ML18033A944

Bus

1A and initiated several

engineered

safety features.

Following the

second

occurrence,

an investigation

was

initiated

which

determined

the

cause

to

be

an

unstable

undervoltage

relay in the circuit protector,

which tripped

when

subjected

to minor vibration.

The defective relay was replaced

and

returned

to the

manufacturer

for evaluation of potential

generic

problems.

,The manufacturer

determined

the probable

cause of failure

to

be

relay

contact

degradation

due

to airborne

contamination

creating corrosion

and pitting of the contacts.

The licensee

also

consulted

NPRDs to determine

whether similar failures of this type

relay

had

occurred

elsewhere

within the

industry.

No similar

failures

were

reported.

Therefore, it was

determined

that this

relay failure was

an isolated

case with no generic implications at

this time.

The inspector

reviewed the above actions

and evaluations

and determined

them to be appropriate.

This item is closed.

(CLOSED)

LER 260/88-19,

Revision 1, Operation

Over Spent

Fuel

Pools

Without

the

Minimum

Number

of

Standby

Gas

Treatment

Trains

Operable.

This

event

involved

a

SBGT train

becoming

when

the

surveillance

period

required .by the

Technical

Specifications

had

expired.

The

SBGT train

was

subsequently

relied

upon to

meet

a limiting

condition for operation

action

statement.

This constituted

an

operation prohibited

by the technical specifications

and resulted in

the reportable

event.

Issues

are discussed

elsewhere

in this report

in

LER 259/88-48

(paragraph

2.j),

and

260/88-35-02

(paragraph 3.s).

The

NRC

inspector

reviewed

the

event

description,

cause

and

corrective actions

and

found that the

LER met reporting standards.

This

LER is closed.

(Closed)

LERs

259/88-22

and

259/88-43,

Unplanned

ESF Actuations

Caused

By Radiation Monitor Upscale

Relay Failure

Both

LERs identify separate

unplanned

ESF actuations

resulting from

failures

of upscale

relay coils in the reactor

and refuel

zones

exhaust

radiation

monitoring circuity.

Both events

resulted

in

isolations

of the Unit 1 primary containment,

refuel

zone,

and

control

room ventilation

systems

and

initiated

SBGT

and

CREV

systems.

The

NRC inspector

reviewed

LERs

259/88-22,

259/88-22

Revision

1,

259/88-43,

and

other

documentation

provided

by the licensee.

The

relay failures

were

due to undersized

relays in the original design.

In both cases,

corrective

maintenance

was initiated to replace

the

failed

relay.

Following repair,

the

radiation

monitor

was

functionally tested

and returned

to service.

The original

24 volt

Potter

Brumfield

KH4690 electromagnetic

relays

were

determined

by

the

licensee

to

be

undersized

when

continuously

operated

at

24

volts.

These

relays

are continuously energized

when in use

and fail

due

to deteriorati'on

resulting

from excessive

heat.

Industry

experience

has

shown that continuously

operated

relays

should

be

rated

at approximately

150K of their planned

operating

voltage.

This condition

had

been previously identified in

GE SIL 189,

dated

July

30,

1976.

In the

above

LERs the licensee

states

that the

failure to implement the

recommendations

of the

GE SIL in a timely

matter

had

been

a contributing factor.

The failure to implement

this SIL was identified in July of 1987.

The licensee

had initiated

procurement

of new

36 volt relays but the

new relay's

had not arrived

at the time of the

two events.

Procurement

delays resulted in the

replacement

relays

not

being

available

for installation until

November 1988.

The

new

upgraded

relays

are

now installed in the

refuel

zone

and reactor

zone exhaust radiation monitors.

As part of

the corrective action to the

LERs the licensee

stated that vendor

information

is

now

reviewed

and

incorporated

as

required

in

accordance

with Volume I of the Nuclear

Performan'ce

Plan.

These

LERs are closed.

(CLOSED)

LER 259/88-48,

Unplanned

ESF Actuation.

This event

involves

two items,

the first was the discovery that the

"C"

SBGT Train was in operation

on December

13,

1988 for no apparent

reason.

The train was secured

and increased

security surveillance of

the

area

was provided.

There

was

on going work in the area.

The

second

item

was

the

discovery,

through

a surveillance

functional

test of the existence

of a undocumented inlet damper in the suction

ductwork of the

"C" train.

The "C" train was declared

inoperable,

the, damper

was

locked

open,

the system

was flow tested

and returned

to service.

A

CARR

BFP

881087

was initiated to investigate

and

correct appropriate

drawings

and procedures.

This

event is associated

with

LER 260/88-19,

Revision

1

(paragraph

2. h)

and

Unresolved

Item 260/88-35-02

(paragraph

3') and

are

discussed

further in this report.

A violation was issued in

NRC

IR 89-33.

The

NRC inspector

reviewed the event in association

with the above

and the

CARR corrective action document.

The actions

proposed

were

adequate

and

were

implemented

in a timely manner.

This item is closed.

(CLOSED)

LER 259/89-02,

Design Error

On EECW,Discharge

Causes

Plant

To Be In An Unanalyzed Condition.

This

LER was associated

with the presence

of seismically unqualified

vitrified clay piping in certain

portions

of the

EECW discharge

flowpaths for various safety related

components.

Contrary to the

requirements

of

FSAR section

10. 10. 2. 2,

EECM piping was

found to

discharge

into non-qualified

24 inch

RCW discharge

headers.

These

24

inch

RCW headers

were routed

from the Reactor Building through the

RHRSW pipe tunnels

where they eventually

became

buried pipe and tied

into

30

inch

headers

constructed

of vitrified clay.

During

a

seismic

event,

these vitrified clay headers

could collapse

and block

,the discharge

flow paths to the affected

components.

The licensee

reported

the condition to the

NRC on February 8,

1989.

The

circumstances

and

events

associated

with this

issue

are

discussed

in

greater

detail

in

NRC

Inspection

Report

259,

260, 296/89-10.

Any regulatory

concerns

associated

with this

issue will be followed as part of the open items identified in that

report.

This section will only address

the technical

resolution of

the non-seismically qualified discharge

piping.

Separate

plant modifications

are

intended to correct the problem by

rerouting

the- three

affected

EECM discharge

paths

to qualified

discharge

paths.

The licensee

has planned and/or performed work for

the following DCN's:

H5120A reroutes

piping associated

with both Unit 1/2 Control

Bay Chillers to qualified Unit 1

EECW 'discha'rge 'pipi'ng.

7

H5121A reroutes

piping associated

with Unit 2 Shutdown

Board

Room Coolers to qualified Unit 2

EECM discharge

path.

H5122A

reroutes

piping associated

with Unit

3 Control

Bay

Chiller 3A to qualified Unit 3

EECM discharge

path.

H5120A

and

H5122A were field complete

as of July 14,

and work on

H5121A

commenced

on July

18.

The

licensee

has

an

outstanding

commitment

(NC08900920021)

to complete all three modifications prior

to tensioning

the

Unit 2 reactor

vessel

head.

The

NRC inspector

considers

that

any concerns

associated

with the technical

resolution

of this issue

are satisfied.

This. item is closed.

(CLOSED)

LER 296/89-02,

Missed

Compensatory

Sampling

Mhile

Conductivity Nonitor Mas Out Of Service.

This

item

involves

the failure to perform

compensatory

reactor

coolant

water

conductivity

sampling

in Unit

3 at eight

hour

intervals

while

local

conductivity

monitor

3-CIT-43-011

was

inoperable,

as

required

by TS 4.6.B. l.c.

This compensatory

sampling

was

required

when

the local monitor was

removed

from service for-

repair and calibration.

Procedure

SDSP 7.9, "Integrated

Schedule

and

Mork Control," did not require

an IE to be performed

on this type of

instrument prior to allowing work to begin.

The

ASOS was not aware

that the monitor would be rendered

inoperable during troubleshooting

and recalibration,

thus,

the required eight hour sampling

was not

performed

for approximately

23 hours2.662037e-4 days 0.00639 hours 3.80291e-5 weeks 8.7515e-6 months .

The licensee

determined

the

root cause

to

be the

inadequacy

in

SDSP 7.9 which did not require

the

performance of'he IE.

Had the IE been performed, all appropri-

ate

personnel

would

have

been

aware

of the

requirement

for

compensatory

sampling.

Procedure

SDSP

7.9

has

been

revised

to

require

specifically

IEs

to

be

performed

for

chemical

instrumentation

equipment

covered

by technical

specifications.

