Text

Forwards marked-up Copy of Insp Repts 50-424/90-19 & 50/425/90-19,Suppl 1,which Documents Insp Teams Review & Conclusions Re Allegations at Time of Insp Exit Meeting on 900817
	ML20129F418
Person / Time
Site:	Vogtle
Issue date:	07/23/1991
From:	Reyes L; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	Vandenburgh C; Office of Nuclear Reactor Regulation
Shared Package
ML20129F106	List: IA-94-208, Partially Released Ltr Re Conversations W/Staff & Ltrs Identifying Allegations Related to Operations & TS Practices at Plant.Review Completed & Findings Documented in Encls; ML20129F103; ML20129F113; ML20129F174; ML20129F222; ML20129F230; ML20129F243; ML20129F264; ML20129F418; ML20129F436; ML20129F452; ML20129F511; ML20129F549; ML20129F581; ML20129F593; ML20129F599; ML20129F606; ML20129F613; ML20129F621; ML20129F628; ML20129F638; ML20129F643; ML20129F648; ML20129F663; ML20129F684; ML20129F694; ML20129F724; ML20129F729; ML20129F735; ML20129F748; ML20129F865; ML20129F990; ML20129G037;
References
	FOIA-94-208 NUDOCS 9610040144
	Download: ML20129F418 (56)
	v • d • e

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UNITED STATES

'o

NUCLEAR REGULATCRY COMMISSION

.

[

CE! ION 11

.

,

y

,P

101 MARIETTA STREET, N.W.

s

ATLANTA, GEOR0dA 30323

,

'+,.....

JUL 2 31991

MEMORANN N F0P,

Chris A. VanDenburgh, Chief, Reactive Inspection Section 2

Vendor Inspection Branch

Division of Reactor Inspections and Safeguards

FROM:

Luis A. Reyes, Director

e

Division of Reactor Projects

SUBJECT:

V0GTLE SPECIAL TEAM INSPECTION - ALLEGATION FOLLOWUP TEAM

DRAFT INSPECTION REPORT (INSPECTION REPORT NOS. 50-424/90-XX

AND50-425/90-XX)

This memorandum refers to the special inspection conducted on August 6

through 17, 1990, at the Vogtle Electric Generating Plant (VEGP).

This

inspection involved a review of several allegations regarding the safe

operation of VEGP and the review of operational activities generally related

'

to the allegations. As discussed in the inspection plan, the inspection was

performed by two separate teams--an operational followup and an allegation

,

followup team.

As decided in a meeting held in Nuclear Regulatory Commission (NRC) head-

quarters on September 26, 1990, the allegation followup team's findings and

conclusions was not included in Inspection Report 50-424/90-19;

50-425/90-19. This information was to be withheld pending the completion of

an Office of Investigation review of the allegations and the inspection

team's conclusions. On January 11, 1991, Inspection Report 50-424,425/90-19

was issued which included the operational followup team findings.

The

remaining issues from the allegation followup team were then left in

Inspection Report 90-XX, pending the completion of Ois' review of the

allegations.

'

On July 9,1991, a meeting was held in Region II, with members of Region

II-DRP, 01, and NRR-PD3-2 and Regional management.

It was determined to

j

issue the remainder of the 50-424,425/90-19 report, except for the following

i

,

issues: 1) 12.3 Missed Containment Isolation Valve Surveillance; 2) 12.4

-

Mode Change With Inoperable Source Range Monitor Nuclear Instrument; 3) 1

2.7 Reliability of Emergency Diesel Generators and their corresponding parts

to the Notice of Violations.

This memorandum forwards a marked up copy of Inspection Report 50-424,

425!90-19, Supplement 1, which documents the inspection team's review and

conclusions regarding the allegations as of the time of the inspection exit

meeting on August 17, 1990

The report has already been reviewed by the

.

Office of Investigation in Region II for information that clight compromise

'

their on going investigations.

The information that was considered

pertinent to these investigations will not be included in the issued report.

l

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9610040144 960827

l

PDR

FOIA

COLAPIN94-200

PDR

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"

I

'

..

Chris A. VanDenburgh

2

JUL 2 31991

l

.

!i

If you have any questions concerning this issue, please contact P. Skinner

!

at Ext.16299 or S. Vias at Ext.15350.

'i

~/

4

'

Luis A. R

Enclosures

~1.

Draft Notice of Violation

i

2.

Draft -Inspection Report

'

50-424,425/90-19 Supplement I

cc w/encis:

!

L. Robinson 01.

i

D. Hood, NRR, PD3-2~

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s.445

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6. Jul

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ioi maasts ta sie<tti.

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anc.mA.etonc A 3o323

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M%II G;

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i

Docket Nos. 50-424 and 50-425

J

'

License Nos. NPF-68 and NPF-81

i

Georgia Power Company

ATTN:

Nr. W. G. Hairston, III

i

Senior Vice President -

.

i

'

4

j

Nuclear operations

q

L

P.O. Box 1295

y

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<

)

Birmingham, AL 35201

Gentlemen:

SUBJECT:

VOGTLE SPECIAL TEAM INSPECTION AND NOTICE OF VIDIATION

1

(NRC

INSPECTION

REPORT

NOS.

50-424/90-19

AND

3

50-425/90-1gg SuppleAw

This refers to the

inspection conducted by an NRC Special p

!

Inspection Team on August 6 through 17, 1990.Jr Tne inspem. ion .Ag'

i

j

included a

review of activities authorized for your Vogtle

3

i

facility.

At the conclusion of the inspection, thePffindings were

}[

i

discussed with those members of your staff identified in the

j

enclosed inspection report.

i

Areas examined during the inspection are identified in the report.

!

Within

these

areas,

the

inspection

consisted

of

selective

examinations of procedures and representative records, interviews

j.

with personnel, and observation of activities in progress.

Based on the results of this inspection, certain of your activities

.

appeared to be in violation of NRC requirements, as specified in

j

the enclosed Notice of Violation (Notice).

i

-1^3 *hatthe inspection concluded that the facility was opera

_

a safe mann..- Q dance with the requirements

perating

l

license, we are conc =..:f

that there

e

veral operational

policies and programs whers we

___ -

laantified. As part of

your response to the

ons identified n

sed Notice,

,

you are a

sted to address each of the weaknesses

___A

l

ion summary.

!

You are required to respond to this letter and Notice and should

,

follow the instructions specified in the enclosed Notice when

preparing your response to the violations.

In your response, you

1

l

should document the specific actions taken and any additional

i

actions you plan to prevent recurrence.

After reviewing your

response to this Notice, including your proposed corrective actions

'

and the results of future inspections, the NRC will determine

whether further NRC enforcement action is necessary to ensure

compliance with NRC regulatory requirements.

-

.

Q.4 4t g

f & LeM-e t %

w e.e U An, 491. d* k, & ~?

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coorgio Pov3r CcIpany

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you should respond to each of the o

a

4

esses are

E d!*ionally,

(The

weaknessesMfied within the report.

ary.)

The response

specifically annotated-in .the InspectioMhe s3gnificance of the weaknesses

.

should address your analysis at these operat

tices do not

,

and your actions toof non-compliance or reduce the marg n

7:'ev^

evolve int

p ant.

In accordance with section 2.790 of the NRC's " Rules of Practice,"

.'

a copy of this letter and its enclosures will be placed in the NRC

Public Document Room.

r

The responses directed by this letter and the enclosed

and Budget as required by the Paperwork Reduction Act of 1980,

Pub. L. No. 96. 511.

Should you have any questions concerning this letter, please

contact us.

4

sincerely,

0

>

i

ne

r

f

-

>' Regional Administrator

<

Region II

,

.

l

Enclosures:

1.

Notice of Violation

Inspection Report 50-424/90-191

2.

50-425/90-19 % (pl* * d 1

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LIMITED DISTRIBUTION - Not For Public Ralesco

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DRAFT - PREDECISIONAL INFORMATION

,-

,

-> %[

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May 9,

1991

.

,,

MEMO

DUM FOR:

Luis A. Reyes, Director

g:~ y

.i

Division of Reactor Projects

fq ',* :

~

-

Region II

.6.'N ', ' '

NY '- " '-4'.:..T + '-

FROM:

Chris A. VanDenburgh, Chief

.1

Reactive Inspection Section - 2

[5EF.('-j ,y..

Vendor Inspection Branch

Division of Reactor Inspection

and

i

'

Safeguards

SUIL7ECT:

VOGTLE

ECIAL TEAM INSPECTION - ALLEG TION FOLLOWUP TEAM

DRAFT

SPECTION

REPORT

(INSPE

ION

REPORT

NOS.

!

50-424/90

x AND 50-425/90-xx)

l

This memorandum refer

to the special

spection conducted on

August 6 through 17, 19

, at the Vogtle

lectric Generating Plant

.

(VEGP).

This inspection nvolved a rev ew of several allegations

regarding the safe operati

of VEGP a d the review of operational

activities generally relate to the

legations.

As discussed in

i

the inspection plan, the ins

ction was performed by two separate

teams--an operational followup nd n allegation followup team. At

the conclusion of the inspect

all of the inspection team's

,

conclusions with respect to the

erations and allegation followup

were discussed with the me

s

GP's staff identified in the

enclosed draft inspection re rt.

I

As decided in a meeting hel in Nuclea

egulatory Commission (NRC)

headquarters on September 26, 1990, the

llegation followup team's

findings and conclusio

have not bee

included in Inspection

Report 424/90-19; 50-4 /90-19. This info

ation has been withheld

pending the completi

of an office of Inve

igation review of the

allegations and the

spection team's conclus

ns. This memorandum

,

forwards a draft

spection report (50-424/

-xx; 50-425/90-xx)

'

s

which documents

e inspection team's revie

and conclusions

regarding the a

egations as of the time of th

inspection exit

'

meeting on Au

t 17, 1990.

The areas e amined during the inspection are iden

fled in the

.

-

inspection eport. As discussed in Inspection Report

-424/90-19;

50-425/90

9, the inspection team concluded that the f

ility was

safely

rated.

However, the inspection identifie

several

'

,

i

instan

s in which the VEGP was not operated in accordance

ith the

,

inten

of

the Technical

Specifications.

In

additio

the

insp ction

identified

several

potential

weaknesses

i

the

'

fa

lities' operational r licies and practicc..

i

LIMITED DISTRIBUTION - Not For Public Release

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DRAFT - PREDECISIONAL INFORMATION

i

Lui

A. Reyes

-2-

'

The ins'pection team's review of the allegations identified severa

additio 1 weaknesses in these operational policies and practice .

These ar

identified in the inspection summary of the enci

ed

,

'

draft insp ction report.

activities app,esults of this inspection of the allegations,

Based on the t

artain

eared to be in violation of NRC require

nts, as

specified in t e enclosed draft Notice of Violation (Notice).

These violations re important because they indicate (

a failure

administrative proc,equirements of the Technical Speci

cations and

to implement the r

9 ures, and (2) the failure to pr vide accurate

d

-

information to the

.

As part of the response to the violations /dentified in the

~

enclosed notice, VEGP should also be requeste

to address each of

'

the concerns listed in t e inspection summa

.

t

Enclosures:

1.

Draft Notice of Violation

2.

Draft Inspection Report 50-434/90-xx; 50-425/90-xx

/

cc:

BKGrimes

EWBrach

.

d

I

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LINITED DISTRIBUTION - Not For Public Release

i

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LIMITED DISTRIBUTION - Not For Public RalGG00

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DRAFT - PREDECISIONAL INFORMATICN

,

N

>

l

NRR/DRIS

RII/DRP

JWilcox*

RAlello*

RStarkey*

MBranch*

9/ /90

8/31/9

8/31/90

I

RII/DRP

RII

RS

NRR/DLPQ

RJI/DRS

I4arner*

MThomas*

NHuneauller*

ylore

9/27/90

/31/90

8/31/90

8/ 1/90

i

RII/DRP

NRR/DRIS

i

RCarroll*

CVanDenburgh

'

8/31/90

9/ /90

'

Se previous concurrences

'

LIMITED DISTRIBUTION - Not For Public Release

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LIMITED DISTRIBUTION - Not For Public R31easo

DRAFT - PREDECISIONAL INFORMATION

ENCLOSURE 1

!

,

NOTICE OF VIOIATION

<

l

Georgia Power Company

Docket Nos. 50-424 and 50-425

Vogtle Electric Generating Plant

License Nos. NPF-68 and NPF-81

Units 1 and 2

3

'

,

l

During an NRC inspection conducted on August 6 through 17, 1990,

,

violations of NRC requirements were identified.

In accordance with

the " General Statement of Policy and Procedure for NRC Enforcement

i

!

Actions," 10 CFR Part 2,

Appendix C (1990), the violations are

!

listed below.

'

i

!

l

A.

10 CFR Part 50.9, " Completeness and Accuracy of Information,"

requires that info mation provided to the NRC by a licensee

"

shall be complete and accurate in all material respects.

I

Contrary to the above,

the licensee provided g curate

information

to

the

inspection

team

on 4heee

separate

,

1

occasions. Although the information was provided in unsworn,

1

oral statements, the information provided was significant to

the licensing process.

The information was provided by

l

licensed operators, supervisors and management concerning

i

information which was within their specific responsibilities.

1

i

l

The d4ve examples were as follows.

(50-424/90-eese-95; 50-

!

90-eas-99)

li - I'L

l

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l

IV'2-

'

1.

Containment

Isolation

Valves:

During

a

Unit

2

,

i

surveillance procedure, the unit shift supervisor (USS)

'

stated, and the operations manager later confirmed, that

j

the containment isolation valves for the hydrogen monitor

system were allowed to be opened without entering the

3

limiting

condition

for

operation

(Ico)

action

requirements for Technical specification (Ts) 3.6.3

because the valves

received an automatic

isolation

.3

signal. The inspection identified that these containment

]

isolation valves were remotely-operated, manual valves

without automatic isolation signals.

(Discussed in

Section 2.2.1.1 of Inspection Report 50-424/90-19; 50-

425/90-19)

2.

Snubber Reduction:

The operations manager stated that,

4

!

after the second Unit i refueling outage (1R2), the

!

modifications to tt:e snubbers were done in conjunction

.

LIMITED DISTRIBUTION - Not For Public Release

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personnel Accountabiliev: The operations manager stated that

the shift superintendents (SSs) reported directly to the

operations manager and that he personally prepared their

performance appraisals.

The inspection identified that the

sss reported to the unit superintendent (Us), and that the US

1

personally prepared the performance appraisals of the Sss.

j

(Discussed ;,n section 2.Af of this inspection report)

O

,

4 y)

Ts 3. 0. 3 Actions:

The unit superintendent indicated that

there were no Operations

Department actions which were

i

,2

anticipated or required within the first three hours of

'

entering the action statement of TS 3.0.3.

