Text

Insp Rept 50-424/87-31 on 870418-0522.Violations Noted: Failure to Prescribe Appropriate Procedures for Performing Maint & Failure to Declare Both RHR Trains Inoperable & Comply W/Tech Spec 3.5.2
	ML20234C279
Person / Time
Site:	Vogtle
Issue date:	06/24/1987
From:	Rogge J, Schepens R, Sinkule M; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	;
Shared Package
ML20234C229	List: ML20234C225; ML20234C238; ML20234C279;
References
	50-424-87-31, NUDOCS 8707060400
	Download: ML20234C279 (12)
	v • d • e

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j 101 MARIETT A ST RE ET. N.W.

ATLANTA, GEORGI A 30323

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Report No.: 50-424/87-31 r

Licensee: Georgia Power Company

P.O. Box 4545

Atlanta, GA 30302

Docket No.: 50-424 License No.: NPF-68

Facility Name: Vogtle 1

Inspection Conducted: April 18 - May 22,1987

Inspectors: L - - 2'I 7

g J., F. Rogge, Senior Resident Inspector, Date Signed

Operatio s . g

[ . hb? O^

R. J. Schepens, Resident Inspector, Operations

6/z9/8 7

Date Signed

pop,

Accompanying Personnel: C. W. Burger

P Holmes-Ray

Approved by: CL]M

M. V. Sinkule, Section Chief

b' 2

Date' Signed

67

Division of Reactor Projects

l SUMMAR

Scope: This routine, unannounced inspection entailed Resident Inspection in

the following areas: Plant Operations, Radiological Controls, Maintenance,

Surveillance, Fire Protection, Emergency Preparedness, Security, Startup Test

Program, Quality' Programs and Administrative Controls Affecting Quality, and

Follow-up On Previous Inspection Identified Items.

Results: Two violations were identified Failure to Prescribe Appropriate

Procedures for Performing Maintenance (MSIV 1HV-3006B) and failure to Declare ;

Both RHR Trains Inoperable and Comply with Technical Specification 3.5.2. ]

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REPORT DETAILS

1. Persons Contacted

Licensee Employees

P.-D. Rice, Vice-President, Vogtle Project Direction

R. H..Pinson, Vice-President, Project Construction

C. W. Whitney, General. Manager, Project Support

G. Bockhold, Jr. ,- GeneralL Manager Nuclear Operations

E. M. Dannemiller, Technical Assistant to General Manager

T. V. Greene, Plant Manager

R. M. Bellamy, Plant Support Manager

P. R. Bemis, Manager of Engineering

C. W. Hayes, Vogtle Quality Assurance Manager

C. E. Belflower, Quality Assurance Site Manager - Operations

+W. E. Mundy, Quality Assurance Audit Supervisor-

'W. C. Gabbard, Regulatory' Specialist .

C. F. Meyer, Operations Superintendent

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R. M. Odom, Plant Engineering Supervisor

M, A. Griffis, Maintenance. Superintendent

G. R. Frederick, Quality Assurance Engineer / Support Supervisor

R. E. Srinnato, ISEG Supervisor

J. F. D' A tico, Nuclear Safety & Compliance Manager

W. F. Kitchens, Manager Operations

V. J. Agro, Superintendent Administration

A. L. Mosbaugh, Assistant Plant Support Manager

M. P. Craven, Nuclear-Security Manager ,

J. E. Swartzwelder, Deputy Manager - Operations !'

P. H. Burwinkel, Engineering Supervisor - HVAC-

Other licensee employees ccntacted included technicians,, supervisors,

engineers, operators, mechanics, inspectors, and office personnel.

Attended Exit Interview

2. Exit Interviews - Unit 1 (30703)

The inspection scope and findings were sumr..srized on May P2,1987 with

those persons indicated in paragraph.1 above. The inspectors described the.

arehs inspected and discussed in- detail the . inspection results. No

dissenting comments were received from the licensee. The licensee ~did not

identify as proprietary any of the materials provided to or reviewed by-

)- the inspector during this inspection. . Region based NRC exit interviews-

were attended during the inspection period by a. resident inspector. This-

inspection closed two unresolved items. The items identified during this

inspection are:

a. Viciation 50-424/87-31-01 " Failure to Prescribe Appropriate

Precedure- 'or Performing Maintenance on safety Related Equipment

(MST" itiv uG6B)" - Paragraph 5.b.(7).

