ML18033A538

by Proceeding

with Unmonitored

Core

Loading of Unit 2.

This

Constitutes

a

Potential

Unreviewed

Safety

Question

and

Com-

promises

Fundamental

Safety Principles

(paragraph

4.a)

259,260,296/89-04-02:

Failure to

Implement the Requirements

of

Procedure

SDSP-27.

1

to

Perform

Adequate

Unreviewed

Safety

Question

Determinations,

as

Evidenced

by Numerous

Inadequacies

in

the

Reviews

of

Fuel

Loading

Procedures

(paragraph

4.b)

259,260,296/89-04-03:

Failure

to

Provide

Adequate

Cross

- Disciplinary

Review

of

Procedu'res

Impacting

Plant

Safety

(paragraph

6)

Three

unresolved

items

were identified.

One

concerned

the adequacy

of the licensee

determination

that the

unmonitored core loading was

not reportable

per

10 CFR 50.72

or 50.73

(paragra'ph

10),

another

concerned

the

procedure

review process- (paragraph

6),

and the third

concerned

the

review

of

Technical

Specification

(TS)

requirements

for core monitoring (paragraph

5).

All of

the

identified violations

and

unresolved

items

must

be

satisfactorily resolved prior to Unit 2 restart.

The

inspection

noted

significant

weaknesses

in the

areas

of fuel

loading operations,

10 CFR 50.59 safety reviews,

review and approval

of procedures,

and

TSs.

The

inspection

also

indicated

that

the

licensee

accepted

without question

the provision of TSs which did

not preclude

unmonitored

core alterations

and

may

have

taken

non-

conservative

and

improper

advantage

of existing

TS

wording

in

performing

unmonitored

core alterations.

As

a

consequence,

this

gives

indication of

a general

licensee

attitude

which appeared

to

emphasize

compliance rather than

safety

in order to

accommodate

the

easiest

option of performing the fuel loading operation.

When the problem was initially identified, the licensee's

assessment

and actions

were considered

to

be nonconservative,

incomplete,

and

inadequate.

Once

the licensee

acknowledged

the full significance of

the

issues

of unmonitored

core

loading,

however,

the

corrective

actions.

taken

were

appropriately

conservative,

thorough,

and

acceptable.

L1

-

1

REPORT

DETAILS

Persons

Contacted

Licensee

Employees:

0. Kingsley, Jr.,

Senior Vice President,

Nuclear

Power

C.

Fox, Vice President

and Nuclear Technical Director

J.

Bynum, Vice President,

Nuclear

Power Production

C. Mason, Acting Site Director

"G. Campbell,

Plant Manager

H. Bounds,

Project Engineer

.

"J. Hutton, Operations

Superintendent

"D. Mims, Technical

Services

Supervisor

G. Turner, Site gual,ity Assurance

Manager

P. Carier, Site Licensing Manager

J. Savage,

Licensing Supervisor

A. Sorrell, Site Radiological

Control Superintendent

Other

licensee

employees

or contractors

contacted

included

licensed

reactor

operators,

auxiliary

operators,

craftsmen,

technicians,

and

quality assurance,

design,

and engineering

personnel.

NRC Attendees

"D. Carpenter

"E. Chri stnot

K. Ivey

"P. Castleman

"Attended exit interview

On January

9

and

10,

1989,

while

NRC managers

were

on site for

a plant

tour

and

schedule

review,

TVA management

made presentations

to the staff

on the root cause

of the

unmonitored

core

loading,

corrective

actions

(short

term

and

long term),

and

plans

for resumption

of fuel loading

activities.

Attachment

A to this report

summarizes

TVA's presentations.

The following persons

were in -attendance:

Licensee

attendees:

0. Kingsley, Jr.,

Senior Vice President,

Nuclear

Power

J.

Bynum, Vice President,

Nuclear

Power Production

C.

Fox, Jr.,

Vice President

and Nuclear Technical Director

J.

Kirkebo, Vice President,

Nuclear Engineering

N. Kazanas,

Vice President,

Nuclear guality Assurance.

R. Gridley, Director, Nuclear Safety

and Licensing

J

~ Robertson,

Manager,

Nuclear

Fuel

C.

Mason, Acting Site Director,

Browns Ferry Nuclear Plant

G. Campbell,

Plant Manager.

G. Turner, Site guality Assurance

Manager

P. Carier, Site Licensing Manager

J.

Savage,

Licensing Supervisor

T. Overlid, Nuclear

Manager's

Review Group

Licensee contractor attendees:

W. Cobean,

TVA Consultant

P'. Marriott, General Electric

D. Janecek,

General

Electric

NRC attendees:

P. Burnett,

Reactor Inspector

D. Carpenter,

Site. Manager

P.

Castleman,

Plant Systems/TVA Projects

E. Christnot,

Resident

Inspector

K. Ivey, Resident

Inspector

A. Johnson,

Project

Engineer

B. Liaw, Director,

TVA Projects Division

W. Little, Section Chief, Inspection

Programs,

TVA Projects Division

A. Long, Project Engineer

E. Marinos,

Branch Chief, Reactor Operations

Branch

F.

McCoy, Assistant Director,

TVA Inspection

Programs

Acronyms used throughout this report are listed in the last paragraph.

Sequence

of Events

After an

extended

shutdown

of over four years

in duration,

Unit

2 fuel

loading

commenced at 9:50 a.m.,

on January

3,

1989.

The reactor

core

included four Source

Range Monitors

(SRMs),

one in each

quadrant,

to provide

neutron

monitoring during fuel

loading.

TS

3. 10

states

that

a minimum count rate of 3 cps is required for SRM operability

unless

other specified conditions are met.

Because

the

SRMs were reading

less

than

3 cps,

the licensee

performed fuel loading in accordance

with TS

3. 10.B. l.b.2.

This

allowed

count

rates

less

than

3

cps

provided

SRN

response

checks

were successfully

performed every

8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months using

a neutron

source,

both fresh

and irradiated

fuel were

b'eing

loaded,

and the core

was

loaded

in

a spiral

sequence.

The

TS does

not specify what

minimum

count rate is acceptable.

As the assemblies

were loaded,

the count rate

on

SRN

C fluctuated

between

0.2 cps

and 0.65 cps,

B indicated

between

zero cps

and 0. 17 cps,

and

A and

D showed

no detectable

count rates.

During the source

checks,

the

SRMs responded

with count rates

on the order of

3 cps.

The observed

responses

with

and without the

sources

indicated

that

the

SRNs

were

operable

and capable of detecting

neutrons,

but at least

three

SRMs were

not continuously

responding

to core

neutrons

from the fuel configuration

~

1

established

during the,.core

reload

process.

The

response

of the fourth

SRM, channel

C,

was also questionable.

Both prior to and immediately following the initiation of the fuel loading

process,

NRC inspectors

questioned

the

licensee

regarding

whether

the

positive reactivity additions

from the fuel assembly

movements

were being

adequately

monitored

to

ensure

core

safety,

since

the

were

not

responding

to neutrons

from the fuel.

