance-of-plant equipment to the proper position following
maintenance. These four instances did not affect plant safety. There was one instance that
involved safety related equipment which is describe <1 below.
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On July 19, 1993, an Auxiliary Operator found the 'B' post accident hydrogen monitor
analyzer panel function selector switch in the ' Span' position instead of the required ' Sample'
position. The hydrogen monitor consists of two analyzer panels located in the Intermediate
Building and remote panels located in the Control Building. Both the analyzer panels and the
remote panels have a function selector switch which has three positions,: 'Zero', ' Span', and
' Sample'. In the ' Sample' mode the hydrogen monitor is aligned to the reactor building to
measure the percent hydrogen gas present. In the 'Zero' mode, the hydrogen monitor is
calibrated by isolating the reactor building flow and aligning a calibration gas. In the ' Span'
mode, the hydrogen monitor is calibrated by isolating the reactor building flow and aligning
two calibration gases. The analyzer panels contain a flow indicator to measure air flow
through the hydrogen monitor.
The licensee found that the analyzer panel function selector switch had been mispositioned by
an I&C technician while performing Surveillance Procedure (SP) 1303-4.23 " Channel Test of i
Reactor Building Post-LOCA Hydrogen Monitor." The I&C technician had gone to analyzer ,
panel 'B' because the flow rate rotameter was improperly indicating full flow. The
technician tapped on the rotameter and the flow returned to the normal range. SP 1303-4.23,
Step 8.4.1.2, states that at the remote panel turn the function selector switch to the ' Span'
position and allow the analyzer to stabilize for 45 minutes. In order to minimize the 45
minute wait, since the technician was at the analyzer panel to correct the flow problem, the
technician decided to select the ' Span' position at the analyzer panel rather than the remote
panel. While completing section 8.4, " Span Adjustment," the technician found several
readings to be low and wrote a Surveillance Deficiency Report (SDR). The SDR had the
technician make the necessary repairs in accordance with the technical manual. During these
repairs the technician placed the function selector switch at the remote panel to the ' Span'
position. SP 1303-4.23, Step 8.4.5.2, restores the remote panel to the required alignment by
placing the function selector switch in the ' Sample' position. Since the repairs required the
technician to place the function selector switch in the ' Span' position at the remote panel, he i
could perform this step as written. Therefore, this restoration step did not signal the i
technician to remember that he had placed the function selector switch to ' Span' at the
analyzer panel.
The inspector determined that the cause of this incident was that the I&C technician did not
properly follow SP 1303-4.23. However, the inspector determined that the safety
significance of this incident was minimal, because the Auxiliary Operator, demonstrating
good attention to detail, noted the error within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the comp'etion of the
surveillance. In addition, if the hydrogen monitor ever had to be used, Oprating Procedure
1105-18, " Containment Hydrogen Monitor," Step 3.1.2.7, would require an operator to take
remote control of the hydrogen monitor and select the ' Sample' position.
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The licensee is currently evaluating the root causes for several instances of I&C technicians ]
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not properly returning equipment to service and have expanded their evaluation to include this
event. NRC review of the results of that evaluation is being tracked as unresolved item.
Therefore, the inspector will evaluate the adequacy of the licensee's corrective actions
addressing this event during review of the unresolved item (Update URI 50-289/93-13-01).
4.0 ENGINEERING (71707)
4.1 Reactor Protection System Bistables Out-of-Tolerance
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l On July 14, 1993, while performing Surveillance Procedure 1303-4.1, " Reactor Protection j
System," (RPS) the licensee found that the power / flow / imbalance bistable on channel 'B' was
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slightly outside the minimum allowable tolerance allowed by the procedure (the as found
bistable setting was 7.156 VDC and the tolerance is 7.181 to 7.571 VDC.) When a bistable
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is found to be out of the tolerance band, the surveillance procedure requires the licensee to
recalibrate the parameter string (power / flow / imbalance string) for the channel being tested.