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b. Violation 50-424/87-31-02 " Fail ure to Declare Both RHR Trains

Inoperable and Comply with Technical Specification 3.5.2" -

Paragraph 7.

c. Unresolved Item 50-424/87-31-03 " Complete Review of Crud Tank

Overpressurization" - Paragraph 4.

d. Inspector Followup Item 50-424/87-31-04 " Review Methodology for

Control of Throttle Valves" - Paragraph 5.b.(1).

3. Licensee Action on Previous Enforcement Matters - Unit 1 (92702)

Not inspected.

4. Unresolved Items - Unit 1(92701)

Unresolved items are matters about which more information is requi.ed to

determine whether they are acceptable or may involve violations or

deviations. One unresolved item identified during this inspection is

discussed in Paragraph 5.b.(4).

5. Operational Safety Verification - Unit- 1 (71707) (93702)

The plant began this inspection period in power operation (Mode 1)

conducting startup testing at the 50% power plateau. On April 23, the

unit achieved 75% power plateau but was forced to reduce power to perform

repairs on the B main feed pump and EHC hydraulic leaks. On April 29, the

! reactor tripped from 74% power on OT-delta-T and returned to Mode 1 on

April 30. The unit remained at the 75% plateau until May 4, when the

reactor again tripped on OT-delta-T. Following this trip a faulty card j

was ide7tified as the cause for both trips. On May 6, the unit returned '

to Mode 1 and performed further testing unt',1 the reactor tripped on

May 9. This trip was due to low steam generator water level following the

10% load upswing test from 65% power in v nich the power overshot to 80%

with insufficient feed capability due to one feed pump out of service. On

May 10, the unit returned to Mode 1 and completed the 75% power plateau

testing on May 12. On May 13, the reactor tripped from 90% power when j

improper maintenance on one main steam isolation valve resulted in valve

closure.

Plant management elected to proceed to cold shutdown (Mode 5) and conduct

a main turbine bearing inspection. On May 19 the unit proceeded from Mode j

5 to hot standby Mode 3 in preparation for reactor restart when plant a

management placed a startup hold while the site responded to contamination

of the demineralized water header problem. The unit was in Mode 3 at the

end of this inspection period. During the inspection a total af ten (10)

ESFAS actuations occurred as follows: Five (5) control room ventilation

isolations from three (3) radiation monitor problems and two (2) chlorine

detector proble.'s; one (1) containment ventilation isolation resulting

from breaker switching; two (2) main feedwater isolations due to Hi-Hi j

steam generator levels; two (2) auxiliary feedwater actuations resulting j

from low suction pressure trip of the main feedwater pumps.

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a. Control Room Activities

Control Room tours and observations were performed to verify that

facility operations were being safely conducted within regulatory

requirements. These inspections consisted of one or more of the

following attributes as appropriate at the time of the inspection.

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Proper Control Room staffing

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Control Room access and operator behavior

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Adherence to approved procedures for activities in progress

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Adherence to Technical Specification (TS) Limiting Conditions

for Operations (LCO)

- Observance of instruments and recorder traces of safety related

and important to safety systems for abnormalities

- Review of annunciators alarmed and action in progress to correct

- Walkdown of main control board; electrical auxiliary control

board; heating, ventilation, and air-conditioning systems

panel; and miscellaneous systems and equipment panel

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Safety parameter display and the plant safety monitoring system

operability status

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Discussions and interviews with the On-Shift Operations

Supervisor, Shift Supervisor, Reactor Operators, and the Shift

Technical Advisor to determine the plant status, plans and

assess operator knowledge

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Review of the operator logs, unit log and shift turnover sheets'

Additional inspections conducted consisted of an in-depth review of

the following clearances:

1-87-1337 RWST Sludge Mixing Pump

1-87-1341 Steam Generator Main Feed Pump "B"

No violations or deviations were identified.

b. Facility Activities

Facility tours and observations were performed to assess the

effectiveness of the administrative controls established by direct

observation of plant activities, interviews and discussions with

licensee personnel, independent verification of safety systems status

and LCO's, licensee meetings and facility records. During these

inspections the following objectives are achieved:

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(1) Safety System Status (71710) -

Confirmation of system

operability was obtained by verification that flowpath valve

alignment, control and power supply alignments, component

conditions, and support systems for the accessible portions of I

the ESF trains were proper. The inaccessible portions are I

confirmed as availability permits. Additional indepth

inspection of the containment isolation system and the essential

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chilled water system was performed to review the system lineup

l procedure with the plant drawings and as-built configurations,

compare valve remote and local indications, walkdowns were

expanded to include hangers and supports, and electrical

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equipment interiors. The inspector verified that the lineup was

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in accordance with license requirements for system operability.

