Entergy Operations, Inc.; Consideration of Issuance of Amendment to Facility Operating License, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing

NUCLEAR REGULATORY COMMISSION

[Docket No. 50-458]

Entergy Operations, Inc.; Consideration of Issuance of Amendment

to Facility Operating License, Proposed No Significant Hazards

Consideration Determination, and Opportunity for a Hearing

The U.S. Nuclear Regulatory Commission (the Commission) is

considering issuance of an amendment to Facility Operating License No.

NPF-47 issued to Entergy Operations, Inc. (the licensee) for operation

of the River Bend Station, Unit 1, located in West Feliciana Parish,

Louisiana.

The proposed amendment would revise various instrumentation

technical specifications by extending the allowable outage times (AOTs)

of the instruments, and by increasing their channel functional

surveillance test intervals (STIs) to quarterly. The amendment also

revises certain technical specification actions to address loss-of-

function concerns associated with the AOT and STI changes.

Before issuance of the proposed license amendment, the Commission

will have made findings required by the Atomic Energy Act of 1954, as

amended (the Act) and the Commission's regulations.

The Commission has made a proposed determination that the amendment

request involves no significant hazards consideration. Under the

Commission's regulations in 10 CFR 50.92, this means that operation of

the facility in accordance with the proposed amendment would not (1)

involve a significant increase in the probability or consequences of an

accident previously evaluated; or (2) create the possibility of a new

or different kind of accident from any accident previously evaluated;

or (3) involve a significant reduction in a margin of safety. As

required by 10 CFR 50.91(a), the licensee has provided its analysis of

the issue of no significant hazards consideration, which is presented

below:

Reactor Protection System (RPS)

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

These proposed changes do not involve a change to the plant

design or operation, they simply involve the frequency at which

testing of the RPS instrumentation is performed and the allowable

outage time (AOT) for instruments. Failure of the RPS

instrumentation itself cannot create an accident. As a result, these

proposed changes cannot increase the probability of occurrence of

any design basis accident previously evaluated.

As identified in NEDC-30851P, these proposed changes increase

the average RPS failure frequency from 4.6 x 10-6/year to

5.4 x 10-6/year. This increase (8 x 10-7/year) is

considered to be insignificant. As identified in the NRC Staff's

Safety Evaluation Report of NEDC-30851P, this increase in average

RPS failure frequency would contribute to a very small increase in

core-melt frequency. The small increase in average RPS failure

frequency is offset by safety benefits such as a reduction in the

number of inadvertent test-induced scrams, a reduction in wear due

to excessive equipment test cycling, and better optimization of

plant personnel resources. Hence, the net change in risk resulting

from these proposed changes would be insignificant. Therefore, these

proposed changes do not result in a significant increase in either

the probability or the consequences of any accident previously

evaluated.

2. The proposed change does not create the possibility of a new

or different kind of accident from any previously evaluated.

The proposed changes do not result in any change to the plant

design or operation, only to the AOT and frequency at which testing

of the RPS instrumentation is performed. Since failure of the RPS

instrumentation itself cannot create an accident, these proposed

changes can at most affect only accidents which have been previously

evaluated. Therefore, these proposed changes cannot create the

possibility of a new or different kind of accident from any accident

previously evaluated.

3. The proposed change does not involve a significant reduction

in the margin of safety.

As identified above, these proposed changes increase the average

RPS failure frequency from 4.6x10-6/year to 5.4x10-6/year.

The NRC Staff's Safety Evaluation Report of NEDC-30851P concluded

that this small average RPS failure frequency increase would

contribute to a very small increase in core-melt frequency. This

small increase in average RPS failure frequency would be offset by

safety benefits such as a reduction in the number of inadvertent

test-induced scrams, a reduction on wear due to excessive equipment

test cycling, and better optimization of plant personnel resources.

Hence, the net change in risk resulting from these proposed changes

would be insignificant. In addition, RBS has confirmed that the

proposed changes to the functional test intervals will not result in

excessive instrument drift relative to the current established

setpoints. Therefore, these proposed changes do not result in a

significant reduction in a margin of safety.

Emergency Core Cooling System (ECCS)

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

These proposed changes do not involve a change to the plant

design or operation, they simply involve the frequency at which

testing of the ECCS actuation Instrumentation is performed and the

allowable outage time (AOT) for instruments. Failure of the ECCS

actuation instrumentation itself cannot create an accident. As a

result, these proposed changes cannot increase the probability of

occurrence of any design basis accident previously evaluated.