An RPS parameter string consists primarily of the nuclear instrument (NI) detector, the
calibration test module which supplies a test input in place of the NI detector, the linear
i amplifier which has a potentiometer used to adjust the gain based on the heat balance
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calculation and, the RPS bistables which will trip the reactor if the corresponding setpoint is
l reached. During testing, the licensee adjusts the linear amplifier (potentiometer) using the :
! installed dial vernier for a gain of 1 (course gain set to 10 and the fine gain set to 00.) i
During the recalibration and troubleshooting of the power / flux / flow bistable, the licensee
found that for a test input of 10 i 0.001 VDC at the calibration test module, they were
obtaining a voltage of 10.030 VDC downstream of the linear amplifier, which corresponds to
a gain of 1.03. The licensee found that the reason the gain was not I was that the dial
vernier had slipped on the potentiometer shaft 2/10 of a turn. This dial setting is only relied
upon during RPS testing. This dial setting is not relied upon to set the gain of the linear
amplifier following a heat balance because direct voltage measurements at the linear amplifier
are used. I
The licensee replaced the dial vernier and retested RPS channel 'B'. During the retest, the !
licensee found that the high flux bistable tripped at 8.644 VDC which is below the tolerance
band of 8.651 VDC to 8.741 VDC. The licensee then recalibrated this bistable. Although
l the high flux trip bistable and the flux / flow / imbalance bistable were out of their tolerance
band, they were well within the Technical Specification limit and therefore RPS channel 'B'
was never inoperable.
The Plant Review Group met and recommended performing a test of the three remaining RPS
channels to determine if a similar shaft slipping problem existed on other potentiometers.
The licensee prepared Special Temporary Procedure 1-93-12, "RPS Linear Amplifier Dial
Test," to provide the controls to set up for the test, and for post-test restoration. 'Using a test
input of 10 VDC i 0.001 at the calibration test module and the dial vernier set for a gain of
one, the licensee obtained a linear amplifier reading of 10.010 VDC for channel 'A',10.013
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VDC for channel 'C', and 10.005 VDC for channel 'D' which are within the tolerance
required by the surveillance procedure. The licensee found that none of the other three
channels had experienced the dial vernier shaft slippage that was discovered on channel 'B'. !
The inspector reviewed the Bailey RPS vendor manual and verified that no guidance was !
provided on setting the linear amplifier gain during RPS parameter string testing. The j
licensee plans to change Surveillance Procedure 1303-4.1 to measure voltage immediately ;
downstream of the linear amplifier and compare it to the test input voltage rather than relying j
on the dial vernier setting. The licensee indicated that when the new revision is performed j
the first time, bistables may be found out-of-tolerance due to the small errors in gain that j
existed. Based on calculations, the licensee does not expect to find any bistables outside the j
Technical Specification band. .
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The inspector concluded that the licensee's evaluation of the shaft slippage problem was j
thorough and Special Temporary Procedure 1-93-12 was good. The licensee's corrective l
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action was adequate to prevent recurrence.
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5.0 PLANT SUPPORT (71707) i
5.1 Radiological Controls
The inspectors examined work in progress to verify proper implementation of health physics
(HP) procedures and controls. The inspectors monitored ALARA implementation, dosimetry !
and badging, protective clothing use, radiation surveys, radiation protection instrument use, i
and handling of potentially contaminated equipment and materials. In addition, the inspectors ;
observed personnel working in RWP areas and verified compliance with RWP requirements. j
l During routine tours, the inspectors verified a sampling of high radiation area doors to be !
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On a sampling basis, the inspector verified that radiological surveys were current and that j
radiological area postings were consistent with these surveys. The inspectors noted no '
discrepancies and concluded that overall radiological controls were good.
5.2 Security
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The inspectors monitored security activities for compliance with the accepted Security Plan
and associated implementing procedures. The inspectors observed security staffing, operation
of the Central and Secondary Alarm Stations, and licensee checks of vehicles, detection and
assessment aids, and vital area access to verify proper control. On each shift, the inspectors
observed protected area access control and badging procedures. In addition, the inspectors
routinely inspected protected and vital area barriers, compensatory measures, and escort
procedures.
The inspectors concluded that the Security Plan was being properly implemented.
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6.0 NRC MANAGEMENT MEETINGS AND OTHER ACTIVITIES I
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6.1 Routine Meetings
At periodic intervals during this inspection, meetings were held with senior plant management
to discuss licensee activities and areas of concern to the inspectors. At the conclusion of the
reporting period, the resident inspector staff conducted an exit meeting with licensee
management summarizing inspection activities and findings for this report period. Licensee j
j comments concerning the issues in this report were documented in the applicable report ;
'
section. No proprietary information was identified as being included in the report. l
,
6.2 TMI-110R Outage Briefing
On July, 21,1993, the licensee briefed NRC management on planned outage activities, ,
outage staffing, outage management, shutdown risk assessment, and various other outage
i
topics. The licensee's slide presentation is provided as an attachment to this inspection
report.