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) The inspector utilized FSAR Table 0.2.4.1 and the following

I piping and instrumentation drawings to conduct the walkdown of

the containment isolation system.

DWG N0. TITLE

1X4DB110 Post Accident Sampling System

IX4DB112 Reactor Coolant System

IX4DB114 Chemical & Volume Control System

IX4DB120 Safety Injection System

1X4DB121 Safety Injection System 4

IX4DB127 Waste Processing System - Liquid

1X4DB140 Nuclear Sampling System - Liquid

1X408143 Containment & Auxiliary Building

l Drains - Radioactive

1 1X4DB174-4 Fire Protection Water System -

1X4DB186-1 Service Air System

, 1X4DB186-4 Instrument Air System

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Procedure 14475-1 " Containment Integrity Verification - Valves

Outside Containment" as well as a technical review to verify it

satisfied Technical Specification 4.6.1.la surveillance

requirements.

The inspector utilized the following piping and instrumentation

drawings to conduct the walkdown of containment integrity.

DWG NO. TITLE

1X4DB120 Safety Injection System

IX4DB121 Safety Injection System

IX4DB130 Spent Fuel Cooling & Purification System

1X4DB131 Containment Spray System

1X4DB132 Miscellaneous Leak Detection

1X4DB159-1 Main Steam System

IX4DB159-3 Main Steam System

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1X4DB186-1 Service Air System

AX4DB190-2 Plant Demineralized Water System

1X4DB213-1 Purification and Clean-up System

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The inspection of the essential chilled water system consisted

of a walkdown of Train "A" to verify proper alignment for

standby readiness per Operations Procedures 13744-1 " Essential

Chilled Water System" and 11744-1 " Essential Chilled Water

System Alignment for Startup and Normal Operation" and Piring

and Instrumentation Drawing Nos. 1X4DB221 and 1X4DB233. Per

Procedure 11744-1 the required position for cooling coil outlet

valves from the ESF coolers is throttled. These valves are

throttled to flow balance the system. The inspector noted that

there is no requirement that these valves be locked in position

once set nor is their throttled position specified (i.e. number

of turns opened). Discussions with Engineering personnel

reveaied that the system is being. flow balanced frequently due i

to these valves being closed by operations to support main-

tenance on the ESF coolers. Pending review of the licensee's

program for controlling throttle valves to address the above

noted comments this item will be identified as IFI 50-424/

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87-31-04 " Review Methodology for Control of Throttle Valves".

(2) Plant Housekeeping Conditions - Storage of material and

components and cleanliness conditions of various areas through-

out the facility were observed to determine whether ' safety i

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and/or fire hazards existed.

(3) Fire Protection - Fire protection activities, staffing and

equipment were observed to verify that fire brigade staffing was

appropriate and that fire alarms, extinguishing equipment,

actuating controls, fire fighting equipment, emergency

equipment, and fire barriers were operable.

(4) Radiation Protection (71709) - Radiction protection activities,

staffing and equipment were observed to verify proper program

implementation. The inspection included review of the plant

program effectiveness. Radiation work permits and personnel

compliance were reviewed during the daily plant tours.

Radiation Control Areas (RCAs) were observed to verify proper

identification and implementation. On May 19, 1987 the licensee

identified that the Crud Tank had pressurized and resulted in

the contamination of the demineralized water header and other

back flushable filter systems. At' the conclusion of this

inspection period, complete assessment of the root cause had not

been completed. Until this review can be completed by both the {

licensee. and NRC it is identified as Unresolved Item I

50-424/87-31-03 " Complete Review of Crud Tank (

Overpressurization", j

(5) Security (71881) - Security controls were observed to verify

that security barriers were intact, guard forces were on duty,

and access to the Protected Area (PA) was controlled in

accordance with the facility security plan. Personnel within

the PA were observed to verify proper display of badges and that

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personnel requiring escort were properly escort.ed. Personnel

within vital areas were observed to ensure proper authorization

for the area. Equipment operability of proper compensatory

activities were verified on a periodic basis.