The inspection

identified that the VEGP management policy and stated practice

required preparations

for a

power

reduction,

including

,

8

informing

the

load

dispatcher within

the

first

hour.

(Discussed in section 2.1.1.3 of Inspection Report 50-424/90-

'

19; 50-425/90-19)

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DRAFT - PREDECISIONAL INFORMATION

1

with preplanned system outages which were required for

i

other preventive or corrective maintenance or testing.

The

inspection

identified that

few of the snubber

<

.

modifications were done jointly with pre-planned system

i

outages.

(Discussed in Section 2.1.1.4 of Inspection

j

,

Report 50-424/90-19; 50-425/90-19)

'

.

Diesel-Generater (w)

=14mhl11tyr

vrco

j

E== v.nq

.ncorrectly counted the number of starts and failurgot

,

1

t

DGs and incorrectly represented the EDG re ability

i

in a

on II presentation on April 9, 199

Although

t

I

the presen

ion was not 'ntended to repr sent a specific

,

number

of

su

ssful

valid

test

s specified

in

Regulatory Guide (

1.108 and T

. 8.1.1.2a, but rather

,

i

to describe the EDG ma

nan

test program and the EDG

,

reliability status, the

was not informed of the

i

incorrect information

il th

C asked for it during

j

the inspection.

he confirmat

of action

(CAL)

i

response and L

nsee Event Report (LER

-006 were also

,

incorrect A cause they were based on t

EDG start

l

inform fon that was compiled for the VEGP pre

ation

I

j

in

e Region II Office.

(Discussed in Section 2.

f

g[

s inspection report)

I

This is a Severity M vel IV violation (Supplement VII).

2

l

B.

Technical Specification 6.7.1.a

requires

that

written

j

procedures be established or implemented for those activities

delineated in Appendix A of Regulatory Guide 1.33, Revision 2,

'

i

February 1978.

l

Contrary to the above, two examples were identified in which

the licensee failed to establish or implement the procedures

for these required activities as follows:

(50-424/90-ac'M? :

19- G

50-425/90-asu-49)

i

19 - 6

!

1.

Administrative Procedure 00150-C, " Deficiency Control,"

i-

states that a deficiency card must be written if the

deficiency involves safety-related components which are

i

to

be

dispositioned

"use-as-is/ repair,"

or

other

conditions

involving safety-related components which

require engineering support or other technical assistance

-

to determine if the component is deficient.

d

on August 17, 1990, the NRC identified that a deficiency

card was not written on re'idual heat removal (RHR) pump

f1B (a safety-related component) to document the pump's

'

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DRAFT - PREDECISIONAL INFORMATION

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i

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degraded conditions which were dispositioned "use-as-is".

(Discussed in Section 2.2 of this inspection report)

'

i

2.

Administrative Procedure 00100-C,

" Quality Assurance

Records Administration," Paragraph

4.1.1.8,

specifies

that

quality

assurance

(QA)

records

will

exh! bit

necessary and appropriate signatures or initials and

[

dates.

.

On August 17, 1990, the NRC identified that the Unit

i

Superintendent incorrectly initialed, dated, and signed

i

a QA record which voided Temporary Change Procedure (TCP)

1802-C-7-90-1 to Abnormal Operating Procedure 18028-C,

" Loss of Instrument Air," with the date of June 12, 1990,

i

in lieu of the actual date (June 15, 1990) on which the

document was signed.

(Discussed in Section Jdf of this

'

i

inspection report)

g.1b

4

{

This is a Severity Level IV violation (Supplement I).

i

!

C=

re. ::, w=n= :. cra.r4

--cI,

corr.cu=

-

Action," requires measures to be established to ensure tha [

.

I

l-

onditions adverse to quality are promptly identified f d

'

c

rected.

In the case of significant conditions adve Ne to

!

qua

, the measures are required to ensure that theguse of

i

the co

tion is determined and corrective action Jd taken to

i

preclude

tition.

/

!

/

Contrary to th

ve, two examples were identified in which

,

preclude repetitl'o/aplement

to determine and

adequate

the licensee fa

d

n as follows:

(50-

corrective actions

l

424/90 .ans- et)

/

!

i

14 - 14

/

j

1.

On August 17,1900, theWC4etermined that the licensee

"

did not identify the foitat and normal use of the 14:0

.

status sheet as one oJA.he causes of the event described

j

in Licensee Event Report (LER) '90,-004, " Failure To comply

With Technical Specification 3.0N< occurs on Entry Into

Mode 6"; therefore, corrective action was not taken to

,

preclude repet'ition of the failure to review Ico-required

actions orAremarks which may be on the back side of the

i

LCo status sheet.

(Discussed in Section '2.4

of this

,

inspect' ion report)

i

/

1-

2.

(chnical Specifications 4.8.1.1.3 and 6.8.2 require hat

all valid or non-valid EDG failures be reported to th

NRC in a special report within 30 days.

In addition,

,

i

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DRAFT - PREDECISIONAL INFORMATION

'

.

Operations Procedure 55038-C, " Diesel Start Log," Section

. 0, requires that all EDG failures shall be re or(ed to

'

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t

RC in a special report.

!

On Augu

17,

1990,

the

NRC

ideptified

that the

j

corrective a

ions taken in response 4o a previous notice

inadequate./ Inspection Report 50-

'

of violation

e

.

424/s7-57 (dated

aber S A987) previously identified

j

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a violation of Tech ' al Specification

4.8.1.1.3,

in

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that, all EDG failuresmo

not reported to the NRC in a

special report. During a re

w of the start records for

i

EDG f1B duringA:he period of

h 21 through June 14,

i

1990, the NBC' identified that EDG

lures had occurred

l

which were'not submitted to the NRC in

special report.

In a difion, the NRC identified that Opera

ns Procedure

.l

5

8-C provided inadequate guidance to i

tify and

classify EDG failures.

(Discussed in Section 2.

this

j

inspection report)

This is a Severity I4 vel IV violation (Supplement I).

!

Pursuant to the provisions of 10 CFR 2.201, Georgia Power Company

is hereby required to submit a written statement or explanation to

i

the U.S. Nuclear Regulatory Commission, ATTN:

Document Control

,

Desk,

Washington,

DC

20555,

with a copy to the Regional

Administrator, Region II, and, if applicable, a copy to the NRC

Resident Inspector within 30 days of the date of the letter

i

transmitting this Notice Of Viel: tic . (;5tia) . This reply should

be clearly marked as a " Reply to a Notice of Violation" and should

include for each violation:

(1) the reason for the violation, or,

if contested, the basis for disputing the violation,

(2) the

corrective steps that have been taken and the results achieved, (3)

the corrective steps

that

will

be taken to avoid

further

violations, and (4) the date when full compliance will be achieved.

If an adequate reply is not received within the time specified in

this Notice, an order may be issued to show cause why the license

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should not be modified, suspended, or revoked, or why such other

action as may be proper should not be taken.

Where good cause is

shown, consideration will be given to extending the response time.

,

FOR THE NUCLEAR REGULATORY COMMISSION

l

Stuart D. Ebneter

Regional Administrator

Region II

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Dated at Atlanta, Georgia

this

day of

199p

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50-424/90-N and 50-425/90-N , $u f pleMe aN I

Report No.:

Licensce:

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 Electric Generating Plant, Units 1 and 2

i

Inspection Conducted:

August 6-17, 1990

1

Team Members:

Ron Aiello - Resident Inspector, Vogtle

Morris Branch - Senior Resident Inspector, Watts Barr

.

'

Robert E. Carroll, Jr. - Project Engineer, DRP, Region II

Larry Garner - Senior Resident Inspector, Robinson

Neal K. Nuneauller - Licensing Examiner, NRR

Larry L. Robinson - Investigator, 01, Region II

Robert D. Starkey - Resident Inspector, Vogtle

Craig T. Tate - Investigator, 01, Region II

Peter A. Taylor - Reactor Inspector, DRS, Region II

McKenzie Thomas - Reactor Inspector, DRS, Region II

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John D. Wilcox, Jr. - Operations Engineer, NRR

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Team Leader:,.

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" Chfis E Va'nDenburgh, SectTon C'hief

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Division of Reactor Inspections and Safeguards

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Office of Nuclear Reactor Regulation

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Appeeved by:

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suf W f

90,

TABLE OF CONTENTS

g

INSPECTION SUMMARY........................................

-t"

T'

1.0

INS PECTION OBJ ECTIVES . . . . . . . . . . . . . . . . . . . . . . . . .

.....

2.0

ALLEGATION FOLLOWUP..................................

Ar

2.1

Improper Installation of FAVA System............

9-

2.2

Operability of Residual Heat Removal Pump.......

41-

2.3

Mi.e.d -Ce..teinsent Isc1; tion '!;1v; Ourveillance.

10

2.4

Msde Cheng; "'ith In:perable Source Rang?

3 -Moni te r Ne:l e r ! n t re :n t . . . . . . . . . . . . . . . . . . . . . .

10-

2 .JF Backdating o f Signatures. . . . . . . . . . . . . . . . . . . . . . . . jkP

2.sr Reportability of Previous Engineered Safety

Y Features Actuation System Load Sequencer

24'

Outages.........................................

3.'

Reliability cf Esergency Olcsci Cenereters. . . . . .

2-

2.,r

Air Quality of Emergency Diesel Generator

6 Starting Air System.............................

34F

b eportability of Previous System Outages........

25'

R

2.F

'2..'7 .240 Intimidation of Plant Review Board Members. . . . . .

33-

2 b L kT Personnel Accountability........................

,3&-

3.0

EXIT INTERVIEWS......................................

)4P

APPENDIX 1 - LIST OF TRANSCRIBED INTERVIEWS. . . . . . . . . . . . . . . JFr

APPENDIX 2 - PERSONS CONTACT 2D. . . . . . . . . . . . . . . . . . . . . . . . . . . .

J1r

APPENDIX 3 - LIST OF ACRONYMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

,$8

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INSPECTION SUMMARY

Recent activities which have occurred at the Vogtle Electric

Generating Plant (VEGP) have raised concerns within the Nuclear

Regulatory Commission (NRC) as to the ability and the deterinination

of the licensee to operate the facility in a safe and conservative

manner. To address this concern, the NRC performed a special team

inspection to deter 1mine if the licensee operates the facility in

accordance with approved procedures and within the requirements and

,

intent of the facility's operating license.

In addition to the

occurrence of specific events, NRC concerns regarding the safe

operation of the facility were heightened with the receipt of

several allegations relating to operational activities at VEGP.

The aggregation of the facts and circumstances associated with the

operational events and the allegations was viewed as a possible

indicator of a non-conservative attitude on the part of the

facility's operating staff which warranted the immediate initiation

I

of special inspection activities.

Specifically, the inspection objectives were to:

1)

Assess the operational philosophy, policy, procedures and

practices of the f acility's operating staff and management

regarding operational safety.

'

2)

Determine the technical validity and safety significance of

{

d cf the allegations and their impact on the safe and

conservative operation of the facility.

These inspection objectives were accomplished by the use of two

inspection teams--an operations followup team and an allegations

followup team.

The offorts of these two inspection teams were

closely coordinated;

however,

they independently pursued the

objectives outlined above.

The operations followup team monitored control room activities on

a 24-hour basis

in order to:

(1)

evaluate the operational

philosophy, policies, procedures, and practices of the operating

.

staff and management and (2) determine if the plant was being

i

operated in a safe and conservative manner in accordance with the

l

facilities' operating license.

!

The allegations followup team verified the technical validity and

safety significance of each ef the allegations.

In addition, with

k

the assistance of the OI staff, this team interviewed * members of

the plant

staff

in order to

determine

(1)

their personal

4

j

involvement and knowledge of the specific allegations and (2) their

practice and understanding of the station operational policies.

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These-interviews were transcribed. Although an OI investigator was

l

assigned to the inspection team to assist during the transcribed

1

interviews, this inspection was not an investigation into th

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intent of the alleaed violations. Nu muesrep.a5 **t

w (:iewen. e<mo not iunst.

,

The inspection substantiated the occurrence of the specific events

described in the allegations.

These events resulted i

t

q

i

examples o

v, lations of regulatory requirements (50-424/90-

D

l

50-425/90-

)2' "" G/^; .. ;^ ; and two of the events were

.

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previously identified as non-cited violations (50-424/90-10-03 and

50-425/90-01-01).

: : , the ine;rrtier did net rubettntirt

d

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-th:t th: : ::t: :' vielstien

vere-perfe n f rith the full-

L,wledge of 'J::Or ::n:; ::nt.

Sie cerclerier See herM rper :

j

' ;vice Of the lic:::ee's recorde

^^d

the Svern testisery ef the

i

j

pe;ple inv;17;d in th: ev::t;.

i:

'

)

inspection

also

identified

that

on

several

oc

s

i

respons

agers and supervisors verbally su

naccurate

information to t e '

tion team durin

spection. Although

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the inspection team was

a out the accuracy of the

i

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information provided

an did n

a basis to conclude or

suspect th

examples were the result o

ss disregard

i

,

atory requirements or individual wrongdoing.

1

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Ioom m.' ' b observations and conclusions of the inspection tem #g::o )

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50-424/90-19; 50-425/90-1C In ,)#-

/

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are detailed in Inspection Report

addition, the bases for these previous conclusions are summarized

below.

4

i

Doerational Policies and Practices

3

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NRC Inspection Report 50-424/90-19; 50-425/90-19 identified several

examples in which the licensee's operational policies and practices

'

had the potential to adversely affect the operation of the

i

facility.

llegation followup team's review of the allegations

identifiedn

additional examples in which the licensee's l

l

operational policies and practices had the potential to adversely

affect the safe operation of the facility 1

"^r err;1:-

{

I

1)

The licensee's method of conducting Plant Review Board (PRB)

meetings had the potential for adversely affecting open

,

i

discussions among the PRB members.

This concern was based on

i

an example in which a PRB voting member felt intimidated and

feared retribution during a PRB meeting because of the

4

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presence of the general manager and the absence of dissenting

opinions in the PRB meeting minutes.

Continued licensee

action is necessary to ensure that PRB members freely and

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openly express their technical opinions and safety concerns.

(Section.2AC)

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2)

The

licensee's

practice

of

signing

and dating

quality

i

assurance records was controlled by administrative procedures;

j

however, there was a confirmed exasple in which a signature

j

was backdated to reflect the actual date of performance.