The

inspector

reviewed

the

above

actions

and

procedure

revision

and

determined

them to be appropriate.

Therefore, this item is closed.

(CLOSED)

LER 296/89-03,

Unplanned

Engineered

Safety

Features

Actuations

Caused

By Voltage Transient

On Electrical Distribution

System.

This

issue

is also

addressed

in IFI 260/88-28-03.

The IFI is

discussed

in detail

and

closed

in paragraph

3.m of this report.

This

LER is closed.

3.

Action o'n Previous

Inspection Findings (92701,

92702)

Q

a ~

(CLOSED)

IFI 260/86-40-03,

Unit 2 only,

IRN

Power

Supply

and

Procedure

Changes

Per SIL-445.

This

item

involves

procedural

enhancements

and

equipment

modifications

suggested

by General

Electric

SIL-445.

During

an

outage

at

an operating

GE/BMR, all positive aiid negative

IRN 3/4

amp

fuses

connected

to the

24 vdc

bus

B were

blown because

of a power

e

surge

caused

by

a switching transient

on the

480V power

supply.

After the positive

3/4

amp fuses

were replaced,

all

IRM channels

were operating

normally.

However,

because

of continued loss of the

negative

power

supply

because

of the

blown fu'ses,

the

IRM channels

remained

and

unable to process

flux signals.

The=blown

negative

fuses

were

only detected

during

surveillance

testing

performed later

because

there

was

no blown fuse indication

on the

control

room

panels.

In view of the

above,

SIL-445

made

the

following recommendations:

Procedural

enhancements

to require functional testing of

SRM

and

IRM channels

to ensure

channel operability.

Replacement

of the

3/4

amp

IRM chassis

fuses

with 1.5

amp

fuses,

and

Modification to provide

a reactor protection

system

INOP trip

in response

to a loss of negative

24V power supply to the

IRMs.

As

a result of the

above

recommendations,

TVA has

performed

the

following actions:

Procedure

O-OI-57D,

"DC

Electrical

System

Operating

Instruction,"

Rev.

3,

Section

5.7. 12

requires

functional

testing of IRMs and

SRMs.

The

above

referenced,

fuse

replacement

and system modifications

have

been

completed,

for Unit 2 only, per

DCN-H1706A and Work

Plan 2583-88.

The inspector

reviewed

TVA's actions

and

determined

them to

be

adequate

to close this

item for Unit 2 only.

Units

1 and

3 will

remain

open pending completion of their respective

modifications,

b.

(CLOSED) IFI 259,

260, 296/86-40-12,

Potential for Overpressurization

of Residual

Heat

Removal

System. Piping.

This

item concerns

a modification installed to

reduce

excessive

pressure

drop

across

a throttling valve in the

RHR system.

This

item

was

reviewed

in

NRC Inspection

Report

259,

260,

296/88-32,

paragraph

9.e which concluded that the engineering

analysis

did not

consider

the design

basis

LOCA,

FSAR Section 14.6.3. 3. 2 where torus

pressure

co'uld

be

as high's

27 psig.

The basic issue is that the

portion of the

RHR system in question is rated at

150 psig, which

may

be

exceeded.

With the modification, which installed

an eight

inch orifice plate

downstream

of throttling valve

FCV-74-73, to

reduce

excessive

pressure

drop

across

that valve,

the

pressure

between

the orifice and the valve could be as high as

143 psig under

normal

conditions

with the torus at

atmospheric.

During

a

LOCA

event,

the torus

could

be pressurized

to

27 psig which would mean

that the piping section could exceed

170 psig, in fact by worse case

calculation 173.8 psig.

0

e

The licensee's

reanalysis

confirms this data.

The

Code of record

for this

system

is

USAS

831. 1.0 - 1967.

Under Section

102.2.4,

Ratings:

Allowance for Variations

from

Normal

Operation,

the

following allowances

are provided:

(1)

Up to

15 percent

increase

above the S-value during ten percent

of the operating period.

(2)

Up to

20 percent

increase

above the S-value

dur ing one percent

of the operating'eriod.

For the

LOCA condition,

section

(2)

above

would apply, which would

allow

a

maximum pressure

of

180 psig during

one percent of the

operating

time.

Since

maximum analysis

pressure

would

be 173.8

psig,

the

system

would be within code allowable.

This consideration

is

consistent

with the

NRC's staff position

on similar

issues

resolved at SgN.

This item is closed.

(CLOSEO) IFI 259,

260,

296/87-FRP-01,

Closeout

TMI Item II.F.1.(3)

For Containment

High Range Radiation Monitors

(CHRRM).

The

item was

opened

in IR 87-33

as

a violation for an

inadequate

design modification package.

The

NRC inspectors

concerns

about the

modification

at

that

time

was 'that it

lacked

engineering

documentation

and calculations

to support evaluation of the adequacy

of the design.

TVA responded

to the item in a letter to the

NRC

dated

January ll, 1988

in which they maintained that design

and

justifications

met regulatory

requirements.

The

NRC concurred with

the

TVA position

and withdrew the

item in a letter to

TVA dated

October

31,

1988.

The October

31 letter also stated that the issue

would

be

identified

as

IFI 259,

260,

296/87-FRP-Ol,

because

of

several

deficiencies

described

with the violation.

The IFI was

opened

in

NRC inspection

report

88-14.

In that report the

NRC

inspector

discussed

the

system

design

and installation

as well

as

the fact that the

system

was not operable at that time and would be

subsequently

followed up in later inspections.

The

NRC inspectors

recent

review of this issue

revealed

that the

licensee

discussions

of system

design

in its violation response

provided reasonable

support in the that system design

met guidelines

,specified

in

NUREG 0737,

met industry standards,

and is therefore

considered satisfactory.

The

licensee

provided

documentation

and

explanation

of concerns

raised

by the

NRC about the implementation of the modification that

were described

in IR 87-33.

These

documents

and discussions,

along

with

NRC inspectors

field observations

of the modification, indicate

that

the deficiencies

involving detector

orientation,

electrical

power

supply capabilities,

post modification retest

requirements,

and functional testing of the completed circuits have

been adequately

resolved.

Several

licensee

programs

exist that

are routinely

monitored

by the

NRC that will ensure the'eturn'to"service

of this

system.

These

programs

include:

LCO Tracking (this

system

is

10

required

by

TS 3. 2. F);

commitments

to complete

TMI items discussed

in

a letter from TVA to the

NRC dated

June

16,

1989; the licensee

return

to service

program,

and, System

Pre-operability

Checklist

(SPOC).

The

NRC inspector considers this item closed.

(OPEN) IFI 259,

260, 296/87-02-06,

Baseline

Walkdown Problems.

This item identified that the diesel

generator

starting air motors

were

not

shown

as part of the starting air system

in

FSAR figure

8.5-2.

The

licensee

prepared

a proposed'hange

to the

FSAR to

correct this

item.

The inspector

reviewed

the proposed

change

to

the

TVA has

requested

and been granted

a temporary

exemption

from

10 CFR Part 50.71(e) for an

annual

update

of the

The

change

wi 11

be

made

in the

July,

1990

update.

TVA is

maintaining

a "living" FSAR containing the proposed

changes

until

the

FSAR is

updated.

The inspector

reviewed

a controlled

copy

of the "living" FSAR in

Document Control

and

found the proposed

changes

in place.

Therefore, this item is considered

acceptable-for

restart

based

on the

temporary

exemption.

The item remains

open

since the original concern

has not been corrected in the

(CLOSED) IFI 259,

260, 296/87-20-02,

IE Notice Closeout

This

item concerns

the process

and

adequacy

of nuclear

experience

review activities.

Specifically, after

an action was identified as

being required at

BFNP based

on the nuclear

experience

review, the

item was

being closed

when the responsible

supervisor

stated that

a

particular action

was

committed to

be

done.

There

was

no followup

after

work was

committed

to

be

done.

The licensee

committed to

revise

the governing

procedures,

BFNP Standard

Practice

BF-21. 17,

Review,

Reporting,

and

Feedback

of Operating

Experience

Items,

and

to perform a gA audit of the experience

review process.

Since

NRC

IR 87-20,

the licensee

has

replaced

BF-21. 17 with

SDSP

15.9,

Nuclear

Experience

Review Program,

and performed the committed

gA audit.

The audit results

were

used in formulating

SDSP 15.9 and

in strengthening

the experience

review program.

There

were

22

IE

Notices

identified in the

gA audit that

were

reopened

based

on

incomplete committments.