(6) Surveillance (61726)(61700) - Surveillance tests were observed

to verify that approved procedures were being used; qualified

personnel were conducting the tests; tests were adequate to

verify equipment operability; calibrated equipment was utilized;

and TS requirements were followed. The inspectors observed

portions of the following surveillance and reviewed completed

data against acceptance criteria:

SURV. NO. TITLE

14000 Mode 1 and 2 Daily Surveillance

14980-101 Diesel Generator 1A Operability Test

14220-101 Main Turbine Valves Weekly Stroke Test

14850-101 Cold Shutdown Valve Inservice Test

24551-101 Containment Hydrogen Monitor Train "A"

Analog Channel Operational Test and

Channel Calibration

During an NRC walkdown of the control room miscellaneous systems

and equipment panel on April 20, 1987, the inspector questioned

the valve alignment of the Train "A" hydrogen monitor. The

isolation valves which are normally closed were found to be in

the open position. The licensee determined that Surveillance

Procedure 24551-1 " Containment Hydrogen Monitor Train "A" Analog

Channel Operational Test and Channel Calibration" had been

performed during the day. This procedure required the opening '

of the Train "A" hydrogen monitor inlet (1HV-2792A and 1HV2791B)

and outlet (1HV-2793B) isolation valves; however, it lacked a

requirement to close the valves when restoring the system. The

licensee is revising Surveillance Procedure 24551 to require

closing the subject valves in the system restoration section.

The hydrogen monitoring system is a closed system designed to j

operate following an accident. The normal alignment for these j

valves following an accident is open. Therefore, the fact that j

these valves were not found to be in the closed position is of

no safety significance. However, the fact that the operators

did not question the position of these valves when the

surveillance procedure work was reported to be complete for the 1

day is an indicator of inattention to detail. This matter was ,

discussed in the enforcement conference held with the licensee i

on May 20, 1987 in Region II.

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(7) Maintenance Activities ('62703) -

The inspec' tor observed

maintenance activities.'to _ verify that correct equipment

clearances were in effect; work' requests 'and fire prevention

work permits, as_ required, were ' issued and being '.followed;

quality control personnel were available for inspection

activities as required; retesting and return of systems to -

service was prompt and correct; TS requirements were being

followed. Maintenance backlog was. reviewed. Major maintenance

activities observed by the. inspector were on the steam generator-

main feed pump "B" and the main turbine bearing inspection.

The inspector conducted an inspection of the maintenance being

performed on Main Steam Isolation Valve (MSIV) 1HV-3006B-as part- )

of a. followup inspection to the reactor trip which occurred on

May 13, .1987 as a result of MSIV 1HV-3006B closing. The

inspector reviewed the following- documents pertaining to the'

event:

DOCUMENT TYPE /NO. TITLE

MWO #1-87-06119 Investigate Hydraulic Leak on MSIV

1HV-30fJB-

Clearance #1-87-1378 1HV-3006B MSIV B for SG #1

Deficiency Card Reactor Trip on Lo-Lo SG #1 Level

1-87-1281 Due to MSIV 1HV-3006B Closure

LCO #1-87-507 MSIV IHV-3006B Blocked Open

Adm. Proc. 00350-C Maintenance Program

Adm. Proc. 00054-C Rules for Performing Procedures

Maintenance Work Order 1-87-06119 was generated on May 11, 1987

identifying a hydraulic leak (possibly a solenoid valve leaking

by). Subsequently, on May 12, 1987, the MSIV Loop- 1 trouble

alarm came in. The shift declared MW0s 1-87-06119 and

1-87-06120 as emergency .in accordance with Administrative

Procedure 00350-C and entered a 3-day LCO on MSIV 1HV-3006B per

TS 3.7.1.5. The OSOS met with the maintenance crew to discuss

the work to be conducted on the MSIV (i.e. block the valve open

and replace the leaking pumpside solenoid valve). Subsequently,-

maintenance proc 2eded with physically blocking open the valve

with a pipe and mechanically removing the pumpside solenoid-

valve. At this point, operations hung clearance.1-87-1378 to

de-energize power to MSIV IHV-3006B' which' actuated the MSIV to

close,.

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During the inspector's review of the above event which included

discussions with appropriate maintenance and operations

personnel, the following discrepancies were noted: 1) The

maintenance work being performed was not a true emergency as

defined in Administrative Procedure 00054-C " Rules for

Performing Procedures" Section 4.3 and Administrative Procedure

00350-C " Maintenance Programs Section 2.6, 2) Maintenance

personnel ' were performing work on safety-related equipment

without written work instructions while the MW0s were still in

work plannirg and had not been issued to the field for work, 3)

Clearance 1-87-1378 was inadequate in that it only addressed the

electrical and not the mechanical side for isolating the MSIV in

the open position. The Train A & B normally locked open manual

manifold isolation valves should have been on the clearance to

be closed prior to de-energizing the power to the MSIV solenoid

valves, and 4) Verbal instructions given by operations to the

maintenance crew was unclear in that the maintenance crew

interpreted block open the valve to mean physically blocking it

open with a pipe, not closing the Train A & B normally locked

open manual manifold isolation valves.