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TMI-110R OUTAGE BRIEFING
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JULY 21,1993
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l TMI-110R OUTAGE
JULY 21,1993
!
AGENDA
-
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l. INTRODUCTION T. G. BROUGHTON
!
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11. OUTAGE PREPARATION / SCOPE R. P. ADAMIAK
111. OUTAGE STAFFING L.M.ZUBEY
IV. OUTAGE MANAGEMENT M.J.ROSS
V. RADIOLOGICAL ISSUES W. E. POTTS
VI. SHUTDOWN RISK ASSESSMENT H.C.CRAWFORD
Vll. OTSGISSUES R. O. BARLEY i
IX. FUELS ISSUES R. T. TROPASSO
X. 9R OUTAGE EXPERIENCE M. A. NELSON
XI. SUMMARY T. G. BROUGHTON
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1. INTRODUCTION
T. G. BROUGHTON
(NO SLIDES)
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II. OUTAGE PREPARATION / SCOPE
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R. P. ADAMIAK
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TMI-110R OUTAGE
j KEY OUTAGE PHASES
.,
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LONG RANGE PLANNING / BUDGET
- !
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) . PREOUTAGE MILESTONES
j . PLAN-OF-PLANNING
.
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. RAMP UP PLAN
. INTEGRATED SCHEDULE
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OUTAGE CLOSEOUT (PREREQUISITE PROGRAM)
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f OUTAGE CRITIQUE / REPORT
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! TMI-110R OUTAGE
,
PREOUTAGE MILESTONES
.
,
. OUTAGE READINESS 1 MONTH PRIOR TO OUTAGE
START (AUGUST 1,1993)
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. TOTAL MILESTONES - PREOUTAGE 48
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l . CATEGORIES: l
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BUDGET 1
-
PLANNING 11 l
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ENGINEERING 5
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SCHEDULING 2
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MATERIAL 7
( -
CONTRACTS 4
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LICENSING 2
-
ADMINISTRATION 13
l FACILITIES
1
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I- . EXAMPLES:
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ENGINEERING RELEASED TO CONSTRUCTION
(OUTAGE READINESS - 6 MONTHS)
I~
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CONSTRUCTION PLANNING COMPLETE
.
(OUTAGE READINESS - 4 MONTHS)
-
ISSUE INTEGRATED SCHEDULE
(OUTAGE READINESS - 3 MONTHS)
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! OUTAGE GOALS
- . RADIOLOGICAL SAFETY
-
EXPOSURE
- SKIN CONS
- CONTAMINATED AREA
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TMI-110R OUTAGE STATISTICS ;
9R VS.10R ;
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CATEGORY 9R 10R
ACTUALS FORECAST
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OUTAGE DURATION (DAYS) 48 38-45
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- PREOUTAGE MILESTONES 20 48
. NUMBER OF PROJECTS 77 53
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. NUMBER OF MAINTENANCE TASKS 1,800 1,400 (+200 Growth)
. MANHOURS 249,000 199,000 (Inc.
. Contingency)
- SUSTAINED DIRECT LABOR PEAK 620 870
- ESTIMATED MAN REM 201 175
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MAJOR WORK ITEMS
PROJECTS / MAJOR MAINTENANCE
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REACTOR REFUELING (CORE OFF LOAD AND UT)
- . ILRT - FRONT END OF OUTAGE
- . OTSG EDDY CURRENT INSPECTION
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. OTSG TUBE SLEEVING - 123 'A' OTSG; 128 'B' OTSG l
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. ISI INSPECTIONS
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.
RC-P CONVENTIONAL SEAL REPLACEMENT (2)
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CRDM FLANGE GASKET REPLACEMENT / BORON
INSPECTION - 18 GASKETS PLANNED
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. REACTOR BUILDING PRESERVATION
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AUGMENTED WORKFORCE -
ACTUAL / PLANNED
STATUS AS OF JULY 9,1993 .
