35986 Federal Register / Vol. 63, No. 126 /Wednesday, July 1, 1998 / Notices

action does not affect nonradiological

plant effluents and has no other

nonradiological environmental impact.

Accordingly, the Commission concludes

that there are no significant

environmental impacts associated with

the proposed action.

Alternatives to the Proposed Action

Since the Commission has concluded

there are no significant environmental

impacts associated with the proposed

action, any alternatives with equal or

greater environmental impact need not

be evaluated. As an alternative to the

proposed action, the staff considered

denial of the proposed action. Denial of

the application would result in no

change in current environmental

impacts. The environmental impacts of

the proposed action and the alternative

action are similar.

Alternative Use of Resources

This action does not involve the use

of any resources not previously

considered in the Final Environmental

Statement for the FNP, ‘‘Final

Environmental Statement related to the

Operation of Joseph M. Farley Nuclear

Plant, Units 1 and 2,’’ dated December

1974 and its Addendum, NUREG–0727,

dated September 1980.

Agencies and Persons Consulted

In accordance with its stated policy,

on June 2, 1998, the staff consulted with

the Alabama State official, Mr. K.

Whatley of the Alabama Department of

Public Health, regarding the

environmental impact of the proposed

action. The State official had no

comments.

Finding of No Significant Impact

Based upon the environmental

assessment, the Commission concludes

that the proposed action will not have

a significant effect on the quality of the

human environment. Accordingly, the

Commission has determined not to

prepare an environmental impact

statement for the proposed action.

For further details with respect to the

proposed action, see the licensee’s letter

dated January 19, 1998, which is

available for public inspection at the

Commission’s Public Document Room,

The Gelman Building, 2120 L Street,

NW., Washington, DC, and at the local

public document room located at the

Houston—Love Memorial Library, 212

W. Burdeshaw Street, P.O. Box 1369,

Dothan, Alabama.

Dated at Rockville, Maryland, this 25th day

of June 1998.

For the Nuclear Regulatory Commission.

Herbert N. Berkow,

Director, Project Directorate II–2, Division of

Reactor Projects—I/II, Office of Nuclear

Reactor Regulation.

[FR Doc. 98–17488 Filed 6–30–98; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Advisory Committee on Reactor

Safeguards Subcommittee Meeting on

Advanced Reactor Designs; Revised

A two-day meeting of the ACRS

Subcommittee on Advanced Reactor

Designs scheduled to be held on

Monday and Tuesday, July 6–7, 1998,

has been changed to a one-day meeting

which will be held on Tuesday, July 7,

1998, beginning at 8:30 a.m., in Room

T–2B3, 11545 Rockville Pike, Rockville,

Maryland. Notice of this meeting was

published in the Federal Register on

Wednesday, June 17, 1998 (63 FR

33102). All other items pertaining to

this meeting remain the same as

previously published.

For further information contact, Mr.

Noel F. Dudley, cognizant ACRS staff

engineer, (telephone 301/415–6888)

between 7:30 a.m. and 4:15 p.m. (EDT).

Dated: June 25, 1998.

Sam Duraiswamy,

Chief, Nuclear Reactors Branch.

[FR Doc. 98–17487 Filed 6–30–98; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Biweekly Notice; Applications and

Amendments to Facility Operating

Licenses Involving No Significant

Hazards Considerations

I. Background

Pursuant to Public Law 97–415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 97–415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from June 8, 1998,

through June 19, 1998. The last

biweekly notice was published on June

17, 1998 (63 FR 33103).

Notice of Consideration of Issuance of

Amendments to Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a

proposed determination that the

following amendment requests involve

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. The basis for this

proposed determination for each

amendment request is shown below.

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 before

action is taken. 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 Chief, Rules and

Directives Branch, Division of

Administration Services, Office of

Administration, U.S. Nuclear Regulatory

Commission, Washington, DC 20555–

0001, and should cite the publication
Federal Register / Vol. 63, No. 126 /Wednesday, July 1, 1998 / Notices 35987

date and page number of this Federal

Register notice. Written comments may

also be delivered to Room 6D22, Two

White Flint North, 11545 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.

The filing of requests for a hearing and

petitions for leave to intervene is

discussed below.

By July 31, 1998, 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 and at the local public

document room for the particular

facility involved. 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 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 a 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 a 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 request 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 20555–0001, Attention:

Rulemakings and Adjudications Staff, or

may be delivered to the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington DC, by the above date. A

copy of the petition should also be sent

to the Office of the General Counsel,

U.S. Nuclear Regulatory Commission,

Washington, DC 20555-0001, and to the

attorney for the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

factors specified in 10 CFR

2.714(a)(1)(i)-(v) and 2.714(d).

For further details with respect to this

action, see the application for

amendment which is available for

public inspection at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, and at the local public

document room for the particular

facility involved.

Detroit Edison Company, Docket No.

50–16, Enrico Fermi Atomic Power

Plant, Unit 1, Monroe County, Michigan

Date of amendment request: January

28, 1998 (Reference NRC–98–0027)

Description of amendment request:

The proposed amendment will revise

Section F and I of the Fermi, Unit 1

Technical Specifications to include

requirements for control of effluents;

dose limits; annual reporting in

accordance with requirements of 10 CFR

50.36a; and numerical guideline criteria

based on 10 CFR 50, Appendix I. Also,

this amendment will correct several

editorial errors.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration using the standards in 10

CFR 50.92(c). The licensee’s analysis is

presented below:

(1) Does the proposed change significantly

increase the probability or consequences of

an accident previously evaluated?

No, the proposed submittal establishes

additional requirements and limits on

radioactive effluent releases. No existing

requirements are deleted. For these reasons,

this proposed change will not significantly

increase the probability or consequences of

an accident at Fermi 1.
35988 Federal Register / Vol. 63, No. 126 /Wednesday, July 1, 1998 / Notices

(2) Will the proposed amendment create

the possibility of a new or different kind of

accident from any accident previously

analyzed?

No, the addition of requirements for

radioactive effluent releases will not cause a

new kind of accident. The additional

requirements involve having a functional

waste system with procedures, submitting an

annual report, and restricting the potential

dose to the public from effluents. These

changes, in themselves, do not require a

different type of operation of systems. Any

new system installed to enable future

discharges will be evaluated at the time of

design.

(3) Will the proposed change significantly

reduce the margin of safety at the facilit y?

No, adding new requirements for

radioactive effluents will not decrease the

margin of safety. Since no existing

requirements are being eliminated, this

change will not reduce the margin of safety

of the facility.

NRC staff has reviewed the licensee’s

analysis and, based on this review, it

appears that the three standards of

50.92(c) are satisfied. Therefore, NRC

staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Monroe County Library

System, 3700 South Custer Road,

Monroe, Michigan 48161.

Attorney for licensee: John Flynn,

Esquire, Detroit Edison Company, 2000

Second Avenue, Detroit, Michigan

48226.

NRC Branch Chief: John W. N.

Hickey.

Detroit Edison Company, Docket No.

50–16, Enrico Fermi Atomic Power

Plant, Unit 1, Monroe County, Michigan

Date of amendment request: January

28, 1998 (Reference NRC–98–0025).

Description of amendment request:

The proposed amendment will revise

the Technical Specifications on access

controls to provide flexibility while

maintaining similar controls over

access. Provisions will be established

for cases where work is performed on

the Protected Area boundary, such that

the boundary temporarily will not meet

the Technical Specification criteria.

Redundancy between Technical

Specifications will be eliminated. Figure

B–1, ‘‘Facility Plan,’’ will be modified to

show the buildings within the Protected

Area, delete locations of the Protected

Area gates and doors, and delete a

building and equipment outside the

Protected Area which are planned to be

removed in the future. Finally, several

editorial corrections will be made.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration using the standards in 10

CFR 50.92(c). The licensee’s analysis is

presented below:

(1) Does the proposed change significantly

increase the probability or consequences of

an accident previously evaluated?

The proposed changes do not involve a

significant increase in the probability or

consequences of an accident. The proposed

changes all involve access control, the

Protected Area boundary, or deletion of

details from a sketch, including a building

and equipment planned for removal, which

are outside the Protected Area. The changes

still require control over the gates and doors

to the Protected Area and that only

authorized individuals will be issued the

Fermi 1 key. Since the changes do not

involve operation of any system,

modifications to any required plant systems,

nor eliminate the requirements for control of

the Fermi 1 key and access points, the

probability or consequences of an accident

will be unaffected.

(2) Will the proposed amendment create

the possibility of a new or different kind of

accident from any accident previously

analyzed?

The proposed changes do not create the

possibility of a new or different type of

accident from any previously evaluated. The

proposed changes will not lead to any

different method of operating any systems,

nor will they create any tests involving plant

systems. The changes only affect the access

control requirements, the Protected Area

boundary, and deletion of details from a

sketch. Changes of who issues the key, how

doors are secured, provisions for temporary

modifications to the boundary, requirements

to observe the Protected Area boundary if

degraded, wording consolidation, and more

accurate building outlines cannot cause a

new or different type of accident. Access

points and the Fermi 1 key are still required

to be controlled. The Boilerhouse and main

unit output transformer are not used to

support the Fermi 1 nuclear facility. Removal

of the Boilerhouse and main unit output

transformer from the drawing will help

facilitate future removal plans, but will not

cause a new or different accident from any

previously evaluated, since they provide no

support to the Fermi 1 nuclear facility. For

these reasons, the proposed changes to the

access control requirements and Figure B–1

will not create the possibility of a new or

different type of accident.

(3) Will the proposed change significantly

reduce the margin of safety at the facility?

The proposed changes do not involve a

significant reduction in the margin of safety.

The changes involve access control, the

Protected Area boundary, and the sketch of

the facility. Doors and gates in the Protected

Area boundary will still be required to be

secured when personnel are not inside. The

keys will still be required to be controlled

and issued only to authorized personnel.

Compensatory measures will be required if

the Protected Area boundary is degraded

such that the requirements are not met.

Therefore, there will not be a significant

reduction in the margin of safety.

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,

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Monroe County Library

System, 3700 South Custer Road,

Monroe, Michigan 48161.

Attorney for licensee: John Flynn,

Esquire, Detroit Edison Company, 2000

Second Avenue, Detroit, Michigan

48226.

NRC Branch Chief: John W. N.

Hickey.

Detroit Edison Company, Docket No.

50–341, Fermi 2, Monroe County,

Michigan

Date of amendment request: June 5,

1998 (NRC–98–0067).

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) 2.1.2 to

incorporate cycle-specific safety limit

minimum critical power ratios

(SLMCPRs) for the core that will be

loaded during the upcoming refueling

outage and update the footnote

associated with the SLMCPR values to

limit applicability of the SLMCPR

values to Cycle 7 operation only.

Basis for proposed no significant

hazards consideration determination:

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:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed license amendment

establishes a revised SLMCPR value of 1.11

for two recirculation loop operation and 1.13

for single recirculation loop operation for use

during Cycle 7 operation. The derivation of

the cycle-specific SLMCPRs was performed

using ‘‘General Electric Standard Application

for Reactor Fuel,’’ NEDE–24011–P–A–13;

U.S. Supplement, EDE–24011–P––A–13–US,

August 1996; and the ‘‘Proposed Amendment

25 to GE Licensing Topical Report NEDE–

24011–P–A (GESTAR II) on Cycle Specific

Safety Limit MCPR.’’ Amendment 25 was

submitted by General Electric Nuclear Energy

(GENE) to the NRC on December 13, 1996.