As identified in NEDC-30936P (Part 2), these proposed changes

increase the calculated average water injection failure frequency

from 1.952x10-5 to 1.992x10-5 per year for Case 5B and

from 1.386x10-4 to 1.401x10-4 per year for Case 5C. This

represents an increase of 4x10-7 for Case 5B (2.0%) and

1.5x10-6 for Case 5C (1.1%), which are well within the

acceptance criteria (4%) provided in NEDC-30936P (Part 2). The small

increase in average water injection failure frequency is offset by

safety benefits such as a reduction in the number of inadvertent

test-induced scrams, a reduction in wear due to excessive equipment

test cycling, and better optimization of plant personnel resources.

Hence, the net change in risk resulting from these proposed changes

would be insignificant. Therefore, these proposed changes do not

result in a significant increase in either the probability or the

consequences of any accident previously evaluated.

2. The proposed change does not create the possibility of a new

or different kind of accident from any previously evaluated.

The proposed changes do not result in any change to the plant

design or operation, only to the AOT and frequency at which testing

of the ECCS actuation instrumentation is performed. Since failure of

the ECCS actuation instrumentation itself cannot create an accident,

these proposed changes can at most affect only accidents which have

been previously evaluated. Therefore, these proposed changes cannot

create the possibility of a new or different kind of accident from

any accident previously evaluated.

3. The proposed change does not involve a significant reduction

in the margin of safety.

As identified above, these proposed changes increase the

calculated average water injection failure frequency from

1.952x10-5 to 1.992x10-5 per year for Case 5B and from

1.386x10-4 to 1.401x10-4 per year for Case 5C. This

increase is well within the acceptance criteria found acceptable in

the NRC Staff's Safety Evaluation Report for NEDC-30936P (Part 2).

This small increase in average ECCS actuation failure frequency

would be offset by safety benefits such as a reduction in the number

of inadvertent test-induced scrams, a reduction on wear due to

excessive equipment test cycling, and better optimization of plant

personnel resources. Hence, the net change in risk resulting from

these proposed changes would be insignificant. In addition, RBS has

confirmed that the proposed changes to the functional test intervals

will not result in excessive instrument drift relative to the

current, established setpoints. Therefore, the proposed changes do

not result in a significant reduction in a margin of safety.

Control Rod Block Instrumentation

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

These proposed changes do not involve a change to the plant

design or operation, only the Allowable Outage Time (AOT) and

frequency at which testing of the Control Rod Block Instrumentation

is performed. Failure of the Control Rod Block instrumentation

itself cannot create an accident. As a result, these proposed

changes cannot increase the probability of occurrence of any design

basis accident previously evaluated.

As identified in NEDC-30851P, Supplement 1, these proposed

changes increase the average Control Rod Block failure frequency

less than 0.06%. As provided in the NRC Staff's Safety Evaluation

Report of NEDC-30851P, Supplement 1, this increase is very slight

and is offset by the safety benefits associated with the proposed

changes to the RPS and Control Rod Block Instrumentation. As a

result, the combined effect of the changes proposed for the RPS and

Control Rod Block Instrumentation requirements should result in an

overall improvement in plant safety. Therefore, these proposed

changes do not result in a significant increase in either the

probability or the consequences of any accident previously

evaluated.

2. The proposed change does not create the possibility of a new

or different kind of accident from any previously evaluated.

The proposed changes do not result in any change to the plant

design or operation, only to the AOT and frequency at which testing

of the Control Rod Block instrumentation is performed. Since failure

of the Control Rod Block instrumentation itself cannot create an

accident, these proposed changes can at most affect only accidents

which have been previously evaluated. Therefore, these proposed

changes cannot create the possibility of a new or different kind of

accident from any accident previously evaluated.

3. The proposed change does not involve a significant reduction

in the margin of safety.

As identified above, these proposed changes increase the average

Control Rod Block failure frequency less than 0.06%. This increase

is very slight and is offset by the safety benefits associated with

the proposed changes to the RPS and Control Rod Block

Instrumentation. As a result, the combined effect of the changes

proposed for the RPS and Control Rod Block Instrumentation

requirements should result in an overall improvement in plant

safety. In addition, RBS has confirmed that the proposed changes to

the functional test intervals will not result in excessive

instrument drift relative to the current, established setpoints.

Therefore, the proposed changes do not result in a significant

reduction in a margin of safety.

