ML20058G449

Insp Repts 50-424/90-10 & 50-425/90-10 on 900331-0518. Violations Noted.Major Areas Inspected:Plant Operations, Radiological Controls,Maint,Surveillance,Security,Quality Programs & Administrative Controls Affecting Quality

ML20058G449
Person / Time
Site:	Vogtle
Issue date:	06/13/1990
From:	Aiello R, Brockman K, Rogge J, Starkey R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20058G414	List: IR 05000424/1990010 ML20058G410 ML20058G449
References
50-424-90-10, 50-425-90-10, NUDOCS 9011130164
Download: ML20058G449 (20)

See also: IR 05000424/1990010

Text

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g NUCLEAR RESULATCRY COMMIGON

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Report Nos.: 50-424/90-10 and 50-425/90-10

Licensee: Georgia Power Company

P.O. Box 1295

Birmingham, AL 35201

Docket Nos.: 50-424 and 50-425 License Nos.: NPF-68 and NPF-81

Facility Name: Vogtle Nuclear Station Units 1 and 2

Inspection Conducted: March 31 - May 18,1990

Inspectors:

7. F. L6 ge, Senior

2W/ b)

sident Inspector

/t-/3-90

Date 5igned

, t . A &/d//

R.~ F.' A' e'

& 6-AS-9f)

Date Signed

lo, AcT,inpenior Resident Inspector

f W1 h)

.~ 'D.'5tarkey, Res)Ceht Inspector

4 'A8* Q/)

~ 04te Signed

T

Accompanied By: L.focine

Approved By: p[# /~U-@

K.J. BrockmglPf 5epr on Chief Date Signed

Dtvision of Reactor Projects

SUMARY

scope: This routine inspection entailed resident inspection in the following

areas: plant operations, radiological controls, maintenance,

surveillance, security, and quality programs and administrative

controls affecting quality.

Results: One cited and six non-cited violations were identified. The cited '

violation was in the area of operations for failure to ensure proper

routing and slope of the RCS temporary level indication tygon tube

(paragraph 2.a). Two NCVs were in the area of operations for failure

to have two source range monitors operable in Mode 6 (Refueling) per

TS 3.9.2 and failure to follow procedure 13320, FHB HVAC System,

resulting))in

3.b.(2)(d .

an inadvertent

Three NCVs were automatic FHB

in the area isolation (paragraph

of maintenance and

surveillance for failure to perform an adequate surveillance

resulting in a violation of TS with respect to the AFW and DG systems 1

(paragraph 3.6.(2)(a)),failuretoconductanASMEClass2ISTperTS '

4.0.5 requirements within the quarterl

ASME Section XI (paragraph 3.b.(2)(g))y time interval

, and failure specified

to conduct a by

required TS surveillance due to an inadequate surveillance task sheet

which resulted in a violation of TS 4.6.1.1.a (paragraph 3.b (2)(h)),

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The sixth NCV was in the area of radiological controls for failure to

monitor a liquid waste release in accordance with the requirements of.

TS3.3.3.9(paragraph 3.b.(2)(f)). ,

Two weaknesses and 'one strength in the area of operations were

identified as follows:

On April 27, 1990, the inspector identified a failure to

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maintain an accurate active License Status Report (para-

graph 2.a).

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On April 15, 1990, the inspector identified a weakness in the

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area of administrative controls of WRT physics testing-for

utilizing a procedure beyond the scope of its intent

(paragraph 4).

The Licensee has taken the initiative to install a one-line RHR

system diagram on one of the procedure trays in each control.

room. . These trays will be modified to allow interchangeability  !

of the' system drawings. . This operatt,r aid should serve to  ;

1 enhance the operator's mental picture regarding system align- i

.ments. The inspectors noticed that red and green banding on '

meter faces are being installed in both control rooms. This-

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should serve to enhance operator response to unexpected

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transients.- Both of these items have- been identified as

-strengths in human factors affecting operations, i

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DETAILS

1. Persons Contacted

Licensee Employees

• J. Aufdenkampe, Manager - Technical Support i

G. Bockhold, Jr., General Manager - Nuclear Plant  !

C. Coursey, Maintenance Superintendent i

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G. Frederick, Safety Audit and Engineering Group Supervisor

H.1 Handfinger, Manager - Maintenance t

• W. Kitchens, Assistant General Manager - Plant Operations

• R. LeGrand, Manager - Health Physics and Chemistry

• G. McCarley, Independent Safety Engineering Group Supervisor

• A. Mosbaugh, Assistant General Manager - Plant Support

. *R. Odom,' Nuclear Safety and Compliance Manager

J. Swartzwelder, Manager - Operations t

• E. Dannemiller, Nuclear Security Manager

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Other~ licensee employees contacted included technicians, supervisors,

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engineers, operators, maintenance personnel, quality control inspectors,

and office personnel.

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-; * Attended Exit Interview

An alphabetical list of acronyms and initialisms is located in the last

l: , paragraph of the inspection report.

2. Operational Safety Verification - (71707)(93702)-

The facility. began this inspection. period with Unit 1 in Mode 5 (Cold i

Shutdown) completing 1R2 and Unit 2 in Mode 1 (Power Operetion) at full

power. On April 13, 1990, VEGP was notified by the NRC Operation Center

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that the ENS would be out of service'for approximately one month.