The probability of an evaluated accident is

derived from the probabilities of the

individual precursors to that accident. The

consequences of an evaluated accident are

determined by the operability of plant

systems designed to mitigate those

consequences. Limits have been established,

consistent with NRC approved methods, to

ensure that fuel performance during normal,

transient, and accident conditions is

acceptable.

The probability of an evaluated accident is

not increased by revising the SLMCPR
Federal Register / Vol. 63, No. 126 /Wednesday, July 1, 1998 / Notices 35989

values. The change does not require any

physical plant modifications or physically

affect any plant components. Therefore, no

individual precursors of an accident are

affected.

The proposed license amendment

establishes a revised SLMCPR that ensures

that the fuel is protected during normal

operation and during any plant transients or

anticipated operational occurrences.

Specifically, the reload analysis demonstrates

that a SLMCPR value of 1.11 (1.13 for single

loop operation) ensures that less than 0.1

percent of the fuel rods will experience

boiling transition during any plant operation

if the limit is not violated.

Based on (1) the determination of the new

SLMCPR values using conservative methods,

and (2) the operability of plant systems

designed to mitigate the consequences of

accidents not having been changed;[,] the

consequences of an accident previously

evaluated have not been increased.

Additionally, updating of the footnote on

the SLMCPR value in Technical Specification

2.1.2 to limit the applicability of the SLMCPR

values to only Cycle 7 operation will not

increase the probability or consequences of

accidents previously evaluated. The updating

of the footnote on the SLMCPR value in

Technical Specification 2.1.2 is an

administrative change that has no effect on

the probability or consequences of accidents

previously evaluated.

Therefore, the proposed TS change does

not involve an increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed license amendment involves

a revision of the SLMCPR from 1.09 to 1.11

for two recirculation loop operation and from

1.11 to 1.13 for single loop operation based

on the results of analysis of the Cycle 7 core

using the same fuel types as in previous fuel

cycles, and updating of the footnote on the

SLMCPR values in TS 2.1.2. Creation of the

possibility of a new or different kind of

accident would require the creation of one or

more new precursors of that accident. New

accident precursors may be created by

modifications of the plant configuration,

including changes in the allowable methods

of operating the facility. This proposed

license amendment does not involve any

modifications of the plant configuration or

changes in the allowable methods of

operation. Therefore, the proposed TS change

does not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

3. The change does not involve a

significant reduction in the margin of safety.

The proposed license amendment

establishes a revised SLMCPR value of 1.11

for two recirculation loop operation and 1.13

for single recirculation loop operation for use

during Cycle 7 operation. The derivation of

the cycle-specific SLMCPRs was performed

using ‘‘General Electric Standard Application

for Reactor Fuel,’’ NEDE–24011–P–A–13;

U.S. Supplement, EDE–24011–P–A–13–US,

August 1996; and the ‘‘Proposed Amendment

25 to GE Licensing Topical Report NEDE–

24011–P–A (GESTAR II) on Cycle Specific

Safety Limit MCPR.’’ Amendment 25 was

submitted by General Electric Nuclear Energy

(GENE) to the NRC on December 13, 1996.

Use of these methods ensures that the

resulting SLMCPR satisfies the fuel design

safety criteria that less than 0.1 percent of the

fuel rods experience boiling transition if the

safety limit is not violated. Based on the

assurance that the fuel design safety criteria

will be met, the proposed license amendment

does not involve a significant reduction in a

margin of safety.

Additionally, updating of the footnote on

the SLMCPR value in TS 2.1.2 will not

decrease the margin of safety for accidents

previously evaluated. The updating of the

footnote on the SLMCPR value in Technical

Specification 2.1.2 is an administrative

change that does not reduce the 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.

Local Public Document Room

location: Monroe County Library

System, Ellis Reference and Information

Center, 3700 South Custer Road,

Monroe, Michigan 48161.

Attorney for licensee: John Flynn,

Esq., Detroit Edison Company, 2000

Second Avenue, Detroit, Michigan

48226.

NRC Project Director: Cynthia A.

Carpenter.

Florida Power and Light Company, et

al., Docket Nos. 50–335 and 50–389, St.

Lucie Plant, Unit Nos. 1 and 2, St. Lucie

County, Florida

Date of amendment request: May 27,

1998.

Description of amendment request:

The request, if granted, would modify

the Technical Specifications to allow

the use of various controlled shift

structures during a 36 to 48 hour work

week. The request will allow the use of

up to 12 hour shifts without routine

heavy use of overtime.

Basis for proposed no significant

hazards consideration determination:

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:

(1) Operation of the facility in accordance

with the proposed amendments would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed amendments will delete the

TS 6.2.2.f. requirement ‘‘. . . to have

operating personnel work a normal 8-hour

day, 40-hour week while the plant is

operating.’’ The proposed change will allow

FPL to implement various controlled shift

structures and durations during a nominal

(36 to 48 hours) work week. The proposed

changes will allow the use of up to 12 hour

shifts without routine heavy use of overtime.

The TS will continue to require the controls

and guidelines for work hours to be

contained in administrative procedures. The

proposed amendments do not involve a

change to any structure, system, or

component that affects the probability or

consequences of an accident previously

evaluated. The proposed amendments are

administrative in nature and do not involve

a significant increase in the probability or

consequences of any accident previously

evaluated.

(2) Operation of the facility in accordance

with the proposed amendments would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed amendments will not change

the physical plant or modes of plant

operation and therefore, will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated. The proposed amendments will

not result in the addition or modification of

equipment for any systems, structures, or

components at St. Lucie.

The proposed changes modify the controls

on working hours for operating personnel

without significantly changing the hours

worked on a weekly or annual basis, and do

not alter the current guidelines on the use of

overtime. The changes are administrative in

nature. Consequently, operation of either unit

in accordance with the proposed amendment

would not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

(3) Operation of the facility in accordance

with the proposed amendments would not

involve a significant reduction in a margin of

safety.

The proposed amendments will delete the

TS 6.2.2.f. requirement ‘‘. . . to have

operating personnel work a normal 8-hour

day, 40-hour week while the plant is

operating.’’ The proposed change will allow

FPL to implement various controlled shift

structures and durations during a nominal

(36 to 48 hours) work week. The proposed

changes will allow the use of up to 12 hour

shifts without routine heavy use of overtime.

The TS will continue to require the controls

and guidelines for work hours to be

contained in administrative procedures. This

will result in fewer operating shift-to-shift

turnovers per day and will allow more

contiguous days off between work shifts. The

net result of longer work shifts will be more

rested crews with better communications

between shifts.

The proposed changes do not alter the

current guidelines on the use of overtime and

will not alter the basis for any TS that is

related to the establishment of, or

maintenance of, a nuclear safety margin.

Consequently, operation of St. Lucie Units 1

and 2 in accordance with the proposed

amendments will not involve a significant

reduction in a margin of safety.
35990 Federal Register / Vol. 63, No. 126 /Wednesday, July 1, 1998 / Notices

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Indian River Community

College Library, 3209 Virginia Avenue,

Fort Pierce, Florida 34981–5596.

Attorney for licensee: M.S. Ross,

Attorney, Florida Power & Light, P.O.

Box 14000, Juno Beach, Florida 33408–

0420

NRC Project Director: Frederick J.

Hebdon.

Florida Power and Light Company, et

al., Docket Nos. 50–335 and 50–389 St.

Lucie Plant, Unit Nos. 1 and 2, St. Lucie

County, Florida

Date of amendment request: June 3,

1998

Description of amendment request:

The request will modify the Technical

Specifications to provide for the use of

an interim periodic method of

monitoring oxygen concentration in the

service waste decay tanks in the event

that continuous monitoring capability is

lost.

Basis for proposed no significant

hazards consideration determination:

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:

(1) Operation of the facility in accordance

with the proposed amendment would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed license amendments are

administrative in nature and will rectify an

inconsistency between Surveillance

Requirement 4.11.2.5.1 and the UFSAR that

was inadvertently created by previous license

amendments. The revisions will reinstate a

previously approved conditional exception to

the explicit terms of the presently stated TS

requirement to continuously monitor the

waste gases in the on service Waste Gas

Decay Tank, and allow limited system

operation using the laboratory gas partitioner

to periodically analyze gas samples in the

event that continuous monitoring capability

becomes inoperable. Limits for potentially

explosive mixtures of waste gases have not

been altered, and explosive gas monitoring

instrumentation does not prevent or mitigate

design basis accidents or transients which

assume a failure of or a challenge to a fission

product barrier. The proposed revisions do

not involve any change to the plant accident

analyses assumptions, and do not involve

accident initiators. Therefore, operation of

either facility in accordance with its

proposed amendment would not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

(2) Operation of the facility in accordance

with the proposed amendment would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed license amendments are

administrative in nature and rectify an

inconsistency between Technical

Specification 4.11.2.5.1 and the UFSAR that

was inadvertently created by previous license

amendments. The revisions will not change

the physical plant or the modes of plant

operation defined in the Facility Licenses.

The changes do not involve the addition or

modification of equipment nor do they alter

the design of plant systems. Therefore,

operation of either facility in accordance

with its proposed amendment would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

(3) Operation of the facility in accordance

with the proposed amendment would not

involve a significant reduction in a margin of

safety.

The proposed license amendments are

administrative in nature and rectify an

inconsistency between Surveillance

Requirement 4.11.2.5.1 and the UFSAR that

was inadvertently created by previous license

amendments. The revisions will reinstate a

previously approved conditional exception to

the explicit terms of the presently stated TS

requirement to continuously monitor the

waste gases in the on service Waste Gas

Decay Tank, and allow limited system

operation using the laboratory gas partitioner

to periodically analyze gas samples in the

event that continuous monitoring capability

becomes inoperable. Limits for potentially

explosive mixtures of waste gases have not

been altered, and explosive gas monitoring

instrumentation does not prevent or mitigate

design basis accidents or transients which

assume a failure of or a challenge to a fission

product barrier. The proposed changes do not

alter the basis for any technical specification

that is related to the establishment of, or the

maintenance of, a nuclear safety margin.

Therefore, operation of either facility in

accordance with its proposed amendment

would 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 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Indian River Community

College Library, 3209 Virginia Avenue,

Fort Pierce, Florida 34981–5596.

Attorney for licensee: M.S. Ross,

Attorney, Florida Power & Light, P.O.

Box 14000, Juno Beach, Florida 33408–

0420.

NRC Project Director: Frederick J.

Hebdon.

Indiana Michigan Power Company,

Docket Nos. 50–315 and 50–316, Donald

C. Cook Nuclear Plant, Units 1 and 2,

Berrien County, Michigan.

Date of amendment requests: March

3, 1998.

Description of amendment requests:

The proposed amendments would

remove the word ‘‘immediately’’ from

the Unit 1 hydrogen recombiner

surveillance requirement 4.6.4.2.b.4 and

revise the Unit 1 and Unit 2 Technical

Specification 3/4.6.4 bases.

Basis for proposed no significant

hazards consideration determination:

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:

In accordance with 10 CFR 50.92, the

proposed changes do not involve a

significant hazards consideration if the

changes do not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated;

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.

Criterion 1

This amendment request does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated. The change removes an

ambiguous word from the technical

specification. It does not physically alter the

recombiner, nor does it adversely impact its

operating characteristics.

The resistance to ground test will continue

to be used to detect circuit faults. However,

with the removal of the word ‘‘immediately’’,

it will be possible to conduct the test near the

ambient temperature, the temperature for

which the 10,000 ohm criterion is applicable.

