ML17250B278

Insp Rept 50-244/93-10 on Stated Date.Violations Noted. Major Areas Inspected:Plant Operations,Radiological Controls,Maintenance & Surveillance,Security,Engineering & Technical Support

ML17250B278
Person / Time
Site:	Ginna
Issue date:	06/22/1993
From:	Lazarus W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML17250B273	List: ML17250B272 ML17250B275 ML17250B278
References
50-244-93-10, NUDOCS 9306300191
Download: ML17250B278 (14)

See also: IR 05000244/1993010

Text

U. S. NUCLEARREGULATORY COMMISSION

REGION I

License: DPR-1S

Inspection Report 50-244/93-10

R. E. Ginna Nuclear Power Plant

Rochester

Gas and Electric Corporation (RG&E)

Inspection:

Inspectors:

Approved by:

April 23 through June 9, 1993

T. A. Moslak, Senior Resident Inspector, Ginna-

E. C.

utson

Resid'ent Inspector, Ginna

wiz~A>

W.

s

C

ef, Reactor Projects Section 3B

INSPECTION SCOPE

Date

Plant operations, radiological controls, maintenance/surveillance,

security, engineering/ technical

support, and safety assessment/quality

verification.

INSPECTION OVERVIEW

y~hi:Add-dy l llg

g

gl

d.R

g,ydyl

'g,

and power ascension

were well controlled.

On May 7, 1993, a controlled power reduction to

42 percent was performed to support maintenance

on a main feedwater pump.

Full power

operation

(approximately

98 percent)

resumed

on May 10,

1993,

and continued for the

remainder of the report period.

Operators demonstrated

attention-to-detail by identifying off-

normal plant conditions.

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d

d 'g

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operations were conscientiously implemented.

ALARAcontrols for work to repair a leak on

a steam generator blowdown drain line were noteworthy.

Maintenance/

rveillance:

Rework to correct excessive

seal leakoff from the "A" main

feedwater

pump

and

replacement

of the "A" service

water pump

was

required

due to

procurement program deficiencies.

However, the root cause analysis performed by the licensee

was thorough.

5ggli~:

No deficiencies

were identified during routine observations of security program

implementation.

En ineerin /Technical

Good engineering support was observed in continuing efforts

to correct feedwater control system instability.

fet

A

essment/

ualit

Verifi tion:

Failure to provide adequate

containment of sparks

generated

by metal grinding during maintenance

was an apparent violation of fire protection

program requirements.

9306300i9i

930624

PDR

ADOCK 05000244

8

PDR

TABLEOF CON'ANTS

VERVIEW

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TABLE OF CONTENTS

1.0

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

1.1

Operational Experiences

1.2

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Control ofOperations..............

1.3

Observations During Power Changes

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2.0

RADIOLOGICALCONTROLS (71707).......................

2.1

Routine Observations ..............................

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3.0

. MAINTENANCE/SURVEILLANCE(62703, 61726)...........

3.1

Corrective Maintenance

3.1.1

"A" Main Feedwater Pump Excessive Seal Leakoff...

3.1.2

Steam Leak Downstream of Valve 5706 in Containment

3.1.3

Pressurizer Power Operated Relief Valve Seat Leakage

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3.2

Surveillance Test Observations

3.3

Engineered Safety Features System Walkdown (71710)

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4.0

-. SECURITY (71707)

4.1

Routine Observations ........

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5.0

ENGINEERING/TECHNICALSUPPORT (71707, 92701)

5.1

"A" Main Feedwater Regulating Valve Oscillations

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6.0

SAFETY ASSESSMENT/QUALITY VERIFICATION (90712, 90713, 92701

40500)

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6.1

"A" Service Water Pump Replacement

6.2

Inadequate Fire Prevention Measures During Maintenance

6.3

Quality Assurance/Quality Control (QA/QC) Subcommittee Meeting

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6.4

Erosion/Corrosion (E/C) Integrated Management Team Meeting....