The

'

}

backdating

of

TCP

1802-C-7-90-1

was

verified

and

was

1

identified as Violation 50-424/90-

~6;

50-4 25/90-aus-et .

I

(Section 24)

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3)

The licensee' practice of not initiating a deficiency card

(DC)

during

troubleshooting

activities

involving

the

questioned operability of the residual heat removal (RHR) pump

prevented a documented engineering evaluation for either the

nuclear service cooling water (NSCW) outlet leak or the

excessive vibration on the RHR motor.

The

fal' lure to

,

implesent this administrative procedure was identified as

.4

1

Violation 50-4 2 4/90 .xas-@2.

(Section 2.2)

si-4

,i,

"; 11::ncer '

the ef 59*= 4 = 4 a0 ==d eaaWal l i a; en,a 4 == af

..

,

completed surveillance procedures was not controlled

y

,

inistrative procedures.

Based on the confusio

hich

i

resu

d

in the missed

curveillance of the

tainment

j'

isolatio

alves and a review of this methodo

additional

atter. tion

acessary to ensure that thg

procedures are

3

2

appropriately c ntrolled and used.

(Sect' ion 2.3)

!

,9)

The licensee's math

to

denti

tive and informational

limiting condition for oper

.ns (Iros) on LCO status sheets

allowed continuation o

he ~I40 required ac.tions on the

reverse side of the

rm.

This me

in conjunction with

,

the operator's c

reed practice of re

wing only the front

'

side of the

status sheets, was one of

oot causes for

l

a non-ci

violation (50-424/90-10-03) conc

ng a mode

l

cha

ich occurred with inoperable source ra

uclear

i

ruments.

The failure to identify this additiona

oot

... la.ni.iii.4 es Vicletie.. 50 4 ?/00 ::

^3.

(!::t _ _

v.u..

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The licensee's method of appraising the performance of the

licensed operators resulted in a potential disincentive for

identifying items which may result in LERs or violations,

j

(Section 2 44)

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DRAFT - PREDECISIONAL INFORMATION

Accuraev of Information

The inspection concluded that during the inspection inaccurate

information was received on several occasions, from responsible

'

managers and operators on topp:s well within the scope of * a ir

specific responsibility. In 4Manstances the initial infor Cisn (

.

supplied was clearly incorrect or inadequately researchef

7e

'

l

inspection team concluded that in each of these example:

th t

licensee officials provided inaccurate, unsworn, oral stat cent s

1

l

concerning information which concerned topics well within ts.24r

responsibilities,

"

i

ho

{

In

the 4ir;t

th:::

cases,

the

inaccurate

information

was

l

significant to the inspection process.

Specifically, (1) if the

i

containment isolation valves received an automatic closure signalthe v

'

if the snubber modifications had been performed in conjunction with

'

other preplanned preventive and corrective maintenance, then the

i

voluntary entries into LCO 3.7.8 would not have been required, end- I

.

'

p) ii the-WRC-was.. accurately informed .cf.-the-aumber of p

51:::

1

and i.ilorse ;' - th: Ere:gency Diesel Cenerator-No.

15 ehich

uwwuu.a dosinii ts d10 h:: ting, ther additiert! tecting rey hr :

j

,.... . .,,. is .4- pries to - th: rele::: .. .... ....... _..... .. ::tden

1etten The inspection team concluded that the failure to provide

accurate information was a violation of the requirements of 10 CFR 50.9 concerning accuracy and completeness of information.

The

inspection identified Violation 50-424/90-

50-425/90-Jsw-99 in j

M - ' ~L

.

this area and ncted the following examples:

'

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!

1)

containment Isolation valves:

During a Unit 1 surveillance

procedure, the unit shif t supervisor (USS) stated, and the

operations manager later confirmed, that the containment

,

.

isolation valves for the hydrogen monitor system were allowed

to be opened without entering the 140 action requirements for

i

TS 3.6.3 because the talves received an automatic isolation

i

signal.

The inspection identified that these containment

isolation valves were remotely-operated, manual valves without

i

I

auta.tatic isolation signals.

(Discussed in Section 2.2.1.1 of

'

Inspect Dn Report 50-424/90-19; 50-425/90-19)

2)

Snubber Reduction:

The operations manager stated that, after

Unit 1 refueling outage IR2, the modifications to the snubbers

were done in conjunction with preplanned system outages which

i

were required for other preventive or corrective maintenance

or testing. The inspection identified that few of the snubber

modifications were done jointly with pre-planned system

i

' outages.

(Discussed in Section 2.1.1.4 of Inspection Report

50-424/90-19; 50-425/90-19)

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DRAFT - PREDECISIONAL INFORMATION

Emeraenev Diesel cenerator Reliability: The licensee's met

researching

information

for

Region

II

prese

ion

{

ng the reliability of the emergency diesel ,

erators

conc

(EDGs) v

nadequate in that there was a 1

of specific

l

. guidance conc

ng the EDG information d

red coupled with

I

inadequate resear

the EDG starti

story.

This method

resulted in providing

omplete

therefore, inaccurate

i

,

info mation to the NRC.

inn

lon, the licensee's response

to the NRC's confi nation

ac

letter (CAL) was based on

,

i

this same inadequate

earch.

In

tion, the subsequent

l

Licensee ' Event

rt (LER)90-006 was

so

inadequately

!

researched.

a result of this method of inv

ation, the

i

NRC was

r informed of the correct operability a

until

j

this

pection.

(Discussed in section 2.7 of this inspe

n

!

ort)

34t

Personnel Accountability: Theoperationsmanagerstatedthatl

!

the shift superintendents (sss) reported directly to the

'

operations manager and that he personally prepared their

i

performance appraisals.

The inspection identified that the

SSs reported to the unit superintendent (US), and that the US

I

personally prepared the performance appraisals of the SSs.

!

(Discussed in Section 3 A 1 of this inspection report)

{

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4pf

ts 3. 0. 3 Actione:

The unit superintendent indicated that

l

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there were no operations Department actions which were

i

anticipated or required within the first three hours of

i

entering the action statement of TS 3.0.3.

The inspection

!

identified that the VEGP management policy and stated practice

!

required preparations

for a power reduction,

including

i

informing

the

load

dispatcher within

the

first

hour.

i

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(Discussed in Section 2.1.1.3 of Inspectica Report 50-424/90-

its 50-425/90-19)

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490

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In summary, the inspection identified these violations and two\\

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inspector followup items. The violations involved: (1) a violation

i

'

of 10 CFR 50.9 in that responsible licensee officials provided

inaccurate information to the NRC during the inspectionf4(2) a

k

,

i

violation of TS 6.7.1.a in that, two examples were identified of

i

the licensee failing to }mplement actions in accordance with

l

administrative procedures /, M _f')

vi:1;ti... .

M 0.7 n,

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W '. 2, 0: iter!:: "Y!, in th:t, tt:0 ::: ;1 : ::: id:nti f'<ed-of

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th; li;;;;;; i pi::: ting in:d:ptt: ::rr :tiv: ::ti;;;.

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DRAFT - PREDECISIONAL INFORMATION

The inspection also identified two inspector

followup items

'

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involving: (I) an unreviewed safety question concerning the use of

the alternate radwaste building, and (2) the lack of operator

guidance concerning the applicable limiting conditions of operation

during engineered safety

features actuation system sequencer

outages.

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DRAFT - PREDECISIONAL INFORMATION

INSPECTION DETAILS

1.0

INSPECTION OIL 7ECTIVES

Recent activities which have occurred at the Vogtle Electric

Generating Plant (VEGP) have raised concerns within the Nuclear

Regulatory Commission (NRC) as to the ability and the determination

of the licensee to operate the facility in a safe and conservative

manner. To address this concern, the NRC performed a special team

inspection to determine if the licensee operates the facility in

accordance with approvad procedures and within the requirements and

intent of the facility's operating license.

In addition to the

occurrence of specific events, NRC concerns regarding the safe

operation of the facility were heightened with the receipt of

several allegations relating to operational activities at VEGP.

The aggregation of the facts and circumstances associated with the

operational events and the allegations was viewed as a possible

indicator of a non-conservative attitude on the part of the

i

facility's operating staff which warranted the immediate initiation

'

of special inspection activities.

Because a non-conservative attitude or operating philosophy may

j

represent a hazard to the health and safety of the public, a

l

special inspection team comprising staff from the Region II Office

and the Office of Nuclear Reactor Regulation (NRR), assisted by

'

staff from the Office of Investigations

(OI), was formed to

determine the individual validity and collective impact of these

allegations on the safe operation of the facility.

The purpose of

the inspection was to determine if the licensee operates the

j

facility in a conservative and safe manner in accordance with

approved procedures,

and the intent and requirements of the

facility's operating

license.

Specifically,

the

inspection

objectives were to:

1)

kssess the operational philosophy, policy, procedures, and

!

practices of the facility's operating staff and managemenc

regarding operational safety.

)

2)

Determine the technical validity and safety significance of

,

each of the allegations and their impact on the safe and

'

conservative operation of the facility.

,

l

These inspection objectives were accomplished by the use of two

inspection teams--an operations followup team and an allegations

followup team.

The efforts of these two inspection teams were

closely coordinated; however,

they independently pursued the

obj a::tives outlined above.

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DRAFT - PREDECISIONAL INFORMATION

3-

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Theoperationsfollowupteammonitoredcontrolroomactivitieson(

a 24-hour basis

in order to:

(1)

evaluate the

operational 8

!

philosophy, policies, procedures, and practices of the operating 3

-

' staff . and management and- (2) determine if the plant was being'

3

.

!

operated in a safe and conservative manner in accordance with the g

!

facility's operating license.

3

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C

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The specific inspection activities of the operations team was

described in Inspection Report 50-424/90-19 and 50-425/90-19

e

g

,.

efforts and conclusions of the allege _ti_ ops followup teams are

i

described in this inspection report. " In addition, this report }

l

8

identifies several violations 2nd $2;;..;i:1 ;;;%..;;;;; in the

l

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licensee's :;;reti;nci pr!!:ler, prr;r-- , and procedures.

9he-

gecificdetails==d h:;i; fe; th; ine;::ti:r trer's r---=ea= are

,

detailed in the sections that follow and in the Inspection Summary.

'

2.0

ALLEGATION FOLIDWUP

'

The

inspection team

reviewed

several

allegations

for

their

[

technical validity and interviewed licensed and non-licensed

!

personnel to determine their personal knowledge and experience

!

regarding these issues.

This portion of the inspection was

j

performed to determine the validity and significance of the

.

allegations.

5:::r:: the elle;stierr errerted *Mt

li--- ed

!

'

t

:: 5:d vieleted the Techair:1 Sp^cificatiene (??) vith the

!

h.:rl:d;: f lieerrer rare;r rrt, *he inepretien t:22 revier:d th: '

cirrrrrtencer e-d retienele fer i=dividerl retic- .

'

The inspection of the allegations included technical reviews of the

licensee's records, logs, and interviews of the personnel involved

in the alleged violations.

Although a transcribed record was not

required for every discussion with the licensee's staff, the

inspection team conducted sworn,

transcribed interviews with

selected individuals in order to document (1) the individual's

personal knowledge and involvement in the alleged violations and

(2) the circumstances and rationale for their individual actions.

Although an of investigator was assigned to the inspection team to

assist during the transcribed interviews, this inspection was not

an investigation into .the intent ofghe alleaed vi.olations. TwaM

ew cos ar 7w Anap Aur.sers 5 m Amp.rc h

p c. m n Cnc ,ys

s,.

r

The interviews were transcribed after the technical evaluations of

the allegations in order to permit a focused interview and to

minimize the length and scope of the transcribed proceedings.

The

transcribed interviews are listed in Appendir 1 in the order they j 4 p

were conducted.

The sworn testimony was th? 5;;;6on wnica the

inspection team reached its conclusion on each of the allegations.

i

These conclusions are prcsented in the material that follows

(Sections 2.1 through W ).

'

2. . B

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DRAFT - PREDECISIONAL INFORMATION

2.1

.Imoroner Installation of FAVA System

i

An allegation indicated that VEGP installed and operated a radwaste

nicrofiltration

system,

known

as

the

FAVA

system,

without

!

j

performing an adequate engineering and safety evaluation (i.e.,10

!

CFR 50.59).

Furthermore, the material configuration, fabrication

.

i

I

and quality of the system did not meet the guidance of Regulatory

l'

Guide (RG) 1.143 and the requirements of the American Society of

Mechanical Engineer's (ASME) Code.

.

1

i

The FAVA system was temporarily installed for removing Niobium-95.

The system was later determined to be better suited for as-low-as-

,

i

'

reasonably-achievable considerations during refueling outage IR2,

i

particularly for removing Cobalt-59 and Cobalt-60. VEGP planned to

i

replace this temporary

modification with a permanent, high-

l

j;

quality, steel system in the future; however, the health and safety

of the public may be jeopardized if a break in the system

i

(resulting in a radioactive release to an unrestricted area)

i

occurred in the interim.

'

,

j

Discussion

i

,

?

!

.

In February 1988, the VEGP experienced difficulty in removing

[

e

,

colloidal Niobium-95 following a reactor shutdown for maintenance

-

i

work.

FAVA Control Systems (FAVA) was hired to help rectify this

problem.

FAVA was selected because of its experience in filtration

!

j'

and domineralization.

The situation was corrected by installing

a 0.35-micron filter system downstream of the existing vendor-

-

i

supplied pre-filters.

However, a large volume of radwaste was

t

i

generated as the 0.35-micron

filters

rapidly exhibited high

i

differential pressure and were required to be changed frequently.

The need to change filters frequently also resulted in additional

!

radiation exposure to Radwaste Department personnel.

!

Upon evaluation of the performance of the 0.35 micron filter

j

system, the Radwaste Department felt that the best approach to the

problem was a back-flush, pre-coat filter system.

However, no

operational data was available for a system of this type in this

j

specific application. FAVA supplied a proprietan Ultra Filtration

System (Model No. SFD/E) for testing purposes in order to evaluate

3

whether or not this was a viable and economic solution to the

problem.

The FAVA system was installed before the Unit I refueling

i

outage and was operated under Test Procedure T-OPER-8801.

The test

j

system kept liquid effluent releases well below TS limits.

on the

basis of an evaluation of test results by the Radwaste, Chemistry,

,

and Engineering Departments, a general work order was initiated to

l

purchase a permanent system.

.

i

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4

DRAFT - PREDECISIONAL INFORMATION

l

<

1

In the early part of 1989, a Quality Assurance (QA) Department

'

audit

identified

a

significant

audit

finding

involving

a

programmatic breakdown in the procurement of the FAVA system and

!

the failure to meet commitments of the Final Safety Analysis Report

i

!