The foregoing is considered to be in violation of Technical

Specification 6.7.la and will be identified as " Failure to

Prescribe Appropriate Procedures for Performing Maintenance on

Safety Related Equipment (MSIV IHV-3006B)".

(8) Emergency Preparation

The Hatch Senior Resident Inspector familiarized himself with

the site for emergency purposes, obtained a badge for unescorted i

access, toured the emergency response facilities and major plant

areas. The inspector reviewed the licensee's emergency plan and

the emergency plan implementing procedures. The review was

limited to a brief familiarization with the documents. Thi s

familiarization was conducted with the Senior Resident

Inspector.

No violations or deviations were identified except as identified in

paragraph (7) above.

6. Followup on Previous Inspection Items - Unit 1 (92701)

(Closed) Unresolved Item 50-424/86-136-01 " Review Inspection Results of

the Licensee's Review of the Containment Combined Leak Rate Surveillance

Calculation for All Penetrations & Valves Subject to Type B and C Tests."

The completed Deficiency Report (No. 1-87-331) was reviewed for root cause

determination and corrective action to prevent recurrence. The licensee's

currective action consisted of but was not limited to the following: 1.)

A review to verify that all Type B and C penetrations per FSAR Table

6.2.4-1 were included in Surveillance Procedure 28916-1 " Containment. Type

A, B, & C Leakage Totalization"; 2.) An update to NRC Report on Reactor

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Containment Building Integrated Leakage Rate Test to reflect the current

leakage rates for all required penetrations including the encapsulation

vessels per Surveillance 28916-101; and 3.) A re-summation of total

leakage in Preoperational Test Procedure 1-300-04 " Containment Local Leak

Rate Testing" which included the encapsulation vessels (Penetrations

36-39) to verify total leakage within acceptance criteria.

Based on the above the inspector has concluded the omission of the

encapsulation vessels from the total leakage calculation was an isolated

case with no safety significance since the re-summed total leakage

including the encapsulation vessels was found to be within the acceptance

criteria; therefore, no violation has occurred and this item is considered

to be closed.

(Closed) Unresolved Item 50-424/87-27-03 " Deportability of CRVI

Actuations As a Result of the Chlorine Gas Monitors". This item was

discussed during an NRC Staff Meeting with the licensee at Vogtle on

April 28, 1987. The licensee agreed to abide by the staff's position to

report CRVI actuations as a result of the chlorine gas monitors. The

inspector also has reviewed corporate deportability guidance to the site

to verify that it included CRVI actuations as a result of toxic gas as

reportable. The licensee is submitting Licensee Event Reports as

appropriate to document all CRVI actuations as a result of the toxic gas

monitors. Based on the above, the inspector has concluded that no

violation has occurred; therefore, this item is considered to be closed.

7. Review of Licensee Event Reports - Unit 1 (90712)(92700)

Licensee Event Reports (LERS) and Deficiency Cards (DCs) were reviewed

for potential generic impact, to detect trends, and to determine whether

corrective actions appeared appropriate. Events which were reported j

immediately, were reviewed as they occurred to determine if the Technical j

Specifications were satisfied.

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Deficiency Card (DC) 1-87-1192 was generated on April 28, 1987, at j

1:45 p.m. This DC identified that the Train A & B RHR Heat Exchanger

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Discharge Valves (1HV-0606 & 1HV-0607) were noted by local observation to !

be only 90-95% of full open. This position was determined from the

positioner card and the fact that the valves had a slight loading pressure i

on the diagram even though the remote hand controllers HIC 0606A and HIC

0607A on the Main control room board had full demand. The OSOS evaluation

for LCO applicability was signed off as not applicable.