The licensee

contended

tha" per

TS

3. 10.B. l.b.2,

SRM response

to core

neutrons

during core

loading

was not

required.

On January

5,. 1989, after approximately

45 fuel assemblies

had

been

loaded

into the core without achieving

any observable

response

on

more than

one

SRM,

an

NRC inspector

again raised the issue of adequate

core monitoring.

The licensee

continued to maintain that it was unnecessary

to require

response

to core neutrons

because it was not required

by TS,

and obtained

concurrence

on this position from the

GE representative

on site.

Licensee

reactor engineers

indicated to the

NRC inspectors that,

due to the length

of the extended

shutdown,

the radioactivity level of the fuel

was

so

low

that the loading of approximately

200 bundles

was anticipated

before the

count rates

on all

SRMs would exceed

3 cps.

The licensee

was

requested

to provide the

NRC inspectors

with

a safety

analysis

supporting

the

unmonitored

core

loading,

including:

(1)

analytical

verification that

the

unresponding

would promptly

and

adequately

reflect

any

significant

adverse

flux trend;

and

(2)

the

calculations of the minimum number of assemblies

required for criticality.

Simultaneously,

the

NRC inspector's

concern

was

escalated

within

NRC

senior

management

and appropriate

NRC Regional

and Headquarters

technical

sections.

On January

5,

1989, at approximately 5:00 p.m., fuel loading

was halted

by

the licensee

pending review and resolution of the

NRC concerns,

with 74 of

the total

764 fuel assemblies

loaded.

Recorded

SRM count rate readings at

the termination of fuel loading

showed

C at 0.8

cps

and

the other

three detectors

reading essentially

zero.

On January

8,

1989,

at approximately

1: 15 a.m.,

the licensee

moved

D

to the location of

IRM

F near

the center of the core

and reestablished

core monitoring'n indication of 35 to 40 cps was attained

and the count

rate

remained stable at that level.

Safety Significance of the Event

Inadvertent criticality is prevented

during core alterations

by the margin

of safety

provided in the core design,

through refueling inter locks,

and

by continuous

core neutron flux monitoring.

Neutron

monitoring is essential

to

ensure

the

prompt detection

of and

operator

response

to

an

inadvertent

criticality.

The

safety

impact

of loading

fuel without the

on

scale

is that if a criticality

l

condition

did

occur, it would

continue

undetected

until flux levels

increased

enough -to bring the

SRM readings

on scale.

In such

an event,

the lack of core

neutron

monitoring could delay actions

to mitigate the

consequences

of an inadvertent criticality accident.

Detector

"response"

to neutrons

must

be distinguished

from

a detector

being

"operable"

per

TS

requirements.

The

were verified to

be

operable

by the periodic source

checks

performed during the fuel loading;

however,

detector

operability

does

not

assure

the

monitoring of core

neutrons.

For monitoring to occur,

an operable

detector

must

be

in

a

geometry

which assures

exposure

to sufficient core flux for the detector

to

be

on scale

and

responding directly to changes

in the

magnitude

of

the neutron

flux'icensee

reactor engineers

indicated to the,NRC inspectors that as few as

eight fuel bundles

could

have

achieved criticality if two control

rods

were withdrawn.

The licensee

loaded

several

multiples of this potential

critical

mass

without

core

neutron

monitoring.

Based

on

additional

licensee

calculations,

criticality could

have

occurred

during the

core

loading

sequence if four worst case

loading errors

had occurred

combined

with the withdrawal of the highest worth control rod.

In this specific

case,

no control

rods

were

withdrawn

and

no loading

sequence

errors occurred.

Consequently,

as discussed

in paragraph

9, the

reactor

was in fact adequately

shutdown.

However, the licensee

did load

74 fuel bundles into the core without continuous

neutron flux monitoring.

This is contrary "to the fundamental

concept of not adding positive reacti-

vity or making

core alterations

without the ability to determine

the

effect of that reactivity addition or alteration

on the core.

Unreviewed Safety Question Determinations

10 CFR 50.59 requires

that the holder of a license authorizing operation

of

a

production

or utilization facility must

receive

prior

Commission

approval

to

make

changes

in the procedures

as described

in the safety

analysis

report if the

proposed

changes

involve

an

unreviewed

safety

question

(USQ).

Two conditions

of

10 CFR 50.59 for which

a

proposed

change

shall.

be

deemed

to involve

an

unreviewed

safety

question

are:

(I) if the

consequences

of

an

accident

or malfunction of equipment

important to safety previously evaluated

in the safety analysis

report

may

be increased;

or (2) if the margin of safety

as defined in the basis for

any

TS is reduced.

The holder of

a license

authorizing

operation

of

a

utilization facility who desires

to

make

a

change

in the

procedures

described

in the safety analysis report which involve an unreviewed safety

question

must submit

an application for a license

amendment.

Site Director

Standard

Practice

(SDSP)

27. 1,

"Evaluations .of

Changes,

Tests,

and

Experiments-Unreviewed

Safety

Question

Determination,"

which

implements

the requirements

of 10 CFR 50.59, requires that

new procedures

or proposed

changes

to existing procedures

'be given

a screening

review to

determine

whether

the

proposed

change

could impact nuclear safety.

This

screening

review process

applies the criteria in 10 CFR 50.59 to determine

if proposed

changes

require

a safety

ev'aluation

or

TS

Change.

If it is

determined

that there

could

be

an

impact

on

nuclear

safety,

a safety

evaluation

of the

proposed

procedure

change

is required.

The

safety

evaluation

determines

whether

a proposed

change

involves

a

US/ or change

to

a

TS and therefore

would require prior NRC approval.

'a

~

Core Loading Procedures

In August 1988,

2-GOI-100-3, "Refueling Operations,"

was approved in

preparation for loading fuel in Unit 2.

This procedure

prescribed

a

fuel

loading

sequence

which did not provide for continuous

source

range

monitoring.

Specifically,

the

procedure

did

not

delineate

a

minimum

acceptable

count

rate

or

assure

that

the

were

responding

to core

neutrons.

The

screening

review of

the procedure,

conducted

per

SDSP 27. 1, indicated that the procedure

could not impact nuclear safety.

Therefore,

no safety evaluation

was

performed to determine

whether the proposed

change

involved a US/ or

required

a

TS change.

The inspectors

consider that

a procedure

which

allows unmonitored positive reactivity additions

does

impact nuclear

safety,

and

consequently

should

have

been

supported

by

a

proper

safety evaluation

as required

by 10 CFR 50.59.

The written justification

on the

screening

review,

which supported

the classification

of 2-GOI-100-3

as

having

no potential

safety

impact,

stated that the proposed

steps

were within the guidelines of

the

TS and

FSAR.

The TSs were

amended

in 1979 to allow fuel loading

with

SRM count rates

less

than

3 cps under certain conditions.

The

inspectors

considered

that

the

licensee's

safety

evaluation

supporting this amendment

was inadequate

as discussed

in paragraph

5.