(FSAR).- Because of that finding, the FAVA system was removed from

service.

In late 1989, the licensee sought to reinstall the FAVA

,

i

system under a temporary modification because colloidal Cobalt-59

i

and Cobalt-60 had to be removed.

The Plant Review Board (PRB)

,

reviewed this temporary modification and several members expressed

]

strong objections to it based on the previous QA audit finding.

>

-

i

Subsequently, a request for engineering assistance

(REA) was

i

submitted and a 10 CFR ' 50.59 safety evaluation was performed in

i

late 1980.

This safety evaluation did not properly address the

i

guidance of Regulatory Guide (RG) 1.143 regarding the use of

polyvinyl chloride

(PVC)

piping.'

Therefore,

another safety

-

evaluation was p rformed in February 1990 to address this issue---

j

particularly with respect to radiation degradation.

'

2

l

The February 1990 safety evaluation specifically stated that the

1

FAVA system did not conform to the criteria of RG 1.143.

This

!

deviation was found to be acceptable for the following reasons:

'

!

The design of the FAVA system had been previously evaluated

'

1)

and found to be adequate in the response to REA VG-9057 dated

l

November 28, 1989 (log SG-8592).

l

2)

The location of the FAVA microfiltration system inside a

i

shielded, watertight vault provided adequate assurance that

any system failures will be contained and would not create the

l

potential for offsite releases of, radioactivity.

.

l

3)

The presence of PVC pipe in the FAVA system,

although

l-

prohibited by RG 1.143, was acceptable because the radiation

i

exposure to the plastic was within acceptable limits for up to

i

6 months based on the following:

!

a)

The amount of PVC piping used was not extensive and was

j

contained on the FAVA filter skid.

1

i

b)

There were no reported leaks or malfunctions during the

approximately 6 months that the FAVA system filter was

'

previously in use,

d

c)

Since the FAVA system filter skid was located within the

demineralizar vault, it would be protected from being

damaged.

)

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DRAFT - PREDECISIONAL INFORMATION

I

d)

On the basis of the assumed' length of time that the PVC

.

!

piping would be used in a radioactive environment and the

activity levels of the effluent at this stage in the

!

liquid radwaste process, the integrated dose to the PVC

'I

piping would be well below the radiation damage threshold

!

for PVC pipe as reported in Electric Power Research

'

i

Institute (EPRI) Report NP-2129, dated November 1981

!

(i.e., 6.5 rad over a 6 month period versus the radiation

j:

damage threshold of 5.0 x lo rad) .

s

i

e)

The PVC pipe would not be subjected to excessive pressure

!

I

conditions since the maximue. available inlet pressure to

i

the filter was so to 100 pounds per square inch gauge

!

(psig) which is well below the maximum allowable working

pressure of 120 psig for the PVC pipe.

,

j

f)

The system could be operated at design-basis conditions

for 182 days before it would exceed the radiation damage

i

threshold. However, under conditions currently existing

at the plant, the expected dose to the PVC piping will

.

j

less than 0.1 percent of the design basis.

i

Although the testamony of one of the PRB members indicated that the

temperature effects on the use of PVC in the FAVA System were not

,

adequately evaluated before the system was installed, the testimony

i

!

of the corporate system engineer indicated that this was considered

i

prior to installation, although not specifically documented in the

safety evaluation.

1

The VEGP general manager subsequently consulted the NRC resident

!

inspector to seek an NRC position with regard to placing this

1-

system - back in service.

This was supplemented by information

{

documenting reasons why it should not be placed in service.

This

'

package was forwarded to Region II and the office of Nuclear

Reactor Regulation (NRR) for review.

In March 1990, following

Region II and NRR concurrence via a telephone conference, the

licensee placed the FAVA system in service with the following NRC

!

stipulations:

I

j

1)

Procedures for operating the FAVA system required an operator

to be in attendance for the entire length of time the systra

,

j

would be in operation.

-

'

i

.

2)

All hoses going to and coming from the FAVA syst en required

!

verification that they met the requirements of RG 1.143.

1

3)

The cover over the FAVA system wa- required to be securely

fastened when the system was in operation to ensure that if a

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_ _ _ _ _ _ _._ -_ __ _______ _ ____

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DRAFT - PREDECISIONAL INFORMATION

-

!

i

spraying leak developed, it would be contained in the concrete

l

vault.

i-

4)

The design of the . walls of the alternate radwaste building

!

(ARB) was required to be evaluated to determine whether or not

!

i

a design modification should be made to reduce the potential

?

j

of wall leakage in the event that a hose leak developed and

j

sprayed its contents on the walls.

,

i

In June 1990, in response to item 4 (above), the licensee revised

Part G of the safety evaluation for the FAVA system.

Part G of the

i

safety evaluation addressed the effect that operation of the FAVA

1

!

system would have on the probability of occurrence or consequences

,

{

of accidents described in the FSAR.

Although there was no

comparable accident analysis in the FSAR that addressed the ARB

4

accidents or the consequences of accidents in the ARB, the FSAR

I

accident analyses (Chapters 15.7.2 and 15.7.3) did describe worst-

case releases of the contents of the recycle holdup tank (HUT) .

l'

The first bounding analysis in Chapter 15.7.2 addressed the release

of the entire gaseous radioactive contents of the HUT to the

{

i

environment at ground level and the second bounding analysis

'

addressed the release of the entire liquid contents of the HUT-

i

through an assumed crack 2n the ARS floor directly into the ground

.

water supply.

In both cases, the 10 CFR Part 100 and 10 CFR Part 1

'

20 limits were not exceeded.

These criteria were consistent with

'

j

criteria provided in NRC Circular 80-18, "10 CFR 50.59 Safety

Evaluations for Changes to Radioactive Waste Treatment System."

'

However, neither of these analyses addressed the potential for wall

spray down and leakage. through the ARB walls and the subsequent

t

release path to the environment.

Therefore, the licensee revised

the safety evaluation in June 1990 to address the consequences of

a hose break on the FAVA system which would result in wall spray

i

l

down and potential leakage to the environment.

,

l

The inspection team's review of the revised Part G of the safety

i

evaluation identified several erroneous assumptions with respect to

the release path and the dilution volumes that could be used in the

analysis of a hose break and resultant wall spray down.

However,

,

the inspection team also found that the design of the FAVA system

(i.e., the use of a system cover) would prevent wall spray down and

that the only potential source for wall spray down and subsequent

leakage was from a hose break in another radwaste system in the

ARB. Therefore, the inspection team concluded that the FAVA system

safety evaluation dated June

1990,

adequately addressed the

.

temporary modification for the installation of the FAVA system;

j

however, the inspection team's review identified an unreviewed

4

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DRAFT - PRIDECISIONAL INFORMATION

safety question concerning the release paths and consequences of a

failure of the other radwaste systems in the ARB.

In' addition, the team noted that in Supplements 3 and 4 of the

Safety Evaluation Report (SER), the NRC staff reviewed and accepted

the design of the ARB and specifically addressed the consequences

,

of a hose break on a radwaste system in the ARB.

However, the SER

supplements addressed the effects of high airborne activities and

puddling and did not address the potential for wall spray down and

,

leakage.

The ARB was installed before the plant was licensed;

t

therefore, the NRC approved the design and use of the ARB in

'

)

Supplements 3 and 4 of the SER.

Thus, there was no requirement to

!

perform another evaluation of the potential effects of hose brcaks

on systems other than the system being installed by the temporary

1

modification (i.e., the FAVA system).

Because the design of the

FAVA system effectively prevented a wall spray down, this was not

a concern that was required to be addressed by the FAVA system

'

safety evaluation.

Nevertheless, now that it has been identified,

l

the consequences of a hose break and wall spray down in the other

ARB radwaste systems must be resolved.

Therefore, this issue will

4

be followed as an inspector followup item pending further review

and evaluation and is identified as:

19 -14

19 - N

'

IFI 50-424/90.x*-07 and 50-4 2 5/ 9 0-xw-02",

'Dotential Unreviewed

Safety Question Regarding Spray Down of the Alternate Radwaste

Building."

.

J

<

Conclusion

Although the FAVA system was originally installed without an

adequate safety evaluation and did not meet the regulatory

3

guidance, the inspection team concluded that the subsequent safety

i

evaluations were acceptable for the system's use.

Serefert, th;

)

inerrtien tem cen:1 d:d that th; ellgeti;n iter ne*

M 1y

(

1

'

r Arte el.i d.

As a result of QA Department's significant audit finding in early

1989 involving a breakdown in procurement and failure to meet FSAR

commitments, the system was removed from service.

Subsequently,

the FAVA system was returned to service following two safety

j

evaluations which adequately addressed the use of PVC piping with

,

respect to radiation degradation and pipe rupture.

Therefore,

these safety evaluations justified the use of the FAVA system, even

though the recommendations of RG 1.143 and ASME Code requirements

were not met. Although the safety evaluations did not specifically

address high-temperature effects, the testimony indicated that

these effects had been considered before the system was installed.

i

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13

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DRATT - PREDECISIONAL INFORMATION

Although the safety evaluation performed in June 1990 at the

request of the NRC Region II Office did not adequately evaluate the

,

!

effects of a wall spray down and wall leakage to an unrestricted

area, this evaluation was not required because the FAVA system has

a protective cover and the use of hoses and effects of hose breaks

(i.e.,

airborne activity and puddling) were addressed in SER

J

Supplements 3 and 4.

I

Regardless of whether the safety evaluation was required to address

the effects of a break in the hoses (which could result in wall

spray down or leakage), the inspection team identified a new

concern involving the use of the ARB because the safety evaluation

inadequately addressed the potential effects of wall spray down

l

from any other source in the ARB owing to erroneous assumptions

concerning the release path and the dilution volumes.

This is a

potentially unreviewed safety question concerning the use of the

<

alternate radwaste building.

2.2

Operability of the Residual Heat Removal Pumo

An allegation indicated that during Unit i refueling outage IR2

'

with residual heat removal

(RHR)

Train A out of service for

maintenance, the Train B RHR pump experienced excessive vibration

and a nuclear service cooling water (NSCW) motor cooler outlet

i

leak.

In addition, TS 3.9.8.1, "RHR and coolant Circulation," was

i

allegedly violated because the operations Department chose not to

declare RHR pump 15 inoperable in an effort to mitigate the impact

,

on the critical work path.

Discussion

TS 3.9.8.1 requires at least one RHR train to be operable and in

operation during Mode 6 (refueling) when the water level above the

top of the reactor vessel flange is 23 feet or more.

Otherwise,

,

'

Suspeed all operations involving an increase in the

reactor decay heat

load

or

a

reduction

in

boron

i

,

concentration of the reactor coolant system (RCS) and

j

!

Immediately initiate corrective action to return the

'

I

required RHR train to operable and operating status as

soon as possible and close all containment penetrations

'

providing direct access from the containment atmosphere

to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

The inspection team verified that during Unit I refueling outage

IR2 with higher than normal vibration measurements on the RHR pump

<

1B and a leak on the NSCC outlet of the Rh motor cooler,

4

operations

Department

personnel

did

not

declare

the

pump

'

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DRAFT - PREDECISIONAL INFORMATION

,

i

inoperable.

This determination was made after consulting with the

on-shift duty engineer from the Engineering Department and was

!

,

based on the determination that the pump would fulfill its intended

j

safety function in Mode 6.

Specifically, the RHR pump was capable

'

!,

of removing decay heat from the partially defueled reactor core.

j

The testimony of the individuals involved indicated that this

.

operability determination was based on the fact that the vibration

readings taken at the inservice test (IST) surveillance points did

not reach the IST Alert levels and were therefore acceptable for

continued service. Although the high vibration readings on the top

end of the RHR pump

were

later determined by

the

vendor

.

(Westinghouse) to be excessive, at the time of the operability

i

-

evaluation, the licensee accepted these values, regardless of their

f

magnitude, because the readings at IST test points were below the

I

Alert levels. The testimony also indicated that, even with a leak

l

on the NSCW outlet of the RHR motor cooler, the motor was receiving

'

j

full cooling water flow and cooling would not have been immediately

i

compromised following a complete NSCW discharge pipe break.

l

Furthermore, the. testimony indicated that the operations Department

j

had : implemented compensatory actions to monitor the vibration

levels and NSCW 1eakage and ensure the continued operability of the

l

pump by stationing an operator at the RHR pump to monitor the

vibration levels and notify the control room if the vibration

levels increased, thus allowing the control room to implement the

,

actions of the limiting condition for operations (140).

The inspection team also noted that in event of a catastrophic

i

!

failure of the RHR pump, all the required actions of TS 3.9.3.I

I

'

(i.e.,

closing all containment penetrations)

could have been

completed within the required 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months time period of the 140 because

$

the Ito for TS 3.9.4, " Containment Building penetrations," was in

-

'

effect during this time period.

This LCO was implemented due to

the movement of irradiated fuel from the core to the spent fuel

,

j

pool.

The 140 required that,

4

The equipment door be closed and held in place by at

least four bolts; at least one door in each airlock be

closed; and each penetration providing direct access from

1,

the containment atmosphere to the outside atmosphere

i

shall be either closed by an isolation valve, blind

.

flange, or manual valve, or be capable of being closed by

'

an operable automatic containment ventilation isolation

j

valve.

4

As a result of the implementation of TS 3.9.4, the only remaining

act' .n for the LCO of 73 3.9.8.1 would have been to close the

}

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DRAFT - PREDECISIONAL INFORMATION

'

containment purge valve which receives an automatic closure signal

'

and could have been isolated within the Iro action times.

!

During the course of this review, the inspection team found that

4

the licensee failed to initiate a deficiency card for either the

!

i

NSCW 1eak or the excessive vibration as required by operations

!

Procedure 00150-C, " Deficiency Control."

This procedure requires

i

4

i

that a deficiency card be written if the deficiency involves

'

i

safety-related components which are to be dispositioned "use-as-

i

is/ repair," or other conditions involving safety-related components

,

which require engineering suppolt or other technical assistance to

i

determine if the component is deficient.

Failure to establish,

implement, and maintain adequate operating procedures represents a

,

1

violation of TS 6.7.1.a.

This item is identified as:

R-G

6

I

.'

VIO 50-424/90_xx=Gt,i..a ;; = /;0 m ;i, " Failure To Establish or

l

Implement Procedures for Required Activities."

h

conclusion

,

!

The inspection team c

luded tht t.5e

11-v. tis, n; n;t frily

'

i

c ertentistrf L;;.;;.3

e operations Department had an adequate

i

engineering basis for accepting the operability of the RHR pump in

!

j

spite of the pump's deficiencies.