The previously observed resistance value that

was lower than 10,000 ohms is not indicative

of a faulted heater circuit. Rather, it is the

result of an elevated heater temperature and

the electrical characteristics of the heater’s

insulating material, magnesium oxide.

Magnesium oxide has a negative electrical

resistance temperature coefficient, and it is

not unusual or unacceptable for the

measured insulation resistance to be less

than 10,000 ohms when the heater

temperature is elevated.

Criterion 2

This proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated. The hydrogen recombiner is used

to mitigate the consequences of an accident,

and it performs no function during normal

operation. The change to the surveillance

requirement removes an ambiguous word

and does not affect the equipment or its

installed configuration. No accident initiators

that might be introduced by this change have

been identified.
Federal Register / Vol. 63, No. 126 /Wednesday, July 1, 1998 / Notices 35991

Criterion 3

This proposed change does not involve a

significant reduction in a margin of safety.

The change removes an ambiguous word

from the T/S. The performance

characteristics for the recombiner are not

affected by this change, and no margin of

safety is impacted.

The resistance to ground test will continue

to be used to detect circuit faults. However,

with the removal of the word ‘‘immediately’’,

it will be possible to conduct the test near the

ambient temperature, the temperature for

which the 10,000 ohm criterion is applicable.

The previously observed resistance values

that were lower than 10,000 ohms are not

indicative of a faulted heater circuit. Rather,

they are the result of an elevated heater

temperature and the electrical characteristics

of the heater’s insulating material,

magnesium oxide. Magnesium oxide has a

negative electrical resistance temperature

coefficient, and it is not unusual or

unacceptable for the measured insulation

resistance to be less than 10,000 ohms when

the heater temperature is elevated.

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 requests involve no

significant hazards consideration.

Local Public Document Room

location: Maud Preston Palenske

Memorial Library, 500 Market Street, St.

Joseph, MI 49085

Attorney for licensee: Jeremy J. Euto,

Esq., 500 Circle Drive, Buchanan, MI

49107

NRC Acting Project Director: Dr.

Ronald R. Bellamy

Northeast Nuclear Energy Company

(NNECO), et al., Docket No. 50–423,

Millstone Nuclear Power Station, Unit

No. 3, New London County, Connecticut

Date of amendment request: May 7,

1998

Description of amendment request:

The proposed revision to the Millstone

Unit 3 licensing basis would address the

addition of the dose from refueling

water storage tank (RWST) back leakage

into the design basis loss-of-coolant

accident (LOCA) analysis and Chapter

15 of the Final Safety Analysis Report

(FSAR).

Basis for proposed no significant

hazards consideration determination:

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:

NNECO has reviewed the proposed

revision in accordance with 10CFR50.92 and

concluded that the revision does not involve

a significant hazards consideration (SHC).

The basis for this conclusion is that the three

criteria of 10CFR50.92(c) are not satisfied.

The proposed revision does not involve an

SHC because the revision would not:

1. Involve a significant increase in the

probability or consequence of an accident

previously evaluated.

The RWST is a standby system during

normal operation, and provides the initial

makeup water supply for the Emergency Core

Cooling System (ECCS) when actuated in

response to a Safety Injection signal. The

RWST supply piping does not interface

directly with the Reactor Coolant System or

associated Reactor Coolant Pressure

Boundary piping. All piping, up to and

including the last isolation valve prior to the

RWST, is rated for pressure exceeding RSS

[recirculation spray system] pump discharge

pressure.

The RWST is a passive tank, vented to

atmosphere. Following swapover to postLOCA recirculation cooling, the RWST is

isolated and is no longer required for

accident mitigation purposes. Back leakage

will collect in the tank and mix with any

remaining volume of water. The temperature

of the mixed fluid will not significantly

exceed the ambient temperature of the

remaining tank volume due to the extremely

low leakage rates involved. Because the tank

is vented to atmosphere, pressurization of the

tank [cannot] occur.

The specific condition of back leakage

through the RWST isolation valves in

combination with a motor operated valve

failure does not contribute to the probability

of a malfunction previously evaluated in the

Safety Analysis Report. In lines that contain

a motor operated valve and result in back

leakage to the RWST, there exists another

valve in series. The other valve is either

another motor operated valve, a check valve,

or a manually operated valve. The most

limiting single failure assumed is the failure

of the lowest leakage series valve to close and

results in the maximum calculated leakage

rate. Certain ECCS check valves are not

subject to single failure consideration and are

therefore credited as the barrier valve against

back leakage.

The back leakage into the RWST results in

sump water entering the RWST when it is at

its minimum level. The RWST now becomes

a radioactive source and contributes a shine

dose to the surrounding areas. The increase

in dose rates onsite will not prevent

operators from remaining in the control room

or from accessing equipment needed to

mitigate the accident.

All piping and valves associated with

RWST back leakage are located in harsh

radiation areas. Backflow from RSS could

increase dose rates in the areas where these

valves are located. Since these areas are

already classified as harsh radiation

environments post LOCA, additional dose

contributions from these pipes would not

adversely impact EEQ [environmental

qualification of electrical equipment] doses

to vital equipment located in these rooms.

Any vital equipment located within would

continue to perform its safety function.

The leakage back to the RWST has no effect

on the ability of the RSS pumps to perform

their design function. The NPSH [net positive

suction head] required by the RSS pumps is

not adversely impacted by the loss of sump

water back to the RWST. The RSS switchover

to cold leg recirculation occurs prior to

reaching a minimum level of 392,000 gallons

in the RWST. Not counting the reactor

coolant system volume, 774,000 gallons of

water is in the sump. QSS [quench spray

system] pumps shut off when the inventory

in the RWST decreases to 93,000 gallons.

Another 303,000 gallons will reach the sump

prior to QSS shutoff. RWST back leakage

displaces approximately 36,000 gallons of

sump water back into the RWST at the end

of 720 hours, leaving more than 1,000,000

gallons, not counting RCS [reactor coolant

system] volume, in the sump. When RSS

switches over to recirculation, at least

774,000 gallons of water will remain in the

sump. After 720 hours, more water resides in

the sump than when RSS is started.

Therefore minimum NPSH requirements will

not be impacted by this leakage.

Post-LOCA back leakage to the RWST has

not previously been included in the

radiological consequence analyses for

Millstone Unit 3. Including this source in

dose assessment increases the consequences

of the accident. NNECO has tested the

associated valves to establish bounding

criteria to be used in the analysis of potential

radiological consequences. The contribution

of the RWST back leakage has been

determined to be 2.1 Rem at the LPZ [low

population zone] and 0.9 Rem at the Control

Room. When combined with the present

LOCA analysis radiological consequences,

the results remain below the previously

analyzed values reported in the FSAR. All

dose estimates reflect the limiting exposure

which, in this case, is Thyroid dose. All

resultant doses are less than 10CFR100 and

GDC [General Design Criterion] 19 limits to

offsite and control room.

Back leakage to the RWST from the

operation of RSS is a result of a LOCA. It

cannot increase the probability of a LOCA.

Therefore RWST back leakage does not

increase the probability of an accident

previously evaluated.

Based on the above, the proposed license

amendment request does not involve a

significant increase in the probability or

consequence of an accident previously

evaluated.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

No new condition potentially impacting

the ability to mitigate the accident is created

by the back leakage. The low leakage rates

from these valves occurs over [an] extended

period of time during which other makeup

water sources can be brought into service to

account for lost inventory, if necessary.

Therefore, the proposed license

amendment request does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. Involve a significant reduction in a

margin of safety.

The resultant dose from back leakage of

ECCS valves to the RWST does not reduce

the Margin of Safety. The offsite and control

room doses, with the addition from RWST

back leakage, remain below the licensing
35992 Federal Register / Vol. 63, No. 126 /Wednesday, July 1, 1998 / Notices

base dose as listed in the SAR [safety analysis

report]. Technical Specification 6.8.4 defines

the basis for the leak reduction program. The

basis for the program is to reduce leakage

outside containment to the maximum extent

possible. The Technical Specifications do not

define the maximum amount of leakage or

the origin of the leakage. The addition of the

back leakage valves to the leak reduction

program does not reduce the Margin of

Safety.

Therefore, the proposed license

amendment request does not involve a

significant reduction in a margin of safety.

In conclusion, based on the information

provided, it is determined that the proposed

revision does not involve an SHC.

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.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, Connecticut, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

P.O. Box 270, Hartford, Connecticut.

NRC Deputy Director: Phillip F.

McKee.

Northeast Nuclear Energy Company

(NNECO), et al., Docket No. 50–423,

Millstone Nuclear Power Station, Unit

No. 3, New London County, Connecticut

Date of amendment request: June 5,

1998

Description of amendment request:

The proposed revision to the Millstone

Unit 3 licensing basis would address a

recent steam generator tube rupture

(SGTR) analysis that was determined to

be an unreviewed safety question. The

SGTR analyses described in the Final

Safety Analysis Report (FSAR) include

an offsite dose analysis and a margin to

overfill analysis. Both of the analyses

have been updated. The offsite dose

analysis was updated to reflect a larger

capacity for the steam generator

atmospheric dump valve, and the

margin to overfill analysis was updated

to reflect a new single failure.

Basis for proposed no significant

hazards consideration determination:

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:

NNECO has reviewed the proposed

revision in accordance with 10CFR50.92 and

has concluded that the revision does not

involve a significant hazards consideration

(SHC). The basis for this conclusion is that

the three criteria of 10CFR50.92(c) are not

satisfied. The proposed revision does not

involve an SHC because the revision would

not:

1. Involve a significant increase in the

probability or consequence of an accident

previously evaluated.

The FSAR Steam Generator Tube Rupture

offsite dose analysis is being updated to

reflect a larger capacity for the steam

generator atmospheric dump valve. The

updated analysis, as well as the current

FSAR analysis, postulate the failure, in the

open position, of the steam generator

atmospheric dump valve associated with the

steam generator with the ruptured tube.

Revising the analyses does not impact the

failure probability of the steam generator

atmospheric dump valve. The SGTR analyses

credit closure of the atmospheric dump valve

block valve to isolate the failed open

atmospheric dump valve. The revised SGTR

analysis uses a larger flow capacity for the

atmospheric dump valve. A larger flow

capacity, without other changes being made,

would increase the consequences associated

with this failure. However, the time credited

for closure of the block valve is being

reduced to 20 minutes after the atmospheric

dump valve fails open, instead of 30 minutes

after the atmospheric dump valve fails open.

A shorter isolation time, without other

changes being made, would decrease the

consequences associated with the

atmospheric dump [valve] failing open. This

faster isolation time more than compensates

for the larger capacity assumed for the

atmospheric dump valve. Therefore, the

revised analyses does not increase the

consequences of a Steam Generator Tube

Rupture. The change is a revision to the

analyses for a steam generator tube rupture

and the description of the analyses in the

FSAR. Changing the analyses and its

description [cannot] cause an increase in the

probability of a steam generator tube rupture.

Therefore, the proposed revision does not

involve a significant increase in the

probability or consequence of an accident

previously evaluated.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The change is to the analyses and FSAR

description of that analyses. The important

changes in the analyses are the increased

capacity of the atmospheric dump valve and

the shorter time utilized for isolation of the

failed open atmospheric dump valve. The

only change in equipment credited in the

analyses is the crediting of the block valve to

close when there is a larger flow through the

valve. The block valve can close under the

postulated accident conditions.