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-Unit 1:

On April 11, 1990, the unit entered Mode 4 (Hot Shutdown). On April 13,

an FWI occurred due to procedural inadequacy. Later that same day, the 4

unit entered Mode 3 (Hot Standby). On April 16, the unit entered Mode 2

(Startup), went critical and comenced. low power physics testing. Later

, that same day, while conducting rod worth measurements, SDB-E rods dropped

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into the core from 170. steps. The reactor was imediately returned ' ,

L critical by pulling SDB-B -(see the strengths and weaknesses paragraph, '

above, and paragraph 4 for details). On April 17, the unit entered

Mode 1. . On April' 21, the main generator was tied. to the grid. - On

l/ April 25, the operator manually tripped the reactor, due to a lowering

l level in SG #2 as a result of a failed closed MFRV, and subsequently 1

entered MoJe 3. Later that same day, following repairs, the unit entered

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Mode 2. On. April 26, the unit went critical, entered Mode 1 and

synchronized to the grid. On May 6, the unit comenced a shutdown to

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Mode 3 for a DMIMS investigation (due to a metallic noise, apparently

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emanating from around a SG tube sheet). To facilit3te troubleshooting, on

May 8, the r.ain turbine and reactor were manually tripped and the unit

entered Mode 3. On May 9, following the DMIMS investigation, the unit

entered Mode 2, went critical, entered Mode 1 and tied to the grid. The  :

unit remained at full power, with the exception of minor power reductions.  !

for maintenance, through the end of this inspection period.

Unit 2:

On AprJI 1,1990, HDP-A tripped causing the unit to exceed its licensed

thermal power limit-(see paragraph 3 for detail 3).. On May 6, the reactor

tripped, due to MSly 3026A failing closed and entered Mode 3.- On May 7.

following repairs, the unit entered Mode 2, went critical, entered Mode 1  :

and tied to the grid. The unit remained at full power with the exception

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of minor. power reductions for maintenance, through the end of this

inspection period,

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.

- Proper Control Room staffing  :

- Control Room access and operator behavior

- Adherence to approved procedures for activities in progress

- Adherence to Technical Specification 1.imiting Conditions for

Operations _ _

- Observance of instruments and recorder traces of safety related and

important to safety systems for abnormalities

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- Review of annunciators alanned and action in progress to correct'

- Control Board walkdowns

- Safety parameter display and the plant safety monitoring system

operability status

- Discussions and interviews with the On-Shif t Operations Supervisor,

Shif t Supervisor. Reactor Operators, and the Shift Technical

Advisor (when stationed) to determine the plant status, plans, and  ;

to assess operator knowledge

- Review of the operator logs, unit logs and shift turnover sheets

On March 21, 1990, a Region 11 Waiver of Compliance from TS 3.0.4

was issued for Unit 1. This waiver was requested due to failures of

the Unit 1 Train A diesel generator and :its associated load

sequencer. This rencered diesel generator operability to be i

questionable.- At the time, the Unit 1 Train B' diesel generator was

out of service for scheduled maintenance and the reactor coolant

system water level was at mid-loop. This one-time waiver allowed

entry into Mode 5 from Mode 6 with diesel generator 1B inoperable

and the operability of diesel generator 1A and its associated load

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sequencer unverified. On April 7,1990, while in Mode 5, Unit 1 '

initiated an RCS drain down from 100% pressurizer level to 195 feet

RCS level for the purpose of repairing the pressurizer spray valves.

On April 8, with indicated RCS level at approximately 202.5 feet, the

PE0 assigned to the tygon tube watch reported an apparent air bubble- ,

in the tube. RCS drain down was imediately stopped while the tygon

tube watch adjusted the tygon tube to clear the bubble. After the

bubble was cleared, the actual RCS level was determined to be

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approximately 194 feet. Operations then initiated refill back to 195

feet.

Procedure 54840-1 Installation And Removal Instructions For. The RCS 4 '

Temporary Level Indication Tygon Tube And The Defeat Of The Residual

Heat Removal Suction Valve Auto Closure Interlock, Revision 1, had a

Temporary Procedure change initiated TCP No. 54840-1-1-90-2, to ,

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permit installation of a tygon tube from the vent valve on the .

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.mid-loop sight glass to the pressurizer vent connection. This was i

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the first time this particular method' of routing the tygon tube had

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been used. TCP No. 54840-1-1-90-2 required that the tygon tube be.

properly routed to prevent kinks, bends or other irregularities which:

could cause incorrect readings and to have a continuous slope up to

prevent air entrapment. However, the routing of the tube was such

that -a loop occurred at the point where the tube was routed over

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the " lip" of a walkway grate. This allowed an air bubble to develop-

, and' invalidate indicated level readings. Further contributing to

the )roblem was the inability of the licensee to verify complete

filling and venting of the system since part of the system consisted

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of hard piping-which_ prevented visual inspection for an air bubble.

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' Failure to. ensure proper routing and slope to the RCS Temporary level

Indication Tygon Tube- constitutes a violation of TS 6.7.1.a. This

item has been: identified as: ,

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VIO 50-424/90-10-01, " Failure To Ensure Proper Routing And Slope'

To The RCS Temporary Level Indication Tygon Tube." l

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On April' 24, L1990, representatives from engineering,' operations.

J1 maintenance, and SCS metc to discuss design proposals for mid-loop

sight glass installation on Unit 2.- Once the design is finalized,

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installation should occur'during 2R1. j

On April 27, 1990, the inspector reviewed the Control Room Active

LicLse Status Report which was dated April 23, 1990. This report

listed all- active license holders, the unit on which licensed, and

any restrictions on the license. The report listed a reactor -

operator who, according to letter ELV-01563 0350, from the licensee

to. the NRC,- dated April 20, 1990, was no longer employed by GPC.

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Further inquiry revealed that the operator had last worked at Vogtle

on April 12,1990, then took accumulated vacation time, and was 1

i- subsequently processed out on May 4,1990.