In addition, the team concluded

i

j

that declaring the pump inoperable would not have impacted the

!

critical work path: the 140 actions would not have been restrictive

because containment (excluding ventilation) had been isolated as

'

required by TS 3.9.4.

The LCO actions would not have prev . ,ad the

,

continuation of refueling activities because the actions to close

!

all containment penetrations providing direct access from the

containment atmosphere to the outside atmosphere would only have

1

required closing the containment purge valve which has an automatic

closure signal.

l

In addition, the inspection team identified that the licensee

violated the station's administrative procedures by failing to

initiate a deficiency card for either the NSCW outlet leak or the

,

excessive vibration on the RNR motor as required by Operations

Procedure 00150-C.

'

t

-?'

"!:rrf cert fr ent I:02: tier V:1:: Ourec ill;; ca

-

,

l

An allegatio

icated that a unit

shift

isor

(USS)

concealed the correc

time for a T

o prevent a forced

i-

shutdown of the unit and to

CFR 50.72 notification to

'

the NRC.

Furthermore, c

ent

. ion valves (CIVs) which

were missed durin

rveillance test sho

va been declared

'

!-

inoperabl

e immediata actions of the TS LCO s

ave been

1

i

at the time the missed surveillance was identifie .

1

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l

DRAFT - PREDECISIONAL INFORMATION

.

i .

ddition, delaying the initiation of the deficiency card (DC) unt 1

t

surveillance had been re-perfonned allowed the licensee

o

i

av (d the immediate actions of the LCO and allowed the un

to

i

)

rema'in in operation and avoid the immediate NRC notificatio .

Discus

on

!

The inspe tion team reviewed the documentation of

a missed

j

surveillanc

on the containment isolation valves

escribed in

a

Licensee Eve

Report (LER)90-001 for which a non-

ted violation

(50-425/90-01- 1) was issued.

The LER identified

hat during the

i

review of mont y Surveillance Procedure 1447 -2,

" Containment

!

Integrity Veriff

tion-Valves outside Contain

nt," the licensee

>

'

discovered that 3

CIVs had been overlook

and had not been

i

tested.

In additi

the valves had not b en tested during the

,

previous two month 1 s surveillances.

Up

identification, the

<

j

operating

shift

re-performed

the

com

ete

surveillance

and

initiated an investigat

n which resulte

in a deficiency card (DC)

!

for the previously misse

surveillance .

i

The LER indicated that t

root cause of the violation was

,

8

personnel error in reviewing t e co

lated surveillance task sheet.

i

In addition, the computer softv

which generated the surveillance

l-

task sheets (STS) has been a

ified so that it is no longer

!

possible to inadvertently

t

n

incomplete listing of the

'

equipment.

Even if an error

imil

to the one which resulted in

only two valves being shown n the S

were to recur, it could only

.

result in either all or n

e of the e

ipment being listed,

i

The

inspection team v rified

that

S

3.6.1.1,

" Containment

!

Integrity," 140 actio

statement requi

d restoring containment

l

integrity within 1

ur or commencing

unit shutdown to hot

standby within the n t 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

A shutdo

required by Technical

-

Specifications wou d have required that th

NRC be immediately

.

notified in accor ance with 10 CFR 50.72.

s

!

'

The inspection aan found that the CIV surveillance requirement of

i

TS 4. 6.1.1. a

ad been completed and approved.

The surveillance

procedure r

ired verification every 31 days that al penetrations

,

i

not capabl4 of being closed by operable containee t automatic

i

isolation / valves and

required to be

closed durin

accident

conditi ns be closed by valves, blind flanges, or d

etivated

)

autom

ic valves secured in their normal positions.

Du ing the

'

the

oncoming

shift

supervisor

noted

th

the

elllance procedure was only partially performed and that 9 of

su

'

t

CIVs on the surveillance procedure had been marked as

ot

pplicable" and had not been performed.

'

a

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TS\\.0.2.a requires that each surveillance requirement be perforud

witft4n the specified time interval

with a maximum allowa ee

exten) ion not to exceed 25 percent of the surveillance inte

In aditition,

Ts 4.0.3

requires

that

failure

to perf

a

al.

survell

nee requirement within the specified time interva

shall

constitut

a failure to meet the operability requirement

for an

LCo.

As

uch

the failure to perform Surveillance R

irement

4.6.1.3.a fk a,ll the CIVs within the surveillance pari

(i.e. , 31

days plus the 25-percent extension) would have co

tituted an

inoperable con' ition of the CIVs.

The oncoming USS

stified that he lacked sufficie

information to

determine if the

omplete surveillance had n

been performed

within the survalliance frequency because he w

not familiar with

the

circumstances under which the surveillance procedure was

performed.

Furthermorg, he lacked sufficient information in the

control room to deterni e if the complete. surveillance procedure

had been performed within

he surveillance ' period. On the basis of

his experience, the CIV su

eillance was mormally performed in its

entirety; therefore, the potential existed that another partial

surveillance procedure had verified the position of the missed

CIVs.

Although

the control room,previously performed surveillances were filed in

these records' era for information only and were

neither controlled nor complete.

'

The USS indicated that the previou stwo monthly surveillances on

the CIVs obtained from this ' file sqre performed incompletely;

however, he did not know whether surveillances on the missed CIVs

had been performed completely under \\some other surveillance

procedure.

This was confirmed when tNe team interviewed the

surveillance coordinator! Who indicated that approximately once a

missed surveillances were performed under different tasks. mo

Upon identificatiop/

\\

of the potential missed surV illances the USS

initiated an inveftigation to determine whether the surve,illances

t

had actually b en missed and,

concurrently,

r -performed the

surveillance w hin three hours. The inspection tea verified that

the discove

time on the deficiency card correctly

flected the

time at whi

it was verified that the previous two s

had been p

formed incompletely.

elllances

Cone

sion

On the basis of the testimony of the USS, the inspection te

,

neluded that the allegation was not fully substantiated becaus

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USS did not conceal the true discovery time of the sis

CIV

surve

ces to avoid a unit shutdown.

The USS indi

that he

was not pre

ed to keep the plant in operation

o prevent NRC

'

notification.

stated that he had

er been given any

indication or instruc

to do " wha

r it takes" to kecp the

unit on line or to avoid N

t

ation of unusual events.

The

USS did not know and co

no

onfirm if the previous CIV

surveillances had been

equately per

ed and believed that the

surveillance co

re-performed within

allowable outage

time; ther

, his actions to initiate an invest

on into the

4

adequ

of the previous surveillance and to concur

re-

"

ora the CIV surveillance procedure were appropriate.

:

":f: chance 9fth Inemable source -Jtanae- Monitor Mucien

f

Instrument

1

An al

gation indicated that the operations

staff allege

'

knowing

violated Technical Specifications (TS) when the uni

as

taken fro

Mode 5 (cold shutdown) to Mode 6 (refueling)

th a

i

source rang monitor (SRM) nuclear instrument inoperable

d that

the prohibite operational mode change was made in orde

o reduce

the critical pat outage time.

!

Discussion

i

'

The inspection team rev

ved the documentati

of the mode change

described in Licensee Eva

Report (LER) 90

04 for which non-cited

'

Violation 50-424/90-10-03

s issued.

e LER indicated that TS 3.0.4 was violated on March

1990,

en Unit 1 entered Mode 6

from Mode 5 with an 140 for Sou

e Ra .ge Channel IN31 in effect to

allow performance of an 18-mont

'hannel calibration.

The LER

indicated that the root cause to

t

avant was personnel error by

the shift superintendent.

The inspection team confir'med that TS 310.4 required that entry

into an operational mode not be made unlesssthe conditions for the

'

i

LCO are met without aeliance on the provisions of the action

.

requirements.

With.one source range monitor inoperable, TS 3.9.2,

" Instrumentation,' could not be satisfied in\\ Mode 6 without

reliance on the' action statement.

'

/

Personnel w[re interviewed to (1) confirm the effect on the outage

schedule /irectly attributed to this TS violation, (2)

etermine

d

whether'it was known at the time of the mode change that

mode-

i

restraining LCO was in effect, and (3) determine the exte t of

phasis on schedule.

,

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_. - _ _ _ _ _ _ . _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ ,

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.

The testimony and a review of the outage schedule confirmed tha

(

ere was a reduction in critical path outage time which w

!

d

ectly attributed to proceeding to Mode 6 before restoring /

the

l

!

SRM

o an operable status.

l

i

.

i

l

The te

isony also indicated that the shift superintendent (SS) and

{

.the uni

shift supervisor (USS) did not recognize that 'a mode-

j

restrain

LCO was in effect at the time of the mode change.

Both

!

the SS and

SS were aware that there was an active 140 on the SRM,

,

but neither

f them had connected the LCO to the mode., restriction.

i

contributing

ctors to the error were that both th

8 and USS had

i

directed their attention to a problem with the

esting of the

)

engineered safet

features actuation system (ES

) sequencer and

i

that the-work whi

had been emphasized to be

1 ding up the mode

!

change was the d ontamination of the

redctor head.

Upon

i

<

j

notification that th

Health Physics Department had cleared the

l

reactor head for work, the SS granted permission to enter Mode 6.

i

The testimony also indic

ed that there 'was no indication of an

,

i

unreasonable emphasis on t

critical path schedule.

Both the SS

-

and USS indicated that they had never been given any indication or

'

!

instruction to do "whatever i

takes" to stay on schedule.

They

also indicated that they did

t , feel undue pressure to stay on

'

I

schedule and, particularly, not

!

, f it meant compromising safety.

,

/

!

The SS admitted that he was initia ly commended for the schedule

'

j

benefits; however, the violation of

he Technical Specifications

'

was not recognized at the time.

Thg SS indicated that he had

initially received some

sitive fee

ack during the morning

.

!

management briefing for

a shift's acco

lishments and later in

!

the briefing the TS vio ation was recognize and discussed. In the

l

SS's opinion the re

nition of the TS

olation negated all

positive feedback.

,

a

]

The inspection t an identified an additional c cern during the

'

inspection conc ning the format and use of the

status sheets.

On the basis

interviews with the SS and USS a

the review of

the format o the 140 status sheets, the inspection

as concluded

i

that both

a format and normal use of this form co

ributed to

this TS v lation.

i

The LC status sheet, is a two-sided form; the section for

quired

j

actio a begins on the front and continues on the back, who

the

"re

rks" section is located.

During the testimony, both t

SS

i

USS indicated that their usual practice, notwithstanding a

an

e

anges, was to review only the front of this form because onk

restorative actions were noted on the back.

In this case, the mode

1

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!

estraint was noted on the back of the form in the " remarks"

se

ion.

,

l

LER 90

4 did not identify the format and use of the LCO st

us

sheet, as

cause of the violation

therefore, corrective

ions

,

have not ye

een i:aken in this regard.

The failure to

dentify

,

and implement dequate corrective actions to preclude repetition is

a violation of l'O.CFR 50, Appendix B,

Actions," and as s(uch will be followed as: Criterion XVI/ " Corrective

VIO 50-424/90-xx-03,

"F

ure To Determine and' implement Adequate

,

s

Corrective Actions."

'

N

!

Conclusion

'

j

onthebasisofthetranscribed,Jnkerviewsandfromitsreviewof

)

,

the outage schedule,

the i

pectionsteam concluded that the

allegation was not fully s

stantiated. N.The testimony indicated

'

that the mode change w

a critical pathsites.

However, the

testimony of the shift

perintendent and the unit shift supervisor

involved indicated that at the time of the mode' change they were

not aware that an'Ir0 was in effect on the SRM a

that a mode

change was pro

ited.

The ins

ion team also concluded that the corrective act

ns for

the

failed to identify that the format and use of th

140

sta

sheets, was one of the causes of the event.

Therefore,

e

f

ure to implement appropriate corrective actions was found to

violation of 10 CFR 50, Appendix. 8, Criterion XVI.

,

2.b pr Backdatino of sianatures

An allegation indicated that a temporary change to Abnormal

Operating Procedure (AOP) 18028-C, "Ioss of Instrument Air," was

not approved within the 14-day requirement of TS 6.7.3.c; and that

the unit superintendent intentionally incorrectly signed and dated

the temporary change to indicate that the TS requirement was

'

satisfied.

,

j-

Discussion

TS 6.7.3.c requires that temporary changes to AOPs which do not

involve changes to the intent of the original procedure be

documented and reviewed in accordance with TS 6.7.2 and approved

4

within 14 days of implementation.

TS 6.7.2 requires that changes

to AOPs be reviewed as stated in administrative procedures and

approved by the Plant Review Board (PRB) and gener11 manager.

Administrative

Procedure

00100-C,

"Cuality Assurance

Records

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_ _ _ _

_ _ _ . . _ . _ . . _ . _ _ _ . _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ .

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j

Administration," Paragraphs 4.1.1.4

and

4 .1.1. 8,

require ~ that

'

i

corrections to Quality Assurance records exhibit necessary and

appropriate signatures, initials, and dates.

!

i

operations Procedure 18028-C, Revision 7, provided operator actions

in the event of a loss of the instrument air system.

A temporary

'

.

j

change to the procedure was initiated on May 29, 1990, to delete

i

the references to the header isolation at 70 psig and the

i

associated actions.

This change was processed in accordance with

Administrative

Procedure

00052-C,

" Temporary

Changes

to

i

Procedures," which allowed the temporary implementation of minor

i

changes to procedures as long as the change was approved by the PRB

j

and signed by,the general manager within 14 days of the temporary

'

change. Therefore, Temporary Change Procedure (TCP) 1802-C-7-90-1

)

was required to be approved by the PRB and signed by the general

,

manager by June 12, 1990.

'

'!

l

The PRB tabled the TCP on June 8,

1990, (PRB meeting 90-81) and

i

j

assigned action to the Operation's Department to void the TCP or

revise the TCP to incorporate the PRB comments.

Revision 8 to

]

Operations Procedure 18028-C was developed to modify valve numbers

and descriptions reflected in Temporary Modificatiors 1-90-006 and

i

2-90-002.

This revision superseded the changes of the TCP.

On

i

June 12, 1990, the PRB approved Revision 8 (PRB meeting 90-82) and

!

the TCP was removed from the control room copies of the procedure.

]

On June 15, 1990, the unit superintendent lined out the operations

.

manager's previous approval of the TCP and marked the TCP form as

!

disapproved by the operations Department.

The date entered on the

!

form was June 12, 1990.

I

j

On June 22, 1990, the PRB secretary initiated Deficiency Card (DC)

1-90-282 which indicated that the unit superintendent incorrectly

i

dated the TCP with the date of June 12, 1990, rather than actual

date of June 15,1990, and DC 1-90-283 which indicated that the TCP

j'

was not processed within the required 14 days (i.e., by June 12,

1990).

