47696 Federal Register / Vol. 62, No. 175 / Wednesday, September 10, 1997 / Notices

By October 10, 1997, 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 located at the HoustonLove Memorial Library, 212 W.

Burdeshaw Street, Post Office Box 1369,

Dothan, Alabama. 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 hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) The nature of the

petitioner’s right under the Act to be

made party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment 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 M.

Stanford Blanton, Esq., Balch and

Bingham, Post Office Box 306, 1710

Sixth Avenue North, Birmingham,

Alabama 35201, attorney for the

licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

presiding Atomic Safety and Licensing

Board that the petition and/or request

should be granted based upon a

balancing of the factors specified in 10

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

For further details with respect to this

action, see the application for

amendment dated September 3, 1997,

which is available for public inspection

at the Commission’s Public Document

Room, the Gelman Building, 2120 L

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

local public document room located at

the Houston-Love Memorial Library,

212 W. Burdeshaw Street, Post Office

Box 1369, Dothan, Alabama.

Dated at Rockville, Maryland, this 5th day

of September 1997.

For the Nuclear Regulatory Commission.

Jacob I. Zimmerman,

Project Manager, Project Directorate II–2,

Division of Reactor Projects—I/II, Office of

Nuclear Reactor Regulation.

[FR Doc. 97–23995 Filed 9–9–97; 8:45 am]

BILLING CODE 7590–01–U

NUCLEAR REGULATORY

COMMISSION

Biweekly Notice; Applications and

Amendments to Facility Operating

Licenses Involving No Significant

Hazards Considerations

I. Background

Pursuant to Pub. L. 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
Federal Register / Vol. 62, No. 175 / Wednesday, September 10, 1997 / Notices 47697

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 August 18,

1997, through August 28, 1997. The last

biweekly notice was published on

August 27, 1997 (62 FR 45452).

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 Freedom

of Information and Publications

Services, Office of Administration, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555–0001, and

should cite the publication date and

page number of this Federal Register

notice. Written comments may also be

delivered to Room 6D22, Two White

Flint North, 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 October 10, 1997, 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
47698 Federal Register / Vol. 62, No. 175 / Wednesday, September 10, 1997 / Notices

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.

Maine Yankee Atomic Power Company,

Docket No. 50–309, Maine Yankee

Atomic Power Station, Lincoln County,

Maine

Date of amendment request: August

15, 1997.

Description of amendment request:

The proposed amendment would revise

portions of the facility Technical

Specifications regarding facility staffing

and training requirements to power

operations. By letter dated August 7,

1997, the licensee certified permanent

cessation of power operations and

permanent removal of fuel from the

reactor vessel. By two letters both dated

August 15, 1997, the licensee has also

submitted a related ‘‘Request for

Exemption from Certain Requirements

of 10 CFR 50.54, Conditions of License,’’

and a ‘‘Request for Approval of the

Certified Fuel Handler Training and

Retraining Program.’’

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 does not:

1. Involve a significant increase in the

probability or consequence of an accident

previously evaluated.

The purpose of the proposed change is to

eliminate the requirements for licensed

operators and a licensed operator training

program and to replace those with certified

fuel handlers and a certified fuel handler

training and retraining program. Since the

plant has permanently ceased operation and

will be maintained in a defueled condition,

the range of accidents for which an operator

needs to be trained has significantly

diminished such that a training program of

the depth and breadth of that required by 10

CFR [Part] 55 is no longer needed. In lieu of

a 10 CFR [Part] 55 licensed operator training

program, a[n] NRC-approved certified fuel

handler training and retraining program will

be implemented. Since this training program

will adequately equip appropriate operations

personnel for fuel handling operations,

including responses to abnormal events/

accidents, there will be no increase in the

probability of these events occurring or in the

consequences of these events. The proposed

changes do not affect plant equipment or the

procedures for equipment operation or

response to abnormal events/accidents.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The purpose of this proposed change is to