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The failure to maintain an accurate active License Status Report is

identified as a progransnatic weakness, inspection reports

50-424/89-25, 50-425/89-29, 50-424/89-27 and 50-425/89-31 identified

similar shortcomings in Reactor Operator License Verification.

One violation was identified. l

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 LCOs, licensee meetings and facility records. During these

inspections the following objectives were achieved:

(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 +

the ESF trains were proper. -The inaccessible portions are

confirmed as availability permitted.

(2) Plant Housekeeping Conditions - Storage of material' and

f components: and cleanliness conditions of various areas -

throughout the facility were observed to detemine whether

safety and/or fire hazards existed. ,

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

equipment were observed to verify that fire brigade staffing was 4

appropriate and that fire alams, extinguishing equipment,

actuating controls, fire fighting equipment, emergency '

equipment, and fire barriers were operable,

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(4) Radiation Protection - Radiation protection activities., staffing ,

and equipment were observed: to verify proper program

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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 were observed to verify proper

identification and implementation.

(5) Security - Security controls were observed to verify that '

security barriers were intact, guard -forces were on duty, and

access to the Protected Area was controlled in accordance with

'the facility- security plan. Personnel were observed to verify

proper display. of badges and that personnel requiring escort

were properly escorted. Personnel within Vital Areas were

observed to ensure proper authorization for the area. Equipment

i operability or proper compensatory activities were verified on a

periodic basis.

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(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 ewipment operability, calibrated equipment was utilized,

and TS requirements were followed. . The inspectors observed

portions c> the following surveillances and/or reviewed

coinpieteo data against acceptance criteria:

Surveillance No, Title

14811-2, Rev.1 Boric Acid Transfer Pumps And

Discharge Check Valves Inservice

Test

14810-1, Rey, 9 TDAFW Pump Check Valve Inservice

Test

- 13610-1, Rev, 9 - Auxiliary Feedwater System

14647-1, Rev.1 SSPS Slave Relay K644 CS Train "B"

Test

14825 1, Rev, 14 NSCW IST-

E 24726-1, Rev. 4 Time History Accelergraph And

SMA-3 Recorder AXT-19905 ACOT And

Channel' Calibration

14830-2, Rev. 2 NSCW Train "B" Check Valve IST

34218-C. Rev,-10 Channel Calibration Of The Gaseous

-Process Monitors

(7) ; Maintenance; Activities- (62703) - The- inspector -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 personnell were available for ' inspection

activities =as required; retesting and return of systems to

service was prompt _and: correct and TS requirements were being

-followed. -The Maintenance Work Order backlog was reviewed.

Maintenance was observed and/or work packages were reviewed for

the following maintenance activities:

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MWO No. Work Description

c 19002424 Repair MSIV 3026B Hydraulic Reservoir

Oil Leak

19002391 Repair EDG Air Start Air Dryer Auto

Drain Trap Leak ,

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19002467 Investigate And Repair SGBD Sample

Valve Position Indication ,

No violations or deviations were identified.

3. Review of Licensee Reports (90712)(90713)(92700)

a. In-Office Review of Periodic and Special Reports

This inspection consisted 'of reviewing the below listed reports to

, detennine whether the infonnation reported. by the licensee was-

. technically' adequate and consistent with the inspector knowledge of l

the material contained within the report. Selected material within

the- report was questioned randomly to verify accuracy and to provide

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a-reasonable assurance that other NRC personnel have an appropriate .

document for their activities.

Monthly Operating Report - The report dated April 11,'1990 was' l

reviewed. The inspector had no coninents. ,

The following special reports were reviewed:

The special report dated May 4,1990 regarding an inoperable TS -

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radiation , monitor at - VEGP was reviewed. The inspector had no

Consnents.

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l The report on structure settlement. dated April 23, 1990 was reviewed.

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The following items were included:

A comparison between the. predicted and measured ' values of

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, total gross settlement of the major seismic Category 1

structures.

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- A comparison between the actual . differential settlements

and the 1 differential . settlements used for design of

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safety-related piping passing between adjacent seismic

Category i structures.

- A comparison- between the basemat deformations due to

measured differential settlement within a structure and the

acceptance criteria.

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- Evidence supporting a reduction- in the frequency of ,

settlement monitoring during the life of the plant. I

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-The total gross settlement of the major seismic Category 1 structures

had not shown significant increase since the sumer of 1987. All I

settlement values were within the estimated maximum total settlement )

for each active marker. The piping _ differentials had been reviewed

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for piping systems passing between the major concrete structures. l

The data and sumaries indicated that the differential settlements

were not reaching the maximum design criteria. Only two cases

exceeded the design criteria in the latest review of the settlement

data, and these did not exceed 30% of the ASME code aliowable.

Basemat deformation for the major Category 1 structures continued to

show deflection of the concrete mat well below the design criteria. .

The data will continue to be reviewed -in the future for total I

settlement, and pipe system differentials. No significant changes I

are anticipated. in the future measured settlement values since all ,

loading conditions for the Category 1 structures have occurred. The '

inspector.has no further. comments.

INP0' evaluation of A. W. Vogtle . Nuclear Plant. The inspector I

reviewed Vogtle's INP0 evaluation dated December 1989. No matters j

that could substantially effect nuclear safety were noted. '

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k b. Deficiency Cards and Licensee Event Reports

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Deficiency ' Cards and Licensee Event Reports were reviewed for-

. potential generic impacti to' detect trends, and to determine whether

corrective' actions appeared appropriate. Events which were-reported y

pursuant . to 10 CFR 50.'72, were reviewed following occurrence to  :

determine: if the technical specifications and other regulatory

requirements were satisfied. In-office review of LERs may result in '

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further: followup to= verify that the stated corre::tive actions have

been completed 'or to identify violations -in addition to' those:

described in-the LER. Each LER was reviewed for enforcement action >

in accordance with.10 CFR Part 2. Appendix C, and where the violat!on'

was not. cited the criteria specified in Section V of the Enforcement

l Policy = were1 satisfied. . Review of DCs was performed to maintain.a

realtime status of deficiencies, detemine regulatory compliance, ,

follow the licensee corrective- actions, and assist as a . basis for

closure of the LER when reviewed. 'Due to the numerous DCs processed,

L only. those DCs which result in enforcement action or. further

L -inspector followup 1with the licensee at the end of the inspection 'are

listed below. The DCs and LERs denoted with an asterisk indicates a

that reactive inspection occurred following the event and prior to

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receipt of the written report.