The resolution of these DCs, the associated PR8 meeting

,

i

minutes, and discussions with the operations manager and Nuclear

Safety and Compliance Department staff indicated that described

deficiencies were acknowledged and confirmed by the Operations

!

Department on July 3,1990, and attributed to personnel error. The

I

TCP form was dated with the date on which the Operations Department

l

decided to void the TCP and not the date on which the original was

j

actually signed.

l

As part of the corrective actions for DC 1-90-282, a TCP record

correction notice was initiated to correctly indicate the date on

'

,

!

which the TCP fo11 was pro..ssed;

however,

tus TCP record

l

correction notice could not be produced--one was subsequently

,

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written on August 14, 1990.

In addition, the operations manager

4

-

counselled the unit superintendent and assigned him to investigate

!

both DCs because he was the most knowledgeable of the deficiencies

and the assignment served to reinforce the reprimand.

The

'

subsequent PRB meeting of June 28,

1990,

(PRB meeting 90-90)

4

determined that the 14-day TS violation addressed in DC 1-90-283

.

was reportable to the VEGP vice president, but not to the NRC.

1

However, the inspection team found that the report to the VEGP vice

i

president was not made.- On August 9, 1990, the PR8 (PRB meeting

} .90-104) confirmed that the report was required.

As of August 17,

!

i-

1990, ~ the licensee had not issued the required . cport to the VEGP

j

vice president; however, the licensee intended to issue the report.

>

,

With respect to the rationale for the unit superintendent's

actions, the inspection team learned (during discussions with the

'

}.

Technical Support Manager) that the PRB secretary told the unit

i

superintendent on June 15, 1990, that the TCP needed to be voided

and 4 DC written for violating the 14-day requirement of TS 6.7.3.

l

As discussed in Section 2.11 of this inspection report, Operations

i

Department personnel are held personally accountable for violations

' -

and LERs

(i.e., there is a direct impact on their bonus pay);

therefore, a reportable occurrence based on this event could have

!

i

adversely impacted the unit superintendent's salary.

l

The testimony of the unit superintendent indicated that he dated

the TCP with the date (June 12, 1990) on which the PR8 disapproved

I

it and not the date on which it was actually signed (June ' 15,

i

1990).

Additionally, the unit superintendent had no recollection

l

of any discussions on June 15, 1990, regarding violation of the 14-

l

day TS requirement. He indicated that he never considered the 14-

day requirement despite his previous

knowledge and training

i

concerning this requirement and the June 12, 1990, expiration date

'

indicated on the TCP form.

i

!

The testimony of the PRB secretary

indicted that during a

!

discussion with the unit superintendent on June 15, 1990, she

'

identified the need to void the TCP, as well as the need to write

'

!.

a DC for violating the 14-day TS requirement.

Therefore, the

inspection team was concerned about whether the TCP was voided

j'

before or after the PRB secretary identified the need to void the

!

TCP and initiate a DC.

In order to resolve this discrepancy, the

'

inspection team discussed the discrepancy with the PRB secretary on

August 16,

1990.

In addition to earlier testimony, the PR8

,.

'

secretary indicated that during her discussions concerning the TCP

'

with

the

unit. superintendent

on

June

15,

1990,

the

unit

'

superintendent had indicated that the TCP had already been voided

j

aar13 : in the day.

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conclusion

i

l

On the basis of the testimony, the inspection team concluded that

backdating to avoid a violation of the 14-day TS requirement was

I

not '0117

="h-"-ti tted

idditi^=,

the

r^ cer- th t ibis

T=

i

practice

m-

a plant-wide problem,==-

et

"11 y =nh=* =-t itt ed .

However, the inspection team did confirm that TCP 1802-C-7-90-1 had

i

!

been dated incorrectly; this was a violation of Administrative

i

Procedure 00100-C, " Quality Assurance Records Administration,"

!

Paragraphs 4.1.1.4 and 4.1.1.8 and will be followed as:

!

19-l's

tq -G

VIO 50-424/90-xx-@2' and 50-425/90-mm-er, " Failure to Establish or

\\

'

1

l

Implement Procedures for Required Activities."

l

g y Reportability of Previous Enaineered Safety Features Actuation

System Load Seauencer outaaes

!

<

An allegation indicated that the operations Department incorrectly

l

used a 72-hour shutdown requirement when one of the two ESFAS load

sequencers was previously inoperable.

It was also indicated that

l

VEGP had taken no action to ensure that the past occurrences were

,

l

identified and reported to the NRC as required by 10 CFR 50.73,

j

despite newly acquired information that deenergizing an ESFAS

i

sequencer required entry into the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months limiting condition for

.

operation (Iro) action requirements of TS 3.0.3.

In addition, the

l

possibility existed that the 140 for TS 3.0.3 (i.e. , 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months to hot

standby)

were exceeded when

the

sequencers

were

previously

.

i

deenergized for maintenance and testing. This concern was based on

(1) the

lack of a specific TS for the

sequencers,

(2)

the

'

operations Department historically linking the sequencer outages to

!

the emergency diesel generator (EDG) 140 of TS 3.8.1.1.b (78 hours9.027778e-4 days 0.0217 hours 1.289683e-4 weeks 2.9679e-5 months

to hot standby), (3) a limited review of past maintenance work

,

i

orders (MW0s) indicated possible sequencer deenergiration; and

(4) comments by the engineering staff that the sequencers had been

'

previously deenergized.

Discussion

l

There are two ESFAS sequencers for each unit--one for each 4.16-

kilovolt (kV) emergency bus.

Each sequencer is activated by one of

e

!

two conditions, undervoltage (UV) on the associated emergency bus

!

or a respective train's safety injection (SI) signal. Upon receipt

i

of either or both of the initiating signals, each sequencer will

j

perform all or part of the following functions:

Start the associated EDG.

'

I

i

j

Stop any test sequence in progress.

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Strip the associated emergency bus of all loads (UV

j

i

only).

close the associated EDG breaker (UV only).

[

Energize

the

associated

train's

engineered

safety

features (EST) loads as determined by the initiating

j

signal.

l

Each ESFAS sequencer contains three levels of UV detection and

system response, as well as the power supply for this UV circuitry.

j

Four potential transformers monitor the emergency bus voltage for

}

these three levels of degraded bus voltage (Invel 1, 5 70 percent:

Level 2, 5 86 percent; and 14 vel 3, 5 88.5 percent) and furnish an

analog signal to three sets of four bistables located in one of the

five sequencer cabinets.

4-

.

!

Level _1 is the " loss of voltage" and Level 2 is the " degraded

!

voltage" which is referred to in TS Table 3.3-2,

Items 6.d, 8.a.

i

and 8.b.

As these TS items (applicable in Modes 1 through 4) do

i

not address the loss of all four channels in Level 1 or in Level 2

i

(as would be the case when the sequencer is deenergized), TS 3.0.3

would apply if such a loss were to occur.

It should be noted,

i

however, that if the sequencer were deenergized, it could not

respond to a safety injection signal either.

Therefore, there

would be only one automatic safety injection actuation channel

!

(i.e., associated with the unit's unaffected sequencer) and Item

i

1.b of TS Table 3.3-2 (6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to hot standby) would be the most

l-

limiting 140.

1

';

Discussions with the operations manager, the assistant general

i

manager-plant support, and system engineers for the ESFAS and

,

sequencers confirmed that the Operations Department historically

'

i

linked the sequencer outages to the emergency diesel generator

2

(EDG) 140 of TS 3.8.1.1.b (78 hours9.027778e-4 days 0.0217 hours 1.289683e-4 weeks 2.9679e-5 months to hot standby) . Although the

)

applicability of TS Table 3.3-2 and TS 3.0.3 to sequencer outages

l

had been recently identified, past sequencer outages were not

reviewed.

Therefore, with the assistance of the licensee, the

i

inspection team reviewed the completed MWOs which were performed on

'

the sequencers on Units 1

and

2,

as well

as the

i

Instrumentation and control (I&C) , Engineering, and Operations

1

Department surveillance tests.

-

,

)

The review of completed MWOs did identify several instances where

,

i

the work performed would most likely require the sequencers to be

deenergizedt however, the associated unit was found to have not

,

been in Modes 1, 2, 3, or 4 at the time the work was performed.

.

Somewhat = related to this concern, the review did identify two

4

-

LIMITED DISTRIBUTION - Not For Public Release

j

25

,

!

.

4

.

-

.

e

__m. _ _ . _ _ _ _ _ _ _ _ _ _ _ _ ,

-

- -

. - - -

_ -_

_

_.

_ _

__

_ _ _.__ _ _ _ _ __ _ __ _ _ __ ._

_

=

, .

1

i

LIMITED DISTRIBUTION - Nat For Public R21ecco

.

DRAFT - PRIDECISIONAL INFORMATION

,

.

occurrences (March 4 and June 17, 1987) where the Unit 1 Train B

l

sequencer was inoperable during the change of sequencer controller

card A (SI4T A4-3) .

Specifically, when the controller card was

4

j

removed, both the automatic SI function and UV function for the

sequencer were rendered inoperable.

Because the unit was in Mode

i

3 (hot standby) during these two occurrences, the sequencers and

the ESFAS were required to be operable per TS 3.3.2.

However, the

i

associated 140 status sheets (1-87-354, dated March 4, 1987 and 1-

I

87-566, dated June 17, 1987) only recognized TS I40 3.8.1.1.b as

being applicable to the outage.

Despite the fact that 140s

associated with TS Table 3.3-2 (Item 1.b) and TS 3.0.3 were not

,

recognized,

these TS were not violated since the system was

-

j

. restored within 30 minutes and 10 minutes, respectively, f1F

i

' fditie ,

2- ran unit rerrin:d in 5:t Ot:nry,

pert:tility "adar

_

4

j

^ 07

00.72 w 00.73

n;t r:';;ir:d [i.:., there

t'

pr:r

.

r=A n<-+ 6 ihil: 1. 2 TS LOO (10 OFP 50.?2) ner we

G. piani.-aken

I

j

te hat utandhy er : :::Olt ;f e TO LOO (10 m 50. H ) F.

1

,

i

similar to the MWO review, the inspection team's review of related

I&C, Engineering, and Operations Department's surveillance tests

l

did not find any examples of the sequencers or the ESFAS being

l

deenergized in Modes 1 through 4.

Completed 18-month ESFAS channel

i

calibrations, EDG tests, and ESTAS tests were verified as having

i

been done in Modes 5 and 6.

Completed quarterly testing of the

i

ESFAS Auto SI K610 slave relay, which removed the automatic SI

signal to the sequencer, were verified to be performed within time

i

limits allowed by TS 3.3.2.

All other sequencer testing that used

l

installed test circuitry is automatically bypassed on an SI or UV

signal.

4

i

In

addition to

the

inspection

team's

review

of

MWos

and

,

l

surveillance test

procedures,

the

system engineers

for the

i

sequencers and ESFAS [as well as the nuclear steam supply system

(NSSS) supervisor) were asked if they knew of any time in which the

4 -

sequencers were deenergized in Modes 1 through 4.

None of these

'

engineers remembered any such occurrences.

)

'

A

review

of

applicable

operator

treining

material

(System

Description 8b for Engineered Safety Features System Sequencers)

.

'

revealed that there was no reference to ESFAS TS 3.3.2, just those

for the diesel and other power sources and distributions (i.e., TS

i

3.8.1.1,

TS 3.8.3.2,

TS 3.8.2.1,

TS 3.8.3.1,

and TS 3.8.3.2.).

This finding, along with the March 4 and June 17, 1987, occurrences

l

discussed

above,

indicates

that

the

operations

Department

historically has not linked sequencer outages to the 140s of TS 3.3.2 or TS 3.0.3.

Nevertheless, discussions with the operations

~ manager and the licenced operators on shift indicated that although

,

no

written

guidance

or

TS

interpretation

existed

for

the

,

LIMITED DISTRIBUTION - Not For Public Release

,

26

,

!

-

,

'

_

-

.

_ . . .

_ _

_ _ _ .

. _ _ _

__

. - _ _ _ _

__

___ _

_ _ _ _ _ _ _ . _ _ _ - _

-

.

!

>

..

<

LIMITED DISTRIBUTION - Not For Public Rslease

.

l

,

DRAFT - PREDECISIONAL INFORMATION

l

sequencers,

the

operations

Department

staff

would

currently

!

consider all applicable TS requirements, including TS 3.3.2 and

j

3.0.3.

!

Conclusion

The 140 actions of TS Table 3.3-2,

"ESFAS Instrumentation," are

]

applicable for determining the operability of ESFAS components;

'

however,-if a load sequencer is not operable, the more restrictive

i

requirement of TS Table 3.3-2, TS 3.0.3, or the affected system Iro

'

i

should be considered.

Although the EDG LCO of TS 3.8.1.1.b had

been used for sequencer outages in t.he past, the allegation's

'

concern of possibly exceeding the 140 for TS 3.0.3 when the

.

sequencers

were

previously

deenergized yM

not

5:

';11y

,

}

c'2:t: .tiatad. loa!nR*1co.

}

Because there is no specific TS for the sequencers and considering

,

j

(1)

their unique interaction with numerous other systems and

j

equipment, and (2) the varying degrees in which related failures,

!

i

maintenance work, and surveillances can affect the sequencers'

j

associated functions, the inspection team concluded that additie.31

!

l

guidance for the operators is warranted.

Therefore, this issue

l

,

will be followed as an inspector followup item pending further

'

review and evaluation and is identified as

-

I

N-If

39- 6

IFI 50-424/90-M and 50-425/90-xm-94, "Iack of Operator Guidance

I

concerning the Leo Actions Applicable During ESFAS Sequencer

1

'

outages."

1

{

.-

2:11211!ty -'

h-==~

v nine.1 r:eneratera

i

An allega

n indicated that VEGP counted the number of starts

failures

o

e

EDGs

incorrectly

and

misrepresente

is

e

l

information know

in (1) a verbal presentation to

C,

(2)

a formal response to

Region II confirmation o

ction letter

l

(CAL), and (3) LER 90-00 ,

evision 0, issued

owing the March

j-

20, 1990, event involving fa

es of the

f1A.

In addition, it

j

was alleged that VEGP attempte

use the EDG reliability

issue with Revision 1, and delayed

-006, Revision 1, in order

.

j

to avoid drawing attention to

ese inco

et representations.

1-

i

Discussion

]

The

spection team reviewed the following:

i

LIMITED DISTRIBUTION - Not For Public Release

27

,

e

G

w--,

.