The

inspector

reviewed

SDSP

15.9,

Revision

5 and determined

that

it provides

adequate

guidelines

and

checks

to ensure

that nuclear

experience

review action

items

are

tracked until completion

and-

acceptance

of

the

required

action.

Section

6.3 of

SDSP

15.9

requires

the responsible

supervisor to retain Attachment

G, "Closure

of

NER Item," until the action

items

are fully implemented.

When

completed

and

returned

to

the

BFN

Site

Licensing

Manager,

Attachment

G tracking

on

the

NER

and/or

TROI

data

base will

be closed.

The licensee's

action

on this item were responsive

and

acceptable.

This item is closed.

f.

(CLOSED)

IFI

260/87-33-02,

Fail ure

of

Drywel 1

Control

Air

Isolation Valves to Fail Closed

Upon Loss of Air.

During

performance

of Restart

Test

Procedure

-032,

the

drywell

control air suction

valves

(FCV-32-62

and -63) failed "as is" upon

loss of control air instead of failing closed,

as

was intended.

The

licensee

determined

the

cause of this malfunction to be the improper

implementation

of an

equipment modification intended to upgrade

the

solenoid

valves for environmental

qualification.

In addition,

the

following related

problems

were also noted:

Drawings

1-47E610-32-2

(Units

1

8 3) and 2-47E610-32-2 (Unit 2)

incorrectly depicted these

valves

as diaphragm valves.

Drawing 1-47E610-32-2

(Units

1

8 3) erroneously

indicated,

in

Note 8, that the air supply for these

valves in both units

was

drywell control air.

Th'ese

valves

were

found to

be missing

from

FSAR Table 7.3-1,

"Pipelines Penetrating

Primary Containment."

To correct

these

problems,

the licensee

has

completed

the following

actions:

The Unit

2 valves

have

been

replaced

per

ECN W0690

and work

plan 2353-88,

and were successfully

retested

on May 22,

1989 per

2-BFN-RTP-032,

CN-08.

Drawing

2-47E610-32-2

has

been

revised to identify accurately

the valves

as air operated-vane

drive motor plug valves.

Amendment

6 to

the

incorporated

these

valves

into

Table 7.3-1.

The

inspector

reviewed

the

above

completed

actions

and

determined

them to address

adequately

the identified problems

as

they pertain

to Unit 2.

Therefore,

this item is closed for Unit 2, but will

remain

open for Units j. and

3 pending

completion of the necessary

hardware

modifications

and drawing deficiencies.

In addition,

the

licensee

had also reported

the valve malfunction in

LER 260/87-07.

This

LER is closed in paragraph

2.c of this report.

g.

(CLOSED) IFI 260/87-37-03,

Reactor Mater Level Sensing

Lines.

This item involves questions

pertaining to licensee's

resolution of

the

February

13,

1985 reactor water level

mismatch

event.

General

Electric

had performed

a review of the event

and submitted

a report

to

TVA containing conclusions,

determination

of probable

cause,

and

recommendations.

This

report

was

an

attachment

to

GE letter

G-ER-6-333,

dated

August 21, 1986.

The cause of the above event

was

determined

to

be the rigid instrument piping system which would not

permit adequate

movement

upon thermal

growth of the reactor vessel.

To correct

the problem, the rigid instrument piping system in Unit 2

has

been

replaced

with a modified flexible system in accordance

with

ECN

E-2-P7131,

which completed all modifications to Reactor

Water

Level

Instrumentati on

necessary

to

suppor t

Unit

2

restart.

Therefore, this item is closed for Unit 2 only.

During the review of documentation

associated

with the installation

of the flexible instrument piping system,

additional

concerns

were

observed.

The

new installation is

comprised

of 1 inch diameter

stainless

steel

piping, with spring-can

hangers utilized to provide

. the desired 'flexibi.lity.

As Unit.,2 is currently in a cold shutdown

condition,'ost-modification

testing

could verify only the "cold"

settings

on

the

spring-cans.

The

"hot" settings

can

only

be

verified

when

the reactor

reaches

or nears

operating

temperature.

The

review of the

documentation

provided,

and conversations

with

licensee

personnel

indicate that,

due to the cold position of the

spring-cans

and the actual

length of the springs,

the expected

hot

position setting will be

adequate

when

the reactor

achieves its

expected

thermal

growth.

However, at present,

there are

no plans to

verify physically

these

hot settings

and engineering

calculations

have not been provided to support the above conclusion.

. A- second

concern

involves preventive

maintenance.

The previously

referenced

GE report contains

a statement

which reads

as follows:

"The drywell instrument line piping in all

3 units appeared

to have

been

designed

for flexibility, but was

a rigid system.

The rigid

system

appears

to

have

evolved

from the years

of operation

and

absence

of plant maintenance

of small

diameter piping."

When the

inspector

questioned

licensee

personnel

as to what actions

had been

or

would

be

taken

to

address

GE's

assessment,

no

evidence

was

provided

to

indicate

that

the

question

of maintenance

and/or

periodic inspections

had

been

considered

nor were

any

such actions

anticipated

in the future.

It should

be

noted that ISI programs

cannot

be

relied

upon

for these

lines,

as

ASME Section

XI

specifically

excludes

ISI requirements

for 1" diameter

and

under

piping

systems.

Resolution of these

concerns

involving preventive

maintenance

and verificaton of hot spring-can

hanger

settings will

be

tracked

as

inspector

followup item IFI 259,

260, 296/89-35-01.

This is not

a Unit 2 restart

item but programmatically it should

be

, addressed

during the

power ascention

testing,

prior to full power

operation.

(CLOSED) IFI 260/87-42-03,

Core Spray

ECN L2003 Closeout.

This

item concerned

work performed

under

ECN

L2003 which involved

licensee

action in response

to Generic Letter 84-11,

Inspection of

BWR Stainless

Steel

Piping.

This

ECN was to replace

304 series

stainless

steel

piping in system

75,

Core Spray, with carbon steel

to reduce

the potential for intergranular stress

corrosion cracking.

The IFI noted that

several

3/4

inch drain

and test lines

were

omitted from the

ECN and would thus remain stainless

steel.

0

13

The

NRC inspector

reviewed Safety Evaluation

L2003, Revision

3 dated

October

28,

1988,

which addressed

IGSCC

concerns

for small

bore

piping in the

Core Spray

system.

Generic Letter 88-01,

NRC Po'sition

on

IGSCC in

BWR Austenitic Stainless

Steel

Piping, states

that this

GL superseded

the requirements

of GL 84-11.

It further states that

the requirements

of GL 88-01

do not apply to piping less

than four

inches

nominal diameter

regardless

of code classification.

Based

on the revision of the

SE

and

guidance

of GL 88-01,

the

licensee

does

not intend to replace

the 3/4 inch drain

and test

lines

of

system

75.

This is consistent

with the

NRC's staff

interpretation of the

IGSCC requirements.

This item is closed.

(OPEN) IFI 259,

260, 296/88-04-04,

Single Failure Criteria Involving

Emergency

Core

Cooling

Systems

Identified as

Part of the Restart

Test Program.

This

inspector

followup

item

involved

a

licensee

identified

condition

where single failure design criteria

was not applied to

the design

of* subsystem

280, Battery Boards,

and subsystem

231,

and

the 480 Volt AC SDBD.

The finding was documented

on

CARR BFP 880067,

Revision

1.

These

two issues

represent

significant

examples

of

design

program. deficiencies.

These

and

other

examples

of single

failure violations are

discussed

elsewhere

in the report along with

the corrective

actions

to improve the design control program.

This

IFI involves only equipment modifications associated

with

CARR

BFP

880067.

CARR

BFP 88067 discusses

two

DC power systems:

the

250V

DC battery

supply that the

TS refer to as station unit batteries,

and the

250Y

shutdown

board batteries.

The unit batteries

supply certain safety

such

as

HPCI-valves

and containment isolation valves.

The

SD battery

board supply provides control

power for the load shed

logic of the 4160V AC shutdown boards.

The

CA(R stated

the loss of the

250V

DC Unit Battery Board

1 would

result in the loss of the

DC control

power for the load shed logic

features

to the 480V AC SDBDs

SDBD lA (Unit 1, Div 1)

SDBD 1B (Unit 1, Div 2)

and in the

loss of core

spray logic for Unit 1, Division 2.

This

violated single failure because

two divisions

were affected

by one

failure.

Loss of this load

shed

feature

during certain accident

conditions

will result

in overloading

the

associated

diesel

generator.