(1) -The following Deficiency Cards were reviewed:

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(a) *0C 1-90-0190, " Pressurizer Liquid Space Temperature

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Exceeded Maximum Cooldown Of 200 Degrees F In A One Hour i

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Period."

On April 6, 1990, while in Mode 5, Unit 1 began a

pressurizer cooldown to allow maintenance to be performed 1

on the pressurizer spray valves. The pressurizer liquid

temperature decreased from approximately 350 degrees F to

110 degrees F in one hour which exceeded the maximum

allowable cooldown rate of 200 degrees F in a one hour

period.- This transient was analyzed by Westinghouse and in

their letter, CP-14826, dated April 9, 1990, they

determined that pressurizer integrity . had not been '

compromised by the transient. This event will be further

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followed up when submitted as a LER.

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(b) DC 1-90-0127, " Incorrect Procedures Results In inadequate

TS Surveillance On Valve 1-HV-8801B."

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On January 8, 1990, valve 1-HV-88018, SI System - BIT

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Discharge . Valve, was tested using procedure 14825-1, Rev.

12. The procedure incorrectly listed the safety position

l of the valve as closed. Consequently, the valve was stroke

o tested to the incorrect position during performance of a TS ,

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required' surveillance. This event will be further followed- '

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up when submitted as a LER.

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L (c) DC~1-90-0149, "RHR Suction Relief Valves May Not Be

Properly Sized- To Provide Cold Overpressure Protection."

On March 28, 1990, Vogtle wa's informed by Westinghouse of'a-

l potential problem with use of the RHR suction . relief valves

for cold overpressure. protection when RCS temperature is

greater than 160 degrees F. The potential: problem concerns-  ;

the capability of these relief.. valves to mitigate a low

temperature overpressure event associated with a reactor

coolant pump start when RCS temperature is greater than 160

degrees F 'and the RCS. is water solid. This potential-

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problem will be further followed up when submitted as a t

LER.

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(d) DC 1-90-0204, " Inadvertent Feedwater Isolation Signal l

l: Received Due To Procedural Inadequacy." l

j On April 13, 1990,. during performance of OPS 14703-1,

p Reactor -Trip Bypass Breaker Undervoltage Trip TA00T, an

inadvertent Feedwater Isolation signal was received. This-

H event was due to an inadequate procedure and resulted in an

unplanned ESFAS actuation. This event will be further

followed up when submitted as a LER.

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L (e) *DC 1-90-0215 " Shutdown Bank 'E' Dropped From 170 Steps

l_ During Low Power Physics Testings."

For details of this event, see paragraph 4. This event

will be followed up when submitted as a LER.

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(f) DC 1-90-0218. " Operator- Error And' Inoperable Containment

Temperature Computer Point Lead To Missed Surveillance."

On April 18, 1990, during review of procedure 14001-1,

Operations Shift and Daily Surveillance logs, it was noted

that containment 'C' level temperature per ERF point T7502

was recorded as 0.027 degrees F and the average temperature

per ERF point UT7501 as 57.9 degrees F. ERF point T7502 .

had been inoperable since April 11, 1990. Proteus point

T2502, which receives the same input as ERF ' point T7502,

was reading correctly. It was subsequently determined that

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the last successful surveillance for containment air

temperature was performed on April 16. 1990. Technical

Specification 4.6.1.5 requires that an average containment

temperature shall be determined once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This  ;

event will be followed up when submitted as a LER.

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(g) DC 1-90-0237, "Three Containment Integrity Isolation Valve

Were Not Verified Closed As Required By TS."

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Valves 1-1222-X4-017, 018.. and _019 were added to

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containment penetration flanges during 1R2. These valves

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were not added to procedure 14475-1, Containment Integrity

Verification - Valves Outside Containment. During a

routine review -of. Safety > Related Locked Valve Worksheets,  !

'it was discovered that these valves were required to be

verifi'id for containment integrity. The valves were then

promptly verified closed, capped - locked and hold tagged. *

However,' this verification and locking had exceeded the

time allowed by TS' 4.6.1.1. This event will be further

s followed up when submitted as .a LER.

(h). *DC- 1-90-0255, "Mispositioning Of Main Feed Reg Valve Local

Positioner Controller Causes Valve Closure and Subsequent

Manual Reactor Trip."

On April 25,1990, a manual reactor trip was initiated _ by

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operators due:to decreasing water level in 'SG #2. Level

decrease was due to the accident mispositioning by

maintenance -personnel of the Feed Reg '/alve Local

Positioner Controller which caused the valve to close

thereby isolating feedwater SG #2. This event wil1 be

further followed up when. submitted as an LER.

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(i)_ *DC 2-90-0034,. " Heater Drain Pump Trip Causes Reactor To

Exceed Licensed Thermal Power Limit."