--w

vge

~

_

-,

-

,n-.

_ _ - . . . . - - - . - - . - . - - . .

.

. . . - . -

...

-

~

. - . . . - . - . . . . - -

.

1l.

.

i

LIMITED DISTRIEUTION - Nat For Public R31eC03

.

DRAFT - PREDECISIONAL INFORMATION

!,

,

1)

VEGP presentation in the Region II office ~ on April 9,

1990

concerning the site area emergency event of March 20, 19

.

his presentation is identified as Enclosure 2 to the Re

on

i

meeting summary letter of May 14, 1990.

/

1

~2)

VEG 1etter dated April 9,

1990, in response to the Region II

4

i

cont

ation of action letter (CAL) dated March 23,'1990.

}

!

j

3)

LER 90-006, issued April 19, 1990, to report the site area

emergen

event of March 20, 1990.

>

i

These documentsMnd the following procedures describe the EDG

operability.statu

and the licensee's program for recording EDG

!

start

information

and

the

EDG

surveillance

test

frequency

i

4

requirements:

\\

Procedure 550 8-C,

" Diesel Start.Ing"

l

?

'

Procedure

1314

-1,

"EDG

Operation

for

Maintenance

'

Troubleshooting

Maintenance Testing"

^

Procedure 14980-1,

EDG Operability Test"

i

\\/

e

l

The licensee indicated in a tran

arency used during the Region II

l

presentation that there were 18 s

cessful starts on EDG f1A and 19

i

successful starts on EDG $15

tyeen the loss-of-offsite-power

'

event (March 20,

1990) and the pXesentation to Region II of

-

April 9, 1990. The inspectiop' team r

iewed the EDG start logs and

j

the detailed EDG start records complet

during the performance of

i

Surveillance Procedures 13145-1 and 1498

-1.

The inspection team's

i

review of these records in'dicated that th

a were 31 EDG f1A and 29

EDG #1B attempted startpi

Two of the EDG

A and eight of the EDG

f1B starts involved pr#blems or failures.

EDG f1A there were a

'

i

total of 29 successful starts and on E

013 there were 21

i

problems or failurm/

successful starts.

However,

there were s veral intermittent

s during the EDG f18 start

tempts.

Although

there were 29

6

sequential starts

n EDG f1A,

the

inspection team,7 ccessful,

!

identified that there were on1

12 successful,

j

sequential staits of EDG $15 during this time peri

.

i

!

TS 4.8.1.1.

.a requires that each EDG be demonstrate

operable in

l

accordane

with the periodicity specified in TS Tabl

4 . 8-1 by

l

Verifyl

that the EDG starts and assumes rated fra

ency and

'

voltage in accordance with the EDG surveillance test

This

,

surve lance test required a minimum run time of 1 hou

at a

.

.desi

ated load. The inspection team found that at the time

the

l

pre entation to the NRC, the operability test of the EDGs had een

s

cessfully demonstrated two times.

In addition, the EDGs

ad

l

LIMITED DISTRIBUTION - Not For Public Release

28

j-

i-

.

- . ,

n...

. , . . . .

-

n -.

. . . , - ,

,.

___

_ _ _

_ ..-_ _ _ . . _ _ _ _ _ _ _. _ __ _ _

_ _ __._.

_

._

i

..

.

LIMITED DISTRIBUTION - Not For Public Ralesso

.

DRAFT - PREDECISIONAL INFORMATION

i

uccessfully passed four operability tests before Unit i entere

j

e 2.

Therefore, the EDGs were reliable and operable before t e

i

j

presentation.

\\

/

-

The WRC Region II Office was

not verbally informed of the

l

j

incomp'lete information regarding the number of EDG starts until

'

1

June 11

1990 (approximately 2 months after.the presentation).

.

j

Although

evision 1 of LER 90-006, dated June 29, 1990, correctly

!

)

identifiedsthe number of sequential, successful EDG starts from the

i

j

and of the'saintenance test program

(i.e.,

the first successful

!

operability

st per TS 4. s . l .1. 2 a ) until the issuance of LER 90-

i

006, Revision

dated April 19, 1990, this revision (June 29,

,

j

1990) did not address the number of EDG starts that should have

been cited in th presentation, in the VEGP letter in response to

the CAL, and in

R 90-006, Revision

O.

The correct number of

,

j

sequential, ' success ul starts for EDG #18 was 12 and not 19 as

!

indicated in the pres tation. Therefore, the NRC was not informed

l

of the correct informa ion in a timely manner.

1

The information presente

to the NRC did not coopletely describe

.

the problems and failures k at occurred with EDG $18. However, the

i

testimony indicates that t

general manager's intention was to

'

i

demonstrate that the problem

involving the immediate trip of EDGs

identified during and follow

gfthe March 20,

1990 event were

corrected prior to Unit 1 start

Therefore, a compliation of the

.-

total number of successful sta' s

(i.e.,

a start that did not

immediatelytrip)wasanimp/

ortan

factor in his presentation.

s

i

l

The testimony also indica

d that

e unit superinten: lent (Us)

i

researched the EDG starti

history fo the NRC presentation based

-

on a request from the ge

ral manager.

e general manager did not

ask the US to prepare a complete descri

ion of the EDG starting

i

history. specifica11 ' the general manage requested a summary of

l

only the successfu

starts--the informati

concerning the EDG

j

problems and fallu s was not requested.

In ddition, the Us used

j

the unit reactor

rator logs instead of the E

operating logs to

compile the EDG

tarting history.

The reactor perator logs did

!

not contain a

etailed description of problems

failures which

!

occurred dur g the EDG starts.

The US did not

ceive specific

j

guidance co erning the type.of EDG starts that he w

requested to

l

summarise.

In addition, the testimony indicated that the original

assumptigns and EDG $1B start information used in the

esentation

,

were alp 6 used in the VEGP response to the CAL, and in

R 90-006

issuedApril 19, 1990.

i

,

.

The [nspection team's review of ' the Unit 1 EDG's reliabil

f

and

o

rability status between March 21 and June 14, 1990, raise

the

11owing additional concern.

The review was performed to ve

fy

l

LIMITED DISTRIBUTION - Not For Public Release

29

.

t

'

4

!

'

.

,

!

,.

. ,

, . . . , . . , ,

, , _ .

._.

_ . _ _

. - - .

_

. _ _ _ .

- --

_ _ _

,

,.

,

LIMITED DISTRIBUTION - Not For Public R31ocS3

DRAFT - PREDECISIONAL INFORMATION

t at all EDG failures were identified and classified as eit

r

va id or non-valid and were reported to the NRC as required b TS 4.8. .1.3 and TS 6.8.2.

The inspection team discovered tha

the

foll

ing f ailures during starts of EDG f1B had not been clas ified

as val d or non-valid and, consequently, had not been repofted to

the NRC ursuant to TS 4.8.1.1.3 and TS 6.8.2.

EDG S art

Date

Remarks

/

1-90-13

3/22/90

EDG

trfp,

high-

temperature

lube

oil.

Ma i n 't e na nce

troubleshooting test.

/

1-90-134

3/23/90

EDG / trip,

low

jacket

water

pressure.

M ,4

i nt a nance

troubleshooting test.

2

.

1-90-136

3 24/90

EDG intentionally stopped

due to alarmed condition,

high

jacket

water

temperature.

Maintenance

troubleshooting test.

1-90-157

5/23/ 0

EDG

trip,

high

jacket

water

temperature

M a i nt e na nce

troubleshooting test.

i

<

1-90-160

5/23/90

END G

trip,

1ow

4

tutbocharger

-161

o i'

pressure.

Main

nance

-162

troub e. shooting test.

\\

'

1-90-164

5/23/90

EDG

trip

high

jacket

'

water

-163

temperature.

Maintenance

troubleshoot

g test.

These in

.ction findings were discussed with the e

ineering

support

anager who agreed that these types of failures

ave not

'

been

ported.

The licensee committed to have all E

start

recor s reviewed for any unreported failures.

,

Th

inspection team also four.. that a violation -as previou ly

i entified for the failure to report all EDG failures in Inspecti

LIMITED DISTRIBUTION - Not For Public Release

30

.

e

l

- - - -

--- . -

- -

-

. ~ - - .

- . - . - - -_----..- _- -. - _ - .

- -.. -- - - - . - -

.

~

LIMITED DISTRIBUTION - Not For Public R310000

]

DRAFT - PREDECISIONAL INFORMATION

Re

rt 50-424/87-57 dated November 1987.

Although the failur

to

,

repo

all EDG failures is a violation of TS 3.8.1.1.3

d TS

,

6.8.2,

the inspection team concluded that the failure

s the

'

result

inadequate implementation of corrective a

ions to

j

prevent re urrence of a violation and, as such, is a vjdlation of

l

10 CFR 50 Ap ndix 8, Criterion XVI, " Corrective Actiop," and will

be followed a

-

VIO 50-424/90-xx-

" Failure to Determine and Implement Adequate

,

j

Corrective Actions.

,

/

j

Conclusion

/

1

/

i

The. allegation that VEGP i

orrectly counted the number of starts

and feilures of the EDGs a

knowingly misrepresented the EDG

'

reliability

in

order

to

s1 cad

the

NRC

was

partially

.

substantiated.

On the basis of

u sworn testimony and its review

j

of EDG records, the. inspection te

concluded that the Region II

1

presentation was not intended to r

resent a specific number of

,

j

successful valid tests as specified in

G 1.108 and TS 4. 8.1.1.2a,

i

but rather to describe the EDG maintenan

test program and the EDG

!

reliability status.

Nevertheless, the i

ction team concluded

,

i

that the NRC was not informed of the incor et information until

the NRC asked for it dh[ ring the inspection.

e lack of specific

guidance concerning

he EDG information desi

d,

coupled with

inadequate researc

of the EDG starting histo

resulted in

1

,

providing incomp te and therefore inaccurate infotsation to the

]

NEC.

The CAL r ponse and LER 90-006 were also inco

ct because

they were bas

on the IDG start information that was c

iled for

!

the VEGP pr antation in the Region II Office. The inspec

on teaa

!

concluded

at the failure to provide accurate information

o the

]

NRC was a violation of 10 CFR 50.9 requirements and wil

be

l

follow

as:

.

l

VI

50-424/0-xx-05; 50-425/90-xx-05, " Failure to provide Accurate

ornation to the NRC."

r

,

7 6 y Air Quality of Emeraency Diesel Generator Startina Air System

\\

,

An allegation indicated that VEGP had no basis for its conclusions

i

regarding the air quality of the EDG starting air system and

misrepresented the air quality in the licensee's written response

,

to the CAL.

I

f

Discussion

i

The ir. pection team reviewed the maintenance records and deficiency

cards associated with Unit 1 EDG starting air system.

The team

i

!

LIMITED DISTRIBUTION - Not For Public Release

i

31

'

!.

!

t

'

.

'

,

9

I

e

-

.

, - ,

-

.

-

-. --

.

. ..

--

-

-.

. _ ._

._.__ --

- - -

.

q.-

-

,

4

} .

LIMITED DISTRIBUTION - Not For Public Release

DRAFT - PREDECISIONAL INFORMATION

,

was established when preoperational tests were ini

,

4

on Unit 1 in November 1986.

t

but' not on a scheduled frequency. Dewpoint measurements w

this date,

'

'

part of 1988,

During the latter

!

established to measure the EDG starting air system dewpoi

i

{

current PM progran required checking the dewpoint monthly, cleanin

!

The

the air dr

!

addition, yer condensing units, and cleaning the fan motors.

i

'

operating Procedure 11882-1,

"Outside Area Rounds,"

i

In

dryers be blown down on a daily basis untilrequired that the E

i

)

i

noisture.

they were free of

The inspection team verified that the plant equipment

,

operators blev

down the air systems on each shift during the

i

!

performance of their rounds.

!

A review of the Unit 1 EDG maintenance history records indicat d

i

that the majority of the devpoint measurements taken were withi

e

specifications.

i

.

There were instances,

n

i

asasurements were above specifications.however, when the dewpoint

j

primarily attributed to problems with (1)

These conditions were

i

instruments,

(2) system air dryers beingthe dewpoint measuring

l

extended pwriods of time, and

out of service for

system following maintenance. (3) repressurizing the EDG air start

i

The inspection team reviewed maintenance records associated with

s

internal inspection of the EDG air start system air receiveran

micron control _ air system filter inspection and replacement

,

5-

the replacement of the dowpoint measuring instrument with an EG6

, and

analyzer.

Following the loss of offsite power event of March 20

i

G

1990, the control air system instrument lines were disconnected f

maintenance troubleshooting and functional tests of Calcon sensors

,-

or

The system engineers associated with this work stated that no

evidence of internal noisture or corrosion was noted during

.

i

inspection and calibration of the calcon sensors or the cortrol

system instrument lines when this equipment was disconnected for

air

.

maintenance troubleshooting and testing.

!

j

Conclusion

'

!.

-

The inspection team concluded that the licensee did have an

adequate basis to assess the quality of the EDG starting air

,

{

'

system.

inspection of EDG air start system components for degr

1

addition,

the

PM program dewpoint readings have shown more

a

on.

In

The allegation that VEGP did not have a basis for th

i

er.

concerning

EDG

air

start

system

quality

was

not

-M4y

i

-4.i.uil-i A

C #id *4 C D

a ements

I

y

LIMITED DISTRIBUTION - Not For Public Release

'

32

um * A L m. eta & Guy as@v

yQWidLC. INhgebm.

.

.

i

!

- , . . _ _

-

_

.

. . .

_

.

- , _

-

. - _ -

-.

. - - - - .

- -

- - - _ - - - . . . . _ _ . - - . - - -

.- -

!

c

!

j.

-

t

.

i

j

LIMITED DISTRIBUTION - Not For Public Release

'

DRAFT - PREDECISIONAL INFORMATION

M

y Reportability of Previous System Outages

l

f

\\<

!

An allegation indicated that VEGP failed to immediately notify the

>

{

NRC as required by 10 CFR 50.72 when VEGP identified that both

j

trains of the containment fan coolers (CFCs) had been previously

i

j

inoperable at the same time on Unit 1.

.

I

Discussion

i

i.

The inspection team's review of plant records indicated that this

!

I

!

condition occurred when EDG f1A was declared inoperable when tape

(used when the EDG was being painted) was found on the EDG fuel

j

rack.

The tape kept the fuel injector piston from moving and

i

injecting fuel into the EDG.

With EDG f1A inoperable,

the

.

equipment associated with the Train A was also inoperable.