The

480V

AC shutdown

boards 'are

supplied

from the associated

4160V

AC

SDBDs.

The resolution of this problem

was the

reassignment

of the

250V

DC

control logic power supplies

of the 480V AC shutdown boards lA, 2A,

1B and

2B from the unit batteries

to the 4160V AC SDBDs,

250V DC SDBD

batteries

(SB-A, SB-B, SB-C,

and SB-D).

Now with the failure of a single

DC control

power source

such

as

SB-D, only the associated

4160V

AC board, its diesel

generator,

and

the

480V

AC

boards

fed

from them

would

be affected,

thereby

preserving single failure design criteria.

TYA implemented

the resolution of this problem

by performing work

associated

with ECN's

E-2-P7117

and

E-2-P7124

which reassigned

the

source of normal

480V SDBD control power feeds.

The inspector

reviewed

the

documents

provided

by the

TVA licensing

section

for

closure

of this

issue.

The

implementing

work

instructions

appeared

satisfactory.

The

NRC inspector

observed that

the safety evaluation

associated

with this modification specified

no

TS

changes

would

be

necessary.

The

10 CFR 50.59 evaluation

did

state

that the

bases

for auxiliary electrical

equipment,

section 3.9

of the

TS would need to

be revised.

A review by the inspector of

the

complete

TS

revealed

that this plant modification

caused

a

confusing relationship

between

two different limiting conditions.

TS 3.9.8.4 entitled "Operation with Inoperable

Equipment", requires

initiation of

and orderly shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months if a 4160V

AC

shutdown

board

and

any

480V

AC emergency

power

shutdown

board

are

at the

same

time.

This condition,

loss of a 4160

and

a

480V

AC board, will occur anytime

a

250V

DC shutdown board is made

inoperable.

This is

the result of the recent modification that

aligns

the

250V

DC shutdown

boards

to supply control power to both

of the associated

AC shutdown boards.

TS 3.9.B.8

addresses

the

loss of a

250V shutdown battery or its

associated

battery

board,

and permits

continued reactor

operation

for up to five days if a

250V shutdown battery or battery board is

inoperable.

Since

the loss of any

250V

DC shutdown

board

now always results in

the

loss of control

power,

to both

a 4160V AC and a 480V

AC board,

TS 3.9.B.4,

and

3.9.B.8 are conflicting with one another.

The

NRC inspector

considers

this IFI as

open until the conflict

between

TS 3.9.B.8

and

3.9.B.4V is

resolved.

This is

a Unit 2

restart

issue.

(OPEN)

IFI

259,

260,

296/88-05-06,

Potential

Single

Failure - Two Sets

of

Two

Dampers

From

Two Trains

are

Actuated

Thru One Relay.

This

issue

represents

a

design

deficiency that pertains

to the

secondary

containment

isolation

system.

Four ventilation

dampers

located

in the

equipment

bay

(Drawing 47E865

Damper¹

1-FC0-64-65A,

B,

C and D),

between

the inner and outer equipment doors,

were found

to close

on

an initiation signal

from either of two trains of the

SBGT.

The

two signals

actuate

the

same

single relay which closes

.. =

all

four

dampers.

Failure of this relay

would prevent

proper

operation of all four dampers.

15

This

item

was identified

as

a result of the Restart

Test

Program

(RTP-65-SBGT)

and

documented

by

TVA

on

CAQR

BFT 880186.

The

resolution of the

CAQR was to rework the

system

design

and return

the

system

to

one that

meets

design

requirements

pertaining

to

single failure.

The

work will not

be

performed until the next

refueling cycle.

Therefore

the

system

hardware is to remain

and

certain

compensatory

actions

will

be

,implemented

to

provide

assurance

that

the

system

functions will be

met.

This action is

effectively

a use-as-is

disposition to

a non-conforming condition,

at least

for the interim period before

the

hardware

is reworked.

The

use-as-is

disposition

is

a

design

output;

a modification of

design criteria

and therefore

requires

a

10 CFR 50.59 review.

The

NRC inspector

could find no 50. 59 review associated

with the

CAQR.

TVA was notified of the inspectors

concern

on August 8, 1989.

This IFI wi 11

remain

open until a documented

10 CFR 50. 59 evaluation

is provided or performed.

Review of other

CAQR's despositioned

as

rework or use-as-is

should

be performed to ensure that

a trend of such oversights did not exist

in

resolutions.

The

current

CAQR

program

is

documented

in

SDSP

3. 13,

Revision 2, "Corrective Actions,"

and in

concert

with

NEP 6.6,

Revision

1,

"10 CFR 50.59

Evaluations,"

provide

reasonable

assurance

that recently dispositioned

CAQR's are

not vulnerable to problems that existed in the earlier program.

k.

(CLOSED)

IFI

259,

260,

296/88-21-04,

Deficiencies

Identified

During Retest of LOP/LOCA C.

This item involves the failure of the

RHR

Pump

2A breaker to close

automatically during performance

of LOP/LOCA test

C in July 1988,

as

required

by procedure

2-BFN-RTP-L/L-C, Revision

2.

Upon discovery,

an unsuccessful

attempt

was

made to manually close the breaker

from

the control

room.

The following actions

were then taken:

Voltage

measurements

taken in the breaker

control

compartment

revealed

that the positive side of the

250V

DC close signal

was

present

up to the breaker position switch.

=

The

breaker

was

removed

from

its

compartment

for

troubleshooting.

All components

associated

with the charging,

closing,

and tripping circuits were

checked,

revealing nothing

that would indicate

a lack of continuity in the positive

250V

DC closing circuit.

Secondary

disconnect

HG2 was observed to stick slightly and

showed

signs

of arcing.

The pin

was

cleaned

with contact

cleaner

and exercised

several

times to eliminate the sticking.

Inspection of the breaker

compartment

revealed

the guide rail

in the bottom center of the

compartment

to be bent.

The rail

was straightened

and proper alignment verified.

16

The

breaker

was

racked

back into its

compartment,

tested

several

times,

and observed

to -be functioning correctly.

Subsequent

licensee

evaluatio'n

of the

above actions

and findings

determined

the probable

cause

of the failure of the breaker to be

slight misalignment

(due to the bent guide rail) in conjunction with

the sticking

MG2 pin.

These actions

and evaluations

were documented

on

TE -05, Maintenance

Request

908524,

and CA(R-BFP880518.

As previously reported

in

NRC

IR 88-24, at the request of NRC, the

PM records for this breaker

were reviewed.

It was revealed that

had not been

performed

on this breaker

in three years.

The failure

to perform the

required

PM on this breaker

and

on safety-related

.

4. 16

KV breakers,

i'n general,

resulted in the issuance

of Violation

259,

260,

296/88-24.-08.

Therefore,

as this specific breaker

has

been

adequately

addressed,

and

as

programmatic

corrective

actions

regarding

on

4. 16

KV breakers

is

being tracked

by the

above

violation, this item is closed.

(CLOSED)

IFI

259,

260,

296/88-21-05,

Vaulting of Completed

and

Approved Test Results.

This concern

was originally identified by the inspector

during the

continuous

observation

of the

RTP.

The inspector

reviewed

SDSP 2.5,

equality

Assurance

Records,

and

noted that completed

gA records

may

be, stored

up to

30

days in fire resistant

metal file cabinets.

The

NRC inspector

observed

that

RTP

gA records

were being maintained in

fire resistant

metal file cabinets

with restricted

access.

This

item is closed.

During

the

above

review,

an

additional

concern

was

observed.

SDSP 2.5,

Revision

9,

page

nine

contains

a

note

specifying

requirements

for temporary

storage

of gA records..

There is one set

of requirements

for records

being temporarily stored for 60 days or

less,

and

a second set of requirements

for records

beings temporarily

stored

for more

than

60

days.

TVA Topical

Report

TVA-TR75-1A,

Revision

10,

Table

17D-2, Sheet

7 makes

no allowance for temporarily

storing

gA records for periods

in excess

of 60 days.

This concern

is identified

as IFI 259,

260,

296/89-35-02,

pending resolution of

this

potential

conflict

between

the

commitment

and its

implementing

procedure

and

should

be

addressed

prior to restart

of Unit 2.

(CLOSED) IFI 260/88-28-03,

Spurious

RPS Trips Associated

With

Alternate

Power Supply and Circuit Protectors.

NRC

IR 88-28 identified the concern that

many spurious

RPS trips

were being actuated

by the

RPS circuit protectors.