DIRECT 2 SUPPORT
10R
ACTUAL
^
PLANNED
1,400
1,200 mae:iess*** wi+-:*m = = + = :: m e + = * = + +=tm:4:en
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- ff$@$$$$$$$$@#$$O@$$$@$$$4Wy:$$$
WEEK BEGINNING
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GENERAL MAINTENANCE CONTRACTOR
8R, 9R,10R COMPARISONS
- DURATION (DAYS)
!
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8R 58
9R 49
10R 38
l
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CRAFT #S
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7 8R . -
373
!
9R .
457
10R . .
l625
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AVG CRAFT EMPLOYMENT (WEEKS)
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11
10R - 5
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_ . _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _____
-
_4 __a, ___________e_ - - _ _ _ _ _ _ _ _ _ _ _ -
- - . - . - . - - - - - . - - _ _ - - - - - - . -.
'
TMI-1 10R OUTAGE
i
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AUGMENTED WORKFORCE
l
UE&C-CATALYTIC
625 .
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C $ AFT SUPPORT STAFF
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NON-CRAFT SUPPORT
i TURBINE CRAFT 134 -
105 OTHER CRAFT
,
197
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TOTAL POPULATION
PEAK W/O 9/20
1
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TMI-1 10R OUTAGE
~
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CRAFT FORECAST
.
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. . . . . . . . . .
- pygggggigD;li 40WIBoARD!! EdDAO;!! !!TMi!fsks5RisNC5D;!
aCRAELy ~
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-]NEED[ lif/2/9 f lljBAVEllERSM ~lMORK$FISi!W ~
UASBESTOS!! WORK 5RSl! if22!! 15! if80/20)! l[60l!
i , IB'OlkERslAKERS!l 73l b h$fShll lho
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[silsdtl51Cl4Spl ji24!! jiMi [56/50] l[75;!!
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INCLUDES UE&C AND TURBINE ONLY
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_ - - - , . _ - - - _ _ . - - . _ - - - - - - . - - - - - _
_....__-.- ___.
_ - _. _ - - . - - - . _ - -. .
-
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GENERAL MAINTENANCE CONTRACTOR .
'
UE&C-CATALYTIC
10R OUTAGE CRAFT SUPPORT STAFF SUMMARY
TMI EXPERIENCE
60%
67 UE&C CA"A_Y"C N -0VSE
'O8.8.1SEC00D
NEW PERSONNEL
i
-
14%
' '
3 AN" SEC00D
_
OC EXPERIENCE
26%
,88 "0"A_ 310s EC"D 8"A=
NOTE: 80% OF PROJECTED STAFF ARE ON SITE OR CONFIRMED 7/2/93
4
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.
.
,
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10 R OUTAGE TRAINING PLAN ~
- * _IN PROCESSING - (NEW HlRE - 5 DAYS, EXPERIENCED WORKER - 2 DAYS)
SECURITY BADGING / PSYCHOLOGICAL SCREENING-MMPl/
j RESPIRATOR TRAINING AND FIT TEST /WHOLE BODY COUNT /
1
G.E.T.- INCLUpES FFD, HEAT STRESS, CONFINED SPACE
HEARING CONSERVATION /RWP/ ADVANCED RAD WORKER TRAINING *
.
- SPECIALTY TRAINING - AS REQUIRED
FIRE WATCH / SWITCHING AND TAGGING / WELDER QUALIFICATIONS /
RIGHT-TO-KNOW/ ASBESTOS & LEAD HANDLING /FLAMMABLES CONTROL
l CRAFT SAFETY TRAINING [TTIS/ MATERIAL TRACEABLITY/ LIFTING & HANDLING /
l . CRANE USE/ SOLDERING, RAVCHEM SPLICING, TUBING & FITTING, PIPE END
i g PREP TOOLING
,
e
'
- JOB SPECIFIC TRAINING -PER INTEGRATED SCHEDULE
l OTSG TUBE SLEEVING AND PLUGGING *
OTSG ROGERS (ROBOTICS)*
RV HEAD REMOVAL / INSTALLATION
. RCP SEALS *
! CRD REBUILDS
MOTOR OPERATED VALVE / BREAKER REBUILD
- JOB SUPERVISORS TRAINING (3 DAYS - 17 MODULES)
- CRAFT FOREMEN ORIENTATION - INCLUDES: LESSONS LEARNED / PLANT
- ORIENTATIONIENTATION
- USE OF MOCKUPC
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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OUTAGE MANAGEMENT
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OUTAGE MANAGEMENT AND CONTROL ORGANIZATION
< . RESEMBLE OPERATING CYCLE
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OUTAGE CONTROL CENTER (OCC) -
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IMPROVED FACILITY (*)
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STAFFED 7 DAYS P/WK,24 HOURS P/ DAY (*)
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ENSURE WORK ACTIVITIES ADHERE TO SCHEDULE
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ENSURE PLANT CONDITIONS, CONTROLS ARE
! .