Therefore, the proposed revision does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Involve a significant reduction in a

margin of safety.

The revised analyses reduces the time

available to the Operators to isolate the failed

open atmospheric dump valve from 30

minutes to 20 minutes. The actions required

are unchanged. The twenty minutes allows

sufficient time for the Operators to both

recognize the failure of the atmospheric

dump valve and to close the block valve.

However, reducing the available time to the

Operators from 30 minutes to 20 minutes

represents a reduction in the margin for error

available to the Operators and thus

represents a reduction in the margin of

safety. The reduction in the margin of safety

is not significant since the twenty minutes

allowed by the analysis is still significantly

above the typical ten minute minimum

assumed response time for Operator actions

performed in the control room. In addition,

Operator training provides assurance that the

twenty minute time limit is met.

Therefore, the proposed revision does not

involve a significant reduction in a margin of

safety.

In conclusion, based on the information

provided, it is determined that the proposed

revision does not involve an SHC.

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.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, Connecticut, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut.

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

P.O. Box 270, Hartford, Connecticut.

NRC Deputy Director: Phillip F.

McKee.

Northeast Nuclear Energy Company

(NNECO), et al., Docket No. 50–423,

Millstone Nuclear Power Station, Unit

No. 3, New London County, Connecticut

Date of amendment request: June 6,

1998.

Description of amendment request:

The proposed revision to the Millstone

Unit 3 licensing basis relates to

operation of the supplementary leak

collection and release system (SLCRS)

after a postulated accident. Specifically,

the proposed revision to the Final Safety

Analysis Report (FSAR) would address

(1) the manual actions required to trip

the non-nuclear safety grade fans and

time requirements for control room

ventilation realignment, and (2) the

input assumptions and results of the

new loss-of-cooling accident/control rod

ejection accident analyses.

Basis for proposed no significant

hazards consideration determination:
Federal Register / Vol. 63, No. 126 /Wednesday, July 1, 1998 / Notices 35993

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:

NNECO has reviewed the proposed

revision in accordance with 10CFR50.92 and

has concluded that the revision does not

involve a significant hazards consideration

(SHC). The basis for this conclusion is that

the three criteria of 10CFR50.92(c) are not

satisfied. The proposed revision does not

involve [an] SHC because the revision would

not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The potential condition of radioactive

effluent bypassing the isolated boundary in

the Supplemental Leak Collection and

Release System after an accident cannot

contribute to the probability of an accident

previously evaluated. The leakage is caused

by a postulated failure of the non-nuclear

safety grade exhaust fans within the SLCRS

boundary to trip after a safety injection

signal. Operator action is needed to verify

that the fans in question are tripped within

a predetermined time delay after the accident

in order that credit can be taken in the

radiological dose analysis for the isolation of

this source.

The proposed operator action will verify

that the power to the fan motors is

terminated, which cannot create any

conditions leading to a new accident. The

verification will augment the procedure to

minimize the consequences of the accident

itself. The trip circuits of the fan motors do

not interface with safety systems.

The consequences of the limiting design

basis accidents have been evaluated with the

additional bypass leakage. The doses for the

Exclusion Area Boundary, Low Population

Zone and Unit 3 Control room remain below

the previously calculated and approved

licensing values. The calculated doses for the

Technical Support Center are higher than

previously approved, but below the

radiological acceptance criteria of GDC

[General Design Criterion] 19.

Therefore, the proposed license

amendment does not involve a significant

increase in the probability or consequence of

an accident previously evaluated.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

There are no conceivable conditions,

created by the proposed operator action, that

may lead to the possibility of a new accident.

Interruption of power to the exhaust fans is,

in itself, a part of accident mitigating activity.

The proposed activity cannot create an

adverse environment where a possibility of a

new accident has to be considered.

The breakers used to de-energize the fans,

control only the fan motors and no other

equipment. Clear labeling ensures that no

safety equipment is inadvertently deactivated. The revised ventilation system

operating procedure will clearly specify the

order of steps and confirmatory indicators

necessary for safe shutdown of the exhaust

fans. The equipment operator will be briefed

before proceeding to open the breakers to the

affected fan motors. To minimize the

possibility of an error, this step will be done

early in the sequence of procedural steps

performed to re-align the control room

ventilation system to the filtration/

recirculation mode of operation after an

accident.

Therefore, the proposed license

amendment does not create the possibility of

a new or different kind of accident from any

accident previously evaluated.

3. Involve a significant reduction in a

margin of safety.

In considering the impact of the proposed

revision on the margin of safety, as defined

in the Technical Specifications, the impact

on the design basis analysis of the fission

product barriers must be evaluated.

The proposed operator action to trip the

fans is done as part of personnel protective

actions after a major accident, which is to

stop the distribution of radioactive iodine

into the vital areas through the ventilation

system within a predetermined time. The

maintenance of the fission product barriers is

not affected by this action. This potential

source of radioactivity associated with the

ventilation fans discharging through the

closed SLCRS boundary dampers has not

been considered previously in the dose

analysis. Including this source results in a

small increase in the gamma and beta doses

to the Technical Support Center. The GDC 19

limits for protection of personnel in the vital

areas however, are not violated. The

calculated doses to EAB/LPZ [exclusion area

boundary and the low population zone]

zones and to the control room vital area

remain below the current licensing base

values.

Therefore, the proposed license

amendment request does not involve a

significant reduction in a margin of safety.

In conclusion, based on the information

provided, it is determined that the proposed

revision does not involve an SHC.

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.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, Connecticut, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut.

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

P.O. Box 270, Hartford, Connecticut.

NRC Deputy Director: Phillip F.

McKee.

Pennsylvania Power and Light

Company, Docket Nos. 50–387 and 50–

388; Susquehanna Steam Electric

Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of amendment request: April 23,

1998.

Description of amendment request:

The amendment would update the

operating licenses such that the

corporate name of Pennsylvania Power

and Light Company ‘‘be changed to

PP&L, Inc.’’

Basis for proposed no significant

hazards consideration determination:

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:

1. Will the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

No. This request involves an

administrative change only. The Operating

Licenses (OLs) are being changed to reference

the new corporate name of the licensee. No

actual plant equipment or accident analyses

will be affected by the proposed changes.

Therefore, this request will have no impact

on the possibility of any type of accident:

new, different, or previously evaluated.

2. Will the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

No. This request involves an

administrative change only. The OLs are

being changed to reference the new corporate

name of the licensee. No actual plant

equipment or accident analyses will be

affected by the proposed change and no

failure modes not bounded by previously

evaluated accidents will be created.

Therefore, this request will have no impact

on the possibility of any type of accident:

new, different, or previously evaluated.

3. Will the change involve a significant

reduction in a margin of safety?

No. Margin of safety is associated with

confidence in the ability of the fission

product barriers (i.e., fuel and fuel cladding,

Reactor Coolant System pressure boundary,

and containment structure) to limit the level

of radiation dose to the public. This request

involves an administrative change only. The

OLs are being changed to reference the new

corporate name of the licensee.

No actual plant equipment or accident

analyses will be affected by the proposed

change. Additionally, the proposed change

will not relax any criteria used to establish

safety limits, will not relax any safety

systems settings, or will not relax the bases

for any limiting conditions of operation.

Therefore, this request will not impact

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
35994 Federal Register / Vol. 63, No. 126 /Wednesday, July 1, 1998 / Notices

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Osterhout Free Library,

Reference Department, 71 South

Franklin Street, Wilkes-Barre, PA 18701.

Attorney for licensee: Jay Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street NW.,

Washington, DC 20037.

NRC Project Director: Robert A. Capra.

Philadelphia Electric Company, Docket

No. 50–171, Peach Bottom Atomic

Power Station, Unit 1, York County,

Pennsylvania

Date of application for amendment:

March 2, 1998

Brief description of amendment: This

proposed amendment will revise the

Peach Bottom Atomic Power Station,

Unit 1, Technical Specifications (TS) to

include requirements for control of

effluents and annual reporting in

accordance with the requirements of 10

CFR 50.36a.

Basis for proposed no significant

hazards consideration determination:

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:

a. Does the proposed amendment involve

a significant increase in the probability or

consequences of an accident previously

evaluated?

The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated because the proposed changes do

not impact the SAFSTOR status of Unit 1 or

the design of any plant system, structure, or

component (SSC). These changes are

administrative in nature. They do not affect

security at Unit 1 or the potential of

radioactive material being released.

Inspections for potential liquid and gas

effluents have previously been established.

These changes ensure the requirement for

procedures and reporting are listed in TS.

Therefore, these proposed changes do not

increase the probability or consequences of

an accident previously evaluated.

b. Does the proposed amendment create

the possibility of a new or different kind of

accident from any accident previously

evaluated?

The proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated because implementation of the

proposed changes do not involve any

physical changes to plant SSC or impact the

SAFSTOR status. The changes are

administrative in nature. Therefore, the

possibility of a new or different kind of

accident from any accident previously

evaluated is not created.

c. Does the proposed amendment involve

a significant reduction in a margin of safety?

The proposed changes do not involve a

significant reduction in a margin of safety

because the proposed changes do not affect

the plant SAFSTOR status. Because proposed

changes are administrative in nature, they do

not involve a question of safety. These

changes involve reporting and adding a

requirement that procedures be in place for

effluent monitoring. 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 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment requests

involve no significant hazards

consideration.

Local Public Document Room

location: Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Education

Building, Walnut Street and

Commonwealth Avenue, Box 1601,

Harrisburg, Pennsylvania 17105.

Attorney for licensee: J. W. Durham,

Sr., Esquire, Sr. V.P. and General

Counsel, PECO Energy Company, 2301

Market Street, Philadelphia,

Pennsylvania 19101.

NRC Branch Chief: John W. N.

Hickey.

Public Service Electric & Gas Company,

Docket No. 50–354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of amendment request: May 13,

1998.

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) 3/4.10.8,

‘‘Inservice Leak and Hydrostatic

Testing,’’ to delete the requirement for

an operable High Drywell Pressure trip

function. Specifically, TS 3.10.8.a is

being revised to remove the reference to

the Secondary Containment Isolation

Actuation Instrumentation trip function

2.b.

Basis for proposed no significant

hazards consideration determination:

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:

(1) The proposed changes do not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed TS revisions will continue to

allow the performance of inservice leak and

hydrostatic testing at a reactor coolant

temperature of greater than 200 degrees

Fahrenheit but less than or equal to 212

degrees Fahrenheit while considering the

plant to remain in Operational Condition 4;

however, the requirement to have an

operable ‘‘High Drywell Pressure’’ Secondary

Containment Isolation trip function during a

leak or hydrostatic test is being deleted. This

change will not have an impact on the

consequences of an accident previously

evaluated since the tests will continue to be

performed nearly water solid and with all

control rods fully inserted. The stored energy

in the reactor core and coolant will continue

to be very low and the potential for causing

fuel failures with a subsequent increase in

coolant activity will continue to be minimal.

The remaining restrictions provided in

Special Test Exception 3.10.8 requiring

Secondary Containment Integrity and

Filtration, Recirculation and Ventilation

System (FRVS) operability will continue to

provide assurance that potential releases into

secondary containment will be restricted

from direct release to the environment. With

the reactor coolant continued to be limited to

212 degrees Fahrenheit, there will be little or

no flashing of coolant to steam, and any

release of radioactive materials will be

minimized.

