Text

Safety Insp Rept 50-289/93-14 on 930622-0731.Violations Noted.Major Areas Inspected:Plant Operations,Maint, Engineering,Radiological Control & Security as Related to Plant Safety
	ML20056F429
Person / Time
Site:	Crane
Issue date:	08/11/1993
From:	Rogge J; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	;
Shared Package
ML20056F425	List: ML20056F424; ML20056F427; ML20056F429;
References
	50-289-93-14, NUDOCS 9308270218
	Download: ML20056F429 (47)
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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

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Approved by:

f[Jr(1 F. Rogge, Chief// 9

(/ Date'

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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.

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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 . . . . . . . . . . . . . . . . . . . . . . . .

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1.1

Licensee Activities

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1.2

NRC Staff Activities

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2.0

PLANT OPERATIONS (71707) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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3.0

M AINTENANCE (61726, 62703, 71707) . . . . . . . . . . . . . . . . . . . . . . . .

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3.1

Maintenance Observations

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3.2

Surveillance Observations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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3.3

Emergency Diesel Generator Fuel Oil Ink Due to an Inadequate

Procedure (Violation 50-289/93-14-01)

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3.4

Post-Accident Hydrogen Monitor Switch Out of Position (Update URI

50-289/93- 13-01 ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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4.0

ENGINEERING (71707)

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4.1

Reactor Protection System Bistables Out-of-Tolerance . . . . . . . . . . . . .

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5.0

PLANT SUPPORT (71707) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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5.1

Radiological Controls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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5.2

Securi ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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6.0

NRC MANAGEMENT MEETINGS AND OTHER ACTIVITIES

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6.1

Routine Meetings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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6.2

TMI-l 10R Outage Briefing . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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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

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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 0.0117 hours 6.944444e-5 weeks 1.5981e-5 months

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

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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 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months later. The Plant Review Group determined that 3.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months 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

10 CFR 50.73.

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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

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

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inspector agre.:d that it could be possible to make the repair even under the difficult

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

determinations and the reasonableness of the required actions the licensee was assuming could

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

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

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lubricating oil leakage had occurred prior to the run on July 1,1993, the diesel was operable

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

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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

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cooling, the event was reponable under 10 CFR 50.73. The licensee submitted Licensee

Event Report 93-006-00 on August 2,1993.

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The inspector concluded that the licensee's immediate corrective actions in response to this

event were appropriate. In addition, following discussion with the inspector, the licensee

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

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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

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Reactor Building Post-LOCA Hydrogen Monitor." The I&C technician had gone to analyzer

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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

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could perform this step as written. Therefore, this restoration step did not signal the

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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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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

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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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On July 14, 1993, while performing Surveillance Procedure 1303-4.1, " Reactor Protection

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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

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bistable setting was 7.156 VDC and the tolerance is 7.181 to 7.571 VDC.) When a bistable

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

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amplifier which has a potentiometer used to adjust the gain based on the heat balance

calculation and, the RPS bistables which will trip the reactor if the corresponding setpoint is

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reached. During testing, the licensee adjusts the linear amplifier (potentiometer) using the

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installed dial vernier for a gain of 1 (course gain set to 10 and the fine gain set to 00.)

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.

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

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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'.

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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

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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

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on the dial vernier setting. The licensee indicated that when the new revision is performed

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the first time, bistables may be found out-of-tolerance due to the small errors in gain that

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existed. Based on calculations, the licensee does not expect to find any bistables outside the

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Technical Specification band.

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The inspector concluded that the licensee's evaluation of the shaft slippage problem was

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thorough and Special Temporary Procedure 1-93-12 was good. The licensee's corrective

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action was adequate to prevent recurrence.

5.0

PLANT SUPPORT (71707)

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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

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and badging, protective clothing use, radiation surveys, radiation protection instrument use,

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and handling of potentially contaminated equipment and materials. In addition, the inspectors

observed personnel working in RWP areas and verified compliance with RWP requirements.

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During routine tours, the inspectors verified a sampling of high radiation area doors to be

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locked as required.

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On a sampling basis, the inspector verified that radiological surveys were current and that

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radiological area postings were consistent with these surveys. The inspectors noted no

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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.

1

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__ _

.-

.

1

.

9

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6.0 NRC MANAGEMENT MEETINGS AND OTHER ACTIVITIES

I

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

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.

,

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

topics. The licensee's slide presentation is provided as an attachment to this inspection

i

report.

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TMI-110R OUTAGE

JULY 21,1993

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INTRODUCTION

T. G. BROUGHTON

!

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

i

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OTSGISSUES

R. O. BARLEY

IX.

FUELS ISSUES

R. T. TROPASSO

X.

9R OUTAGE EXPERIENCE

M. A. NELSON

XI.

SUMMARY

T. G. BROUGHTON

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KEY OUTAGE PHASES

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PREOUTAGE MILESTONES

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START (AUGUST 1,1993)

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OUTAGE GOALS

RADIOLOGICAL SAFETY

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EXPOSURE

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LESSONS LEARNED FROM PREVIOUS OUTAGES

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REACTOR TRIPS AT RESTART

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FULL POWER OPERATION

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8R, 9R,10R COMPARISONS

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UE&C-CATALYTIC

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TMI EXPERIENCE

60%

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NEW PERSONNEL

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NOTE: 80% OF PROJECTED STAFF ARE ON SITE OR CONFIRMED 7/2/93

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CRAFT SAFETY TRAINING [TTIS/ MATERIAL TRACEABLITY/ LIFTING & HANDLING /

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JOB SUPERVISORS TRAINING (3 DAYS - 17 MODULES)

CRAFT FOREMEN ORIENTATION - INCLUDES: LESSONS LEARNED / PLANT

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USE OF MOCKUPC

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OUTAGE MANAGEMENT CONTROL

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WESTINGHOUSE LEAD TEST ASSEMBLY INSPECTION

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CORE DESIGN

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GADOLINIA FUEL RODS

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MARK-B9 FUEL DESIGN

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NO NEUTRON SOURCES

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CONTROL ROD CHANGES

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9R OUTAGE EXPERIENCE

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M. A. NELSON

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SPECIAL REVIEW OF SELECTED INFREQUENTLY

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PERFORMED / CRITICAL EVOLUTION PROCEDURES

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MANAGEMENT DECISION FOR SPECIAL REVIEW

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TABLE-TOP VALIDATION OF PROCEDURES

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OWNER / TECHNICAL / USER / HUMAN FACTORS

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37 PROCEDURES REVIEWED AND REVISED

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NO MAJOR TECHNICAL PROBLEMS IDENTIFIED

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NUMEROUS MINOR PROBLEMS CORRECTED

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EXAMPLES OF TYPICAL IMPROVEMENTS

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STEPS WITH MULTIPLE ACilONS WERE SPLIT

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COMPONENT IDENTIFICATION WAS STANDARDIZED

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ACTIONS IN NOTES AND CAUTIONS WERE MADE INTO STEPS

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MATERIAL REORGANIZED TO *!!NIMlZE REFERENCING AND

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SUMMARY

T. G. BROUGHTON

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ML20056F429

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