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ADHERED TO, INCLUDING, COMPLIANCE WITH FUEL
l. PROTECTION CRITERIA
,
-
SHIFT SUPERVISOR LOCATED IN THE OCC
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B&W COORDINATOR
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. HEAD WORK
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" . OTSG WORK
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COORDINATION WITH B&W HOME OFFICE
J' . REPORTS TO OUTAGE MANAGER
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(*) DENOTES CHANGE
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. MAINTENANCE SHIFT COORDINATOR ,
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!. (MAINTENANCE WORKING TWO 10 HR. SHIFTS) .
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- IMPROVED COMMUNICATIONS BACK SHIFT
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(PARTICUI.ARLY IN AREA OF CRITICAL PATH
{, WORK AND WORK IMPACTING SAFETY)
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24 HOUR 7 DAY WEEK (*) ;
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ADDITIONAL CAPABILITIES (*) [
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. LEVEL il SCHEDULE *
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WORK LIST, SYSTEM COMPONENT LISTS, STATUS
REPORT
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IMPROVED QUICK ACCESS TO ALL DATA FOR
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OUTAGE TEAM
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4 OUTAGE MANAGEMENT CONTROL
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! . ISSUED' OUTAGE MANAGEMENT EXPECTATIONS THAT
! PLACES EMPHASIS ON THE FOLLOWING: (*)
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PERSONNEL / EQUIPMENT
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!l ; - QUALITY OF WORK
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. INCREASED EMPHASIS ON FUEL PROTECTION ,
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- e CRITERIA (*)
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DECAY HEAT
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CONTAINMENT CLOSURE
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STRICT ADHERENCE TO INTEGRATED SCHEDULE
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PRE OUTAGE TRAINING
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MEETING TO STATUS AND CONTROL OUTAGE (*)
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KICKOFF DETAILED MEETING ON PRINCIPLES FOR
j FIRST TWO WEEKS OF SCHEDULE
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MEETING AT 0630 AND 1600 EACH DAY
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(*) DENOTES CHANGE
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KEY MANAGEMENT PERSONNEL SHIFT COVERAGE (*)
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i - ASSIGNED TO SHIFT TO ENSURE MANAGEMENT j
I EXPECTATIONS DURING CRITICAL PERIODS
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STA ROUND THE CLOCK COVERAGE
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SENIOR NUCLEAR ENGINEER IN ADDITION TO
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NUCLEAR ENGINEERING STAFF DURING FUEL
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ACTIVITIES
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- CONTROL OF WORK HOURS
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EXTENDED TO BALANCE OF PLANT WORK
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}~ RADIOLOGICAL CONTROLS
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j - STAFFING
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j - STAFF AUGMENTED WITH CONTRACTORS
j - DEDICATED RAD CON TO MAJOR JOBS
. TRAINING
!
l - CONTRACTOR TESTING !
