Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

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

[[Page 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.

[[Page 35988]]

(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

[[Page 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.

[[Page 35990]]

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.

[[Page 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 post-LOCA 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

[[Page 35992]]

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:

[[Page 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 de-activated. 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

[[Page 35994]]

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

[[Page 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 deg.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

[[Page 35996]]

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

[[Page 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.

[[Page 35998]]

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 deg.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

[[Page 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

[[Page 36000]]

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 +17\1/2\ 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.

[[Page 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 U\235\.

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.

[[Page 36002]]

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