On April 1,1990, with Unit 2 at 100% power, Heater Drain

Pump ' A' tripped due to the Heater Drain Tank high level

dump- valve not opening on a high level condition. A

secondary plant transient occurred which resulted in a ,

primary power excursion. Reactor power peaked at 3590 MWTH

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which exceeded the licensed limit of 3411 MWTH. This event l

will be further followed up when submitted as a LER. I

(j) DC 2-90-0037 Surveillance 14030-2 Not Completed Within

Required TS Time Limit."

Surveillance 14030-2, Power Range Calorimetric Channel

Calibration, was successfully completed on April 10, 1990.

This was a daily required surveillance. The next daily

surveillance was completed on April 11, 1990, but exceeded

the maximum allowable interval of '30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> by 51 minutes.

This event ~ will be further followed up when submitted as a

LER.

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(k) *DC 2-90-0043, " Inoperable Reactor Coolant Systec Leakage

Detection Systems Result In Entry To TS 3.0.3." >

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On May 1 '1990, with the containment air cooler condensate i

flow-rate out of service, both the Containment Atmosphere

Gaseous , and Particulate ' Radioactivity monitors were' '  :

rendered inoperable while perfoming a background count.

'

This deficiency resulted in ' entry into TS 3.0.3. This-

event will be further followed up when submitted as a LER. g

'a

(1) *DC 2-90-0049, " Reactor Trip On Lo Lo Steam Generator Level

,

Due To' MSIV Fast Closure."

.

'

On May 6,1990, a reactor trip occurred due to a lo lo

level in SG #3 caused by a. fast closure of MSIV 3026A. The

MSIV fast closure was due to a failure of the AX-1 relay in

the keep-oaen circuitry of the MSIV. This_ event will be-

further followed up when submitted as a LER.

(2) The following LERs were reviewed and closed.

(a) 50-424/90-02, ' Rev. O, " Procedure u lnadequacy Leads To ,

Inadequate Surveillance Tests."

1:

p As a follow-up corrective action: to LERs 50-424/88-28-01

'

and 50-424/88-31-01, plant. personnel were reviewing- test

procedures for adequacy. During the course of this review .

< on February 15, 1990, a system engineer discovered that the

Train C AFW system actuation relay K266 had not been

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properly tested in .accordance with TS 4.7.1.2.1.b.1 8- 2.

The SS was advised and an LCO was entered. The required

testing was performed and Train C of the AFW System was i

returned to service. Continuing the review on February 16,  !

1990, the system engineer discovered that the automatic DG 1

'

electrical trips that are bypassed upon loss.of voltage on

the emergency bus, concurrent with a SI actuation-signal,

'

were not tested to verify that they would actually bypass

'in accordance with TS 4.8.1.1.2.h.6.c. The SS was advised

and an LC0 was entered. .The appropriate testin!) was .

,

performed and both DG trains were returned to serv'ce. The

procedures did not'specify adequate testing for the DG' -

automatic ' electrical trip bypass functions or the Train C l

AFW actuation relay. These procedural. inadequacies

subsequently caused these events. The appropriate ',

L procedures have been revised for Unit 1 and will be revised

for. Unit 2. 'This licensee identified violation is not

being cited because criteria specified in Section V.G.1 of

the' NRC Enforcement Policy were satisfied. In order to

track this item, the following is established. - 3

L NCV 50-424/90-10-02, " Failure To Perform An Adequate

u ~ Surveillance Results In A Violation Of TS

.- 4.7.1.2.1.b.1 & 2 And .4.8.1.1.2.h.6.c With Respect To

I The AFW And DG System - LER 1-90-02."

-(b)*bO-424/90-03, Rev. O, " Missing Seismic Bolts On .

'

Transformers Leads To Unit Shutdown."

L On February 23, 1990, a system . engineer found 16

transformer core clamp bolts' missing on seismically . 1

qualified switchgear. The switchgear was .deenergized to 1

' replace the missing bolts which.resulted in a CIV being

~

deenergized Af ter the' four hour. time ? period for

reenergizing the valve - had expired,- unit shutdown was  ;

initiated asa required by the Technical. Specifications.

Although - the bolts were replaced -and the CIV was

reenergized before shutdown was completed,~ plant management.-

! , elected to complete the shutdown and enter into a planned 4

l refueling outage approximately four hours early. The cause

of this event is attributed- to an installation error during

the construction phase of the plant. The missing clamp

bolts were installed and all similar transformers in Units

l' and 2 were inspected with no 'other bolts found to be

missing. The inspector has .no further comments. 1

~

(c) 50-424/90-04, Rev. O, " Failure To Comply With Technical

Specification 3.0.4 Occurs On Entry Into Mode 6."

L On February 1,1990, a failure to comply with TS 3.0.4

L occurred when Unit 1 entered Mode 6 from Mode 5. Prior to

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entering Mode 6, an LCO had been initiated for Source Range ,

Channel IN31 to allow performance of an 18 month channel I

calibration. Althoegh this LCO remained in effect, the SS "

signed off on the applicable procedure to indicate he had -l

reviewed the LCO Book for. impact on entering Mode 6 and I

that approval was granted to change status from Mode 5 to I

Mode 6. After entry into Mode 6 the SS recognized that TS j

3.9.2 required two %Ms to be operable in Mode 6 and that a l'

S' , failure to compi.v with TS 3.0.4 had occurred. No innediate

action was reodired since the action requirements of TS 3.9.2 were satisfied. The root cause for this event is-

1. considered to be cognitive personnel error by the Shift

.E Superintendent. The Shif t . Superintendent has been

counseled and a copy of this LER will be placed in the

.

Operations Required Reading Book. This licensee identified

b violation is not being cited because criteria specified-_in '

l

Section V.G.1 of the NRC Enforcement Policy were satisfied.

In crder to track this item, the following is. established. >

!