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- ADVANCED RAD WORKER TRAINING
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i, . PLANNING ,
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INTEGRATION OF RAD CON INTO WORK PACKAGES
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INTEGRATED SCHEDULE
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STANDARD SHIELDING PACKAGES r
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IODINE CONTINGENCY WINnnw j
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4 . CONTAMINATION CONTROL
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INCREASED DECONTAMINATION CREWS
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INCREASED RAD CON / TRAINING TO COACH
j WORKERS IN DRESSING AND UNDRESSING
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VALETS TO ASSIST IN DRESSING AND UNDRESSING
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HIGH RISK / SPECIALIZED JOBS ARE MINIMlZED COMPARED
- TO 9R OUTAGE
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RADIOLOGICAL CONTROLS
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- COLLECTIVE DOSE GOAL 175 PERSON-REM (SRD)
j MAJOR ADDITIONAL TASKS
OTSG 30-33
,
TUBE SLEEVING 10-12
i TUBE PLUGGING 10-13
] EDDY CURRENT 5.5-7.5
l MISC. 4-5
l FUNCTIONAL MAINTENANCE 30
- REFUELING 25
l OUTAGE INCREMENT 20
SCAFFOLD 15
j IN SERVICE INSPECTION 12
i
j RCP SEALS 6
l RB PRESERVATION 6
1
1 DOSE ESTIMATE 180-200 PERSON-REM (SRD)
i HIGHEST INDIVIDUAL 1.25 REM (SRD)
i SKIN CONTAMINATION GOAL <8.0/10K RWP-HRS
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V. SHUTDOWN RISK ASSESSMENT ,
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_. . . _ . _ _ . . _ _ _ . . - . . . _ . . _ _ - _ .
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OUTAGE RISK
IMPLEMENTATION ,
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. LITERATURE SEARCH AND REVIEW {
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NRC (AEOD);iNPO;NSAC - SHUTDOWN EVENTS :
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NUMARC 91-06 INDUSTRY GUIDELINE !
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B&WOG - SHARE RESPONSES / APPROACHES i
!
. OUTAGE MANAGEMENT EXPECTATIONS l
l
-
MANAGEMENT STRATEGY ;
-
PLANNING AND SCHEDULING l
-
IMPLEMENTATION
-
CRITIQUE PROCESS '
.
OUTAGE FUEL PROTECTION CRITERIA l
-
PRE-OUTAGE - PLANNERS AND SCHEDULERS ,
'
- PRG COMPARE SCHEDULE
-
OUTAGE - SCHEDULE CHANGES
- SAFETY ASSESSMENT i
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- OUTAGE OVERSIGHT
.
OUTAGE CHECKLIST FOR CONTROL ROOM (OP 1104-4)
.
IMPROVE EMERGENCY PROCEDURES FOR SHUTDOWN
.
DEVELOP SHUTDOWN DISPLAYS
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DEVELOP DAILY STATUS SHEET
.
E-PLAN - REVIEW SHUTDOWN EALS
-
PERFORM TRAINING
.
STA FULL-TIME SHIFT COVERAGE I
- 28 - l
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j OUTAGE RISK- -
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j OUTAGE FUEL PROTECTION CRITERIA
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i . GENERIC CRITERIA FOR ANY OUTAGE < 250 F
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} . IDENTIFY PLANT CONDITIONS
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- 16 IDENTIFIED ,
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10 CONCERNED WITH SHUTDOWN
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! . CRITICAL SAFETY FUNCTION APPROACH FOR EACH PLANT
l CONDITION
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REACTIVITY
l -
HEAT REMOVAL
1 - INVENTORY
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CONTAINMENT '
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ELECTRICAL DISTRIBUTION
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! . DEFINE ADEQUATE SAFETY MARGIN FOR FUEL
i
i -
DESIRED LEVEL OF PROTECTION
i -
MINIMUM LEVEL OF PROTECTION
! -
DIRECTOR LEVEL APPROVAL TO GO TO MIN
l -
IMMEDIATE ACTION IF REDUCED BELOW MIN
!
!
f . IDENTIFY OUTAGE MILESTONES AND ACTIVITIES IN EACH
- PLANT CONDITION
!
1
i . IDENTIFY CONCERNS WHICH SHOULD BE EVALUATED WHILE
.
IN EACH PLANT CONDITION
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- - 29 -
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. . .._ __ _ . ._ __ _ __ _ _ _ _. . . _ _ _ _ _ _ .-. . _ _ . . . . .
. .
.
.