This issue

was

also discussed

in IR 89-11.

The

RPS for each of the three

BFN units is divided into two trip

systems

(A and

B) and both systems

are provided with a

MG set.

The

MG sets

are

powered

from the

480

V auxiliary power system.

Each

unit also

has

a

single

maintenance

power

supply (alternative

e

,17

transformer)

that

can

be aligned to either

RPS distribution system

A

or

B,

but not at the

same

time.

Circuit protectors. are provided

between

the

output

of the

MG sets

and

the

breakers

for. the

associated

RPS distribution

bus,

and

between

the

output of the

regulating

transformer

and

the

connection

switches

to the

distribution.

The circuit protectors will open to disconnect

the

RPS distribution

on

under

voltage,

over voltage, or under frequency

conditions.

After the Unit 3

loss of power event

on March 7,

1989 (see

IR

89-11,

paragraphs

4 and 8) the licensee's

system engineers

issued

a

report

on

RPS circuit protector

performance

and

made

recommendations

for minimizing or eliminating circuit protector problems.

The

NRC inspector

reviewed the licensee's

report

and discussed

the

status

of the

recommended

actions

with the

cognizant

system

engineer.

The

NRC inspector verified that the licensee

performed

the following actions:

o

Operating

and

PM instructions

were revised to minimize the time

that

RPS buses

are left on the alternate

supply transformer,

o

Testing

of the

MG set

voltage

regulator

potentiometers

was

performed

and

instructions

for their

inspection

and

cleaning were enhanced.

o

Modifications

to

improve circuit protector reliability were

initiated.

The cognizant

system

engineer

stated

that these

actions

were taken

to

provide

better

performance

of the current circuit protector

design

and

to minimize the

chances

for spurious

trips.

System

engineering

also

requested

that

ONE reevaluate

the basis for the

circuit protector relaying setpoints,

reevaluate

the current

use of

time

delays

in

the circuit protectors,

and

perform

a safety

evaluation

to determine if the Unit 1 and Unit 3 circuit protectors

could

be

bypassed

unti 1 unit refueling.

These

actions

were not

complete at the

end of this reporting period.

The system engineer

stated

that these

evaluations

could

be

used to enhance

the current

systems

performance

but are

not

necessary

for it to provide its

intended function.

The

NRC

inspector

concluded

that

the

licensee

had

adequately

addressed

the

concerns

raised

by this item,

had taken actions

to

preclude

recurrence

of spurious trips,

and

were actively pursuing

actions

to

enhance

the current design.

This item is closed.

In

addition,

the

licensee

reported

the Unit 3

RPS loss of power in

LERs 296/89-03

and 259/88-18.

These

LERs are closed in paragraph

2

of this report.

18

'CLOSED)

IFI 259,

260,

296/88-32-02,

Diesel

Generator

Overspeed

RTP Test.

This item was originally identified by the licensee

and involved a

review of the

system

82,

DGs,

RTP test

procedure

results.

This

review indicated that

the

section of the procedure

involving the

overspeed

test

of the

3A

DG

was either

not performed

or

was

inadequately

documented.

A decision

to perform the

overspeed

test

on

3A

DG was

made

and section

5. 7,

Load Run,

Load Acceptance

Test,

and Miscellaneous

Tests,

Data Sheet

7.21 of RTP-082

was performed

on

October

27,

1988.

The

NRC inspector

reviewed

data

sheet

7.21 .and

noted that

the

overspeed

test of the

3A

DG was successful.

This

item is closed.

(CLOSED)

259,

260,

296/87-02-05,

Ambiguous

Surveillance

Intervals.

This item involved

by plant technical

once

per operating

old.

This applied

at all times.

the fact that certain surveillance tests

required

specifications

to

be performed at a frequency of

cycle had performance

dates

as

much as four years

to some

systems that were required to be operable

This

condition

was identified

because

of the duration

of this

shutdown period for the Brown's Ferry Units started

May 1985 and the

wording of the plant's

custom

TS.

Standardized

TS generally specify

18

months

as

a refueling

and operating

cycle.

This permits

the

application

of period

extensions

as

well

as

a bounded

period of

time.

TVA evaluated its surveillance

testing

program in its response

to

this

URI with the stated

intent of identifying tests

scheduled

on

a

once

per operating

cycle frequency that would more prudently be on

an

18 month frequency..

The investigation,

completed

in June

1987

indicated

44 survei llances

should

have their frequency

upgraded.

The

tests

were

primarily

on

secondary

containment

systems

and

control

room 'emergency ventilation.

The review also determined that

the tests

identified for

upgrade

had been performed in the prior 18

months.

TVA also

revised

SDSP 12.7,

"Systems Pre-operability Checklists," to

require

review of once-per operating cycle surveillances

to evaluate

whether

the

SI

needs

to

be reperformed prior to declaring

a system

operable.

The inspector

reviewed

the issue

and

TVA's corrective actions,

and

found that

the

program for scheduling

of the

once

per operating

cycle

surveillances

had

been

effective

as

evidenced

by

the

successful

reperformance

of previously

identified SI's

as

they

approached

an

18 month period since their last performance.

SDSP

12.7

was

reviewed with no

comment.

The

NRC inspector also reviewed

the

plant's

book

of

TS interpretations

for"the "purpose

of

determining if an official TVA position

had

been

documented

on the

19

once

per

operating

cycle

issue.

The

inspector

found

an

interpretation

discussion

on

TS wording of surveillance

frequencies

but the discussion

did not include the "once per cycle" issue.

This

point

was

brought

to

the

attention

of l.icensing

personnel.

Licensing

responded

with assurance

that

an 'expansion

of the

interpretations

would

be

considered

to

ensure

long

term

and

consistent

understanding

of the frequency issue until appropriate

TS

changes

were approved.

The

inspector

occurred

as

a

actions

taken

high level of

be

operable.

performance

of

closed.

determined

that

no violation of

NRC 'equirements

result of long surveillance

periods.

However,

the

by

TVA arq

expected.

to remain in place to ensure

a

confidence. will be maintained

in systems

required to

This

confidence will be obtained

by the successful

regularly scheduled

surveillance tests.

This item is

(CLOSED)

259,

260,

296/87-26-02,

Adequacy

of

Sampling

Program for Resolution of IEB 79-14,

Phase I Deficiencies.

This

item

involves

the

question

as

to whether

TVA's proposed

sampling

program

would

be

adequate

to resolve

concerns

regarding

pre-1985

walkdowns of piping supports

pertaining to IEB 79-14.

TVA

proposed

to perform walkdowns of 60 supports

to determine as-built

configurations,

'and

then perform evaluations

of these configurations

to determine

support

adequacy.

The results of this sampling program

were

intended

to provide

an acceptable

level of confidence

in the

Phase

I walkdowns

performed prior to 1985.

Subsequent

reviews

by

NRC staff have

determined

that the

proposed

sampling

program would

not

provide

adequate

assurance

as

to the

accuracy

of pre-1985

walkdowns.

Therefore,

TVA has

been directed to perform 100K of the

Unit

2

Phase II walkdowns

and

subsequent

engineering

evaluations

prior to restart.

These

directions

are contained in NRC letters to

TVA dated

March 25,

1988 and June

19,

1989.

Because it has

been

determined

that TVA's proposed

sampling program

cannot

be utilized in conjunction

with the overall

IEB 79-14

program,

and

as

future

NRC reviews

of the results

of the

100%

walkdown

and

evaluation will be performed

as part of the overall

assessment

of TVA's 79-14 program, this item is closed.

(Closed)

259,

260,

296/88-28-05,

Failure

to Report

Loss

of

Cooling Water to Diesel Generators.

During

an overheating

event

associated

with the

3C

DG that occurred

on September

29,

1988, the licensee

operated

the

DG for surveillance

testing with, no cooling water available.

The north

and south

EECW

headers

had

been unintentionally isolated at

an earlier date

from

the four Unit 3

DGs

due to a valve alignment problem resulting from

a

known

drawing

discrepancy.

The

condition

which

would

have

resulted

in both divisions of safety-related

electrical

equipment

failing went undetected

for three

days.

'A contributing factor to

this

problem

was the

absence

of either local or remote

EECW flow

20

instrumentation.

Violatjon 259,

260,

296/88-28-01

was

issued

to

document the licensee's

failure to maintain configuration control.

During, the

licensee's

subsequent

evaluation

of the event it was

determined

that

no report to the

NRC was required per 10 CFR 50.72

or 10 CFR 50.73.