6.5

Corporate Management Changes

6..6

Perlodlc Reports.................................

6.7

Licensee Event Reports.............................

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7.0

ADMINISTRATIVE(71707, 30702, 94600)

7.1

Backshift and Deep Backshift Inspection

7..2

Exit Meetings

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DETAILS

1.0

PLANT OPERATIONS (71707)

1.1

Operational Experiences

At the beginning of the inspection period, the plant was nearing completion of the annual

refueling

and

maintenance

outage.

Reactor

coolant

system

heatup

to normal

operating

temperature (547'F) was completed on April24.

Operational testing in hot shutdown included

main steam

safety valve and control rod drop testing.

All requirements

for startup were

completed on April25, and criticality was achieved at 12:11 p.m.

Startup physics testing was

completed on April25 at 11:33 p.m. Steam plant startup was conducted and the main generator

was closed on the grid at 11:56 p.m. on.April26. Full power, approximately 98 percent, was

reached at 2:01 p.m. on May 1. On May 7, a controlled power reduction to approximately 42

percent was conducted to support corrective maintenance on the "A"main feedwater pump. Full

power operation resumed on May 10, and continued for the remainder of the inspection period.

No significant operational challenges occurred during the inspection period.

1.2

Control of Operations

Overall, the inspectors

found the R.- E. Ginna Nuclear Power plant to be operated

safely.

Control room staffing was as required.

Operators exercised control over access

to the control

room.

Shift supervisors

maintained authority over activities and provided detailed turnover

briefings to relief crews.

Operators adhered to approved procedures

and were knowledgeable

of off-normal plant conditions.

The inspectors reviewed control room log books for activities

and trends,

observed

recorder

traces for abnormalities,.assessed

compliance with technical

specifications,

and verified equipment availability was consistent with the requirements

for

existing plant conditions.

During normal work hours and on backshifts, accessible

areas of the

plant were toured.

No operational inadequacies or concerns were identified.

1.3

Observations During Power Changes

The inspectors observed portions of the reactor startup, physics testing, and power ascension

upon completion of the refueling outage,

as well as the power reduction

and escalation

performed in support of main feedwater pump maintenance.

These operations were conducted

in a professional

manner.

Communications

and supervisory involvement-were noteworthy

'trengths.

No deficiencies were noted.

2.0

RADIOLOGICALCONTROLS (71707)

2.1

Routine Observations

The inspectors periodically confirmed that radiation work permits were effectively implemented,

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dosimetry was correctly worn in controlled areas

and dosimeter

readings

were accurately

recorded, access to high radiation areas was adequately controlled, survey information was kept

current, and postings and labeling were in compliance with regulatory requirements.

Through

observations ofongoing activities and discussions with plant personnel, the inspectors concluded

that radiological controls were conscientiously implemented.

'articularly noteworthy was the planning and coordination to minimize worker exposure when

troubleshooting the steam leak downstream ofvalve 5706. Valve 5706 is located in containment

and is in a high radiation area (10-12 rem/hour) during power operations.

Prior to permitting

personnel entry into this area, the radiological control personnel

assured

that personnel were

equipped with the proper dosimetry, aware of low dose waiting areas,

and briefed on the scope

and nature of individual assignments.

3.0

MAINTENANCE/SURVEILLANCE(62703, 61726)

3.1

Corrective Maintenance

3.1.1

"A"Main Feedwater Pump Excessive Seal Leakoff

A major overhaul was performed on the "A" main feedwater pump (MFP) during the 1993

refueling outage.

On April 26, following pump startup during plant power escalation,

seal

leakoff from the "A"MFP was observed to be excessive.

This leakoff, nominally five gallons

per minute (gpm), is directed to a seal water drain tank and pumped back to the main condenser.

Seal water leakoff from the "A" MFP, however, was estimated to be approximately 60 gpm.

This exceeded

the capacity of the seal water drain tank pump, and the excess flow (estimated

at 20-30 gpm) was dumping from the tank overflow to a floor drain.