FSAR Section 7.5.4. 1, which documents

the design

basis

of the

SRMs,

states

that

neutron

detectors

shall

be provided which result in

a

count

rate

of

no

less

than

3

cps with all

control

rods .fully

inserted.

The

FSAR was not updated

when the

TSs were changed,

so the

change

in TS 3. 10 was clearly in direct contradiction with the

requirement.

When

2-GOI-100-3

was written in

1988,

the trained initiator and

qualified reviewer performing the procedure

review failed to identify

the contradiction

between

the

and

TS.

Additionally, review of

the

applicable

portions

of the

and

TS,

as

required

by the

SDSP

27 '

screening

review apparently

did not include

a review of

the

TS Bases

and

SER,

which require

core

monitoring.

Due to the

inadequate

screening

review,

a safety

evaluation

was

not performed

and

a potential

unreviewed safety question

was not identified.

The inspectors

consider

that

had

an appropriate

10 CFR 50.59 safety

evaluation

of

2-GOI-100-3

been

performed

when

the

procedure

was

written, the necessity

for adequate

core

neutron monitoring

should

t

I

have

been identified

and the contradiction

between

the

FSAR and

TS

should

have

been resolved.

In addition,

an

adequate

evaluation

of 2-GOI-100-3

should

have

questioned

the applicability of TS 3. 10.B. l.b.2, which

allows fuel

loading with

SRN count

rate

levels

less

than

3

cps

provided that both fresh

and irradiated fuel are loaded.

Irradiated

fuel should provide adequate

minimum flux'levels for core monitoring,

to meet the intent of TS 3. 10.B. l.b.2.. In this case,

the irradiated

fuel should

have

been considered

equivalent to fresh fuel due to the

decay

of the

neutron

levels

in the irradiated

fuel

during the

extended

shutdown.

Performance

of

an

unmonitored

core

loading is considered

to 'be

a

potential

Unreviewed

Safety Question'n

that it may

increase

the

consequences

and/or probability of an accident

previously evaluated

in the

SAR and

may

reduce

the margin of safety

as defined

in the

basis for TS 3. 10.B. l.b.2 (see

paragraph

5 of this report).

The fact that the licensee

began

an unmonitored core loading without

performing

a

proper

evaluation

and

obtaining prior

NRC approval

as

required

by

was identified

as

apparent

violation

260/89-04-01.

b.

Programmatic

Assessment

NRC inspectors

reviewed

the

adequacy

of the licensee's

program for

unreviewed safety question determinations.

The

NRC

reviewed

13 screening

reviews

(performed

between

June

and

December

1988)

associated

with refueling

procedures

and

their

revisions.

Several

errors

were

noted.

In

some

cases,

these

errors

were

not

in accordance

with SDSP-27. 1, while in other

cases,

the

errors

appeared

to violate the intent of both

SDSP-27.

1

and

10 CFR 50.59 to ensure that proposed

changes

do not adversely

impact nuclear

safety.

The following deficiencies

in implementation of SDSP-27.

1 were noted:

1)

Section

6.2.3

and Attachment

B

"Yes,"

"No " or "N/A" response

screening

review form be given

support

that

conclusion."

In

assessed,

there

were

a total of

these:

of SDSP-27.

1 require that

each

to the three

questions

on the

"sufficient justification... to

aggregate,

among

the

13

SRs

39 justifications required.

Of

Two were left blank.

Six merely stated that the proposed

changes

did not affect

nuclear

safety,

but

did

not

provide

any

supporting

analysis

or other

expl'anation.

The

inspectors

consider

that these

changes

could have affected nuclear safety.

Eleven

were

incomplete

and/or

included what the inspector

considered

to be illogical assessments

of the issues

being

screened.

Many of the justifications failed to answer

the

question of whether the issues

could impact nuclear safety.

Several

justifications

stated

that

the

changes

had

no

safety

impact

because

they were administrative

in nature,

but this

assumption

was

not

supported.

A

number

of

discrepancies

were

observed

between

the

information

provided

on

different

screening

forms

for

the

same

procedures.

Fourteen

were identified as

N/A as allowed by SDSP-27. 1.

Six were considered

to be satisfactory.

From

the

above

categorization

of justifications',

additional

analysis

and justification

should

have

been

performed

for

approximately half of the questions

reviewed.

For those

pro-

posed

changes

which could not be categorically

shown to have

no

impact

on safety,

a

safety

evaluation

(SE)

should

have

been

performed.

The

inspector

considers

that

for the

screening

reviews

(SRs)

assessed,

12

SEs

should

have

been

performed in

accordance

with the provisions of SDSP-27. 1.

2)

One activity which is included

in the screening

review process

is the

requirement

for screeners

to list

and

TS sections

researched

in conjunction with their

reviews.

There

are five

locations

on

each

screening

review form to list the applicable

research

documentation,

resulting

in

a total

of

65

research

citations required for the

sample of 13 SRs.

Of those

65:

Two were left blank

Four listed "ALL" as having been

reviewed (both

TS

& FSAR)

Six stated that

no

TS and/or

FSAR section applied.

It is not clear

from the inspectors

review of the

SRs that the

FSAR was properly researched

and reviewed

as required.

3)

On

the

SRs

reviewed

by the

NRC inspector,

there

were five

instances

where

the

wrong

box

("NO" instead

of "N/A") was

checked

as defined in SDSP-27. 1.

These errors were all made

in

response

to question

1 of the screening

review form:

"Does the

proposed

change

involve

a

change

in the facility (or plant

operating characteristics)

from that described

in the

FSAR and

which could

impact

nuclear

safety?"

In accordance

with the

guidelines

of Attachment

B of

SDSP-27. 1,

in

each

of these

instances

the "N/A" box should

have

been

checked

as the changes

did

not

involve

changes

to either

the facility or plant

operating characteristics.

The

deficiencies

described

above

indicated

that

SR preparers

neither

strictly

nor

consistently

adhered

to

the

requirements

of

SDSP-27. 1.

The resulting determinations

that

no safety

evaluations

were

necessary

appeared

to incorrectly bypass

the

mechanism

established

by SDSP-27.

1 to

ensure

that

proposed

changes

receive

the

appropriate

review regarding

safety impact., In summary, it is considered

that the licensee's

threshold

for performance

of safety

evaluations

based

on

the

screening

review

process

is too

high.

This

may result

in

a superficial

evaluation of

nuclear

safety

consequences.

Additionally, it is also

considered

that

the

numerous

deficiencies 'noted

in the

SRs indicate

a weakness

in the

diligence

with

which

the

screening

reviews

are

- performed.

These

concerns're

considered

to

have contributed to

a fai lure to perform

a

safety evaluation

to dete'rmine if a

US/ exists

as cited

in Violation

260/89-04"01.

r

10 CFR 50,

Appendix

B, Criterion

V requires

that activities affecting

quality shall

be

accomplished

in accordance

with documented

procedures.

The

failure

to

implement

the

requirements

of

10 CFR 50

Appendix

B,

Criterion

V and

procedure

SDSP-27. 1 for

unreviewed

safety

question

determinations,

as indicated

by the .numerous deficiencies

iden-

tified by the

NRC inspectors

in the screening

reviews of the fuel loading

procedures,

was identified as Violation 259,260,296/89-04-02.