The licensee's

basis for this conclusion

was that

no Technical

Specifications

were violated

and that the hydrostatic

testing

was

not

normally

performed

during

power

operations.

However,

10 CFR 50.73 (a)(2)(v) requires

that the 'licensee

report

any

event

or condition that

alone

could

have

prevented

the

fulfilment of a safety function needed to mitigate the consequences

of

an

accident.

In this

case all four

DGs would have

overheated

when called

on to perform their function.

During discussions

held

with the

licensee

the

inspector

was

informed that

an

LER would

be

submitted.

The

licensee

subsequently

submitted

LER 296/88-,007

dated

December

30,

1988,

to

cover

the

event.

This

LER

was

classified

by the licensee

as

an voluntary informational report and

submitted approximately three

months after the event occurred.

The

NRC inspector

determined

that

a violation did occur, i.e.

the

licensee

failed to report the event to the

NRC within 30 days

as

required

by 10 CFR 50.73.

Violation 259,

260,

296/89-27-03

was

subsequently

issued

by the

NRC

for three

separate

examples

of the licensee's

failure to submit

a

required

LER in

a timely matter.

Since this failure constitutes

another similar failure, it will be included

as

a fourth example of

Violation 89-27-03.

The unresolved

item is therefore

closed

and any

corrective actions will be followed as part of the violation.

This

item is closed.

(CLOSED)

URI 259,

260,

296/88-33-03,

Unauthorized,

Undocumented

and

Inadequate

Maintenance Activity.

This

item

involved

a field observation

by

a

NRC inspector

of

fasteners

that

displayed

improper thread

engagement

on

a bolted

flange connection

.

The work on the connection

was later found to

have

been

performed

by an improper expansion of scope of an existing

maintenance

request

that

was being worked in the

same

area.

When

notified by the

NRC, the licensee initiated

CARR

BFP 881020

which

documented

the conditions.

TVA corrected

the fasteners

by replacing

them with bolts

of the

proper

length.

Work was

performed

by

maintenance

request

A803275.

TVA investigated

the occurrence

to consider if this type of improper

work control

was

a trend.

This effort involved use of the

CARR

trending

program.

The results indicated

no trend existed

because

no

similar

occurrences

were

documented within the previous

6 months.

Corrective actions

involved training of maintenance craft personnel

in the

importance

of work scope

and control.

The apparent

poor

communication

between

the field craft and the work supervisor

(who

was contacted

prior to performing the additional work) was stressed

as

the

root

cause.

The

corrective

actions

are

considered

appropriate.

The

NRC inspector,

after review of

CARR Trending

and previous

NRC

documented

findings,

considers

no violation of

NRC requirements

occurred

and that this occurrence

was isolated

and not the result of

a programmatic deficiency.

This item is closed.

(CLOSED) URI 260/88-35-02,

Missed 'SBGT Surveillances.

This

item

involved

a

SBGT train

becoming

when

the

surveillance

period

required

by the

technical

specifications

had

expired.

The

item

was

considered

unresolved

at the

time the

inspection

report

was

issued

because

the operability status

of

redundant

trains of the

SBGT was not readily available.

Subsequent

investigations

by

TVA revealed

that

a

second train of

SBGT was

because

its onsite

power supply

was

not available

(DG

B

was inoperable).

Mith two of the three

SBGT trains

inoperable,

the requirements

for

secondary

containment

stated

in

TS 3.7.B. 1.b

were

not

met.

Secondary

containment

was required for ongoing fuel pool activities

(Ref.

TS

3'.C.4).

This

event

was

reported

as

an operation

prohibited

by

TS in

LER 260/88-19,

Revision

1.

The information provided in the

LER is the

basis for resolving

URI 260/88-35-02

and concluding that a violation

of

NRC

requirement

did occur.

This

URI is considered

closed

and

upgraded to a licensee identified violation.

The

violation,

NCV

259,

260,

296/89-35-03,

is

considered

a

.

licens'ee-identified

violation

and

is

not

being

cited

because

criteria specified

in section

V.G. 1 of the

NRC enforcement

policy

were satisfied.

TVA investigation of the events

indicated that when SI 4.2.A-12 was

performed

on

November

29,

1988,

the

SI steps

associated

with train

"C" were

marked

N/A as

allowed.

The

SOS

acknowledged,

by his

signature

on the SI that acceptance

criteria were incomplete.

The SI

scheduling

section

identified the

need for train "C" testing'nd

placed

the

item on

a schedule

requiring performance prior to fuel

load.

The

impact

on the operability of the

SBGT was not effectively

passed

on to the operating shifts.

TVA attributed this to a lack of

a

formalized

process

for tracking the inoperability status

of

equipment for which SI's

had been partially completed.

The corrective

actions

taken

by

TVA was to develop

a formalized

process

for documenting

and tracking the operability status

of TS

required equipment.

22

This program

process

consisted

of revising the procedure

describing

the

conduct

of testing

(PMI-17. 1)

to

provide

directions

for

documenting

the conditions that prevent

the. completion

of'

SI and

ensuring that the

SOS

and the

STA are notified.

The

STA would then

be required to enter the condition in the

LCO tracking system.

PMI-15. 10,

"Tracking of Limiting Conditions for Operations,"

was

developed

and

implemented

to assist

the

SOS with the tracking of

components.

The

NRC inspector

reviewed

the corrective

actions

and

determined

that

the

concepts

and

program

were

adequate

and

implemented

in a

timely manner

and should prevent

a recurrence

of the event.

(CLOSED)

259,

260,

296/89-08-03,

Loss

of

Approximately

200,000 Gallons

From the Condensate

Mater Storage

Tank.

This

item involved

an event which occur red

on February

10,

1989,

during which the level of the Unit 1 condensate

storage

tank dropped

from

an indication of 26.7 feet at 4:00

a.m.

to

an indication of

10. 1 feet at 7:30 a.m..

This corresponded

to a loss of approximately

200,000

gallons

of water

in 3.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

Although the release

of

water

was

not monitored,

the activity was within 10 CFR 20 limits.

This loss of water

was not acted

upon until the evening of the

same

day, indicating

a lack of awareness

of the status of plant systems

by the control

room personnel.

The failure to adequately

maintain

an

awareness

of plant

systems

status

is considered

a violation of

plant

procedure

PMI-12. 12,

Conduct

of Operations,

section

4.3,

"Shift Personnel

Conduct," which states:

The operator

at the controls

and the immediate supervisor

must

be

continuously

alert

to plant

conditions

and activities

affecting plant operations,

including conditions

external., to

the plant

such

as grid stability, meteorological

conditions,

and

change

in support equipment status;

operational

occurrences

should

be anticipated;

alarms

and off-normal conditions

should

be

promptly

responded

to;

and

problems

affecting reactor

operations

should

be corrected in a timely fashion.

This failure to follow procedures

is identified

as Violation 259,

260, 296/89-35-04,

Failure

to

Respond

in

a

Timely

Manner

to

Off-Normal

Conditions.

An

NRC

NOV will be

issued

rather

than

. classifying it

as

'licensee

identified" with

no

NOV since

the

violation

was self disclosing.

The unresolved

item

URI 259,

260,

296/89-08-03 is closed

and upgraded to a violation.

(CLOSED) URI 259,

260, 296/89-11-02,

Single Failure Criteria.

This

item

involved

the

TVA design

control

process.

The

NRC

I

P

23

inspectors

concern

arose

over

the fact that several

single failure

criteria violations

in areas

of mechanical,

electrical, civil and

I8C design

had

been identified by Restart Testing

and other

means in

recent

months.

TVA investigated

each

issue

and

several

CA(Rs

resulted.

Corrective actions

or suitable

compensatory

measures

were

proposed for each

case.

t

CARR

BFT 880186

provided

a discussion

of the root cause

and the

recurrence

control

program for identification of single failure

deficiencies.

In the root cause

analysis

of the

CARR, several

conditions

adverse

to quality

had

been

identified concerning failure to meet single

failure criteria as identified below:

SCRBFNNEB8604 - Loss of 250VDC Battery

BD 2 Causes

Loss of

Three

U-3 Core Spray

Pumps.

SCRBFNNEB8607 - Loss of 250VDC Battery Concurrent with Recirc

Discharge

Break Results in Only One

Core Spray

Loop.

SCRBFNEEG8654 - Loss of Paralleled

Diesels 1/2

D 8

3D Causes

Loss of SBGT Trains

B 8

C (Common to all

units).