On

May 7,

1993,

the

licensee

commenced

a controlled

power

reduction

to

support

troubleshooting and repair of the "A" MFP.

When power had been reduced to less than 50

percent, the pump was secured,

cooled down, and disassembled

for inspection.

The cause of

the excessive

seal leakoff was found to be that the bolts that attach the breakdown bushing to

the pump casing were improperly sized.

The function of the breakdown bushing is similar to

that of a labrynth seal.

The hold-down bolts for the breakdown bushing were found to be too

long and prevented

the bushing from seating

against the pump casing.

This allowed high

pressure water inside the pump to be directed to the outside of the bushing, thus bypassing the

pressure

reducing channels

and traveling directly out the shaft penetration in addition to the

normal seal leakoff.

Investigation revealed that the improperly sized hold down bolts had been part of a breakdown

bushing replacement kit that had been installed during the recent pump overhaul.

The problem

had not been obvious during the overhaul reassembly,

because the bushing was brought nearly

in contact with the casing (0.017 inch clearance with the hold down bolts bottomed out in the

casing), and an 0-ring seal at the opposite end of the breakdown bushing absorbed the resultant

slight amount ofplay. No other significant problems were discovered during pump disassembly

and inspection.

The original breakdown bushing hold down bolts were subsequently

used in

pump reassembly.

The pump was returned to service on May 10, 1993.

A licensee performed root cause analysis revealed that procurement of improperly sized bolts

had caused the problem. This error, in turn, was attributed to 1) inaccuracies

in the vendor's

manual,

and 2) a change from procurement by individual part numbers to procurement by

assembly number (which does not specify the length of the hold down bolts). The licensee took

appropriate corrective actions which included returning to the practice of ordering replacement

parts by individual part numbers, and upgrading the vendor technical manual.

The inspector observed portions of the maintenance activities conducted on the "A"MFP and

noted no deficiencies.

The inspector reviewed the maintenance work package, generated under

work order 9320797,

"Feedwater Pump A - Repair Excessive

Seal Leakage."

Work was

conducted in accordance with maintenance procedure M-11.24, "A"Main Feed Pump Inspection

and Maintenance," revision 8, effective date March 30, 1991. The inspector concluded that the

work package adequately directed and documented, this maintenance and had no further questions

in this area.-,

3.1.2

Steam Leak Downstream of Valve 5706 in Containment

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On April30, 1993, while viewing the containment video monitor, on-shift operations personnel

observed intermittent wisps of steam in the vicinityof the "B" reactor coolant pump.

Although

the source of leakage could not be directly observed,

containment leakage detection

systems

confirmed that it was small in size, and containment air particulate monitor activity confirmed

that it was not reactor coolant system leakage.

Subsequent

containment entry determined that

the leak was from a pipe cap downstream ofvalve V-5706 (steam generator "B" blowdown line

drain valve).

Due to high radiation levels (10-12 rem/hour) in the area while at full power,

repair efforts were deferred for one week to coincide with the power reduction that was required

for the main feedwater pump repair (see section 3.1.1). Temporary repair using a special'clamp

assembly injected with a sealant material was successful in stopping the leak. Further corrective

action willrequire cold shutdown conditions.

The inspector observed

the steam leakage

as viewed on the control room video monitor and

concluded

that operations

personnel

had been very alert to have noted the condition.

The

inspector considered that deferring complete repair on the basis of ALARAconsiderations

was

appropriate,

and had no additional concerns on this matter.

3.1.3

Pressurizer Power Operated Relief Valve Seat Leakage

At 9:56 a.m. on April 25, 1993, main control board annunciator F-19, "Pressurizer PORV

[power operated relief valve] Outlet High Temperature 145'F," alarmed.

The PORVs (valves

PCV-430 and -431C) are designed to automatically open on abnormally high reactor coolant

system

(RCS) pressure,

directing coolant to the pressurizer

relief tank (PRT) and thereby

reducing pressure.