All of the

SRs

assessed

were

approved

by personnel

who were officially

designated

as

"Approvers."

Each

of these

personnel

had

successfully

completed

an

eight-hour training

course

in the

USED process

and

were

current in their

required

annual

requalification training.

Also, review

of the

USED training material

determined

that the required

information

from SDSP-27. 1,

10 CFR 50 '9,

and

other

NRC

and

industry

guidance

is

presented

during training courses for SR Preparers

and Approvers.

Per

SOSP-27. 1, the Approvers are charged with the responsibility to review

the

responses

to the

SR questions

and the associated

justifications for

technical

adequacy,

and to indicate their approval of the.SRs.

It appears

that the

standard

of technical.

adequacy

enforced

by 'the first level of

supervisory

review has not been sufficient.

Technical Specification

and

Bases

Adequacy

The inspectors

reviewed the technical

adequacy

of the specific

TS sections

used

by the licensee

as

a basis for conducting

unmonitored fuel'oading.

Technical

Specification

3. 10.B. 1 required that during core alterations,

other

than

a

complete

core

removal,

two

SRNs shall

be operable

in or

i

adjacent to any quadrant

where fuel or control rods are being moved.

For

an

SRM to be considered

operable,

the following shall

be satisfied:

TS 3.10.B.1.b.l:

The

SRNs shall

have

a

minimum of

3 cps with all

rods fully inserted

in the core, if one or more fuel

assemblies

are

in the core, or

'S 3. 10.B. l.b.2:

During

a full core reload where both irradiated

and

fresh fuel is being loaded,

SRMs (FLCs)

may have

a count rate of less

than

3 cps provided that the

SRMs are

response

checked at least

once

every

8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months with a neutron

source until- greater

than

3 cps

can

be

maintained,

and provided also that the

core is loaded

in

a spiral

sequence

only, or

TS 3. 10.B. l.b.3:

During

a full core reload where both irradiated

and

fresh fuel are being loaded,

four irradiated fuel assemblies will be

placed adjacent to each

SRN to establish

a count rate of greater

than

3 cps,

provided

each

SRM is functionally tested prior to adjacent

fuel loading,

a neutron

response

is observed

as the adjacent fuel is

loaded,

and the core is loaded

in

a .spiral

sequence

only after the

SRM adjacent

fuel loading.

The provisions of TS Sections

3. 10.B. 1.b.2

and 3. 10.B. 1.b.3 could result

in fuel

loading

sequences

without continuous

monitoring of reactivity

changes

because

of the geometry of the.SRM locations.

As fuel is loaded

in

a spiral pattern

from the center of the core,

the flux is not initially

neutronically

coupled with the

SRM locations

due to attenuation

between

the

fuel

and

the

detectors.

In this

situation,

the

are

not

monitoring

core

neutrons.

Although the provision

of

TS

3. 10.8. l.b.3

should maintain

a

3 cps minimum count rate, this only provides continuous

demonstration

of

SRM operability

and

does

not ensure

monitoring of core

neutrons.

The licensee

had not performed

a safety analysis

to justify

that the

SRNs would promptly respond to

a criticality event if continuous

monitoring was not maintained,

Both the

TS

Basis

and

the

NRC

1979

SER

indicated

the

necessity

for

continuous

core monitoring.

The

TS Basis for Section

3. 10 states

that the

SRNs

are

provided for monitoring

and for guidance of the operator during

refueling operations.

The

TS Basis further states

that

3 cps

on the

is required

to ensure

that the flux is being monitored.

The

SER states

that

one function of the minimum count rate requirements

in the

TS is to

provide assurance

that the

SRN detectors

are close

enough to the array of

fuel assemblies

to monitor core neutron flux levels.

Both

TS 3. 10.8. l.b.2 and

TS 3. 10.B. l.b.3 are only applicable

when the fuel

being

loaded

contains

irradiated

assemblies.

The

SER indicates that the

relaxation of the

3 cps

minimum count rate

applies

only when irradiated

fuel is being

loaded

because

the

neutrons

from spontaneous

fission

and

fission product decay,

and photoneutrons

provide

a minimum neutron flux to

demonstrate

SRM response.

Section

3. 1 of the

SER states

that the presence

of exposed

fuel will ensur'e

the required

minimum flux levels required for

'

10

monitoring.

The

SER further states

that the .loading of a core containing

only fresh'fuel

must

use

lumped

neutron

sources

and

FLCs to

meet

the

normal

3 cps

minimum count rate.

The inspectors

consider that the intent

of the

TS 3. 10.B. 1.b.2

and

TS 3. 10.B. l.b.3

requirements

for irradiated

fuel relates

to an assumed

minimum flux level from the presence

of photo-

neutrons

even

though

a minimum flux level

is

not specified.

Although

the

previously

exposed

fuel

assemblies

being

loaded

in Unit

2 were

"irradiated", the radiation levels

had

decayed

during the

extended

shut-

down

such that the assemblies

should

have

been

considered

equivalent to

new fuel with respect

to neutron levels.

The inspector

concluded

that

loading of fuel,

under

the existing conditions,

would appear

to preclude

the

use of TS 3. 10.B. l.b.2 and

TS 3. 10.B. 1.b.3.

Based

on the assumption

of loading irradiated fuel of sufficient activity

level

to

meet

minimum flux requirements

for monitoring,

the

TS Basis

states

that

a large

number

of fuel assemblies

will not

be required

to

maintain

3 cps'he

licensee's

plan

to

load

200

assemblies

before

achieving

the

specified

minimum count

rate

appears

to contradict this

Basis.

Originally, the

TSs required

fuel loadings to

be monitored with

SRMs or

FLCs reading greater

than

3 cps.

FLCs -were to be used during .fuel loading

until

3 cps could

be achieved

on the

SRMs.

In 1975,

the

TS were changed to allow for a full core unload with SRM or

FLC count rates

of less

than

3 cps.

Also, the requirement

to

have

an

operable

neutron monitor in the quadrant in which fuel was being

moved and

in the adjacent

quadrant

was inadvertently

changed to require

an operable

monitor in the quadrant

of fuel

movement

or the adjacent

quadrant.

Thus

an operable

monitor was

no longer required in the quadrant

in which fuel

was actually being loaded.

In 1979,

TSs were

changed

to allow a full core reload with less

than

3 cps

if the core is loaded in a spiral

sequence.

This loading

method did not

provide continuous

core monitoring.

TVA based their safety evaluation

on

the fact that the

NRC had

approved

a simi.lar change

for Nine Mile Point

Unit 1.

Records

indicated that the

BFNP submission

simply referenced

the

NMP Unit

1

SER without describing

any

differences

between

the

two

reactors

and the required procedures

and process

in carrying out the fuel

loading

and core monitoring.

The

inspectors

considered

the

licensee

safety

analysis

supporting

the

1979

TS

amendment

to

be

inadequate

in that it did not provide

any

minimum count rate.