SCRBFNNEB8612

Loss of ECCS Division I Inverter Power to ATU

Causes

Loss of Automatic Vacuum Relief of

Torus.

NCRBFNMEB8403 - Loss of Offsite Power Concurrent with an

Accident Signal

Causes

Loss of Control

Bay

HVAC.

CA(R BFP880067 - Loss of 250VDC Battery

BD 1 Causes

Loss of

480VAC Load-Shed

Signals to Both U-1 480V Shut

Down Boards

and

Loss of Core Spray

Loop II.

CA(R BFP880154D01,

D02,

D03 - Certain Battery Failures

Concurrent

With

a

LOP/LOCA

Results

in Inadequate

Combination

of RHR Pumps.

These

deficiencies

were

identified

as

a failure to

have

an

appropriate

design

control

to

ensure

compliance

with the single

failure criteria specified in the

The

TVA Design Control

Program

was

a basic

issue that lead to the

development

of the

design

baseline

verification

program.

The

applicability of single fai lure criteria

was

addressed

by the

DBVP

and

is

discussed

for each

system

in the

system's

Design

Basis

Document.

This

increased

visibility of

such

a

fundamental

consideration

currently provides

an

acceptable

action

to prevent

recurrence.

24

The

items identified thus far were early designs

performed

before

enhanced

programs

were developed.

TVA provided the

NRC

a copy of its Single Failure

Design Criteria

Document

(BFN-50-729)

used

in the analysis

of the design of fluid

and mechanical

systems

and subsequent

design

changes.

This document

was

developed

to promote

a general

understanding

of single failure

requirements

'and

was

issued

in June

1987

as part, of the

DBVP.

The

inspector

considers

the

actions

taken

by

the

licensee

to

be

appropriate.

A violation for failure to implement

adequate

design

controls required

by 10 CFR 50 Appendix B, Criterion III is not being

cited

becaus'e

the criteria specified in section

V.G. 1 of the

NRC

Enforcement Policy were satisfied.

This

NCV

(NCV 259,

260,

296/89-35-05)

requires

no response.

The

inspectors

will continue

to monitor

TVA activities in this area.

URI 259,

260, 296/89-11-02 is closed.

(CLOSED) VIO 260/84-34-03,

Core Spray Relief Valves.

This violation involved the failure to test

the

Core

Spray

System

suction

relief

valves

per

ASNE

code

subsection

IWV-3510

requirements.

The

violation

was- previously

discussed

in

NRC

inspection

report

89-19

where

corrective

actions

by the licensee

regarding

ASME Testing

were

found satisfactory

and

documented

therein.

The issue

was not closed at that time because

a

CARR was

open

that

identified

concerns

improper relief valve

sizing.

TVA Site

Licensing

provided

the

NRC with documentation

that the

relief valve sizing

issue

had

been satisfactorily resolved.

The

resolution

involved calculations

of actual'ystem

flow requirements

to ensure

existing relief capacity

was adequate.

The system vendor,

GE,

concurred with the licensee.

Calculations,

GE correspondence,

and other documentation

were

reviewed in the

CA(R 88-07-69 closure

package

and were found complete.

This violation is closed.

(Open)

VIO 296/85-13-01

Failure to

Shut

Down With

Two Reactor

Protection

System Water Level Instruments

Inoperable.

Following

a

NRC inspection

conducted 'to determine

the circumstances

surrounding

the inoperability

of two Unit 3

RPV water level

instruments

(LIS-3-203 A,

B) during

a reactor startup

on February

13,

1985, it was determined

that the responsible

licensee

personnel

failed to

commence

a reactor

shutdown in accordance

with required

actions

stated

in Technical

Specification

3. 1 (Table 3. 1.A).

T.S.

3. 1 states

there

shall

be

two operable

or tripped

systems

for

each trip function.

If the

minimum number of operable

channels

per

trip system

cannot

be

met for both trip systems,

the licensee

shall

initiate insertion of operable

control

rods

and complete insertion

of all operable

rods within four hours.

Even though there existed

sufficient redundant

information which should

have alerted operators

that

two required water level switches

were inoperable,

the licensee

did not shut

down

and continued

power escalation.

The reactor

was'

0

25

eventually

shutdown

on

March

9,

1985

to

conduct

further

investigations

required

by

TVA management

following review of the

-circumstances

associated

with the event.

This resulted

in the

NRC

issuing

a

severity

level II violation with

a civil penalty

(EA-85-13).

The inspector

reviewed

the

licensee's

responses

to the violation

and civil penalty

dated

August 21,

1985

and August 30,

1985.

In

that

response

the licensee

has

stated their inability to determine

explicitly the root cause

for the observed

level mismatch which led

to the

event.

It is suggested

that the level

mismatch

was

most

likely caused

by

a loss. of water. in the

"A" instrument

reference

leg.

Two possible

causeh

are

as follows:

Reference

level

leakage

via identified transgranular

stress

corrosion

cracking

in the line that existed

adjacent

to the

X-28 drywell penetration.

This

cracking,was

found during

post-event

investigation

and

repairs

have

been

made

to the

affected line.

Potential

for the presence

of air bubbles in the "A" reference

leg.

This possibility is

supported

by licensee

engineering

calculations

and

may

have

been

enhanced

by the

above listed

cause.

The presence

of high points in horizontal

runs

and the

number

and character

of restrictions

gives credibility to this

possibility.

Additionally, various activities affecting vessel

water level

and negative pressures

maintained

on the vessel

for

several

days prior to the startup could have contributed to the

introduction of air in the horizontal

runs of the reference

leg

lines.

The licensee's

analysis

of operator actions

pointed, out the

need for

additional

training

in diagnosing

water

level

instrumentation

problems

at off-rated conditions.

A similar condition

had existed

- during

an earlier startup that occurred

on

November

20,

1984

when

operators

also failed to diagnose correctly and fully appreciate

the

condition.

The

inspector

reviewed

'the

licensee's

corrective

actions

for

this violation.

Specifically, the following corrective actions

were

noted:

Both of the

above potential

hardware

problems

should

have

been

corrected

by completion of

ECN E-2-P7131.

This

ECN is related

to

NUREG-0737,

Item II.F.2,

and relocated

the vertical runs of

reference

legs

outside

of the drywell

so

as to minimize the

potential

of erroneous

level

indications

resulting

from the

post-accident

environment

in the

drywell ~

This modification

was

completed

by

TVA during the

second

half of 1988

and is.

covered

in

more

detail

in

NRC

IR 88-32.

The

inspector

noted that during the

ongoing work gC inspection

was included

to verify instrument

line slope criteria were satisfied

and

that the presence of,high points in any horizontal

runs should

not be

a problem.

The

inspector

reviewed

documentation

including

memoranda

and

training

department

lesson .plans

associated

with classroom

and

simulator training.

Lesson

plans included training on the types

and

design

of the

available

level

instruments,

and their

expected

response

during

normal,

off-normal

and accident

conditions.

This

completed

licensee

training was provided to operators,

management,

and

STAs

and

was

intended to enable

them to more rapidly diagnose

water

level indication problems.

Additional training as part of the

planned start-up training program will cover the

same

subjects

and

is scheduled

to be completed

by December

22,

1989.

=The

inspector

examined

copies

of, training records;

a manager

of

licensing

memorandum

dated

March 21,

1989

(R08 890321 878);

and

BFN

Site

equality

Surveillance

Monitoring Report

dated April 14,

1989,

(R22

890414

973).

The monitoring report

was

conducted

by the

licensee

to

independently

verify. closure

of the

commitment

to

provide the training.

The

inspector

reviewed

the

licensee's

Unit

2

Operational

Readiness

Review Interim Report

dated

June

9,

1989.

This review

performed

by licensee

corporate

management

was the first of a two

phase

assessment

of the

readiness

of Browns

Ferry for restart.

Section VI.d covered

reactor vessel

water level

and included various

identified deficiencies

some of which are

as follows:

Interviews with operators

and

STAs indicated

an inconsistent

understanding

of what

was

entailed

in the

reference

leg

modification.

Documented

training

to

operators,

STAs,

and

management

to

enable

them to more rapidly diagnose

level indication problems

did not adequately

cover

the

new water level

reference

leg

installation.

Post modification testing did not verify proper function of the

modified system.

The acceptance

criteria

band specified in the post modification

test

equated

to approximately

27 inches of water.

Significant

indication errors

such

as

trapped air would not

be

cause

for

rejection.

Site licensee

management

has

not yet responded

to this review.