The purpose ofannunciator F-19 is to alert operators that actuation ofeither

or both of the PORVs has occurred.

Coolant leakage past the seats of these valves can also

produce sufficiently high outlet temperature to cause the alarm.

The PORVs do not serve a

safety

function during hot shutdown or at-power

operations.

In these

conditions,

RCS

overpressure

protection is provided by the two pressurizer

safety valves, V-434 and -435.

In

case of failure to close or excessive

seat leakage,

the PORVs can be isolated by individual

upstream block valves (MOV-516 and -515, respectively).

Operators responded by performing the appropriate portions ofalarm response procedure AR-F-

19, "Przr PORV Outlet High Temp 145'F."

PRT temperature,

pressure,

and level were

essentially constant, indicating that a PORV had not actually lifted. However, PRT temperature

was observed to be rising slowly (approximately 10'F in the first 15 minutes), suggesting that

seat leakage had developed from one of the valves.

PORV seat leakage had not been a problem

prior to the refueling outage and no maintenance

had been conducted on the valves during the

outage.

Determination of which valve was leaking was complicated by the fact that the valves

are located in the pressurizer

cubicle in the containment building (making direct inspection

difficult), as well as by the small rate of leakage (calculated to be approximately 0.025 gallons

per minute).

Due to the piping configuration and the low leakage rate, conduction of heat

between the two discharge lines allowed their temperatures to equalize and thus precluded using

PORV downstream temperature to identify which valve was leaking.

Operators attempted to eliminate the leakage by cycling the PORVs (one at a time) with their

associated block valves shut.

In addition to possibly achieving positive seating simply through

the act of being cycled, it was reasoned

that the small amount of coolant flow that would result

from depressurizing

the piping between the block valve and the PORV might clear any material

that could be fouling the seating surface.

This action, however, did not correct the

problem.'ubsequently,

a containment

entry was made to attempt to identify the cause of the high

temperature alarm by direct inspection.

Since leakage through the pressurizer

safety valve's

loop seal drain valves could, by conduction through common piping, raise temperature enough

to cause

the alarm,

these valves were verified to be fully shut.

Results of ultrasonic and

thermographic inspections in the area of the PORVs were inconclusive.

Through discussions

with the valve vendor, it was determined that proper seating may be achieved by reorienting the

valve seat and disc.

This technique was attempted on PCV-430 by rotating the valve stem/disc

90 degrees relative to the seat.

The results, however, were inconclusive, and the technique was

not attempted on the other PORV.

Operators closed the block valves in an attempt to stop the leakage.

Best success

was achieved

by maintaining the block valve for PCV-431C closed.

Although this configuration maint'ained

normal PORV outlet temperature

most of the time, the intermittent high temperature

alarm

condition continued to develop.

As of the close of the inspection period, the licensee was operating with the block valve for

PCV-431C

shut.

The

source of leakage

remained

undetermined,

although

inability to

consistently clear the PORV outlet high temperature alarm by closing both block valves suggests

that all four valves are contributing.

This leakage

does not constitute a direct operational

concern; it is contained within the system designed to collect PORV discharges

and the rate of

leakage is very low. The PORVs are not used during normal plant operations and, ifrequired

due to off-normal conditions, restoration is adequately

directed by the applicable emergency

operating procedures.

At-power operation with one (or both) PORV isolated (that is, with its

respective block valve shut) does not present a safety concern.

This configuration is allowed

by Technical Specifications, based on the requirement for both pressurizer

safety valves to be

operable.

The licensee will continue to monitor this condition and will inspect both PORVs

during the next annual outage.

The inspector considered

these actions to be appropriate.

3.2

Surveillance Test Observations

Inspectors

observed

portions of surveillance

tests

to verify proper

calibration of test

instrumentation,

use of approved procedures,

performance of work by qualified personnel,

conformance

to limiting conditions for operation

(LCOs), and correct

system

restoration

following testing.