TMI 10R PLANT CONDm0N DURATIONS
1993
Denarn0N IUSK STARF De Sep l Oct
(n Im In l4 li ls
TRANSm0N TO COLD SHUTDOWN (4) AW 9/12 9/18 -
f@UCED NANTOR( (5) AW 9/18 9/18 m
LOW LEVE1/MID LOOP (6) HIQiER 9/18 9/25 muuuuuum
FTC FID00ED (7) LOWER 9/25 9/26 as
HANDUNG IRRADIATED FUEL. (8) AW 9/26 9/30 unumus
DEFUE11D (9) LOWER 9/30 10/3 unus
HANDUNG IRRADIATED & NEW FUEL (10) AW 10/3 10/6 unas
FTC FID00ED (7) LOWER 10/6 10/7 mi
DRAIN /0 ECON FTC (11)_ LOWER 10/7 10/8 in
LOW LEVE1/MID LOOP (12) HIGHER 10/8 10/14
COLD SHUTDOWN REDUCED NENTORY (13) LOWER 10/14 10/17 mm
_ - _ _ - - _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ - _ . _ _ . _ _ . - _ _ _ _ _ _ - _ _ _ _ - __.
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VI. OTSGISSUES
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!' BRIEFING FOR NRC
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l OVERVIEW OF OTSG WORK SCHEDULED
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l . OTSG LEAK TESTING - EXTENDED "48 HR. DRIP TEST"
! - EXPECT TO FIND SEVERAL LEAKING EXPLOSIVE AND
i ROLLED PLUGS LEAKING IN LOWER TUBESHEET
l - PREPARED TO ECT INSPECT LEAKING EXPLOSIVE l
l PLUGS l
! - PREPARED TO TIG RELAX, REMOVE, AND EXAMINE ANY !
i LEAKING ROLLED PLUGS
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- WILL REPAIR ALL LEAKING PLUGS
I l
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! -
ROUTINE TECH. SPEC. TUBE EXAMS
l -
SLEEVING ECT EXAMS
i -
I-600 PLUG ECT EXAMS FOR PLUG CRACKING
! INDICATIONS
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. OTSG TUBE PLUGGING
1
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REPAIR DEFECTIVE TUBES IF NOT PRACTICAL TO
4 SLEEVE
i -
REPLACE PLUGS THAT LEAK OR HAVE PRESSURE
j BOUNDARY ECT INDICATIONS
.
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': . ROBOTICS WILL BE USED FOR OTSG ECT, SLEEVING, AND
l PLUGGING IN ALL 4 HEADS PER ALARA. COLD LEG DAMS
i WILL BE USED'
!
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I OTSG TUBE SLEEVING
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OBJECTIVE: TO PREVENT OTSG LANE / WEDGE TUBE LEAKAGE
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- AND FAILURE DUE TO HIGH CYCLE FATIGUE
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! OTSG LEAKER OUTAGE IN MARCH 1990).
!
! METHOD OF ACHIEVEMENT:
.
i . TO INSTALL 80" B&W MECHANICAL SLEEVES OF INCONEL 690
l MATERIAL IN LANE / WEDGE TUBES USING ROBOTICS (B&W'S
j ROGER). l
. FIRST GROUPING INSTALLED IN 9R SLEEVED HISTORICAL
{ LANE / WEDGE LEAKAGE LOCATIONS AND THOSE
LANE / WEDGE TUBE LOCATIONS SUBJECT TO HIGHEST MAIN
l STEAM CROSS FLOW VELOCITIES IN 9R.
- l
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. SECOND GROUPING WILL COMPLETE OTSG TUBE SLEEVING ,
PROGRAM IN 10R. t
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- PL.ANNED SEQUENCE:
- PRE-SLEEVING TUBE EDDY CURRENT EXAM.
- INSTALL / ROLL IN SLEEVES.
,
l - POST-SLEEVING EDDY CURRENT INSTALLATION
j ACCEPTANCE EXAM (ALSO "AS INSTALLED" ECT
BASELINE). WILL ALSO EXAMINE 250 SLEEVES
j INSTALLED IN 9R.
.
l . OVER 1780 B&W MECHANICAL SLEEVES INSTALLED IN
! OTSGS SINCE 1984. NONE PLIJGGED FOR POST-
! INSTALLATION SLEEVE PROBLEMS.
- 33 -
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Vill. FUEL ISSUES
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R. T. TROPASSO )
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FUEL AND VESSEL INSPECTION
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FULL CORE OFFLOAD / RELOAD l
. FULL CORE UT INSPECTION l
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OWNER / TECHNICAL / USER / HUMAN FACTORS
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COMPONENT IDENTIFICATION WAS STANDARDIZED
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MATERIAL REORGANIZED TO *!!NIMlZE REFERENCING AND
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