If relief from the

3 cps

minimum was justifiable,

a

new

minimum

should

have

been

established

which would'e

based

on

signal to noise ratio parameters.

Written relief from the requirement for

3 cps

was obtained,

but

no

new

minimum count rate requirement

was esta-

blished.

This did not relieve

the

licensee

of the responsibility for

providing

adequate

procedures

including specifying

a

minimum acceptable

11

count rate.

Although a count rate of zero literally complies with the

TS

requirements, it is considered

unacceptable

from

a conservative

safety

perspective.

In 1984,

TS were changed to allow loading irradiated fuel around

each

to obtain

a

3

cps

minimum count

rate with

a spiral

loading pattern.

Again, this loading method did not provide adequate

core monitoring.

All three of the

TS revisions singularly and collectively are considered

to

be

non-conservative

and

appear

at variance

with fundamental

core

physics

requirements

to

monitor

core

neutron

population

as

positive

reactivity is added or as core alterations

are

made.

On

January

6,

1989,

following the

termination

of fuel

loading,

the

licensee's

Plant

Manager

informed the

NRC inspectors

that

the

licensee

safety

analysis

supporting

the

1979

TS

amendment

which allowed

core

loading without a minimum

SRM count rate

level of

3 cps

was

inadequate.

The

Plant

Manager

also

acknowledged

that

inadequate

safety evaluations

were provided to the

NRC for the

TS submittal in 1984,

and that management

should

have provided

an in-depth review of the adequacy

of the evaluations

for both the

1979

and

1984 submittals.

The inspectors

questioned

the validity of TSs which allow unmonitored core

alterations.

This is identified

as

Unresolved

Item 259,260,296/89-04-04

pending completion of licensee

aetio'n to generically

review

TSs for this

.

type of concern

and further

NRC review.

6.

Fuel

Load Procedure

Review, Approval,

and Adequacy

The

NRC inspectors

reviewed the following procedures

which controlled the

fuel loading process:

2-GOI-100-3, "Refueling Operations"

TI-147, "Fuel Loading After a Complete

Core Unload"

2-SI-4. 10.8,

"Oemonstration of Source

Range Operability"

TI-14, "Special

Nuclear Material Control"

The following conclusions

were reached

and discussed

with the licensee:

A minimum acceptable

count rate

was

not mentioned,

directly or by

reference,

in the procedure.

b.

The concept of detector

response

as

opposed

to detector operability

was

not adequately

addressed

in the

fuel

loading

procedures

(see

paragraph

3).

Licensee

insensitivity

to

the

requirement

for

and

benefit

of

continuous monitoring was reflected in TI-147, "Fuel

Loading After A

Complete

Core Unload",

step 4.2.8,

which stated

that fuel

movement

12

should

be halted if two or more

SRM readings

double after loading

a

single

fuel

assembly

provided that three

out of four

were

reading

greater

than

3.0

cps without the

Response

Check

Neutron

Source,

prior to loading that bundle.

The

NRC inspectors

did not

consider

this

a

meaningful

precaution

when

the

were

not

continuously

responding

to

core

neutrons.

A large portion of the

core,

many critical masses,

could have

been

loaded

before

the

3 cps

count rate

was achieved.

d.

A primary purpose

of the spiral

loading pattern

being

used

by the

licensee

was

to

ensure

that

no

control

cell

size

(four fuel

assemblies)

flux traps were created

in the loading

sequence.

A flux

trap is

a region of high flux created

in

an unfilled fuel

assembly

location which is surrounded

by fuel.

A fuel assembly

inserted into

a flux trap could

have especially

high reactivity worth.

Although

cell-size flux traps did not occur

in the loading

sequence,

the

NRC

inspectors

determined that fuel

assembly

size flux traps did occur.

This is

due to the

need

to

use

control

rod blade guides during the

loading

sequ'ence.

In the first

36

steps

of the

loading, sequence

'there were two instances

of fuel assemblies

being added to

a position

that was completely surrounded

by fuel.

The

NRC inspectors

'reviewed

the loading error analysis

and

concluded

that

a fuel-assembly

size

flux trap

was within the

bounding

analysis..

The insertion

of

a

single fuel assembly

into a flux trap region of high worth should not

result

in criticality.

Therefore

the

assembly

size

flux traps

allowed by the loading

sequence

did not present

a safety hazard.

NRC

inspectors

observing

the

count

rate

meters

during fuel

loading

noted that

count

rates

on

the

order of

100

cps

occurred

whenever

a fire

alarm

or

medical

emergency

code call

occurred.

Although neutron monitoring capability was lost during these

alarms,

the

procedures

contained

no precaution

to stop fuel

movement

when

this situation occurred.

An. NRC inspector

assessed

the operability of the

by observing

the

response

of the count rate

meters

in the control

room when the

SRMs were checked,

with eight hour frequency,

using

a fixed neutron

source

in

an adjacent

core position.

The

response

was surprisingly

slow,

but

stable

once

complete.

The

inspector

also

checked

the

response

of two channels

with

a stop watch.

The apparent

response

time'onstants

were

34 seconds

and

25 seconds

for channels

A and 8,

respectively.

The

time constant

is the time to reach

63% of the

final

reading.

Review

of

7.5.4.2.4

indicated

that

this

performance

is expected at low count rates

but will improve at higher

- count .rates.

The

fuel

handling

procedures

did

not

include

a

requirement

to confirm that

SRM indications

have stabilized prior to

releasing

the fuel handling grapple.

The

need for such

a provision

was identified to the licensee

by the

NRC inspectors.

Of the

fuel

load

procedures

reviewed,

only TI-14,

"Special

Nuclear

Material Control", which provided the'pecific

fuel

movement

steps,

had

been

reviewed

by the

PORC.

The other procedures,

all of which contained

steps

and precautions

essential

to the safety of the loading 'process

and

13

described

cross disciplinary activities, received only Section Supervisor

and

RSPC review.

The inspectors

reviewed the

TS requirements

applicable to procedure

review

and

approval.

TS 6.8.1.1

itemizes

those

plant activities for which

written procedures

shall

be established,

implemented

and maintained,

and

includes the applicable

procedures

recommended

in Appendix A of Regulatory

Guide 1.33.

Appendix

A includes plant operating

procedures

for refueling

and

core

alterations.

TS 6.8. 1.2

requires

that

each

administrative

procedure

recommended

in Regulatory

Guide

1.33 shall

be

reviewed

by the

PORC,

and all other procedures

required

by Regulatory

Guide 1.33 shall

be

reviewed

in accordance

with

TS Section 6.5.3.

Section

6.5.3

requires

independent

review and cross-disciplinary

review when necessary.

These

TS

provisions

were, implemented

by the

licensee

through

SDSP

7.4,

"Onsite

Technical

Review

and Approval

For Procedures,"

which defines

when cross

disciplinary review is necessary,

and

PMI-7.1, "Plant Operations

Review

Committee."

These

procedures

require

PORC review for the administrative

procedures

of Regulatory Guide 1.33,

and require

a qualified independent

review for other safety-related

procedures.