In the event of a large primary system leak,

the reactor vessel will rapidly depressurize,

allowing the low pressure Emergency Core

Cooling Systems (ECCS) to operate. The

capability of the required ECCS in

Operational Condition 4 remains adequate to

maintain the core flooded under these

conditions. Small system leaks will continue

to be detected by leakage inspections, which

are an integral part of the inservice leak and

hydrostatic testing programs, before any

significant inventory loss can occur. In

addition, the ‘‘High Drywell Pressure’’

Secondary Containment Isolation trip

function (TS Table 3.3.2–1, Trip Function

2.b) provides no additional protection against

the events of concern during the inservice

leak and hydrostatic tests. As a result, these

changes will not increase the probability of

an accident previously evaluated nor

significantly increase the consequences of an

accident previously evaluated.

(2) The proposed change does not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes to Special Test

Exception 3.10.8 contained in this submittal

will not adversely impact the operation of

any safety related component or equipment.

Since the proposed changes involve no

hardware changes and no changes to existing

structures, systems or components, there can

be no impact on the potential occurrence of

any accident due to new equipment failure

modes. The remaining restrictions provided

in proposed Special Test Exception 3.10.8

requiring Secondary Containment Integrity

and Filtration, Recirculation and Ventilation

System (FRVS) operability will continue to

function as required, which will provide

assurance that potential releases into

secondary containment will be restricted

from direct release to the environment.

Furthermore, there is no change in plant

testing proposed in this change request that

could initiate an event. Therefore, these

changes will 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 a margin of safety.

The proposed TS revisions will still allow

the performance of inservice leak and
Federal Register / Vol. 63, No. 126 /Wednesday, July 1, 1998 / Notices 35995

hydrostatic testing at a reactor coolant

temperature of greater than 200 degrees

Fahrenheit but less than or equal to 212

degrees Fahrenheit while considering the

plant to remain in Operational Condition 4;

however, the requirement to have an

operable ‘‘High Drywell Pressure’’ Secondary

Containment Isolation trip function during a

leak or hydrostatic test is being deleted.

Since the reactor vessel head will remain in

place, secondary containment will continue

to be maintained, sufficient isolation

actuation instrumentation will be maintained

and all systems required in Operational

Condition 4 will continue to be operable in

accordance with the TS, the proposed

changes will not have any significant impact

on any design basis accident or safety limit.

Since Hope Creek will still remain capable of

meeting all applicable design basis

requirements and retaining the capability to

mitigate the consequences of accidents

described in the UFSAR, the proposed

changes contained in this submittal were

determined to not result in 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.

Local Public Document Room

location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070.

Attorney for licensee: Jeffrie J. Keenan,

Esquire, Nuclear Business Unit—N21,

P.O. Box 236, Hancocks Bridge, NJ

08038.

NRC Project Director: Robert A. Capra.

Public Service Electric & Gas Company,

Docket No. 50–354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of amendment request: June 12,

1998.

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) Limiting

Condition for Operation (LCO) sections

3.7.1.1, 3.7.1.2, and 3.7.1.3. Specifically,

the proposed changes implement more

appropriate Ultimate Heat Sink (UHS)

limits for river water temperature,

which increases operational flexibility.

In addition, the Station Service Water

System (SSWS) and Safety Auxiliaries

Cooling System (SACS) TS Action

Statements are being revised to provide

additional restrictions on continued

plant operation. These revisions provide

explicit TS guidance, which maintains

SSWS/SACS operating configurations

within design analysis assumptions.

Basis for proposed no significant

hazards consideration determination:

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:

(1) The proposed changes do not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

LCO 3.7.1.3 Changes

The proposed TS revisions related to UHS

involve no hardware changes and no changes

to existing structures, systems or

components. The UHS and supported system

temperature and configuration limits ensure

that the UHS can remove required heat loads

during design basis accidents and transients

with the proposed UHS river water

temperature limits. The proposed UHS TS

ACTION Statements ensure that the plant is

directed to enter a safe shutdown condition

whenever the capability to mitigate design

basis accidents and transients is lost. The

existing UHS TS surveillance requirements to

increase monitoring of the river water

temperature at 82°F adequately ensures that

the actions required at elevated river water

temperature conditions are taken as

appropriate. Since the UHS will still remain

capable of meeting all applicable design basis

requirements and retaining the capability to

mitigate the consequences of accidents

described in the [Hope Creek] HC [Updated

Final Safety Analysis Report] UFSAR, the

proposed changes were determined to be

justified. As a result, these changes will not

increase the probability of an accident

previously evaluated nor significantly

increase the consequences of an accident

previously evaluated.

LCO 3.7.1.1 and 3.7.1.2 Changes

The proposed TS revisions related to

SSWS/SACS operating configuration

restrictions involve no hardware changes and

no changes to existing structures, systems or

components. The additional restrictions

requiring: 1) SACS heat exchanger

operability in one SSWS/SACS pump per

loop scenarios; and 2) assessments of SACS

loop operability when a SSWS loop is

declared inoperable; ensure that the SSWS/

SACS can remove required heat loads during

design basis accidents and transients with

the proposed UHS river water temperature

limits contained in this submittal. The

proposed SSWS/SACS TS ACTION

Statements ensure that the plant is directed

to enter a safe shutdown condition whenever

the capability to mitigate design basis

accidents and transients is lost. Since SSWS/

SACS will still remain capable of meeting all

applicable design basis requirements and

retaining the capability to mitigate the

consequences of accidents described in the

HC UFSAR, the proposed changes were

determined to be justified. As a result, these

changes will not increase the probability of

an accident previously evaluated nor

significantly increase the consequences of an

accident previously evaluated.

(2) The proposed change does not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

LCO 3.7.1.3 Changes

The proposed changes to the UHS TS

contained in this submittal will not adversely

impact the operation of any safety related

component or equipment. Since the proposed

changes involve no hardware changes and no

changes to existing structures, systems or

components, there can be no impact on the

potential occurrence of any accident due to

new equipment failure modes. The system

configuration limits imposed by the UHS

LCO ensure that supported systems can

remove required heat loads during design

basis accidents and transients with the

proposed UHS river water temperature

limits. Furthermore, there is no change in

plant testing proposed in this change request

that could initiate an event. Therefore, these

changes will not create the possibility of a

new or different kind of accident from any

accident previously evaluated.

LCO 3.7.1.1 and 3.7.1.2 Changes

The proposed changes to the SSWS/SACS

TS contained in this submittal will not

adversely impact the operation of any safety

related component or equipment. Since the

proposed changes involve no hardware

changes and no changes to existing

structures, systems or components, there can

be no impact on the potential occurrence of

any accident due to new equipment failure

modes. The system configuration limits

imposed by the SSWS/SACS LCOs ensure

that systems can remove required heat loads

during design basis accidents and transients

with the proposed UHS river water

temperature limits. Furthermore, there is no

change in plant testing proposed in this

change request that could initiate an event.

Therefore, these changes will 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 a margin of safety.

LCO 3.7.1.3 Changes

The proposed changes for the TS related to

the UHS ensure continued capability of the

UHS to mitigate the consequences of design

basis accidents and transients. The UHS

supported systems’ configuration limits and

changes to the operating limits of the UHS

ensure that the UHS can remove required

heat loads during design basis accidents and

transients with the proposed river water

temperature limits. The proposed UHS TS

ACTION Statements ensure that the plant is

directed to: 1) enter a safe shutdown

condition whenever the capability to mitigate

design basis accidents and transients is lost;

or 2) enter a conservatively short period of

continued operation when supported system

redundancy is reduced. Since the UHS will

still remain capable of meeting all applicable

design basis requirements and retaining the

capability to mitigate the consequences of

accidents described in the HC UFSAR, the

proposed changes contained were

determined to not result in a significant

reduction in a margin of safety.

LCO 3.7.1.1 and 3.7.1.2 Changes

The proposed changes for the TS related to

the SSWS/SACS ensure continued capability

of these systems to mitigate the consequences
35996 Federal Register / Vol. 63, No. 126 /Wednesday, July 1, 1998 / Notices

of design basis accidents and transients. The

proposed configuration limits ensure that the

safety-related heat removal systems can

perform their safety functions during design

basis accidents and transients with the

proposed river water temperature limits. The

SSWS/SACS TS ACTION Statements ensure

that the plant is directed to: 1) enter a safe

shutdown condition whenever the capability

to mitigate design basis accidents and

transients is lost; or 2) enter a conservatively

short period of continued operation when

supported system redundancy is reduced.

Since the SSWS/SACS will still remain

capable of meeting all applicable design basis

requirements and retaining the capability to

mitigate the consequences of accidents

described in the HC UFSAR, the proposed

changes contained were determined to not

result in 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.

Local Public Document Room

location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070.

Attorney for licensee: Jeffrie J. Keenan,

Esquire, Nuclear Business Unit—N21,

P.O. Box 236, Hancocks Bridge, NJ

08038.

NRC Project Director: Robert A. Capra.

STP Nuclear Operating Company, Docket

Nos. 50–498 and 50–499, South Texas

Project, Units 1 and 2, Matagorda County,

Texas

Date of amendment request: May 7,

1998.

Description of amendment request:

The proposed amendment would

change the Technical Specifications

(TSs) to reflect reactor coolant system

flow differences between the existing

Model E and the replacement Delta 94

steam generators (SGs). Specifically, it

would (1) add a new reactor core safety

limit figure in TS 2.1.1, Reactor Core

Safety Limits, that shows curves that are

a function of core temperature, power

and operating pressure, applicable to

the Delta 94 SGs, (2) add a footnote in

TS Table 2.2–1, Reactor Trip System

Instrumentation Trip Setpoints, to

specify a new design loop flow rate

applicable to the Delta 94 SGs, and (3)

add a new flow rate requirement to TS

3.2.5, Departure from Nucleate Boiling

(DNB) Parameters, applicable to the

Delta 94 SGs. Related changes to the TS

Bases were also proposed for Bases

2.1.1, Reactor Core Safety Limits, and

Bases 3/4.2.5, DNB Parameters.

Basis for proposed no significant

hazards consideration determination:

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:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed Technical Specification

changes are necessary to reflect new

conditions associated with replacement of

the steam generators. The differences in the

replacement steam generators only require

small changes to parameters modeled in

existing accident analyses. Accident analyses

affected by the replacement steam generator

parameter changes have each been evaluated

to establish that there is no significant change

in the documented results. In cases where an

evaluation was not adequate, new analyses

have been performed to verify that there is

no significant change in the consequences of

the affected accidents.

The Technical Specification changes

specify new requirements (i.e., changed RCS

[reactor coolant system] flow) which support

the new and existing accident analyses. The

accident analysis performed for these new

requirements determined that neither the

probability, nor the consequences, of

accidents previously evaluated in the UFSAR

[Updated Final Safety Analysis Report]

would be increased.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed Technical Specification

changes are necessary to reflect new

conditions associated with replacement of

the steam generators. The differences in the

replacement steam generators only require

small changes to parameters modeled in

existing accident analyses. The replacement

of the original steam generators with new

Model Delta 94 steam generators improves

the structural integrity of the steam generator

tubes. The improved structural integrity of

the new steam generators does not increase

the possibility of a new or different kind of

accident from any accident previously

evaluated such as a multiple steam generator

tube rupture event.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed change does not alter the

manner in which Safety Limits, Limiting

Safety System Setpoints, or Limiting

Conditions for Operations are determined.