eliminate the requirements for licensed

operators and a licensed operator training

program and to replace those with certified

fuel handlers and a certified fuel handler

training and retraining program. This change

ensures the qualifications of operations

personnel are commensurate with the tasks

to be performed and the conditions to be

responded to. This change does not affect

plant equipment or the procedures for

operating plant equipment and, therefore,

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 proposed change is to eliminate the

requirements for licensed operators and a

licensed operator training program to replace

those with certified fuel handlers and a

certified fuel handler training and retraining

program. This change ensures the

qualifications of the operations personnel are

commensurate with the tasks to be performed

and the conditions to be responded to. The

assumptions for a fuel handling accident in

the Fuel Building are not affected by the

proposed changes. Therefore, the proposed

amendment does not involve a 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: Wiscasset Public Library, High

Street, P.O. Box 367, Wiscasset, ME

04578.

Attorney for licensee: Mary Ann

Lynch, Esquire, Maine Yankee Atomic

Power Company, 329 Bath Road,

Brunswick, ME 04011.

NRC Acting Project Director: Ronald

B. Eaton.

Public Service Electric & Gas Company,

Docket Nos. 50–272 and 50–311, Salem

Nuclear Generating Station, Unit Nos. 1

and 2, Salem County, New Jersey

Date of amendment request: August 1,

1997.

Description of amendment request:

The amendments would change

Technical Specification Section 4.2.1 of

Appendix B to the licenses. The changes

include rewording of the section to

generically state that Public Service Gas

& Electric (PSE&G) will adhere to the

Section 7, Incidental Take Statement,

approved by the National Marine

Fisheries Service (NMFS). Removing the

specific requirements of this section

enables PSE&G to utilize relief granted

by the NMFS on a case-by-case basis.

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 changes are administrative in nature

and would in no way affect the initial

conditions, assumptions, or conclusions of

the Salem [Nuclear] Generating Station, Units

1 and 2, accident analyses. In addition, the

proposed changes would not affect the

operation or performance of any equipment

assumed in the accident analyses. Based on

the above information, we conclude that the

proposed changes would not significantly

increase 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 changes are administrative in nature

and would in no way impact or alter the

configuration or operation of the facilities

and would create no new modes of operation.

We therefore conclude that the proposed

changes would not create the possibility of a

new or different kind of accident.

3. The proposed change does not involve

a significant reduction in a margin of safety.

As indicated in the discussion of Criterion

1, the changes are administrative in nature

and would in no way affect plant or

equipment operation or the accident analysis.

We therefore conclude that the proposed

changes would 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
Federal Register / Vol. 62, No. 175 / Wednesday, September 10, 1997 / Notices 47699

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: Salem Free Public Library, 112

West Broadway, Salem, NJ 08079.

Attorney for licensee: Jeffrie J. Keenan,

Esquire, Nuclear Business Unit—N21,

P.O. Box 236, Hancocks Bridge, NJ

08038.

NRC Project Director: John F. Stolz.

Southern Nuclear Operating Company,

Inc, Docket Nos. 50–348 and 50–364,

Joseph M. Farley Nuclear Plant, Units 1

and 2, Houston County, Alabama

Date of amendments request: July 23,

1997.

Description of amendments request:

The proposed amendments would

revise the Technical Specifications

(TSs) by relocating the reactor coolant

system pressure and temperature limits

from the TSs to the proposed Pressure

Temperature Limits Report in

accordance with the guidance provided

by Generic Letter 96–03, ‘‘Relocation of

the Pressure Temperature Limit Curves

and Low Temperature Overpressure

Protection System Limits.’’