Step

4.4 of SDSP 7.4 states

that cross-disciplinary

reviews

shall

be performed

whenever

any of the

following conditions apply:

Steps

in

a

procedure

may affect

equipment

under

another

g'roup's

direct control

'henever

another

group will be required to perform physical actions

not included in previously approved instructions

In

cases

where

parts

of the procedure

are outside

the reviewer's

expertise.

The

above

requirements

indicate that the fuel loading

procedures

should

have

received

cross disciplinary review.

Of particular significance

was

the

fact

that

2-GOI-100-3

did

not

receive

the

appropriate

cross-disciplinary

reviews

by RadCon,

Operations,

Industrial Safety,

PORC

Over sight,

Training,

Vendor

Manual

Coordinator,

Site

Licensing,

Instrumentation

Section,

Mechanical

Section,

and

other

relevant

disciplines.

The

NRC inspectors identified that

none of the procedures

for fuel loading

were classified during the review and approval

process

as safety-related,

despite

the

obvious

safety

implications.

The

inspector

noted

that

Regulatory

Guide 1.33 specifically designates

procedures

for refueling and

core alterations

as safety-related.

The

NRC

inspectors

considered

the

lack of

adequate

review to

be

a

significant contributing factor to the occurrence

of the unmonitored fuel

loading.

Fuel

loading

was of particular safety significance

considering

that

Unit

2

had

been

shut

down .for over

four years

due

to

poor

performance,

the majority of operators

were either newly qualified or had

not recently operated,

and

the condition of the fuel after the extended

shutdown differed from

a typical refueling.

Procedures

to conduct

core

loading should

have

been

given the highest level of review.

Failure to provide cross disciplinary review as required

by TS 6.5.3

and

administrative

procedure

SDSP 7.4,

was identified as

apparent

Violation

259,260,296/89-04-03.

The

NRC inspectors further noted that

an issue involving inadequate

review

and approval

of procedures

had recently

been

raised

in

NRC Inspection

Report

259,260,296/88-36,

but the - licensee's

corrective

actions

focused

only on addressing

the specific procedure

questioned

by the

NRC.

A review

of

TS 6.5. 1

concerning

PORC activities

revealed

an

ambiguously

worded

specification that provided only for

PORC

review of administrative

pro-

cedures

and

emergency

operating

procedures

and did not appear to address

either

PORC overview of potential

unreviewed

safety

questions

associated

with procedures

or

PORC overview of the implementation of the independent

qualified reviewer process.

The adequacy

of the procedure

review process

including the responsibilities

of the

PORC for procedure

review is identi-

fied as Unresolved

Item 259,260,296/89-04-05.

on

SRM Monitoring During Refueling

The inspectors

reviewed

findings in the

area

of

core monitoring during fuel

movement to assess

whether the

licensee

had

to identify

and

evaluate

the

adequacy

of core

monitoring.

NRC

Inspection

Reports

259,260,296/85-43

and

85-44

documented

NRC

concerns

regarding

TS requi rements

for.

SRM count rates during Unit

1 core

unloading.

The specific

issue

involved

an apparent conflict between

TS

3. 10.B. 1, which required

a

minimum

3 cps

count rate for

SRM operability

except

during certain

specific

reloading -conditions,

and

TS

3. 10.B.2,

which allowed the

SRM count rate to drop

below

3 cps during

a complete

core removal.

Although the concern

was identified during

an inspection of

core unloading,

a key concept is that core monitoring is required

as long

as fuel is in the core.

Additionally, the concern brought the ambiguity

of the

TS to the attention of the licensee.

The licensee

committed to

a

reevaluation

of the operability requi rements

in TS 3. 10 for the

SRMs,

and

made

an interim procedure

change

to leave fuel around the

SRMs to maintain

a

minimum count

rate

indication.

This previous

inspection

finding was

identified as IFI 259/85-44-02.

The

same

concern

surfaced

again

during

Unit

3

fuel

unloading,

as

documented

in

NRC Inspection

Report

259,260,296/87-09.

The

inspection

report

again

raised

the

issue

of adequate

monitoring

and reiterated

the

licensee's

commitment to evaluate

the

adequacy

of TS 3. 10.

At that time

the inspectors

questioned

the

adequacy

of

management

oversight

because

this was the

second

incidence of the

same

concern.

Since

the

items

addressed

concerns

with defueling,

they

were inappro-

priately classified

as not affecting fuel load or startup.

Therefore,

the

adequacy

of

TS 3. 10 with respect

to

SRM .operability requirements

had not

been formally evaluated

by the licensee.

4s

15

8.

Other Opportunities

To Identify Unmonitored

Fuel

Loading

As previously discussed,

the licensee

should

have identified the safety

,issue

of unmonitored

core

loading

through

the

performance

of adequate

reviews,

through the procedure

review and approval

process

required

by TS,

through

reference

to the

TS Bases

and/or

SER, or through

adequate

response

to previous

NRC inspection findings.

In addition, other

specific

opportunities

for the

licensee

to identify and correct

the

problem

had also existed.

a.

between

the licensee

and

GE should

have led

to earlier identification of the problem.

In 1987,

as part of the

design

process,

the licensee

began

discussions

with

on

neutron

source

requirements

for

fuel

loading.

Licensee

engineers

were

concerned

that

TS 3.3.B.4, which requires

greater

than

3 cps

on the

SRMs prior to pulling rods to go critical, could not be met because

of the effects of the

long

shutdown

on the fuel.

In May 1987,

recommended

the

use of startup

sources

and

FLCs for fuel loading.

Based

on reference

to the TS, the licensee

did not believe that

FLCs

would

be required

and declined the recommendation

of GE.

GE further

recommended

a change to TS to allow a reduction

in the required

cps

and spiral

loading

around

an

SRM.

The licensee

also rejected this

proposal

as unnecessary

based

on the wording of the existing TS.

.b.

On

December

16,

1988,

GE issued

Nuclear Services

Information Letter

(SIL) No. 478,

"SRM Minimum Count Rate"', which stated that owners of

BWRs which have not operated for an extended

period find that the

signal

is less

than following briefer outages.

The SIL raised

the

concept

of

adequate

core

monitoring.

In particular,

the

SIL

addressed

the

need for establishing

a minimum count rate limit.

The

SIL also

stated

that

SRM monitoring of neutrons

requires

a

minimum

count rate of 0.7 with

a signal

to noise ratio of 20 to 1.

During

the fuel load at

BFNP even

C did not maintain

a count rate level

of 0.7.

The

inspectors

also

observed

that

the

licensee

did not

appear to have

a formal process

to ensure that

GE SILs are adequately

addressed

in a timely manner.

C.

Prior to fuel loading,

the licensee

conducted training for nuclear

engineers,

GE field engineers,

gA personnel,

and operations

personnel

who

would

be

involved

in

the

loading activities.

The training

focused primarily on the actions

to

be taken

and did not cover the

technical

bases

for these

actions.