Isolation Actuation Instrumentation

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

These proposed changes do not involve a change to the plant

design or operation, only the Allowable Outage Time (AOT) and

frequency at which testing of the Isolation Actuation

instrumentation is performed. Failure of the Isolation Actuation

instrumentation itself cannot create an accident. As a result, these

proposed changes cannot increase the probability of occurrence of

any design basis accident previously evaluated.

As identified in NEDC-30851P, Supplement 2, these proposed

changes to the surveillance test interval requirements for the

Isolation Actuation instruments which are common to RPS or ECCS have

a negligible (less than 1%) impact on the average isolation

unavailability when combined with the individual valve failure

probability, and that the changes to the AOTs has [have] less than a

2% impact. The analyses demonstrate that the individual valve

failure probabilities dominate the overall isolation failure

probability. As provided in the NRC Staff's Safety Evaluation Report

of NEDC-30851P, Supplement 2, these proposed changes would have a

very small impact on plant risk. As a result, overall plant safety

is not reduced by these proposed changes.

As identified in NEDC-31677P, the proposed changes to the

requirements for Isolation Actuation instrumentation not common to

RPS or ECCS result in a small decrease of 1.97x10-8/year in the

average isolation failure frequency. As identified in the NRC

Staff's Safety Evaluation Report of NEDC-31677P, the NRC agreed that

these proposed changes are acceptable because the failure frequency

impact is minimal. As a result, overall plant safety is not reduced

by these proposed changes.

The small increase in the average failure frequency of the

instruments common to RPS or ECCS due to the proposed changes to the

Isolation Actuation instrumentation requirements is offset by safety

benefits such as a reduction on the number of inadvertent test-

induced scrams and engineered safety feature actuations, a reduction

in wear due to excessive test cycling, and better optimization of

plant personnel resources. Hence, the net change in risk resulting

from these proposed changes would be insignificant. Therefore, these

proposed changes do not result in a significant increase in either

the probability or the consequences of any accident previously

evaluated.

2. The proposed change does not create the possibility of a new

or different kind of accident from any previously evaluated.

The proposed changes do not result in any change to the plant

design or operation, only to the AOT and frequency at which testing

of the Isolation Actuation instrumentation is performed. Since

failure of the Isolation Actuation instrumentation itself cannot

create an accident, these proposed changes can at most affect only

accidents which have been previously evaluated.

Therefore, these proposed changes cannot create the possibility

of a new or different kind of accident from any accident previously

evaluated.

3. The proposed change does not involve a significant reduction

in the margin of safety.

As identified above, the proposed changes to the requirements

for Isolation Actuation instruments common to RPS or ECCS have a

negligible impact on the average isolation unavailability when

combined with the individual valve failure probability. The analyses

demonstrate that the individual valve failure probabilities dominate

the overall isolation failure probability.

The proposed changes to the requirements for Isolation Actuation

instruments not common to RPS or ECCS decrease their average

isolation failure frequency approximately 1.97x10-8/year.

The small increase in average Isolation Actuation

instrumentation failure frequency of the instruments common to RPS

or ECCS are offset by the safety benefits such as a reduction on the

number of inadvertent test-induced scrams and engineered safety

feature actuations, a reduction in wear due to excessive test

cycling, and better optimization of plant personnel resources. As a

result, the NRC Staff's Safety Evaluation Reports for these BWROG

report concluded that these proposed changes would have a very small

impact on plant risk. In addition, RBS has confirmed that the

proposed changes to the functional test intervals will not result in

excessive instrument drift relative to the current, established

setpoints. Therefore, the proposed changes do not result in a

significant reduction in a margin of safety.

Other Technical Specification Instrumentation

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

These proposed changes do not involve a change to the plant

design or operation, only the Allowable Outage Time (AOT) and

frequency at which testing of the associated instrumentation is

performed. These instruments are designed to mitigate the

consequences of previously analyzed accidents. Failure of these

instruments cannot increase, and is independent of, the probability

of occurrence of such accidents. As a result, these proposed changes

cannot increase the probability of any accident previously

evaluated. As identified in GENE-770-06-01, although not

specifically analyzed, these proposed changes are bounded by the

results of the analyses discussed in Parts I through IV of this

request. Such analyses have shown that the safety function failure

frequency is not significantly impacted by similar proposed changes.

In addition, any increase in the probability of failure of these

instruments to perform their required functions would be offset by

safety benefits such as a reduction in the number of inadvertent

test-induced scrams and engineered safety features actuations, a

reduction in wear due to excessive equipment test cycling, and

better optimization of plant personnel resources. Therefore, these

proposed changes do not result in a significant increase in the

probability or the consequences of any accident previously evaluated

2. The proposed change does not create the possibility of a new

or different kind of accident from any previously evaluated.