Changes in parameters assumed in safety

analyses associated with replacement of the

steam generators have been analyzed and

new Technical Specification limits are

proposed. The new limits proposed for SL

[Safety Limit] 2.1.1, ‘‘Reactor Core’’; Table

2.2–1, ‘‘Reactor Trip System Instrumentation

Trip Setpoints’’; and LCO [Limiting

Condition for Operation] 3.2.5, ‘‘DNB

[Departure from Nucleate Boiling]

Parameters’’ maintain or improve the margin

of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the standards of

10 CFR 50.92(c) are satisfied. Therefore,

the NRC staff proposes to determine that

the request for amendments involves no

significant hazards consideration.

Local Public Document Room

location: Wharton County Junior

College, J. M. Hodges Learning Center,

911 Boling Highway, Wharton, TX

77488.

Attorney for licensee: Jack R.

Newman, Esq., Morgan, Lewis &

Bockius, 1800 M Street, N.W.,

Washington, DC 20036–5869.

NRC Project Director: John N.

Hannon.

Wisconsin Public Service Corporation,

Docket No. 50–305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of amendment request: June 1,

1998.

Description of amendment request:

The proposed amendment would revise

the minimum steam generator (SG) tube

roll expansion distances for the F* and

elevated F* (EF*) repair criteria that

were approved in Amendment 129.

Basis for proposed no significant

hazards consideration determination:

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:

The proposed change was reviewed in

accordance with the provisions of 10 CFR

50.92 to show no significant hazards exist.

The proposed change will not:

(1) Involve a significant increase in the

probability or consequence of an accident

previously evaluated.

The changes to the minimum engagement

lengths for F* and EF* do not change any of

the conclusions of the original F* and EF*

analyses. The technical justification for the

repair criteria has not changed due to

changes in the engagement lengths. The

calculated engagement lengths continue to

preclude tube pullout and rupture during all

postulated conditions. Based on the geometry

of the Model 51 SG, tube rupture type release

rates are not expected for a postulated failure

at an F* or EF* repair location. Engagement

lengths were calculated such that structural

integrity of the repaired tube meets the RG

[Regulatory Guide] 1.121 requirements.

Therefore, application of the new F* and EF*

distances will not increase the probability of

an accident previously evaluated.

The new calculated engagement lengths

continue to preclude primary to secondary

leakage during all conditions. Leakage for

both F* and EF* remains negligible at normal

operating conditions. The amount of leakage

expected at faulted conditions from F* and

EF* repaired tubes remains a small

percentage of the maximum allowable leak

rate during a[n] SLB [steamline break] and is

considered negligible. Therefore, it can be

concluded that leakage will be restricted

such that off-site doses will not exceed a

small fraction of 10 CFR part 100 and control
Federal Register / Vol. 63, No. 126 /Wednesday, July 1, 1998 / Notices 35997

room doses will not exceed GDC [General

Design Criterion] 19 criteria. Therefore, the

proposed change to the F* and EF* distances

will not increase the consequences of an

accident previously evaluated.

(2) Create the possibility of a new or

different kind of accident from any

previously evaluated.

Implementation of the proposed changes in

F* and EF* distances does not introduce any

significant changes to the plant design basis.

As with the original acceptance of the

amendment for using the original F* and EF*

criteria, use of the proposed F* and EF*

engagement lengths will not introduce a

mechanism that will result in an accident

initiated outside of the tubesheet crevice

region. As previously discussed, the

structural integrity of F* and EF* tubes will

be maintained during all plant conditions.

Any hypothetical accident as a result of tube

degradation in the tubesheet crevice region of

the tube will be bounded by the existing tube

rupture analysis. Therefore, implementation

of the proposed engagement lengths for F*

and EF* will not create the possibility of a

new or different kind of accident.

(3) Involve a significant reduction in the

margin of safety.

The calculation for the new F* and EF*

minimum engagement lengths used the same

methodology as the original F* and EF*

analysis. The only change was the assumed

normal operating primary to secondary

differential pressure. The new assumed

differential pressure is the design differential

pressure for the KNPP [Kewaunee Nuclear

Power Plant] SGs. The calculation for the

engagement lengths continues to use the

appropriate safety factors from RG 1.121. The

revised F* and EF* engagement lengths

continue to preclude tube pullout at all plant

conditions and to maintain the structural

integrity of the tube. Additionally, primary to

secondary leakage during all plant conditions

is precluded as described in the preceding

sections. Since the structural and leakage

integrity is not changed by the proposed

changes in engagement length, the margin of

safety is not significantly reduced.

Additionally, use of the F* and EF* repair

criteria will decrease the number of tubes

removed from service by plugging or repaired

by sleeving. Since both plugging and sleeving

reduce reactor coolant flow margin,

implementation of the F* and EF* repair

criteria helps to maintain that flow margin.

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.

Local Public Document Room

location: University of Wisconsin,

Cofrin Library, 2420 Nicolet Drive,

Green Bay, WI 54311–7001.

Attorney for licensee: Bradley D.

Jackson, Esq., Foley and Lardner, P.O.

Box 1497, Madison, WI 53701–1497.

NRC Acting Project Director: Ronald

R. Bellamy.

Previously Published Notices of

Consideration of Issuance of

Amendments to Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing

The following notices were previously

published as separate individual

notices. The notice content was the

same as above. They were published as

individual notices either because time

did not allow the Commission to wait

for this biweekly notice or because the

action involved exigent circumstances.

They are repeated here because the

biweekly notice lists all amendments

issued or proposed to be issued

involving no significant hazards

consideration.

For details, see the individual notice

in the Federal Register on the day and

page cited. This notice does not extend

the notice period of the original notice.

Duke Energy Corporation, Docket Nos.

50–269, 50–270, and 50–287, Oconee

Nuclear Station, Units 1, 2, and 3,

Oconee County, South Carolina

Date of amendment request: February

2, 1998, as supplemented February 18,

1998.

Description of amendment request:

The proposed amendments would

revise the wording to specify refueling

outage surveillances. The changes

clarify that these surveillances are to be

performed on an 18-month frequency

and need not be constrained to refueling

outage conditions.

Date of publication of individual

notice in Federal Register: February 10,

1998 (63 FR 6784).

Expiration date of individual notice:

For comments February 24, 1998; For

hearing March 12, 1998.

Local Public Document Room

location: Oconee County Library, 501

West South Broad Street, Walhalla,

South Carolina.

Duke Energy Corporation, Docket Nos.

50–269 and 50–287, Oconee Nuclear

Station, Units 1 and 3, Oconee County,

South Carolina

Date of amendment request: June 4,

1998.

Description of amendment request:

The proposed amendments would

revise Technical Specification 4.17.2 to

allow continued operation with certain

steam generator tubes that exceed their

repair limit as a result of tube end

anomalies. This action temporarily

exempts these tubes from the

requirement for sleeving, rerolling, or

removal from service until they are

repaired during or before the next

scheduled refueling outages for the

respective unit.

Date of publication of individual

notice in Federal Register: June 17, 1998

(63 FR 33097).

Expiration date of individual notice:

For comments July 1, 1998; For hearing

July 17, 1998.

Local Public Document Room

location: Oconee County Library, 501

West South Broad Street, Walhalla,

South Carolina.

Tennessee Valley Authority, Docket

Nos. 50–259, 50–260 and 50–296,

Browns Ferry Nuclear Plant, Units 1, 2

and 3, Limestone County, Alabama

Date of application for amendments:

June 6 and December 11, 1996, April 11,

May 1, August 14, October 15,

November 5 and 14, December 3, 4, 15,

22, 23, 29 and 30, 1997, January 23,

March 12 and 13, April 16, 20 and 28,

May 7, 14 and 19, and June 2, 1998.

Brief description of amendments:

Conversion to Standard Improved

Technical Specifications (TSs).

Supplements requested less restrictive

changes to the planned conversion.

These changes involve (1) plant-specific

application of generically approved

methodology supporting extended

instrument surveillance intervals and

allowed outage times, (2) operating

practice to treat secondary containment

as a single zone, (3) TS changes to

support installation of a Power Range

Neutron Monitoring System, Average

Power Range Monitor and Rod Block

Monitor TS improvements, and the

Maximum Extended Load Line Limit

analysis, (4) TSs to specify reactor

vessel water level should be greater than

the top of the irradiated fuel, (5) reflect

plant-specific design condition that

excludes average U–235 enrichment, (6)

all spiral off-load procedures and adopt

revision to Surveillance Requirement

(SR). Also, changes to (1) SR relating to

core reactivity difference between actual

and expected critical rod configuration,

(2) calibration frequency for local power

range monitors and (3) an alternate SR

for Unit 3 for position verification of the

low pressure core injection cross tie

valves.

Date of publication of individual

notices in the Federal Register: June 1,

1998 (63 FR 29763), and June 12, 1998

(63 FR 32252).

Expiration dates of individual notices:

July 1, 1998 (63 FR 29763) and July 13,

1998 (63 FR 32252).

Local Public Document Room

location: Athens Public Library, South

Street, Athens, Alabama 35611.
35998 Federal Register / Vol. 63, No. 126 /Wednesday, July 1, 1998 / Notices

Tennessee Valley Authority, Docket

Nos. 50–260 and 50–296, Browns Ferry

Nuclear Plant, Units 2 and 3, Limestone

County, Alabama

Date of application for amendments:

October 1, 1997, as supplemented

October 14, 1997, March 16, April 1 and

28, May 1 and 20, 1998.

Brief description of amendments:

Change Technical Specifications to

allow operation at the uprated power

level of 3458 MWt which represents a

power level increase of 5 percent.

Date of publication of individual

notice in the Federal Register: June 9,

1998 (63 FR 31533).

Expiration date of individual notice:

July 9, 1998.

Local Public Document Room

location: Athens Public Library, South

Street, Athens, Alabama 35611.

Notice of Issuance of Amendments to

Facility Operating Licenses

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application

complies with the standards and

requirements of the Atomic Energy Act

of 1954, as amended (the Act), and the

Commission’s rules and regulations.

The Commission has made appropriate

findings as required by the Act and the

Commission’s rules and regulations in

10 CFR Chapter I, which are set forth in

the license amendment.

Notice of Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for A Hearing in

connection with these actions was

published in the Federal Register as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the applications for

amendment, (2) the amendment, and (3)

the Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, the Gelman Building, 2120 L

Street, NW., Washington, DC, and at the

local public document rooms for the

particular facilities involved.

Boston Edison Company, Docket No.

50–293, Pilgrim Nuclear Power Station,

Plymouth County, Massachusetts

Date of application for amendment:

March 25, 1998, as supplemented on

April 8, and May 5, 1998.

Brief description of amendment: The

amendment modifies the Pilgrim

Nuclear Power Station Technical

Specification Section 3.6.A.1 with

respect to the monitoring requirements

for the vessel flange and adjacent shell

differential temperature during heatup

and cooldown and removes the 145

°Fahrenheit differential temperature

limit.

Date of issuance: June 19, 1998.

Effective date: As of the date of

issuance, to be implemented within 30

days.

Amendment No.: 175.

Facility Operating License No. DPR–

35: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: April 28, 1998 (63 FR 23304).

The May 5, 1998, letter provided

clarifying information that did not

change the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated June 19, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts

02360

Commonwealth Edison Company,

Docket Nos. 50–373 and 50–374, LaSalle

County Station, Units 1 and 2, LaSalle

County, Illinois

Date of application for amendments:

July 15, 1996, as supplemented on June

19, 1997, and February 2, 1998.

Brief description of amendments: The

amendments relocate requirements

related to fire protection from the

Technical Specifications (TS) to the

Updated Final Safety Analysis Report.