The licensee

reported that

field engineers

did informally question

the lack of core monitoring,

but were convinced

by TVA personnel

that the planned methodology

was

acceptable

because

of the wording of the approved

TS.

Adequate

engineering

design

review should also

have identified the problem

of unmonitored

core

loading.

The

licensee's

Nuclear

Fuels

Department

performed the core design

and safety analyses,

and developed

the full core

loading

pattern

used

as

a

basis

for the detailed

loading

sequence.

Nuclear

Fuels

reviewed

the

fuel

assembly

transfer

forms for technical

adequacy

and concurred that they were acceptable

to safely load fuel.

The

16

safety

analyses

per formed for the core

were. directed

toward confirming

that the core for Cycle

6 would operate

within the thermal limits of TS

and the bounds of the analyzed accidents

in the

FSAR.

The safety of the

refueling

operation

was

not

addressed

by the

licensee's

Nuclear

Fuels

Department

in any of the design

documents

reviewed

by the

NRC inspectors.

The

inspectors

considered it significant that

the

licensee

made

a

conscious

decision to apply

TS 3. 10.B. l.b.2 to load fuel with less than

3

cps

based

on loading irradiated fuel.

Licensee

engineers

made

statements

to the inspectors

on several

occasions

which demonstrated full awareness

that the irradiated

fuel being

loaded

was the equivalent of fresh fuel

with respect

to

neutron

levels.

In fact,

licensee

engineers

indicated

that

a decision

was

made

not to apply

TS

3. 10.B. l.b.3,

and

load four

irradiated

assemblies

around =each

SRM to achieve

a

3 cps

SRM count rate,

because

the

licensee

did not believe

that

the

neutron

levels of the

irradiated assemblies

was sufficient to comply with the TS.

The inspector

considered

that the licensee

took nonconservative

and

improper

advantage

'of the

TS wording.

g

emphasize

compliance rather than safe

9.

Licensee

Immediate

Corr ective Actions

In each of the various opportunities to have identified and corrected this

basic safety concern,

the licensee

seemed

to accept

without question

the

provisions of the

TSs which allowed

unmonitored

core alterations.

As

a

conse

uence

this

ives indication of a general attitude which appears

to

ty.

Upon termination of fuel loading

on January

5,

1989,

the licensee

took the

following immediate corrective-actions

to assure

adequate

shutdown

margin

would be maintained:

Licensee

nuclear

engineers

analyzed

the

shutdown

margin of the

74

fuel

assembly

configuration,

conservatively

approximated

in

the

calculations

by

a symmetric pattern of 76 assemblies.

The calculated

SDM with 76 assemblies

correctly

loaded

and all control

rods fully

inserted

was

7.59

% delta k/k, which represented

a very substantial

safety

margin.

The calculated

SDM with

76

assemblies

correctly

loaded

and

a single

stuck rod withdrawn was 2.65

% delta k/k, which

remained substantially

more conservative

than .the

SDM of 0.38

% delta

k/k required

by TS.

The

SDM remained

safely above the

TS limit even

with three

worst-case

loading errors

assumed

and

one

stuck

rod

withdrawn.

The calculations

predicted criticality would occur only

with the

assumption

of four worst

case

loading errors

plus

a stuck

rod.

The

NRC inspectors

concluded that

adequate

core safety margin

had been analytically demonstrated.

'b.

A core verification was performed to confirm that the core

was loaded

as

designed

and analyzed.

No discrepancies

were identified.

An

NRC

inspector

reviewed

the

licensee's

process

for administratively

assuring that the core

was loaded in accordance

with the design.

The

following documents

were reviewed in the assessment:

17

TVA Memorandum, Verification of Browns Ferry Nuclear Plant Unit

2

(BFN2),

Cycle

6

Core

Loading

Fuel

Assembly

Transfer

Forms

(FATFs) BFN-2-21,

November 28,

1988.

TVA Memorandum,

Unit 2 Cycle

6 Final

Core Loading Pattern,

November

17,

1988.

BCD-3&5 (Revision 2/ July 21,

1988),

Fuel Cycle Report,

Volume

I,

Fuel

Cycle Design,

for Browns Ferry Nuclear Plant Unit 2,

Cycle 6.

SAS-366

(Revision 0/ July 1988),

Fuel

Cycle Report,

Volume II,

for Browns Ferry Nuclear:Plant

Unit 2,

Cycle 6, Trans'ient

and

'ccident Analysis (Reconstituted

Core).

TVA-RLR-002 (Revision 2/ July 1988),

Reload

Licensing

Report,

Browns Ferry Nuclear Plant Unit 2, Cycle 6.

TI-14 (Revision 8),

Special

Nuclear Material Control,

Appendix

B, Fuel

Assembly Transfer

Form.

The

inspector

concluded

that

there

was

a

continuous trail

of

documentation

of the licensee's

activities from the outset of

core design to the designation

of each fuel assembly

by unique serial

number to be installed into each

unique core location.

The inspector

.

concluded

that

the

core

was

loaded

in the configuration that

was

designed

and analyzed.

c.

Control

rod

movement

was

inhibited 'by

removing

power

to

the

directional control valves.

d.

D,

which

was

reading

zero

cps in its original installed

core

position,

was

moved to

IRM F location near the center of the core to

provide flux monitoring and trip capability.

A stable

count rate of

35-40 cps

was obtained.

e.

Source

range monitor operability

checks

continued to be performed at

least

every eight

hours

per 2-SI-4. 10.B,

"Demonstration

of Source

Range Monitor System Operability During Core Alterations."

The

NRC

inspectors

reviewed

these

immediate

corrective

actions

and

considered

them

adequate

to maintain

core

safety for the as-terminated

core configuration.

Subsequent

to the

termination

of fuel

loading,

the

licensee initially

persisted

in maintaining that

one

responding

SRM would

be sufficient to

support

continued

fuel

loading,

and

was

supported

by

in this

contention.

The

NRC inspectors

considered

that this proposal

was contrary

to

TS requirements,

as well

as

safe

engineering

practices.

Only after

additional

consideration

and extensive

interaction with the

NRC

and

GE,

was

the decision

made

by the

licensee

to

use

FLCs to provide

redundant

e

18

core

monitor ing

in

each

quadrant, in which fuel

was

being

moved,

as

required

by TS.

Prior to

resuming

fuel

loading,

the

following additional

corrective

actions

were taken

by the licensee:

Redundancy

in neutron

monitoring

was established

through the use of

two FLCs in place of SRM's

A and

B.

Core

loading

and

support

procedures

were appropriately

revised

and

reviewed

by the

PORC.

A review of the revised

procedures

by

NRC

inspectors

reflected that fuel loading procedural

concerns

had

been

adequately

resolved.

Obs'er,vations

by

NRC inspectors

of the

PORC

meetings

indicated that the meetings

were satisfactorily conducted.

Training of refueling

personnel

was

conducted,

including

a critique

of the

event

and training

on

the

procedure

revisions.

The

NRC

inspectors

attended

a

number

of

these

training

sessions,

and

concluded

that

the

licensee

provided

a

good critique

and

lessons

learned

session

with all personnel

involved with fuel

handling

and

provided

a good overview of revisions to the procedures.