During the inspector's review on April 29, 1987, at 1:00 p.m. the

inspector questioned the status of the subject valves and RHR system

operat ility. The inspector determined that the valves were still in the

above noted condition and that MW0s 1-87-05802 & 1-87-05803 had been

written to perform a calibration check of the subject loop such that M/A

station output corresponds with valve position. Also Operations had not

declared RHR inoperable based on the following: 1) The remote hand

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controllers HIC 0606A and HIC 0607A indicated full demand, 2) The Group 1

status monitor lights which illuminates when the valve is at least 95%

closed was not illuminated, and 3) the latest completed ECCS Check Valve i

Cold Shutdown Inservice Surveillance Test Procedure 14896-1 indicated J

acceptable RHR injection line flow rates. The inspector questioned )

operations decision of the RHR system being operable based on the

following concerns: 1) With the subject valves identified as being

partially closed Technical Specification Surveillance Requirement Section 4.5.2b(2) which states, " Verify that each valve (manual, power

operated, or automatic) in the flow path that is not locked, sealed or

otherwise secured in position, is in its correct position", had not been

demonstrated, and 2) With-the valves in the 90-95% open position would tha

RHR System be capable of delivering 3788 gpm with a single pump running, j

in the cold injection mode from the RWST per Technical Specification,

Surveillance Requirement Section 4.5.2h(3).

During the performance of MW0s 1-87-05802 and 1-87-05803 the following

was determined: Train A RHR Heat Exchanger Valve (1HV-0606) was found to

have a cable identification tag restricting the movement of the bellows

assembly in the valve controller I-P. After repositioning the cable

identification tag the valve was returned to the full open position.

Train B RHR Heat Exchanger Valve (1HV-0607) was found to have a positioner

out of calibration. This affected the valve opening stroke and prevented

the valve from opening to the full open position. The controller was

recalibrates which returned the valve to the full open position. These

repairs were completed on April 29, 1987, at 10:00 p.m. and the valves j

were returned to the full open position. i

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Subsequently, Engineering conducted a review to determine if the TS

minimum flow requirement was satisfied. Utilizing the RHR System

Preoperational Test Data Engineering generated a valve position versus

indicated flow curves. These curves in conjunction with the most

conservative value for the as-found valve position of 90% open were used

to determine the as-found pump flow. Using these curves Engineering

determined the Train A flow rate to be 3762 gpm (.7% below TS limit) and q

Train B flow rate to be 3686 (2.7% below TS limit). l

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The inspector reviewed the RHR System Preoperational Test Data contained i

on Data Sheets 7.5 & 7.6 in Preop 1-3BC-01. The flows recorded in the

preop were 4158 gpm for Train A and 4064 for Train B. The latest

completed Surveillance Procedure 14896-1 was reviewed which indicated

Train A to have an actual flow rate of 3875 on February 3,1987,- and

Train B to have an actual flow rate of 4050 gpm on January 9, 1987. Also,

a review was conducted of all MWO's performed on the subject valves to try )

to determine how long this condition could have existed. Based on the '

inspector's review it was not possible to determine if the condition

existed during the latest completed surveillance; however, it was

concluded that between the preop which was completed on February 24, 1986,

for Train A and February 23, 1986 for Train B there was a change in flow

rate due to these valves being partially closed.

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In conclusion, based on the licensee's engineering evaluation using the

conservative figure of the valve only being open 90%, the calculated flows ;

delivered by each RHR preop would have been marginally .below the TS i

minimum of 3788 gpm.

The foregoing is considered to be in violation of Technical Specification 6.7.1.a and will be identified as Violation 50-424/87-31-02 " Failure to

Declare Both RHR Trains Inoperable and Comply with Technical Specification 3.5.2" and is being evaluated for possible escalated' enforcement action.

8. Management Meetings - Unit 1 (30702)

The Resident Inspectors attended a meeting at Vogtle with the NRC Staff

and the licensee on April 28, 1987, to discuss Vogtle's startup history.

Mr. A. F. Gibson, Director-Division of Reactor Projects and

Mr. V. L. Brownlee, Acting Deputy Director - Division of Reactor Projects

were in attendance from the Regional Office.

On May 11, 1987, the Resident Inspectors attended a presentation on Vogtle

Electric Generating Plant given by the licensee to Commissioner

Mr. F. M. Bernthal and his Technical Assistant Mr. J. F. Meyer.

Mr. M. L. Ernst, Deputy Administrator and Mr. V. L. Brownlee, Acting

Deputy Director - Division of Reactor projects were in attendance from the.

Regional Office.

On May 20, 1987, the Resident Inspectors attended an NRC Enforcement

Conference held with the licensee in Region II. The subject was proper

System / Component alignments and attention to detail. The licensee's

presentation discussed six (6) component alignment events and a summary of

both short and long term term corrective actions. While it was noted that

the licensee could not identify the root cause in some of the component

alignment events, the licensee hoped that the establishment of additional

measures for valve manipulations would be helpful in identifying the root ;

cause for future cases should one occur, t

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ML20234C279

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