The following surveillances were observed:

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Performance Test (PT)-34.0, "Startup Physics Test Program," revision 27, procedure

change notices (PCNs) 93T-591, 599, effective date April 16, 1993, observed April25,

1993

P

PT-34.1, "InitialCriticality and ARO [all rods out] Boron Concentration," revision 18,

PCN 93T-576, effective date April 16, 1993, observed April25, 1993

PT-12.2, "Emergency Diesel Generator 1B," revision 73, effective date May 27, 1993,

observed June 2, 1993

"The inspector determined

through observing this testing that operations

and test personnel

adhered to procedures,

test results and equipment operating parameters

met acceptance criteria,

and redundant equipment was available for emergency operation.

3.3

Engineered Safety Features System Walkdown (71710)

The inspector conducted a detailed walkdown of a representative

sample of accessible portions

of the safety injection (Sl) and containment spray (CS) systems.

Primary emphasis was placed

on inspection of system material conditions for items that might degrade plant performance.

Items examined included installation of hangers and supports, housekeeping,

material condition

of valves, correct valve positions, and component labeling.

No safety-significant deficiencies

were noted.

The inspector concluded that the material condition of the SI and CS systems was

satisfactory and verified that accessible portions of the systems were properly aligned.

4.0

SECURITY (71707)

4.1

Routine Observations

During this inspection period, the resident inspectors verified that x-ray machines and metal and

explosive detectors were operable, protected area and vital area barriers were well maintained,

personnel were properly badged for unescorted or escorted access,

and compensatory

measures

were implemented when necessary.

No unacceptable

conditions were identified.

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6.

5.0

ENGINEERING/TECHNICALSUPPORT (71707, 92701)

5.1

"A"Main Feedwater Regulating Valve Oscillations

Inspection report 50-244/92-08 discussed

advanced digital feedwater control system (ADFCS)

instabilities experienced

during the automatic mode of operation:

This problem was first

encountered

after startup from the 1992 refueling outage and, despite some improvement as a

result of troubleshooting efforts, persisted

throughout the operating cycle.

During the 1993

refueling outage, the "B" feedwater regulating valve was overhauled and the associated

control

air system was modified as further attempts to improve ADFCS stability.

During power ascension, ADFCS demonstrated good stability in controlling "B" steam generator

(SG), but developed increasingly poor characteristics

in controlling feedwater flow and water

level in the "A" SG. At full power, oscillations in feedwater flow on the order of 10 percent

were occurring over a period of several minutes.

Although normally stable within these limits,

occasionally the oscillations would rapidly increase in frequency and magnitude, to the point that

operator action was required to regain control.

Extensive troubleshooting

has been performed to determine the cause of ADFCS instability in

controlling the "A" feedwater regulating valve.

Actions have included feedwater regulating

valve lubrication and packing

adjustments,

ADFCS gain changes,

and

extensive

system

performance monitoring. It was found that satisfactory system performance could be achieved

by operating with the "A" feedwater regulating valve in manual control and its bypass valve in

automatic control. While this mode of operation adequately controls feedwater flow variations

that occur under steady state conditions, operator action would be required to maintain steam

generator water level during plant transients.

The inspector observed good engineering support and involvement in the efforts to troubleshoot

ADFCS instability.

6.0

SAFETY ASSESSM<22fT/QUALITYVERIFICATION(90712, 90713, 92701, 40500)

6.1

"A" Service Water Pump Replacement

As discussed in inspection report 50-244/93-03, the cause of excessive load on the "A" service

water pump motor was a suspected problem with the pump. During the refueling outage, testing

with the "D" service water pump motor installed on the "A", service water pump confirmed this.

The "A"service water pump was subsequently replaced.

Post-maintenance

testing demonstrated

that both motor and pump parameters

were within their expected values.

During examination

of the original pump internals, the licensee determined that the impeller first stage discharge was

oversized in diameter (1P/8 inches, as opposed to the required 10'A inches).