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 removal of the Reactor

Coolant System (RCS) pressure temperature

(P/T) limits from the Technical

Specifications (TSs) and relocation to the

proposed Pressure Temperature Limits

Report (PTLR) in accordance with the

guidance provided by Generic Letter (GL) 96–

03 is administrative in that the requirements

for the P/T limits are unchanged. The P/T

limits proposed for inclusion in the PTLR are

based on the fluence associated with 2775

MW [megawatts] thermal power and

operation through 36 effective full power

years (EFPY). GL 96–03 requires that the P/

T limits be generated in accordance with the

requirements of 10 CFR [Part] 50,

Appendices G and H, documented in an

NRC-approved topical report incorporated by

reference in the TSs. Accordingly, the

proposed curves have been generated using

the NRC-approved methods described in

WCAP–14040–NP–A, Revision 2, and meet

the requirements of 10 CFR [Part] 50,

Appendices G and H. TS 3.4.10.1 will

continue to require that the RCS pressure and

temperature be limited in accordance with

the limits specified in the PTLR. The NRCapproved methodology for generating the P/

T limit, WCAP–14040–NP–A, Revision 2,

will be specified in TS 6.9.1.15 and NRC

approval will be required in the form of a TS

Amendment prior to changing the

methodology. Use of P/T limit curves

generated using the NRC-approved methods

described in WCAP–14040–NP–A, Revision

2, as specified in TS 6.9.1.15, will provide

additional protection for the integrity of the

reactor vessel, thereby assuring that the

reactor vessel is capable of providing its

function as a radiological barrier.

TS 3.4.10.3 for Farley Nuclear Plant (FNP)

Unit 1 and Unit 2 provides the operability

requirements for RCS low temperature

overpressure protection (LTOP). Specifically,

TS 3.4.10.3 requires that two residual heat

removal (RHR) system suction relief valves

(RHRRVs) be operable or that the RCS be

vented at RCS cold leg temperatures less than

or equal to 310°F. GL 96–03 recognizes that

RHRRVs do not have variable pressure lift

setpoints and states that those plants that rely

on the RHRRVs for LTOP should continue to

address the LTOP requirements in the TS.

Consistent with GL 96–03, the Farley Unit 1

and Unit 2 requirements for LTOP will be

retained in TS 3.4.10.3.

Based on the above evaluation, the

proposed changes are administrative in

nature and do not involve a significant

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.

As stated above, the proposed changes to

remove the RCS P/T limits from the TSs and

relocate them to the proposed PTLR is an

administrative change. Consistent with the

guidance provided by GL 96–03, the

proposed P/T limits contained in the

proposed PTLR meet the requirements of 10

CFR [Part] 50, Appendices G and H, and were

generated using the NRC-approved methods

described in WCAP–14040–NP–A, Revision

2. The proposed changes do not result in a

physical change to the plant or add any new

or different operating requirements on plant

systems, structures, or components with the

exception of limiting the number of operating

RCPs [reactor coolant pumps] at RCS

temperatures below 110°F. Limiting the

number of operating RCPs below 110°F

results in a reduction in the [delta]P between

the reactor vessel beltline and the RHRRVs,

thereby providing additional margin to limits

of Appendix G. Provisions are made to allow

the start of a second RCP at temperatures

below 110°F in order to secure the pump that

was originally operating without interrupting

RCS flow. The LTOP enable temperature

exceeds the minimum LTOP enable

temperature determined using the NRCapproved methods described in WCAP–

14040–NP–A, Rev. 2, thereby providing

additional assurance that the LTOP system

will be available to protect the RCS in the

event of an overpressure transient at RCS

temperatures at or below 310°F. Using the

methods contained in WCAP–14040–NP–A,

Rev. 2, the minimum boltup temperature for

the reactor vessel flange region is 60°F which

is less than the design limits of the fuel

cladding. Administrative controls require a

minimum RCS temperature of 68°F when

fuel is loaded in the reactor vessel to protect

against brittle failure of the fuel cladding,

and also require that the component cooling

water (CCW) temperature be maintained

between 60°F and 105°F during refueling

operations, thus reducing the potential for

the RCS temperature to be less than the

minimum boltup temperature specified in

the proposed PTLRs.