The TS sections to be relocated are: 3/

4.3.7.9, Fire Detection Instrumentation;

3/4.7.5, Fire Suppression Systems; 3/

4.7.6, Fire Rated Assemblies; and

6.1.C.4, Fire Brigade Staffing. The

amendments also replace License

Condition 2.C.(25) for Unit 1 and

License Condition 2.C.(15) for Unit 2.

Date of issuance: June 10, 1998.

Effective date: Immediately, to be

implemented within 60 days.

Amendment Nos.: 127 and 112.

Facility Operating License Nos. NPF–

11 and NPF–18: The amendments

revised the operating licenses and the

Technical Specifications.

Date of initial notice in Federal

Register: September 25, 1996 (61 FR

50340). The June 19, 1997, and February

2, 1998, supplements clarified the

license conditions by providing specific

approval dates for previous fire

protection safety evaluations. This

information was within the scope of the

original application and did not change

the staff’s initial proposed no significant

hazards consideration determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated June 10, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Jacobs Memorial Library,

Illinois Valley Community College,

Oglesby, Illinois 61348

Consumers Energy Company, Docket

No. 50–255, Palisades Plant, Van Buren

County, Michigan

Date of application for amendment:

March 13, 1998, as supplemented March

30, 1998.

Brief description of amendment: The

amendment revises the auxiliary

feedwater system technical specification

to allow two auxiliary feedwater flow

control valves in one train to be

inoperable for up to 72 hours.

Date of issuance: June 10, 1998.

Effective date: June 10, 1998.

Amendment No.: 183.

Facility Operating License No. DPR–

20: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: April 22, 1998 (63 FR 19967)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated June 10, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Van Wylen Library, Hope

College, Holland, Michigan 49423–3698

Detroit Edison Company, Docket No.

50–341, Fermi 2, Monroe County,

Michigan

Date of amendment request: May 20,

1998 (NRC–98–0099)

Description of amendment request:

The amendment revises the action

specified in Technical Specification

3.1.3.1, ‘‘Control Rod Operability,’’ by

changing the action statements

associated with the scram discharge

volume vent and drain valves to align

with those in the NUREG–1433,

Revision 1, ‘‘Standard Technical
Federal Register / Vol. 63, No. 126 /Wednesday, July 1, 1998 / Notices 35999

Specifications General Electric Plants,

BWR/4.’’

Date of issuance: June 12, 1998.

Effective date: June 12, 1998.

Amendment No: 120.

Facility Operating License No. NPF–

43: Amendment revises the Technical

Specifications.

Public comments requested as to

proposed no significant hazards

considerations (NSHC): Yes (63 FR

29254 dated May 28, 1998). The notice

provided an opportunity to submit

comments on the Commission’s

proposed NSHC determination. No

comments have been received. The

notice also provided for an opportunity

to request a hearing by June 29, 1998,

but indicated that if the Commission

makes a final NSHC determination, any

such hearing would take place after

issuance of the amendment.

The Commission’s related evaluation

of the amendment, finding of exigent

circumstances, and final determination

of no significant hazards consideration

are contained in a Safety Evaluation

dated June 12, 1998.

Local Public Document Room

location: Monroe County Library

System, Ellis Reference and Information

Center, 3700 South Custer Road,

Monroe, Michigan 48161.

Attorney for licensee: John Flynn,

Esq., Detroit Edison Company, 2000

Second Avenue, Detroit, Michigan

48226.

NRC Project Director: Cynthia A.

Carpenter.

Duke Energy Corporation, et al., Docket

Nos. 50–413 and 50–414, Catawba

Nuclear Station, Units 1 and 2, York

County, South Carolina

Date of application for amendments:

May 22, 1998.

Brief description of amendments: The

amendments revise Surveillance

Requirement Section 4.4.3.3 of each

unit’s Technical Specification to be

consistent with the plant design;

specifically, deleting the reference to

manual transfer of power supply from

normal to emergency.

Date of issuance: June 17, 1998.

Effective date: As of the date of

issuance.

Amendment Nos.: Unit 1—166; Unit

2—158.

Facility Operating License Nos. NPF–

35 and NPF–52: The amendments

revised the Technical Specifications.

Public comments requested as to

proposed no significant hazards

consideration: Yes. (63 FR 29759 dated

June 1, 1998). That notice provided an

opportunity to submit comments on the

Commission’s proposed no significant

hazards consideration determination.

No. comments have been received. The

notice also provided for an opportunity

to request a hearing by July 1, 1998, but

indicated that if the Commission makes

a final no significant hazards

consideration determination, any such

hearing would take place after issuance

of the amendments.

The Commission’s related evaluation

of the amendments, finding of exigent

circumstances, and final no significant

hazards consideration determination are

contained in a Safety Evaluation dated

June 17, 1998.

Local Public Document Room

location: York County Library, 138 East

Black Street, Rock Hill, South Carolina

29730.

Attorney for licensee: Mr. Paul R.

Newton, Legal Department (PB05E),

Duke Energy Corporation, 422 South

Church Street, Charlotte, North Carolina

28242.

NRC Project Director: Herbert N.

Berkow.

Florida Power Corporation, et al.,

Docket No. 50–302, Crystal River Unit

No. 3 Nuclear Generating Plant, Citrus

County, Florida

Date of application for amendment:

March 20, 1998.

Brief description of amendment: The

amendment revised the Improved

Technical Specification 5.6.2.8 to reflect

the current schedule for performing the

required reactor coolant pump flywheel

inspection.

Date of issuance: June 8, 1998.

Effective date: June 8, 1998.

Amendment No.: 167.

Facility Operating License No. DPR–

31: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: May 6, 1998 (63 FR 25110).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated June 8, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida

34428.

Florida Power and Light Company,

Docket Nos. 50–250 and 50–251, Turkey

Point Plant Units 3 and 4, Dade County,

Florida

Date of application for amendments:

November 22, 1996, as revised and

replaced February 2, 1998.

Brief description of amendments: The

amendments revise the Technical

Specifications (TS) to allow for the

installation of a temporary fuel oil

storage and transfer system in order to

maintain the operability of one Unit 3

emergency diesel generator during the

performance of a required surveillance

to clean the permanent fuel oil storage

tank.

Date of issuance: June 9, 1998.

Effective date: June 9, 1998.

Amendment Nos.: 197 and 191.

Facility Operating Licenses Nos. DPR–

31 and DPR–41: Amendments revised

the TS.

Date of initial notice in Federal

Register: February 25, 1998 (63 FR

9604).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated June 9, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Florida International

University, University Park, Miami,

Florida 33199.

North Atlantic Energy Service

Corporation, et al., Docket No. 50–443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request: March

23, 1998.

Description of amendment request:

The proposed change would revise the

Seabrook Station Technical

Specifications (TSs) to add a new TS

3.0.5 that would provide an exception to

TSs 3.0.1 and 3.0.2 to allow the

performance of required testing to

demonstrate the operability of the

equipment being returned to service or

the operability of other equipment.

Date of issuance: June 16, 1998.

Effective date: As of its date of

issuance, to be implemented within 60

days.

Amendment No.: 57.

Facility Operating License No. NPF–

86: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: April 22, 1998 (63 FR 19972)

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated June 16, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Exeter Public Library,

Founders Park, Exeter, NH 03833.

Northeast Nuclear Energy Company, et

al., Docket No. 50–336, Millstone

Nuclear Power Station, Unit No. 2, New

London County, Connecticut.

Date of application for amendment:

December 8, 1997.

Brief description of amendment: The

changes modify the Technical

Specifications to resolve several

compliance issues by rewording of the

text, changing terminology, correcting a
36000 Federal Register / Vol. 63, No. 126 /Wednesday, July 1, 1998 / Notices

mode applicability, correcting a

formula, updating the Design Features

section, and updating the Bases section

to reflect the changes.

Date of issuance: June 16, 1998.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 216.

Facility Operating License No. DPR–

65: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 28, 1998 (63 FR 4319).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated June 16, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, Connecticut, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut

Pacific Gas and Electric Company,

Docket Nos. 50–275 and 50–323, Diablo

Canyon Nuclear Power Plant, Unit Nos.

1 and 2, San Luis Obispo County,

California

Date of application for amendments:

February 14, 1997, as supplemented by

letters dated October 9, 1997, March 31,

1998, and April 15, 1998.

Brief description of amendments: The

amendments revised the combined

Technical Specifications (TS) for the

Diablo Canyon Power Plant, Unit Nos. 1

and 2 to change the surveillance

frequencies from at least once every 18

months to at least once per refueling

interval (nominally 24 months) for (1)

eight slave relays, (2) 20 electrical

system tests, (3) one electrical Bases

change, and (4) five miscellaneous tests.

Date of issuance: June 5, 1998.

Effective date: June 5, 1998, to be

implemented within 90 days from the

date of issuance.

Amendment Nos.: Unit 1—126; Unit

2—124.

Facility Operating License Nos. DPR–

80 and DPR–82: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: March 26, 1997 (62 FR 14466).

The October 9, 1997, March 31, 1998,

and April 15, 1998, supplemental letters

provided additional information and

did not change the staff’s initial no

significant hazards consideration

determination. The Commission’s

related evaluation of the amendments is

contained in a Safety Evaluation dated

June 5, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: California Polytechnic State

University, Robert E. Kennedy Library,

Government Documents and Maps

Department, San Luis Obispo, California

93407

Philadelphia Electric Company, Docket

Nos. 50–352 and 50–353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of application for amendments:

January 27, 1998.

Brief description of amendments:

These amendments revise Table 3.6.3–1

of the Technical Specifications by

removing the isolation time for the high

pressure coolant injection turbine

exhaust valves and adding a notation

that the isolation is not required.

Date of issuance: June 16, 1998.

Effective date: As of date of issuance,

to be implemented within 30 days.

Amendment Nos.: 129 and 90.

Facility Operating License Nos. NPF–

39 and NPF–85. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: March 11, 1998 (63 FR 11921).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated June 16, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464.

Power Authority of the State of New

York, Docket No. 50–286, Indian Point

Nuclear Generating Unit No. 3,

Westchester County, New York

Date of application for amendment:

June 25, 1997, as supplemented by letter

dated June 2, 1998.

Brief description of amendment: The

amendment changes the Technical

Specifications (TSs) to allow for up to

+171⁄2 steps of control rod misalignment

when power is greater than 85%.

Date of issuance: June 17, 1998.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 180.

Facility Operating License No. DPR–

64: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 27, 1997 (62 FR 45461)

The June 2, 1998, supplement

provided a clarification to the wording

of the TSs and did not change the staff’s

proposed finding of no significant

hazards consideration. The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated June 17, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10610.

Power Authority of the State of New

York, Docket No. 50–286, Indian Point

Nuclear Generating Unit No. 3,

Westchester County, New York

Date of application for amendment:

September 3, 1997.

Brief description of amendment: The

amendment changes the Technical

Specifications (TSs) by revising the

number of hours operating personnel

can work in a normal shift. The

proposed amendment also contains

some administrative changes to the TS.

Date of issuance: June 17, 1998.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 181

Facility Operating License No. DPR–

64: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 22, 1997 (62 FR

54875).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated June 17, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10610.

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of application for amendments:

February 13, 1998 (TS 97–03).

Brief description of amendments: The

amendments change the Technical

Specifications by adding a new Limiting

Condition for Operation 3.7.1.6 that

addresses the requirements for the main

feedwater isolation valve functions

required by the Sequoyah Nuclear Plant

accident analysis.