'NCV 50-424/90-10-03, " Failure To Have Two Source Range I

Monitors Operable In Mode 6 Per TS 3.9.2 Resulting In-

A TS 3.0.4 Violation - LER 1-90-04."

(d)*50-424/90-05, Rev. O, " Personnel Error Leads To Fuel

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Handling Building Isolation." i

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/. On March '14, .1990, the Unit SS observed ~ that both Train A

l and 8 of the FHB Post Accident Ventilation Systems were

L operating and that the indicator lights associated with the ,

l' actuation handswitches for both trains showed that .an s

actuation .had occurred. Previously, only Train A had been ,

in service with Train B' in Standby. - Since Train B had not

been started as. part of a preplanned sequence during: i

testing -or reactor. operation, this was detennined to be an j

automatic . actuation ' of ESF equipment.

'

An~ automatic FHB

q isolation and associated alarms will' occurfon either a .high  !

. radiation signal or a loss of negative pressure in the FHB. l

..

During.this event,-no alarms were detected by control room ,

'

L personnel. Upon investigation, it was determined .that a

licensed reactor operator had failed to verify that the low

differential pressure actuation signal was blocked - as

required by procedure, when Train A was put'into service on

March 13, 1990. This. cognitive personnel error, in

conjunction with opening.of the doors to the FHB. probably

resulted in the actuation. The operator will be counseled  ;

regarding the importance of procedural compliance and other

licensed operators will be made aware of this event. The

licensee . identified violation is not being cited because

criteria specified in Section V.G.1 of the NRC Enforcement.

Policy were sr.tisfied. In order to track this item,.the

following is established.

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NCY 50-424/90-10-04, " Failure To. Follow Procedure

13320 Fuel Handling Building HVAC System, iW5ulting In

An. Inadvertent Automatic FHB Isolation - LER 90-05." l

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(e)*50-424/90-06,Rev.0,"LossOfOffsitePowerLeadsToSite

'

,. Area Emergency."

On March 20,-1990 Unit I was in a refueling outage and

Unit 2 was operating at 100% power. . At 9:20 a.m. EST, the i

'

driver of: a fuel truck in the switchyard backed into a 'l

support for the phase "C" insulator for the Unit 1 RAT 1A. i

The insulator and line fell causing a phase to ground l

-fault. Both-Unit 1 RAT 1A and Unit 2 RAT 2B High Side and <

Low Side breakers tripped, causing a loss of offsite power '

1

to the emergency busses. Unit 1 DG1A and Unit 2 DG2B

'

started, but DG1A tripped. This resulted'in a loss of RHR

v to the reactor core since the Unit-1 Train B RAT and DG1B ,

L were out of service for maintenance. A SAE was declared i

L' and the . site ' Emergency Plan was implemented. The RCS

heated up to 136 degrees F from 90 degree F before the DG

was emergency started and RHR restored. The. initial

notifications were not made within,15 minutes due to the

"

loss of power to the ENN. At l10:15 a.m. EST, the SAE was

~[ downgraded to an Alert after onsite power was restored, n

The direct cause of this. series of events-was a cognitive

'

personnel error. The truck driver failed- to use proper

'

backing procedures and hit a support, causing the phase to

ground fault and LOSP. The most problable cause ofithe

0G1A trip was ther intermittent actuation of the DG jacket

water temperature switches. Corrective actions- include

L- strengthening ' policies for control of vehicles, extensive 7

testing of the DG, replacement of suspect.DG temperature i

l switches, and improvements in the-ENN system. This-event

resulted in the dispatching ofLa NRC AIT and, subsequently. -

an III. The results of their investigation will be

,

published in c NUREG on June 8,1990. The inspector has no

L further coments.

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(f) 50-424/90-07, Rev. 0,. " Inadequately Monitored Liquid

Effluent Release."

1

I On March 16, 1990, an LCO for liquid radwaste effluent line

'

radiation monitor 1RE-0018 was exited after completion of

' maintenance and satisfactory testing. At the time the LCO l

was exited, monitor 1RE-0018 was still isolated from '

service by isolation valves closed under a clearance.' -This

rendered the monitor functionally inoperable. On March 17,

a liquid effluent release was made with 1RE-0018 isolated.

,

TS 3.3.3.9, Action 37 for 1-RE-0018 required the analysis  ;

i of two independent samples of the effluent in accordance

with TS'4.11.1.1.1 and at least two technically qualified

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members of the facility staff must independently verify the

release rate calculations and discharge line valving.

Since the radiation _ monitor was erroneously considered I

operable, only one sample was analyzed. This event was  ;

discovered on March 20, by personnel reviewing data from i

,

the Process and Effluent Radiation Monitoring System

computer. The cause of this event was cognitive personnel

error. A contributing cause was procedural inadequacy.

The Unit SS terminated the LCO without verifying the l

clearance was released. Corrective actions include 1

counseling the Unit SS and revising Procedures 13216-1 and

2 " Liquid Waste Release," and 10008-C, " Recording Limiting

Conditions For Operation."- This licensee identified '

violation is not being cited because criteria specified in -

Section V.G.1 of the NRC Enforcement Policy were satisfied.

In order to track this item, the following is established. .

NCV 50-424/90-10-05, " Failure To Monitor A Liquid

Waste' Release In Accordance With The Requirements Of

TS 3.3.3.9 - LER 1-90-07."

'

(g) 50-424/90-08, Rev. O, " Procedural Inadequacy Leads To A

Missed ASME Section XI' Valve Test."

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On March 21,.1990, while Unit'l was in Mode 6 at mid-loop

operation, it was discovered that a valid stroke time test

L was not performed for- BIT discharge isolation valve ,

1HV-8801B during the. previous quarterly inservice . tet...