As stated in the above response,

implementation of the proposed changes do

not result in a significant increase in the

probability of a new or different accident

(i.e., loss of reactor vessel integrity). The RCS

P/T limits will continue to meet the

requirements of 10 CFR [Part] 50,

Appendices G and H, and will be generated

in accordance with the NRC approved

methodology described in WCAP–14040–

NP–A, Rev. 2. Therefore, the proposed

changes do not result in a significant increase

in the possibility of a new or different

accident from any previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The margin of safety is not affected by the

removal of the RCS P/T limits from the TSs

and relocating them to the proposed PTLR.

The RCS P/T limits will continue to meet the

requirements of 10 CFR [Part] 50,

Appendices G and H. To provide additional

assurance that the P/T limits continue to

meet the requirements of Appendices G and

H, TS 6.9.1.15 will require the use of the

NRC-approved methodology described in

WCAP–14040–NP–A, Rev. 2, to generate P/T

limits. The RCS LTOP requirements will be

retained in TS 3.4.10.3 due to use of the

RHRRVs for LTOP, consistent with the

guidance provided by GL 96–03. The LTOP

enable temperature exceeds the LTOP enable

temperature determined in accordance with

the NRC-approved methodology, thus

protecting the RCS in the event of a low

temperature overpressure transient over a

broader range of temperatures than required

by WCAP–14040–NP–A, Rev. 2.

Administrative procedures preclude

operation of the RCS at temperatures below

the minimum boltup temperature for the

reactor vessel head, thus precluding the

possibility of tensioning the reactor vessel

head at RCS temperatures below the

minimum boltup temperature. Operation of

the plant in accordance with the RCS P/T

limits specified in the PTLR and continued

operation of the LTOP system in accordance

with TS 3.4.10.3 will continue to meet the

requirements of 10 CFR [Part] 50,

Appendices G and H, and will therefore,

assure that a margin of safety is not

significantly decreased as the result of the

proposed changes.

Based on the preceding analysis, SNC

[Southern Nuclear Operating Company, Inc.]

has determined that removal of the RCS P/

T limits from the TS and relocation to the

proposed PTLR will not significantly

increase the probability or consequences of

an accident previously evaluated, create the

possibility of a new or different kind of

accident from any accident previously

evaluated, or involve a significant reduction

in a margin of safety. SNC therefore

concludes that the proposed change meets

the requirements of 10 CFR 50.92(c) and does
47700 Federal Register / Vol. 62, No. 175 / Wednesday, September 10, 1997 / Notices

not involve a significant hazards

consideration.

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:

Houston-Love Memorial Library, 212 W.

Burdeshaw Street, Post Office Box 1369,

Dothan, Alabama 36302. Attorney for

licensee: M. Stanford Blanton, Esq.,

Balch and Bingham, Post Office Box

306, 1710 Sixth Avenue North,

Birmingham, Alabama 35201. NRC

Project Director: Herbert N. Berkow.

Tennessee Valley Authority, Docket No.

50–390 Watts Bar Nuclear Plant, Unit 1,

Rhea County, Tennessee

Date of amendment request: June 20,

1997 (TS–97–004).

Description of amendment request:

The proposed amendment would be an

administrative change that would revise

the analytical methodology used to

determine the low temperature

overpressure protection (LTOP) event

heatup and cooldown curves. This

revised methodology would be

incorporated by reference in the Watts

Bar Nuclear Plant (WBN), Unit 1

Technical Specification (TS) 5.9,

‘‘Reporting Requirements,’’ Section

5.9.6, ‘‘Reactor Coolant System (RCS)

Pressure and Temperature Limits Report

(PTLR),’’ upon approval for use by the

U.S. Nuclear Regulatory Commission

(NRC). The revised methodology

extends the current LTOP requirements

through the end of 7 effective full power

years (EFPY). The only technical change

being proposed is the substitution of the

7 EFPY American Society of Mechanical

Engineering (ASME), Appendix G,

heatup and cooldown curves adjusted

by ASME Code Case N–514, ‘‘Low

Temperature Overpressure Protection’’

in place of the current 1.5 EFPY curves

as the bounding curves for the LTOP

setpoints. This change will not impact

the current 10 CFR 50, Appendix G,

pressure/temperature (P/T) limit curves

used for heatup and cooldown that are

based on 7 EFPY.