Date of issuance: June 8, 1998.

Effective date: As of the date of

issuance to be implemented no later

than 45 days after issuance.

Amendment Nos.: Unit 1–232; Unit

2–222.

Facility Operating License Nos. DPR–

77 and DPR–79: Amendments revise the

technical specifications.

Date of initial notice in Federal

Register: April 22, 1998 (63 FR 19979).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated June 8, 1998.

No significant hazards consideration

comments received: No.
Federal Register / Vol. 63, No. 126 /Wednesday, July 1, 1998 / Notices 36001

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 101 Broad Street, Chattanooga,

Tennessee 37402.

Tennessee Valley Authority, Docket No.

50–390 Watts Bar Nuclear Plant, Unit 1,

Rhea County, Tennessee

Date of application for amendment:

April 29, 1998.

Brief description of amendment: The

requested changes would allow,

temporarily, both trains of hydrogen

igniters to be declared inoperable for up

to 72 hours.

Date of issuance: June 9, 1998.

Effective date: June 9, 1998.

Amendment No.: 10.

Facility Operating License No. NPF–

90: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: May 7, 1998 (63 FR 25243).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated June 9, 1998.

No significant hazards consideration

comments received: None.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga,

TN 37402

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50–346, Davis-Besse Nuclear Power

Station, Unit 1, Ottawa County, Ohio

Date of application for amendment:

August 26, 1997.

Brief description of amendment: This

amendment changed Technical

Specification (TS) Section 3/4.2, ‘‘Power

Distribution Limits.’’ The departure

from nucleate boiling parameters

limiting condition for operation was

modified due to an industry

notification.

Date of issuance: June 11, 1998.

Effective date: June 11, 1998.

Amendment No.: 222.

Facility Operating License No. NPF–3:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 8, 1997 (62 FR 52590)

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated June 11, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of Toledo, William

Carlson Library, Government

Documents Collection, 2801 West

Bancroft Avenue, Toledo, OH 43606.

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50–346, Davis-Besse Nuclear Power

Station, Unit 1, Ottawa County, Ohio

Date of application for amendment:

August 26, 1997.

Brief description of amendment: This

amendment revises Technical

Specification (TS) Section 3/4.6.1.3,

‘‘Containment Systems—Containment

Air Locks,’’ and the associated bases.

The limiting condition for operation and

the surveillance requirements were

modified. The application also proposed

a change to TS Bases 3/4.9.4, ‘‘Refueling

Operations—Containment

Penetrations.’’ That bases change was

approved by letter dated March 19,

1998.

Date of issuance: June 11, 1998.

Effective date: June 11, 1998.

Amendment No.: 223.

Facility Operating License No. NPF–3:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 22, 1997 (62 FR

54876)

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated June 11, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of Toledo, William

Carlson Library, Government

Documents Collection, 2801 West

Bancroft Avenue, Toledo, OH 43606

Virginia Electric and Power Company, et

al., Docket Nos. 50–280 and 50–281,

Surry Power Station, Units 1 and 2,

Surry County, Virginia

Date of application for amendments:

December 18, 1997.

Brief Description of amendments:

These amendments revise the Technical

Specifications (TS) to clarify the

terminology used for describing

equipment surveillances performed on a

refueling interval frequency, and to use

consistent wording.

In two cases the proposed changes are

denied. These two exceptions, TS

4.6.A.1.b and 4.6.C.1.e, do not include

required specific Mode restrictions and

could not be approved at this time. If

appropriate revisions are submitted,

these two exceptions could be found to

be acceptable at a later time.

Date of issuance: June 11, 1998.

Effective date: June 11, 1998.

Amendment Nos.: 213 and 213.

Facility Operating License Nos. DPR–

32 and DPR–37: Amendments change

the Technical Specifications.

Date of initial notice in Federal

Register: May 6, 1998 (63 FR 25118).

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated June 11, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Swem Library, College of

William and Mary, Williamsburg,

Virginia 23185.

Virginia Electric and Power Company, et

al., Docket Nos. 50–280 and 50–281,

Surry Power Station, Units 1 and 2,

Surry County, Virginia

Date of application for amendments:

November 5, 1997, as supplemented

January 28, 1998 and May 12, 1998.

Brief Description of amendments:

These amendments permit an increase

in the maximum allowable fuel

enrichment for core reloads from 4.1 to

4.3 weight percent U235.

Date of issuance: June 19, 1998.

Effective date: June 19, 1998.

Amendment Nos.: 214 and 214.

Facility Operating License Nos. DPR–

32 and DPR–37: Amendments change

the Technical Specifications.

Date of initial notice in Federal

Register: December 31, 1997 (62 FR

68320)

The January 28 and May 12, 1998

submittals provided clarifying

information that did not affect the initial

no significant hazards determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated June 19, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Swem Library, College of

William and Mary, Williamsburg,

Virginia 23185

Virginia Electric and Power Company, et

al., Docket Nos. 50–280 and 50–281,

Surry Power Station, Units 1 and 2,

Surry County, Virginia

Date of application for amendments:

March 25, 1998.

Brief Description of amendments:

These amendments revise the Technical

Specifications to change certain

management titles. There is no change

in duties or responsibilities proposed.

Specifically, the Station Manager’s title

is changed to Site Vice President. The

title of Assistant Station Manager

Operations and Maintenance is changed

to Manager-Operations and

Maintenance. The title of Assistant

Station Manager Nuclear Safety and

Licensing is changed to Manager-Station

Safety and Licensing.

Date of issuance: June 19, 1998.

Effective date: June 19, 1998.

Amendment Nos.: 215 and 215.
36002 Federal Register / Vol. 63, No. 126 /Wednesday, July 1, 1998 / Notices

1The Company’s former name was ‘‘First Empire

State Corporation,’’ and the name change to ‘‘M&T

Bank Corporation’’ became effective on May 29,

1998. The Company filed the Form 8–A, effective

on May 27, 1998, and mentioned below, under the

Company’s old name.

Facility Operating License Nos. DPR–

32 and DPR–37: Amendments change

the Technical Specifications.

Date of initial notice in Federal

Register: May 6, 1998 (63 FR 25119) The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated June 19, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Swem Library, College of

William and Mary, Williamsburg,

Virginia 23185.

Wisconsin Public Service Corporation,

Docket No. 50–305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of application for amendment:

October 13, 1997, supplemented on

February 10, 1998.

Brief description of amendment: The

amendment involves miscellaneous

changes to the TS to (1) relocate

information to the Updated Safety

Analysis Report (USAR), (2) delete

redundant information, (3) incorporate

new references, (4) delete incorrect

references, (5) correct errors, and (6)

augment existing requirements.

Date of issuance: June 9, 1998.

Effective date: June 9, 1998.

Amendment No.: 137.

Facility Operating License No. DPR–

43: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: April 8, 1998 (63 FR 11926).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated June 9, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of Wisconsin,

Cofrin Library, 2420 Nicolet Drive,

Green Bay, WI 54311–7001

Yankee Atomic Electric Company,

Docket No. 50–29, Yankee Nuclear

Power Station, Franklin County,

Massachusetts

Date of application for amendment:

September 5, 1997 and March 30, 1998.

Brief description of amendment:

Revises Technical Specifications and

bases in order to allow loads of up to 80-

tons to travel over the spent fuel pool.

Date of issuance: June 17, 1998.

Effective date: June 17, 1998.

Amendment No.: 149.

Facility Opertating (Possession Only)

License No. DPR–3: Amendment revised

the Technical Specifications.

Date of initial notice in Federal

Register: October 22, 1997 (62 FR

54879) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

June 17, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Greenfield Community

College, 1 College Drive, Greenfield,

Massachusetts 01301

Dated at Rockville, Maryland, this 24th day

of June 1998.

For The Nuclear Regulatory Commission.

Elinor G. Adensam,

Acting Director, Division of Reactor Projects—

III/IV Office of Nuclear Reactor Regulation.

[FR Doc. 98–17352 Filed 6–30–98; 8:45 am]

BILLING CODE 7590–01–P

SECURITIES AND EXCHANGE

COMMISSION

Issuer Delisting; Notice of Application

to Withdraw From Listing and

Registration; (GST

Telecommunications, Inc., Common

Shares, Without Par Value) File No. 1–

12866

June 24, 1998.

GST Telecommunications, Inc.

(‘‘Company’’) has filed an application

with the Securities and Exchange

Commission (‘‘Commssion’’), pursuant

to Section 12(d) of the Securities

Exchange Act of 1934 (‘‘Act’’) and Rule

12d2–2(d) promulgated thereunder, to

withdraw the above specified Security

(‘‘Security’’) from listing and

registration on the American Stock

Exchange, Inc. (‘‘Amex’’ or

‘‘Exchange’’).

The reasons cited in the application

for withdrawing the Security from

listing and registration include the

following:

The Board of Directors of the

Company, at a meeting held on March

4 and 5, 1998, unanimously approved

resolutions to withdraw the Security

from listing on the Amex and, instead,

to list such Security on the Nasdaq

National Market (‘‘Nasdaq’’). The Board

of Directors, after lengthy deliberation,

determined that, since all other

telecommunications companies in the

Company’s industry segment have their

shares listed for trading on Nasdaq, it

would be in the best interest of the

Company and its shareholders to have

the Security listed on Nasdaq rather

than the Amex.

The Company has complied with

Amex Rule 18 by notifying the Amex of

its intention to withdraw its Security

from listing by letter dated March 27,

1998. The Amex replied by letter dated

April 7, 1998, advising it would not

interpose any objection to such

withdrawal. The Amex suspended

trading of the Security at the close of

business on Monday, April 13, 1998,

and the Security commenced trading on

Nasdaq on Tuesday, April 14, 1998. The

Company has filed an amended

registration statement on Form 8–A to

register the Security under Section 12(g)

of the Act.

By reason of Section 12(g) of the Act

and the rules and regulations

thereunder, the Company shall continue

to be obligated to file reports with the

Commission under Section 13 of the

Act.

Any interested person may, on or

before July 16, 1998, submit by letter to

the Secretary of the Securities and

Exchange Commission, 450 Fifth Street,

N.W., Washington, D.C. 20549, facts

bearing upon whether the application

has been made in accordance with the

rules of the Exchange and what terms,

if any, should be imposed by the

Commission for the protection of

investors. The Commission, based on

the information submitted to it, will

issue an order granting the application

after the date mentioned above, unless

the Commission determines to order a

hearing on the matter.

For the Commission, by the Division of

Market Regulation, pursuant to delegated

authority.

Jonathan G. Katz,

Secretary.

[FR Doc. 98–17436 Filed 6–30–98; 8:45 am]

BILLING CODE 8010–01–M

SECURITIES AND EXCHANGE

COMMISSION

Issuer Delisting; Notice of Application

to Withdraw From Listing and

Registration; (M&T Bank Corporation

(Formerly First Empire State

Corporation), Common Stock, $5.00

Par Value) File No. 1–9861

June 24, 1998.

M&T Bank Corporation 1

(‘‘Company’’) has filed an application

with the Securities and Exchange

Commission (‘‘Commission’’), pursuant

to Section 12(d) of the Securities

Exchange Act of 1934 (‘‘Act’’) and Rule

12d2–2(d) promulgated thereunder, to

withdraw the above specified security

(‘‘security’’) from listing and registration

on the American Stock Exchange, Inc.

(‘‘Amex’’ or ‘‘Exchange’’).