In the

process of investigating the installation of the tape, VEGP

!

identified that this condition existed during a period when the

i;

Train B containment fan coolers were also in a degraded condition

j

for maintenance.

i

'

During the performance of Surveillance. Procedure 14623-1, Train B

i

containment fan cooler (CFC) 1-1501-A7-003 failed to start in slow

!

1

speed.

140 1-90-560 was initiated at 0115 hours0.00133 days 0.0319 hours 1.901455e-4 weeks 4.37575e-5 months on June 19, 1990,

J

and maintenance on the CFC was initiated.

The CFC was returned to

j

operable status on June 19,1990, at 1415 hours0.0164 days 0.393 hours 0.00234 weeks 5.384075e-4 months .

Approximately 9

hours later [on June 19,1990, at 2359 (Lco 1-90-562)), EDG f1A was

i

determined to be inoperable because the tape had been installed on

i-

the fuel rack.

On July 17, 1990, VEGP issued LER 90-014 to

j

identify the previously unrecognized violation of the 140 in

accordance with 10 CFR 50.73.

'

Conclusion

,

!

Based upon the fact that VEGP did not become aware that both trains

,

of CFCs were simultaneously inoperable until after the Train B CFC

!

fan had been returned to service, the immediate notification

j

requirements of 10 CFR 50.72 were not applicable.

The allegation

that VEGP failed to immediately notify the NRC upon discovery of

the previously degraded c

of the CFCs was not .Se44y

,

e , - ,ondition

>

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j

g.1 W Intimidation of Plant Review Board

(PRB) Members

i

An allegation indicated that Plant Review Board (PRB) members were

allegedl

f

meeting.y intimidated and pressured by the general manager in a PRB

j

The meeting occurred in February 1990, to determine the

}

j

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DRAFT - PREDECISIONAL INFORMATION

acceptability of the safety analysis for the installation of the

FAVA microfiltration system.

Discussion

T.i

j

k

As discussed in Section ht" of this inspection report, several

,

!

safety evaluations were performed for the installation of a

'

temporary modification which installed the FAVA microfiltration

i

system.

Discussions with PRB members indicated that during the

t

review of these safety evaluations,

various PRB members had

1

expressed

reservations

on

several

occasions

concerning

the

4

acceptability of the installation of the FAVA system.

!

.'

Despite these reservations, the inspection team's review of the PRB

Meeting minutes

associated

with

this

temporary modification

j

identified few instances of the PRB members documenting their

i

dissenting opinions.

Specifically,

PRB meeting 90-15

(dated

'

February 8, 1990) documented one PRB member's negative vote and

i

dissenting opinions regarding the acceptability of exempting the

temporary modification

from

regu'.atory

requirements

and

the

adequacy of the system's safety evaluation.

PRS Meeting 90-28

,

(dated March

1,

1990)

indicated that information and issues

.

regarding the FAVA system's safety analysis were presented to the

j

PRB and that the general manager solicited written comments and

!

questions from other members for resolution.

The only other

l

example was in PRB meeting 90-32 (dated March

6,

1990) which

!-

identified a dissenting opinion related to the acceptability of

j

votina on the FAVA system installation when the PRB member who

raised the initial questions and concerns on the operation of the

i

FAVA system was not present.

!

Discussions with the PRB members indicated that during the various

PRB meetings concerning the installation of the FAVA system, the

i

PRB members felt intimidated and pressured by the presence of the

'

l-

general manager at the PRB meeting.

The sworn testimony confirmed

that on one occasion an alternate voting member felt intimidated

4

and feared retribution or retaliation because the general manager

i

j

was present at the meeting and the PRB member knew the general

1

j.

manager wanted to have the

temporary modification approved.

j

However, the testimony also indicated that the.PRB member did not

'

alter his vote and felt comfortable with how he had voted.

In

a

l

addition, the PRB member was not aware of any occasions on which he

or any other PRB member had succumbed to intimidation or feared

i

!

retribution.

!

'

The incpection team verified that the general manager was informed

following this meeting that several PRB members viewed his presence

i

as intimidating.

As a result, on March

1,

1990, the general

,

!

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manager set with all PRB members to reiterate the member's duties

and responsibilities.

He specifically told the members that his

presence at PRB meetings must not

influence them and that

!

alternates should be selected who would feel comfortable with this

i

responsibility.

He

also

addressed

the

difference

between

professional differences of opinion and safety or quality concerns,

and their respective methods for resolution.

conclusion

,

The inspection team concluded that in on* case a PR

felt intimidated and feared retribution because the general manager

l

was present at the PRB meeting.

However, this member 6ala nor W

change his vote in response to this pressure and the general

manager met with the PRB to allay fears.

Based on the testimony,

i

the inspection team concluded that retribution did not occur.

Nevertheless, this confirmed event and the absence of dissenting

'

opinions in the PRB meeting minutes indicate that there was a

'

potential for an adverse affect on open discussions at the meeting,

t

The licensee needs to ensure that PRB members

freel

express their technical opinions and safety concerns.y and openly

K Personnel Accountability

As a result of several comments and questions by the licenced

operators to the inspection team, the team reviewed the method used

to rate the

supervisors. performance of the shift superintendents and unit shift

Discussion

The o

rations manager stated that the shift superintendents (SSs)

"

e SSs re

ed to

su

int ndent

s

and

a

i

US personally prepared the performance appraisals of the 8 s.

The personnel accountability system, first used in 1989, was a pay-

i

for-performance methodology. Annual pay increases and e percentage

of the Operations Department bonus were dependent on their ratings

i

in accountability categories.

subdivided into performance categories.Each accountability category was

,

.

Most of the performance

i

categories vern based upon group performance.

Once these are

i

'

eliminated,

any differential

in pay will

result from eight

performance categories. Implementation of the plan in 1989 could

t

result in up to an $8,000-a-year difference 'n bcnus pay to a shift

i

superintendent.

The performance categories and their relative

weights are:

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i

Personnel safety

4.1%

.

Regulatory compliance

10.2%

.

}

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ESFAS actuation

12.2%

!

Reactor trips

10.2%

'

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MWO performance

4.1%

Special projects

a.2%

.

'

Personnel development

30.6%

=

l

Training

.

20.4%

i

Therefore, 51 percent will be associated with personnel development

,

{

and training and 32.6 percent will be associated with the number of

]

LERs, and violations (i.e. , regulatory compliance (10.2 percent),

.

ESFAS actuation (12.2 percent) and reactor trips (10.2 percent)].

'

Conclusion

'

t

The

inspection

team

concluded

that

there

was

a

potential

disincentive for identifying items which may result in LERs or

,

'

violations.

In addition, the inspection team concluded that the

'

operations manager provided incorrect or inadequately researched

!

j

information to the inspection team.

The inaccurate information

)

i

concerned whether the operations manager personally performed the

performance appraisals of shift superintendents.

The information

i

a

-was not very important because the inspection team did not use the

3

information as the basis for a significant inspection finding.

{

j

inspection team concluded that this

The

failure to provide accurate

]

-

information was an example of a violation of the 10CFR 50.9

i

requirements to provide accurate information to the NRC and will be

{

followed as:

14-1L-

Ig-12.

'

VIO 50-424/90-aoeHPJ; 50-425/90.xx-65, " Failure to Provide Accurate

,

Information to the NRC."

'

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3.0

EXIT INTERVIEWS

1

The inspection scope and findings were summarized on August

'

1990, with those persons indicated in Appendix 2.

17,

i

The inspection

team described the areas inspected and discussed in detail the

inspection results.

The

licensee

made

numerous

dissenting

comments.

materials provided to or reviewedThe licensee did not identify as proprietary any o

,

i

inspection.

by the inspector during this

]-

,

.

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APPENDIX 1

LIST OF TRANSCRIBED INTERVIEWS

DATE

TIME

PERSON

,

8/14/90

904 hours0.0105 days 0.251 hours 0.00149 weeks 3.43972e-4 months

George Bockhold

911 hours0.0105 days 0.253 hours 0.00151 weeks 3.466355e-4 months

Jim Swartzwelder

1023 hours0.0118 days 0.284 hours 0.00169 weeks 3.892515e-4 months

Harvey Handfinger

1026 hours0.0119 days 0.285 hours 0.0017 weeks 3.90393e-4 months

Bill Diehl

1109 hours0.0128 days 0.308 hours 0.00183 weeks 4.219745e-4 months

Mike Horton

1335 hours0.0155 days 0.371 hours 0.00221 weeks 5.079675e-4 months

Mike Chance

1136 hours0.0131 days 0.316 hours 0.00188 weeks 4.32248e-4 months

Jimmy Paul cash

1338 hours0.0155 days 0.372 hours 0.00221 weeks 5.09109e-4 months

Dudley Carter

1529 hours0.0177 days 0.425 hours 0.00253 weeks 5.817845e-4 months

Bruce Kaplan

1625 hours0.0188 days 0.451 hours 0.00269 weeks 6.183125e-4 months

Greg Lee

1800 hours0.0208 days 0.5 hours 0.00298 weeks 6.849e-4 months

Jeff Gasser

8/15/90

906 hours0.0105 days 0.252 hours 0.0015 weeks 3.44733e-4 months

Allen Mosbaugh

937 hours0.0108 days 0.26 hours 0.00155 weeks 3.565285e-4 months

Ernie Thornton

1009 hours0.0117 days 0.28 hours 0.00167 weeks 3.839245e-4 months

John Gwin

1048 hours0.0121 days 0.291 hours 0.00173 weeks 3.98764e-4 months

Steve Waldrup

1335 hours0.0155 days 0.371 hours 0.00221 weeks 5.079675e-4 months

Jerry Bowden

1452 hours0.0168 days 0.403 hours 0.0024 weeks 5.52486e-4 months

John williams

1637 hours0.0189 days 0.455 hours 0.00271 weeks 6.228785e-4 months

Carolyn Tynan

1730 hours0.02 days 0.481 hours 0.00286 weeks 6.58265e-4 months

John Williams

!

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APPENDIX 2

PERSONS CONTACTED

Licensee Employees

J. Aufdenkampe, Manager Technical Support

G.

Bockhold, Jr. , General Manager Nuclear Plant

D. Carter, Shift Superintendent

J. Bowden, Work Planning

,

J. Cash, Unit Superintendent

M. Chance, Senior Engineer, Engineering Support

,

S. Chesnut, Technical Support

i

C. Coursey, Mair.cenance Superintendent

l

W. Diehl, Shif t Supervisor, Operations

,

G. Frederick, Safety Audit and Engineering Group Supervisor

1

J. Gasser, Shift Superintendent, Operations

L. Glenn, Manager. - Corporate Concerns

D. Gustafson, Maintenance Engineering Supervisor

i

J. Gwin, Corporate Systen Engineer

H. Handfinger, Manager Maintenance

K. Holmes, Manager Training and Emergency Preparedness

!

M. Horton, Manager Engineering Support

.

,

B. Kaplan, Senior Engineer, Engineering Support

G. Lee, Plant Engineering Supervisor, Operations

R. LeGrand, Manager Health Physics and Chemistry

W. Lyons, Quality Concerns Coordinator

G. McCarley, Independent safety Engineering Group Supervisor

C. McCoy, Vice-President, Georgia Power Company

R. Mcdonald, Executive Vice-President, Georgia Power Company

D. Moncus, Outage and Planning

A. Mosbaugh, VEGP Staff

'

R. Odom, Nuclear Safety and Compliance Manager

A. Rickman, Senior Engineer - Nuclear Safety and Compliance

i

L. Russell, Independent safety Engineering Group, SONOPCO

M. Shelbani, Senior Engineer

!

C. Stinespring, Manager Plant Administration

S. Swanson, Outage and Planning Supervisor

J. Swartzwelder, Manager Operations

-

E. Thorton, Shif t Supervisor, Operations

I

E. Toupin, Oglethorpe Power Corporation

.

C. Tynan, PRB Secretary

i

S. Waldrup, Planning and Scheduling Supervisor

J. Williams, Shift Superintendent, Operations

,

Attended exit interview, August 16, 1990.

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APPENDIX 2

PERSONS CONTACTED (continued)

NRC Employees Who Attended Exit Interview

!

R. Aiello, Resident Inspector - Vogtle

-

B. Bonser, Senior Resident Inspector - Vogtle

M. Branch, Senior Resident Inspector - Watts Bar

R. Carroll, Project Engineer - RIIK. Brockman, Chief, Reacto

N. Huneauller, Reactor Engineer - NRRL. Carner, Senior R

i

0. Matthews, Project Director - NRR

>

J. Milhoan, Deputy Regional Administrator - RII

R. Starkey, Resident Inspector - VogtleL. Reyes, Directo

P. Taylor, Reactor Inspector - RII

M. Thomas, Reactor Inspector - RII

C. VanDenburgh, Section Chief - NRR

J. Wilcox, Operation Engineer - NRR

,

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-

APPENDIX 3

LIST OF ACRONYMS

i

AOP

Abnormal Operating Procedure

ARB.

Alternate radwaste building

ASME'

American Society of Mechanical Engineers

CAL

Confirmation of action letter

i

CFC

Containment Fan Cooler

j

CFR

Code of Federal Regulations

CIV

containment isolation valve

DC

Deficiency card

DRP

Division of Reactor Projects

EDG

Emergency diesel generator

i

EPRI

Electric Power Research Institute

'

ESF

Engineered safety features

ESFAS

Engineered safety features actuation system

TSAR

Final Safety Analysis Report

NUT

Holdup tank

I&C

Instrumentation and controls

IFI

Inspector follovup iten

IST

Inservice test

kV

Kilovolt

LCO

Limiting condition for operation

LER

Licensee Event Report

NWO

Maintenance work order

NRC

Nuclear Regulatory Commission

NRR

Nuclear Reactor Regulation

NSCW

Nuclear service cooling water

NSSS

Nuclear steam supply system

OI

Office of Investigations

PM

Preventative maintenance

PRB

Plant Review Board

psig

Pounds per square inch gauge

PVC

Polyvinyl chloride

QA

Quality Assurance

RII

Region II Office

RCS

Reactor coolant system

l

REA

Request for engineering assistance

RG

Regulatory Guide

,

'

l

RHR

Residual heat removal

SER

Safety Evaluation Report

j

SI

safety injection

i

SONOPCO

Southern Nuclear Operating Company

SRM

Source range monitor

SS

shift superintendent

.

'

SSS

Shift support supervisor

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APPENDIX 3

'

LIST OF ACRONYMS (continued)

STS

Surveillance task sheet

TCP

Temporary change to procedure

TS

Technical Specification

USS

Unit shift superintendent

UV

Undervoltage

VEGP

Vogtle Electric Generati!.g Plant

VIO

Violation

,

i

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ML20129F418

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