This impeller had been installed in February 1992, during a pump rebuild. The deficiency had

not been identified at that time,because

the procedure

did not require measurement

of the

impeller discharge.

The deficiency had not been identified during licensee receipt inspection for

the same reason.

The inspector noted that the licensee had specified the diameter of the first

stage impeller discharge on the purchase order for the impeller in question.

Based on the part

'umber,

the correct part for a Worthington service water pump had been supplied; however,

Ginna service water pump impellers had been modified in 1970, with the discharge diameter

being reduced from 11~/8 inches to 10% inches to reduce pump output.

As of the end ofthe inspection period, the licensee was evaluating this situation for 10 CFR Part

21 applicability. Corrective action included modifying both QA receipt inspection and applicable

maintenance procedures to include measurement ofimpeller discharge diameters.

The inspector

concluded that the licensee's

actions in troubleshooting problems with the "A" service water

pump had been thorough and that corrective actions for problems identified in the course of this

activity were appropriate.

6.2

Inadequate Fire Prevention Measures During Maintenance

On June 7, 1993, while in the intermediate building as part of a routine plant tour, the inspector

observed

sparks entering the intermediate building from underneath

an infrequent access

fire

door to the turbine building. The door, S-36, is located approximately three feet from 1) the

reactor trip breaker panel and 2) one of the two rod control power cabinets.

The inspector

considered that a metal grinding operation in the turbine building was the most likely cause of

the sparks.

The inspector noted that no measures

were in place to contain the sparks and that

no fire watch was present.

The inspector proceeded

to the turbine building via one of the

normal access

doors to investigate the source of the sparks.

He determined that grinding had

been in progress

at about the base of the door in question,

but that this work had been

completed.

The inspector informed licensee

management

and operations

personnel of this

observation.

As a result,

the licensee promptly stationed

a fire watch in the intermediate

building in the vicinityof door S-36.

Although no work was in progress, fire watch coverage

is required for 30 minutes following the completion of work in accordance with administrative

procedure A-905, "Open Flame, Welding and Grinding Permit (Hot Work Permit)."

Subsequent investigation revealed that the work, welding and grinding a shim on the door sill,

had been part of a maintenance package to replace door S-36. A welding and grinding permit

had been obtained in accordance with procedure A-905. The inspector noted, however, that the

requirements of this procedure

had not been effectively implemented.

Specifically, step 3.3

requires that the work area be inspected

by Fire Protection and Safety personnel prior to

commencement ofwork. Among the conditions to be verified by this inspection is that openings

into adjoining areas within a thirty-fivefoot radius be closed and/or that a second fire watch be

established.

Although door S-36 was to remain closed throughout the welding and grinding

operations,

compensatory

requirements for the gap between the door and the sill (i.e., closure

with fire retardant cloth on the intermediate building side or posting of an additional fire watch

in the intermediate building) were not specifically addressed.

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Technical specification 6.8.1 states that, "Written procedures shall established, implemented, and

maintained covering...fire protection program implementation."

In that procedure A-905 is an

implementing procedure of the fire protection program, failure to satisfy the requirements ofA-

905 prior to commencing work on door S-36 represents

a violation of technical specification 6.8.1 (50-244/93-10-01).

6.3

Quality Assurance/Quality Control (QA/QC) Subcommittee Meeting

On June

8,

1993,

the inspector

attended

the quarterly

meeting of the RG&E QA/QC

subcommittee.

The meeting addressed

the overall status, findings, and trends identified in audit

programs,

surveillance

programs,

and the various corrective action systems.

- Substantial

progress has been made in establishing cause codes to facilitate consolidation and analysis ofdata

gathered through the corrective action systems..

Using this approach,

a detailed analysis was

performed by the Quality Performance department of Audit Finding Corrective Action Reports,

Corrective Action Reports,

Identified Deficiency Reports,

Non-conformance

Reports,

QA

Observation Reports, and QA Surveillance Observation Reports compiled from January

1, 1992

through March 31', 1993.