^

The valve had been stroke time tested to its closed

position,-instead"of to its safety position which is open.

Since a valid surveillance ' test did not exist, this-

resultedz in a surveillance required by TS 4.0.5 and ASME  !

3 Section XI not being satisfied. _ The cause.of this event

,

was procedural -inadequacy. The safety; position for valve

1HV-8801B was incorrectly identified in the surveillance - .

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procedure as closed. The procedural inadequacy occurred as.

a result of a typographical error.in a . procedure revision

approved on December 12, 1989. On discovery of the-invalid

surveillance, valve IHV-8801B was declared inoperable. The

surveillance procedure was corrected and the surveillance  ;

w' was satisfactorily. perfonned. - Valve 1HV-8801B was -

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c" ' subsequently returned to operable status on March 29, 1990.

This licensee identified violation is not being cited

because criteria specified in Section V.G.1 of the NRC

Enforcements Policy were satisfied. In order to track this- 1

item, the following is established.

NCV 50-424/90-10-06, " Failure To Conduct An ASME Class

2 IST Per TS 4.0.5 Requirements Within The Quarterly

!. Time Interval Specified By ASME Section XI - LER

L 1-90-08."

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(h)' 50-425/90-01, Rev. O, " Misleading Task Sheet Leads To

inadequate Technical Specifications Surveillances."

<

On. January 3,1990, a surveillance to verify containment -

integrity- was completed and reviewed. The surveillance '

verified that valves 21204U4293 and 21204U4324 were closed

,

, and secured. Subsequently, on February 1,1990, the '

surveillance was repeated and valves 21204U4293 and ,

21204U4324 were again verified to be closed and secured. '

On February 28, the surveillance was again performed.

DuringthereviewbytheSS,henotedthatfortheprevious

month s surveillance, only 2 of the 41 valves and flanges r

listed in : the. associated procedure were addressed. He

initiated an investigation which determined that all 41

line items should have been -verified on January 3 'and

l February 1, as required by TS 4.6.1.1.a.. This

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specification requires.that containment penetrations- which

'are not closed by automatic isolation valves be verified

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closed:and secured at least once per 31 days. Therefore,

the surveillances performed on January 3 and February 1.-

failed to meet the requirement- of TS 4.6.1.1.a. The

principal reason for the missed surveillances was the

. format of :the Surveillance Task Sheets which resulted in

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cognitive personnel errors since they were led to believe ,

that- only two valves were required' to be surveilled. ~ By

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April 29, Surveillance Task Sheets were revised to either

, list all_ components to be surveilled or none at all, unless

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special conditions exist which would make a partial listing-

appropriate. This licensee identified violation-~is not

being cited because criteria specified in Section V.G.1 of ,

the NRC Enforcement' Policy were satisfied. In order to '

track this item,.the following_is1 established. 4

y NCV- 50-425/90-10-01,. " Inadequate. Surveillance Task

.

. Sheets' Lead To A Missed Surveillance Resulting In A'

Violation Of TS 4.6.1.1.a - LER 2-90-01."

(1) *50-425/90-02, Rev.-0, " Unit 2 Recctor Trip From Unit 11 i

Reserve Auxiliary Transformer Feeder Line Fault."

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On March 20,1990, a Unit 2 generator primary differential ,

relay energized causing a generator trip, turbine trip, and J

reactor trip. The relay energized upon a phase to ground

h fault which occurred when a truck operator backed' into as

b' support for Unit 1230kv-phase C " switcher" feeder line for

Unit 1 RAT A. The 230ky.line ~came .in contact- with' the

ground causing a fault which also tripped Unit 1 RAT A and

Unit 2 RAT B. Diesel Generator 2B started automatically

+

and restored. power to emergency bus 2BA03. A loss of power

to certain non-1E busses resulted in a trip of Reactor

Coolant-Pumps 2 and'4. Normal operating procedures were- '

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entered after the unit was stabilized in Mode 3. Although -j

the initiating event was a truck backing into the Unit 1

230kv support pole, causing a phase to ground fault, Unit 2

should not have tripped on the ground fault current

recorded during this event. Investigation determined the i

cause for the actuation on -the generator primary

, differential ' relay was an incorrect tap setting on the

variable ratio current transformers located on the i

generator main output breakers. A contributing cause was i

the failure to test the relay to verify that it was

receiving the proper voltage and current signals from the

current transformers. The tap settings for the current  ;

transfonners were corrected and the remaining protective

relays will be tested to verify they are receiving proper  !

voltage and current signals. The inspector has no further

'

comments.

4. Plant Startup From Refueling '(71711)

The inspectors ascertained whether systems disturbed or tested during 1R2

were returned' to an operable status before plant startup by walking down

' portions of the feedwater system, reactor coolant system and emergency -

,

power. The inspectors also conducted a tour of the containment building

?*!

with operations and' health physics personnel. The containment RHR and CS

sumps. were verified to. be free of foreign material that could possibly

prevent the systems from performing their intended safety function. The  ;

inspectors also verified that the flow path- from the refueling canal to  !

containment lower > level was not obstructed by ensuring that the refueling

canal bottom drain; flanges were removed. The inspectors witnessed

_ portions 1of plant startup. heatup, approach to criticality, and core

physics testing following the outage to verify compliance with approved

procedures.

On April 15, 1990, operators and reactor engineers commenced: low power i'

physics testing following refueling outage '1R2. Rod worth measurements-

were inl progress using the rod swap technique. The worth of SDB-B. the  !

reference bank, had been detennined. Data had been collected for control

i banks'B and D:and SDB-A, C, D, and E. SDB-E was being withdrawn to the

full out position, 228 steps Reactor power was being maintained at 3.5-8

amps by inserting . SDB-B. While attempting to withdraw SDB-E from 170

steps, all four rods in the group dropped. The reactor became subcritical

with power stabilizing at 1.3-10 amps. Negative- reactivity was i

compensated for by withdrawing SDB-B _ to a reactor; power of 1.5-8 amps .