Once

the licensee

acknowledged

the safety significance, of the

issue of

unmonitored

core

loading,

the

immediate

corrective

actions

taken

were

acceptable.

10.

Reportabi lity

Licensee

Reportable

Event

Determination

89-2-004,

issued

on

January

6,

1989, classified the termination of fuel loading due to lack of monitoring

as

non-reportable

per

10 CFR 50.72,

50.73,

or plant

implementing

procedures.

The basis

for this

assessment

was that all applicable

TS

requirements

were satisfied

throughout

fuel loading,

and that the plant

had

been

analyzed

for the

fuel

loading

method

as

described

in

TS

3. 10.B. l.b.2.

The

NRC inspectors

consider

this event to be reportable

under

10 CFR 50.73

and that there

may also

be basis for reportabi lity

under

10 CFR 50.72.

The

adequacy

of the repor tabi lity determination for

this event

was identified as

Unresolved

Item 260/89-04-06

pending

review

of the licensee's

basis of not reporting it under

10 CFR 50.72

and pending

licensee disposition pursuant to

10 CFR 50.73.

ll.

Exit Interview (30703)

0

The

inspection

scope

and findings were

summarized

on January

13,

1989,

with

those

persons

indicated

in

paragraph

1

above.

The

inspectors

described

the areas

inspected

and discussed

the inspection findings listed

below.

Proprietary

material

was

reviewed

by the inspectors

but was not

retained.

Dissenting

comments

were not received

from the licensee.

Inspection

Findings:

\\

Apparent

Violation 260/89-04-01:

Fai lure to Obtain

NRC Approval

Prior to Proceeding

with Unmonitored

Core

Loading of Unit 2, which

Constituted

a Potential

Unreviewed Safety Question

and

Compromised

Fundamental

Safety Principles (paragraph

4.a)

Apparent

Violation 259,260,296/89-04-02:

Programmatic

Failure

to

Implement the

Requirements

of

and

Procedure

SDSP-27.

1

Unreviewed

Safety Question

Determinations,

as

Evidenced

by Numerous

Inadequacies

in the

Reviews of Fuel

Loading

Procedures

(paragraph

4.b)

Apparent

Violation 259,260,296/89-04-03:

Failure

to Provide

Cross

Disciplinary Review of Procedures

Impacting Plant Safety

(paragraph

6)

Unresolved

Item

259,260,296/89-04-04:

Review

of

TS

Requirements

(paragraph

5)

Unresolved

Item

259,260,296/89-04-05:

Procedure

Review

Process

Adequacy (paragraph

6)

Unresolved

Item

259,260,296/89-04-06:

Adequacy

of the

Licensee

.

Determination

that

Loading

Fuel without

Core

Monitoring

was

not

reportable

per

10 CFR 50.72 or 50.73 (paragraph

10).

12.

List of Acronyms

BFNP

CFR

GOI

IFI

NRC

PMI

RCNS

RSPC

SAR

SDM

SDSP

Browns Ferry Nuclear

Power Plant

Boiling Water Reactor

Code of Federal

Regulations

Counts

Per

Second

Enforcement Action

Fuel Assembly Transfer

Form

Fuel

Loading Chamber

Final Safety Analysis Report

General Electric

General

Operating Instruction

Inspector

Followup Item

Intermediate

Range Monitor

Nine Mile Point

Nuclear Regulatory

Commission

(NRC Office of) Nuclear Reactor Regulation

Plant Manager Instruction

Plant Operations

Review Committee

Quality Assurance

Response

Check Neutron Source

Responsible

Section

Procedure

Coordinator

Safety Analysis Report

Shutdown Margin

Site Director Standard

Pract'ice

20

SR

TI

TS

TVA

USQ

USQD

Safety Evaluation

Report

Surveillance'nstruction

(General .Electric) Service Information Letter

Screening

Review

Source

Range Monitor

Technical Instruction

Technical Specifications

Tennessee

Valley Authority

Unreviewed Safety Question

Unreviewed Safety Question Determination

ATTACHMENT A

SLIDES

FROM TVA PRESENTATION

TVA/NRC MANAGEMENTMEETING " JANUARY 9, 1989

AGENDA

~OP C

I.

= INTRODUCT ION

SPEAKER

O. KINGSLEY, JR

II.

'BFN RELOAD ASSESSMENT

~ HISTORY - TECH SPECS

4 LOADINQ SEQUENCE J. BYNUM

~ ROOT CAUSE

~ OE SUPPORT

P. MARRIOTT

~ CORE REACTIVITY

J.BYN UM

III.

RESUMING FUEL LOAD

~ RELOAD PLAN AND CORE MONITORING

~ RELOAD SEQUENCE / QE TECH OVERVIEW

I3. CAMPBELL

P. MARRIOTT

IV. ADDITIONALTVA INITIATIVES

~ SHORT TERM TECH SPEC ASSESSMENT

~ PLANNED TECH SPEC ASSESSMENT

~ SECTION 6 TECH SPEC

~ EXPERIENCE REVIEW

~ VENDOR INTERFACE

CAMPSKLL/OYERLID

N. KAZANAS

P. CARIER

C. FOX

C. MASON

V.

SUMMARY

O. KINI3SLEY, JR

I L AOEND.CKT

TVA/NRC MANAGEMENTMEETING

JANUARY 9, 1989

INTRODUCTORY REMARKS

BY O. D. KINGSLEY, JR.

PL~INTRO.CHT

pe/ERG MANAGEMENTM~~~~~~

JANOARY 9, 1989

t

~FN RELOAo assessors~

BY J. R. BYNUM

FL ASMT.CHT

4

l

TECH SPEC HlSTORY

/

~ 1S73 - TWO SRM's OR FLC's

ONE IN QUADRANT BEING LOADED

AND ONE IN ADJACENT QUADRANT

~ 1976 - TWO SRM'e OR FLC's ONE IN

QUADRANT BEING LOADED~ONE

IN ADJACENT QUADRANT, LESS

T.HAN 3 CPS ALLOWED DURING FULL

CORE UNLOAD

1979 - TWO SRM's OR FLC's, LESS THAN

3 CPS ALLOWED IF FULL CORE RELOAD

AND CORE LOADED IN SPIRAL

SEQUENCE ONLY

~ 1984 - TWO SRM's OR FLC's, FOUR ASSEMBLIES

MAY BE LOADED ADJACENT TO SRMS TO

ESTABLISH A GREATER THAN 3 CPS AND

FUEL LOADED IN SPIRAL SEQUENCE

AF.TER ADJACENT. FUEL LOADING

FL TS1.CHT

CORE LOADING SEQUENCE HISTORY

~ LOADED USING SRM's AND FUEL

LOADING CHAMBERS

~ MAINTAINEDGREATER THAN 3 CPS

~ U3 CYCLE 6 - USED PRESENT

METHOD

r

FL CLH.CHT

h

RELOAD SEQUENCE DEVELOPMENT