The proposed changes do not result in any change to the plant

design or operation, only to the AOT and frequency at which testing

of the associated instrumentation is performed. As a result, these

proposed changes can at most affect only accidents which have been

previously evaluated. Therefore, these proposed changes cannot

create the possibility of a new or different kind of accident from

any accident previously evaluated.

3. The proposed change does not involve a significant reduction

in the margin of safety.

As identified in GENE-770-06-01, although not specifically

analyzed, these proposed changes are bounded by the results of the

analyses discussed in Parts I through IV of this request. Such

analyses have shown that the safety function failure frequency is

not significantly impacted by similar proposed changes. In addition,

any increase in the probability of failure of these instruments to

perform their required functions would be offset by safety benefits

such as a reduction in the number of inadvertent test-induced scrams

and engineered safety features actuations, a reduction in wear due

to excessive equipment test cycling, and better optimization of

plant personnel resources. As a result, these proposed changes will

reduce overall plant risk. In addition, RBS has confirmed that the

proposed changes to the functional test intervals will not result in

excessive instrument drift relative to the current, established

setpoints. Therefore, these proposed changes do not involve a

significant reduction in a margin of safety.

Technical Specification Changes Relating to Loss-of-Function Issues

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed changes associated with the ``loss-of-function''

checks ensure a plant configuration which would have the capability

to automatically actuate the respective system/valve(s). These

instruments are designated to mitigate the consequences of

previously analyzed accidents. Failure of these instruments cannot

increase, and is independent of, the probability of occurrence of

such accidents. As a result, the proposed changes cannot increase

the probability of any accident previously evaluated. The proposed

changes do not involve a change to the plant design or operation and

do not degrade the capability of the system(s) to perform its

required function. Further, these functions or tripped channel(s) in

an isolation logic are not considered as initiators for any

accidents previously analyzed. Therefore, these changes do not

significantly increase the consequences of any accident previously

evaluated.

2. The proposed change does not create the possibility of a new

or different kind of accident from any previously evaluated.

The proposed changes do not result in any change to the plant

design and no new mode of plant operation is introduced. As a

result, the proposed changes can at most affect only accidents which

have been previously evaluated. Therefore, the proposed changes do

not create the possibility of a new or different kind of accident

from any accident previously evaluated.

3. The proposed change does not involve a significant reduction

in the margin of safety.

The proposed changes do not involve a significant reduction in a

margin of safety since the required safety function of the

inoperable channel(s) will be fulfilled. The allowable Outage Time

(AOT) for several trip functions have been increased but only in

conjunction with the incorporation of the loss-of-function check

which ensures a plant configuration which would have the capability

to automatically actuate the respective system/valve(s). Therefore,

the proposed changes do not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

The Commission is seeking public comments on this proposed

determination. Any comments received within 30 days after the date of

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

expiration of the 30-day notice period. However, should circumstances

change during the notice period such that failure to act in a timely

way would result, for example, in derating or shutdown of the facility,

the Commission may issue the license amendment before the expiration of

the 30-day notice period, provided that its final determination is that

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received.

Should the Commission take this action, it will publish in the Federal

Register a notice of issuance and provide for opportunity for a hearing

after issuance. The Commission expects that the need to take this

action will occur very infrequently.

Written comments may be submitted by mail to the Rules Review and

Directives Branch, Division of Freedom of Information and Publications

Services, Office of Administration, U.S. Nuclear Regulatory Commission,

Washington, DC 20555, and should cite the publication date and page

number of this Federal Register notice. Written comments may also be

delivered to room 6D22, Two White Flint North, 11555 Rockville Pike,

Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays.

Copies of written comments received may be examined at the NRC Public

Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC

20555.

The filing requests for hearing and petitions for leave to

intervene is discussed below.