The results of this analysis provided definite guidance as to where

management

attention could be directed to reduce negative trends,

Through

this attendance,

the inspector concluded

that the licensee's

Quality Performance

organization is continuing to refine its techniques of performing safety assessments

and quality

verification to increase its effectiveness

as a management tool.

6.4

Erosion/Corrosion (E/C) Integrated Management Team Meeting

On June 1, 1993, the inspector attended the quarterly meeting ofthe E/C Integrated Management

Team.

Attendees

included

management

representatives

from corporate

engineering,

site

maintenance,

site

technical

engineering,

operations,

chemistry,

and

materials

inspection

departments.

The

meeting

agenda

addressed

the

scope

and

results

of nondestructive

examinations performed on large and small bore piping, elbows,

and tees during the 1993

outage, revisions to procedures implementing the E/C program, and proposed E/C activities for

the 1994 outage.

The inspector concluded that the licensee is actively carrying out an E/C program to identify and

correct potential pipe thinning problems to improve plant safety and reliability.

6.5

Corporate Management

Changes

Effective June

14,

1993, Mr. Will McCoy relinquished his responsibilities

as Department

Manager, Quality Performance, to serve on a Corporate Strategic Task Force addressing future

RG&E business

strategies.

In his absence,

the following organizational

changes

have been

made:

Thomas A. Marlow, formerly Superintendent,

Ginna Production, became Department

Manager, Quality Performance

Richard A Marchionda,

formerly Superintendent,

Support Services,

Ginna, became

Superintendent,

Ginna Production

Steven T. Adams, formerly Technical Manager, Ginna, became Superintendent,

Support

Services, Ginna

Terry R. Schuler, formerly Operations Manager, Ginna, became Technical Manager,

Ginna

Terry A. White, formerly Operations Assistant, Ginna, became Operations Manager,

Ginna

6.6

Periodic Reports

Periodic reports

submitted by the licensee

pursuant

to Technical Specification 6.9.1 were

reviewed.

Inspectors verified that the reports contained information required by the NRC, that

test

results

and/or

supporting

information were

consistent

with design

predictions

and

performance specifications,

and that reported information was accurate.

The following report

was reviewed:

Monthly Operating Report for April, 1993

No unacceptable

conditions were identified.

6.7

Licensee Event Reports

~

~

A hcensee event report (LER) submitted to the NRC was reviewed to determine whether details

were clearly reported, causes were properly identified, and corrective actions were appropriate.

The inspectors also assessed

whether potential safety consequences

were properly evaluated,

generic

implications were

indicated,

events

warranted

onsite

follow-up, and

applicable

requirements of 10 CFR 50.72 were met.

The following LER was reviewed (Note: date indicated is event date):

~

93-002, Steam Generator Tube Degradation Due To IGA/SCC, Causes Quality Assurance

Manual Reportable Limits To Be Reached (April 4, 1993)

The inspector concluded

that the LER was accurate

and met regulatory requirements.

No

unacceptable

conditions were identified.

7.0

ADMIMSTRATIVE(71707, 30702, 94600)

7.1

Backshift and Deep Backshift hispection

During this inspection period, a backshift inspection was conducted on May 7, 1993.

Deep

backshift inspections were conducted on the followingdates: April24, 25,.May 2, 8, 9, 29, 31,

and June 6, 1993.

7.2

Reit Meetings

10

At periodic intervals and at the conclusion of the inspection, meetings were held with senior

station management

to discuss the scope and findings of inspections.

The exit meeting for

inspection report 50-244/93-09 (emergency preleredness program, conducted May 17-21, 1993)

was held by Mr. Craig Gordon on May 21, 1993.

The exit meeting for followup inspection

conducted for inspection report 50-244/93-08 (service water system outage maintenance, initial

inspection conducted April 12-16, 1993 and followup inspection conducted May 24-28, 1993)

was held by Mr. Harold Gregg on May 28, 1993.

The exit meeting for inspection report 50-

244/93-10 was held on June 9, 1993.5

0:

0