(104.5 steps). SDB-E was exercised per operations procedure 18003-C _ Rod 1

Control System Malfunction. . The SS, Shift ' Superintendent and Reactor

. i

Engineer declared. special test exception TS 3.10.2 still applicable. No

procedural' or .TS violations were noted by the inspectors. However,

subsequent rod testing should have been accomplished via a troubleshooting

procedure, = not the procedure used to detennine rod bank worth- or rod

- , control system malfunction. The control room annunciator for rod control

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urgent failure lead operators to one of the power cabinets (SCDE) where a

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regulation failure for- power to movable gripper coils for SDB-E was

diagnosed. A weakness w:: identified in the area of administrative

controls for utilizing a procedure beyond the scope of its-intended use.

5. . A11egetions

A11egatica Ril-89-A-0097, Alleged discrimination for Using the Licensee's 1

. . Quality Concern Program. l

Concern l

l

The alleger expressed three concerns to the NRC: (1) That he had been laid

off in retaliation for submitting a quality concern; (2)Thatindividuals l

who used the Quality Concern Program were usually laid off first during

reductions in force, ~ and; (3) that another employee's confidentiality had j

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been violated.

,. Discussion

3

The licensee -investigated the alleger's concerns and reported their

results - to the alleger in separate letters. The NRC reviewed the

licensee's actions and responses during this reporting period.

Conclusion

At the conclusion of these investigations, the first two concerns

expressed to the NRC by the alleger were not substantiated. The third

concern was substantiated -by the licensee as an unintentional breech of

'

confidentiality. The alleger acknowledged that he was satisfied with both .

the results of the licensee's review and corrective actions. The results 4

of' the NRC investigation do not differ from that of the licensee's. The ,

licensee's- QJality Concern Program thoroughly,-investigated = all concerns

and material issues raised by.the alleger's quality concern. The alleger

-was, interviewed on numerous occasions over a two year period, had his

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deposition taken, and- had the opportunityf to repeat his concerns in a

meeting - with1 NRC representation. 'The Quality Concern Program also

interviewed all material- witnesses named by the alleger and conducted

technical investigations on the condition of all hardware questioned by

the alleger. The licensee's actions in this matter were clearly extensive

and thorough, and all of the alleger's concerns were adequately addressed.4

This allegation is considered closed. ,

,

6.. Evaluation of Licen:ee Quality Assurance Program Implementation - (35502)

,

A mid-SALP. review was conducted during the April 5,1990, QPPR ' meeting.

Each SALP category was evaluated by reviewing inspection reports, past *

SALP findings, the OIL, licensee corrective actions to NRC findings, and.

the input from the resident inspectors.

l No significant trends were identified in any of the SALP categories that ,

would require a change to the MIP; however, the results of the IIT

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inspection need to be evaluated with respect to this decision. Comand

and control during emergency conditions continues to be a concern. This

was demonstrated by the untimely notifications during the March 20, 1990

Site Area Emergency. Security performance will continue to be closely

scrutinized to identify the impact of the new Security Manager.

7. ExitInterviews-(30703)

The inspection scope and findings were summarized on May 18, 1990 with

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

the areas 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 seven Licensee Event Reports and one allegation. The

items identified during this inspection were:

VIO 50-424/90-10-01, " Failure To Ensure Proper Routing And Slope To

The RCS Temporary Level Indication Tygon Tube" - paragraph 2,a.

SCV 50-424/90-10-02 " Failure To Perfonn An Adequate Surveillance

Results In A Violation Of TS 4.7.1.2.1.b.1 & 2 And 4.8.1.1.2.h.6.c

With Respect To The AFW And DG Systems" - paragraph 3.b.(2)(a).

NCV 50-424/90-10-03, " Failure To Have 2 Source Range Monitors

Operable In Mode 6 Per TS 3.9.2 Resulting In A TS 3.0.4 Violation" -

paragraph 3.b.(2)(c). -

NCV 50-424/90-10-04 " Failure To Follow Procedure 13320 Fuel

Handling Building HVAC: System, Resulting In An Inadvertent' Automatic

FHBisolation"-paragraph 3.b.(2)(d).

NCV 50-424/90-10-05, " Failure To Monitor A Liquid Waste Release

In Acccrdance With The Requirements Of TS 3.3.3.9" - paragraph

3.b.(2)(5).

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NCY 50-424/90-10-06, " Failure To Conduct An ASME Class 2 IST Per TS '

4.0.5 Requirements Within The Quarterl I

ASME Section XI" - paragraph 3 b.(2)(2)y Time Interval Specified By

.

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NCV 50-425/90-10-01, " Inadequate Surveillance Task Sheets Lead To A

Missed Surveillance Resulting In A Violation Of TS 4.6.1.1.a" -

paragraph 3.b.(2)(h).

l 7. Acronyms And Initialisms

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ACOT Analog' Channel Operability Test

c AFW Auxiliary Feedwater System

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SMA Seismic Motion Accelemoter

SRM Source Range Monitor

SS: Shift Supervisor

SSPS Solid State Protection System

TAD 0T Trip Actuating Device Operational Test ,

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TCP Temporary-Change of Procedure

TDAFW Turbine Driven AFW Pump ,

TS Technical Specification '

c, VEGP. Vogtle Electric Generating Plant

VIO Violation

WRT Work Request Tag

1R2 Second Refueling Outage - Unit 1

2R1 First Refueling Outage - Unit 2 ,

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