By May 26, 1994, the licensee may file a request for a hearing with

respect to issuance of the amendment to the subject facility operating

license and any person whose interest may be affected by this

proceeding and who wishes to participate as a party in the proceeding

must file a written request for a hearing and a petition for leave to

intervene. Requests for a hearing and a petition for leave to intervene

shall be filed in accordance with the Commission's ``Rules of Practice

for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested

persons should consult a current copy of 10 CFR 2.714 which is

available at the Commission's Public Document Room, the Gelman

Building, 2120 L Street, NW., Washington, DC 20555 and at the local

public document room located at the Government Documents Department,

Louisiana State University, Baton Rouge, Louisiana 70803. If a request

for a hearing or petition for leave to intervene is filed by the above

date, the Commission or an Atomic Safety and Licensing Board,

designated by the Commission or an Atomic Safety and Licensing Board,

designated by the Commission or by the Chairman of the Atomic Safety

and Licensing Board Panel, will rule on the request and/or petition;

and the Secretary or the designated Atomic Safety and Licensing Board

will issue a notice of hearing or an appropriate order.

As required by 10 CFR 2.714, a petition for leave to intervene

shall set forth with particularity the interest of the petitioner in

the proceeding, and how that interest may be affected by the results of

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

following factors: (1) The nature of the petitioner's right under the

Act to be made party to the proceeding; (2) the nature and extent of

the petitioner's property, financial, or other interest in the

proceeding and (3) the possible effect of any order which may be

entered in the proceeding on the petitioner's interest. The petition

should also identify the specific aspect(s) of the subject matter of

the proceeding as to which petitioner wishes to intervene. Any person

who has filed a petition for leave to intervene or who has been

admitted as a party may amend the petition without requesting leave of

the Board up to 15 days prior to the first prehearing conference

scheduled in the proceeding, but such an amended petition must satisfy

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

scheduled in the proceeding, a petitioner shall file a supplement to

the petition to intervene which must include a list of the contentions

which are sought to be litigated in the matter. Each contention must

consist of a specific statement of the issue of law or fact to be

raised or controverted. In addition, the petitioner shall provide a

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

contention and on which the petitioner intends to rely in proving the

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

aware and on which the petitioner intends to rely to establish those

facts or expert opinion. Petitioner must provide sufficient information

to show that a genuine dispute exists with the applicant on a material

issue of law or fact. Contentions shall be limited to matters within

the scope of the amendment under consideration. The contention must be

one which, if proven, would entitle the petitioner to relief. A

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

subject to any limitations in the order granting leave to intervene,

and have the opportunity to participate fully in the conduct of the

hearing, including the opportunity to present evidence and cross-

examine witnesses.

If a hearing is requested, the Commission will make a final

determination on the issue of no significant hazards consideration. The

final determination will serve to decide when the hearing is held.

If the final determination is that the amendment requests involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

request for a hearing. Any hearing held would take place after issuance

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

A request for a hearing or a petition for leave to intervene must

be filed with the Secretary of the Commission, U.S. Nuclear Regulatory

Commission, Washington, DC 20555, Attention: Docketing and Services

Branch, or may be delivered to the Commission's Public Document Room,

the Gelman Building, 2120 L Street, NW., Washington, DC 20555, by the

above date. Where petitions are filed during the last 10 days of the

notice period, it is requested that the petitioner promptly so inform

the Commission by a toll-free telephone call to Western Union at 1-

(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union

operator should be given Datagram Identification Number N1023 and the

following message addressed to William D. Beckner, Director, Project

Directorate IV-1: petitioner's name and telephone number, date petition

was mailed, plant name, and publication date and page number of the

Federal Register notice. A copy of the petition should also be sent to

the Office of the General Counsel, U.S. Nuclear Regulatory Commission,

Washington, DC 20555, and to Mark J. Wetterhahn, Esq., Winston &

Strawn, 1400 L Street, NW., Washington, DC 20005, attorney for the

licensee.

Nontimely filings of petitioners for leave to intervene, amended

petitions, supplemental petitions and/or requests for hearing will not

be entertained absent a determination by the Commission, the presiding

officer or the presiding Atomic Safety and Licensing Board that the

petition and/or request should be granted based upon a balancing of the

factors specified in 10 CFR 2.714(a)(l)(i)-(v) and 2.714(d).

For further details with respect to this action, see the

application for amendment dated January 14, 1994, which is available

for public inspection at the Commission's Public Document Room, the

Gelman Building, 2120 L Street, NW., Washington, DC 20555 and at the

local public document room located at the Government Documents

Department, Louisiana State University, Baton Rouge, Louisiana 70803.

Dated at Rockville, Maryland, this 20th day of April, 1994.

For the Nuclear Regulatory Commission.

Robert G. Schaaf,

Acting Project Manager, Project Directorate IV-1, Division of Reactor

Projects III/IV, Office of Nuclear Reactor Regulation.

[FR Doc. 94-10008 Filed 4-25-94; 8:45 am]

BILLING CODE 7590-01-M