Due

to the significance of the

above

licensee

identified deficiencies

and

the

apparent

inconsistencies

between

these

comments

and the

documentation

provided

by site licensee

personnel,

this item will

remain

open pending further review.

(CLOSED)

VIO

296/86-06-06,

RHR/RHRSW/Diesel

Generator

'Inoperability.

This violation resulted

from

a

personnel

error of failure to

recognize

the inoperability of redundant safety systems.

One system

27

was

due to

an ongoing surveillance test

and the

second

system

was

made

when its associated

diesel

generator

was

removed

from service for scheduled

maintenance.

The combined affect

of these

out of service

systems

was

a reduction of RHRSM systems

to

less

than that

required

by the

TS for the

then

current plant

conditions.

TVA identified this violation and reported it in LER 296/86-04.

The

NRC considered

this occurrence

as

unnecessary if corrective action

from

a previous violation (84-26-02)

had been properly implemented.

The

NRC therefore

issued

a

NOV.

'VA responded

to the

NOV in a letter to the

NRC dated

May 1,

1986

and detailed

the corrective actions for the violation.

The

inspector

reviewed

the

TVA compliance

section

documentation

of the

followup and

closeout

of the plant's activities for this

violation.

The

package

was thorough

and

complete.

All corrective

action

commitments

were

found to have

been

completed.

Corrective

actions

included clarification of TS 3. 0. 5

as it applies

to cold

shutdown

conditions

and

development

of shift turnover checklists.

This violation is considered

closed.

(CLOSED) VIO. 259,

260, 296/87-14-02,

CREV Train

B Inoperable.

This violation involved the

CREV system

and the fact that the system

was

determined

to

be

because

air

flow rates

were

inadvertently

set

below the

minimum allowed

by the

TSs.

This

violation

was

discussed

previously

in inspection

report

89-19.

In that report the licensee's

corrective actions

were

reviewed

and

several

a'spects

of the violation were closed.

The following items

were not closed at that time:

1)

The results

of special test

ST 8726 designed to analyze

systems

flows in various

damper line-ups indicated that the

CREV system

flowrates

could

exceed

the

TS

maximum allowable with certain

damper alignments.

2)

The

CREV

system

did not meet the design

content of the

because

significant unfiltered

inleakage

of outside air into

the control

room habitability zone bypassed

the system.

3)

The

acceptance

criteria of the

TS Surveillance

Instructions

could

be met satisfactorily

even

though the

system

would not

perform its intended function because

of unmonitored inleakage.

4)

The related

issue

of the effect of toxic chemical

releases

from accidents

on transportation

routes

near the site did not

appear to meet requirements

of R. G. 1.78.

These

complex

issues

addressing

the ability of the

CREV system to

maintain

the control

room habitability during accident

conditions

has

been

under consideration

by the

NRC and

TVA for a long period of

time.

28

Recent actions

on the remaining items are

as follows:

1)

After *a

review of special

test

ST

8726,

flowrates

were

determined

by the

NRC inspector to

be within TS limits.

This

determination

was

reached

after discussion

with ventilation

system

engineers

and analysis

of test results

summary.

This

item is considered

closed.

2)

The fact that

the

CREV

system

does

not

meet its

intended

function

is

the

specific

topic of

an

amendment

request

submitted

by

TVA

to

the

NRC

in

a

letter

dated

February

14,

1989.

The

change

request

number

265T discusses

in

detail

the deficiencies

of the

system

design

and operation.

Approval of the

change

request or other action will be required

before

Unit 2 startup

to resolve

the

CREV system operability

issue.

3)

The fact that

a

TS surveillance instruction assumes

that system

integrity had

once

been

established

is an acceptable

practice.

The

SI

then verifies that

major

components

of that

system

continue

to

function

as

designed

and

other administrative

programs will preserve

system structural

configurations.

These

programs will ensure

that

system

performance will not degrade

unknowingly.

The

NRC inspector

has

determined after

review of

the

CREVs

system

surveillance

program

that it meets

TS

requirements

as

well

as industry standards

and is considered

satisfactory for this item to be closed.

4)

The issue of toxic gas

releases

near the site was discussed

in

a letter

from TVA to the

NRC dated

June

27, 1989.

In summary,

that letter stated

that

TVA concluded

that the plant

meets

Reg Guide 1.78

as it applies

to

Browns

Ferry Toxic

Gas

Analysis.

The

discussion

within the letter directly

and

clearly addresses

the

NRC concerns.

Resolution of this issue

now rests

with the

NRC.

Since the scope of this issue

exceeds

the

scope

of the original violation and is clearly documented

in the

June

27,

1989 letter,

the

NRC inspector

considers

the

violation regarding

CREVs testing

techniques

as

closed.

This

is not

an acceptance

of the control

room toxic gas analysis

or

habitability issue.

This violation is closed.

(CLOSED)

VIO 259,

260,

296/88-18-02,

Failure to Initiate a CA(R for

the Overloaded

1/2

D DG.

This item involved

a personnel

error leading to

an overload of the

Units 1/2

D

DG which occurred

during

a conduct of a special

test

ST 88-09

in June,

1988.

The

craftsman

was

taking

a reading to-

verify

a

parameter

being

monitored

on

a recorder.

The

actual

over load condition lasted for approximately

30 seconds.

However,

as

a result

of the

event,

no

CARR

was initiated. 'he

inspector

reviewed

SDSP 3.7,

"Correction Action," and

SDSP

3. 13, "Corrective

29

Actions"

and

noted that both procedures

outline activities required

when

CA(Rs are initiated, reviewed,

and closed out..

SDSP 3.7 states

the following in subsection

6. 1. 1:

Confirmed degradation,

damage,

failure,.'alfunction,

or loss of

plant,.structures,

systems,

and

components

that could adversely

affect

the

performance

of

a safety-related

function (i.e.,

nonconformance).

This

would include

but not

be limited to

material

failure,

abnormal

or

unexpected

wear,

manufacturer

defects,

fai lure

to

function

as

intended,

and repetitive

failures.

SDSP 3. 13 states

the following in subsection

6.2.1.F:

Items

which have

been

subjected

to conditions for which they

have

not

been

designed,

unless

done

intentionally

by

an

approved

and

properly

authorized

procedure

such

as

overpressure,

overvoltage,

overheating,

overstressing,

or

environmental

conditions

hazardous

to their function.

This

appears

to

be

an

inconsistency

in

. that

SDSP 3.7 indicates

Confirmed

Damage,

whereas

SDSP

3. 13 indicates

Condition for Which

They

Have

Not Been Designed.

Both SDSPs

are in effect as of the end

of this reporting period.

The

NRC inspector further noted that

SDSP 3.7,

section

2. 1 states

the following:

CARR's initiated

on or after August 16,

1988, shall

be processed

in accordance

with

SDSP

3. 13, Corrective Actions.

The

NRC inspector

also

noted that

SDSP

3. 13, initial revision was dated

August

5,

1988

and that

any future

occurrences

of

a system

or

component

being subjected

to conditions for which they have not been

designed will be initiated

and processed

in accordance

with

SDSP

3. 13.

This item is closed.

4.

Exit Interview (30703)

The inspection

scope

and findings were summarized

on August 16,

1989 with

those

persons

indicated in paragraph

1 above.

The inspectors

described

the

areas

inspected

and

discussed

in detail

the

inspection

findings

listed below.

The licensee

did not identify as proprietary

any of the

material

provided

to

or

reviewed

by

the

inspectors

during this

inspection.

Dissenting

comments

were not received

from the licensee.

Item

Descri tion

259,

260, 296/89-35-01

259,

260, 296/89-35-02

259,

260, 296/89-35-03

IFI, Flexibility of Reactor Water Level

Sensing

Lines, paragraph

3.g.

IFI, Storage of gA Records,

paragraph 3.l.

NCV, Missed SI Results

in a

TS Violation,

paragraph

3.s.

30

259,

260, 296/89-35-04

259,

260, 296/89-35-05

Violation.

Failure to Respond

in a Timely

Manner

to=-

Off-Normal

Conditions,

paragraph 3.t.

NCV, Design Control of Single Failure,

paragraph

3'.

Acronyms

ASME

ASOS

ATU

BF

BFNP

BWR

CAQR

CHRRM

CREVS

DBVP

DCN

GL

IEB

IEEE

IER

IFI

IGSCC

IM

IR

IRM

ISI

KV

LCO

LER

LIV

LOP/LOCA

MG

MMI

NCV

NOV

NPRD

NRC