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 Nuclear Regulatory Commission has

provided standards for determining whether

a significant hazards consideration exists (10

CFR 50.92). A proposed amendment to an

operating license for a facility involves no

significant hazards consideration if 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.

Each standard is discussed below for the

proposed amendment.

(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 LTOP setpoints (identified as the cold

overpressure mitigation system (COMS) for

WBN), adjusted for instrument inaccuracy,

pressure differential, and setpoint overshoot

by the scaling and setpoint documents

(SSDs), ensure that the 10 CFR 50, Appendix

G P/T [pressure and temperature] limits

based on 7 EFPY are not exceeded by more

than the provisions of ASME Code Case N–

514, and therefore, ensure that the RCS

integrity is maintained.

The change does not modify the RCS

pressure boundary, nor make any physical

changes to the facility design, material,

construction standards, or setpoints. The

LTOP enabling temperature based on TS

3.4.12, ‘‘Cold Overpressure Mitigation

System (COMS),’’ is [less than or equal to]

350 degrees F and is more conservative than

a value of 271.1 degrees F (RTNDT + 90

degrees F) based on 7 EFPY. This

temperature would be acceptable based on

NRC Branch Technical Position-Reactor

Systems Branch (BTP–RSB)–5.2,

‘‘Overpressurization Protection of

Pressurized Water Reactors While Operating

at Low Temperatures.’’ The LTOP enabling

temperature remains unchanged by this

proposed amendment. The probability of a

LTOP event occurring is independent of the

P/T limits for the RCS pressure boundary;

therefore, the probability of an LTOP event

occurring remains unchanged.

The calculation of the P/T limits in

accordance with approved regulatory

methods based on 7 EFPY provides assurance

that reactor pressure vessel fracture

toughness requirements are met and the

integrity of the RCS pressure boundary is

maintained. LTOP setpoints based on 1.5

EFPY P/T limits have provided margin such

that a pressure excursion exceeding the 7

EFPY limits would not exceed the 1.5 EFPY

limits. This margin between the 7 EFPY

curves and the LTOP setpoints is maintained

by changing the bounding curves for the

LTOP setpoints to 7 EFPY curves adjusted by

the provisions of ASME Code Case N–514.

The only technical change being made is the

bounding curves which provide the basis for

the current LTOP setpoints.

The use of theoretical fluence for

generating the P/T curves to be used for the

first 7 EFPY is appropriate and was

submitted July 31, 1995, with the WBN Unit

1 PTLR, Revision 4 and WCAP–13829,

Revision 2, ‘‘Heatup and Cooldown Limit

Curves for Normal Operation for Watts Bar

Unit 1.’’ The present 7 EFPY curves are

generated using a theoretical value for

fluence calculated by Westinghouse in

accordance with NRC approved methodology

since WBN had no surveillance capsule data

available at the time of plant startup. This

value for fluence is conservative, and the

actual fluence to the intermediate shell

forging (the controlling beltline material) is

expected to be significantly less than the

theoretical value used to generate the initial

7 EFPY curves since WBN is transitioning to

a low-leakage core. The LTOP bounding

curves are based on 7 EFPY curves adjusted

in accordance with ASME Code Case N–514

which were generated using the same

theoretical fluence as used for the P/T curves.

The significance of using the theoretical

value of fluence in generating these curves is

the additional margin that exists between the

7 EFPY theoretical curves and curves that

would be generated using actual fluence

values from capsule data. This additional

margin reduces the significance of changing

the LTOP basis from the 1.5 EFPY curves to

the 7 EFPY curves adjusted for ASME Code

Case N–514.

This change does not adversely affect the

integrity of the RCS such that its function in

the control of radiological consequences is

affected. In addition, the change does not

affect any fission barrier. The change does

not degrade or prevent the LTOP power

operated relief valves (PORVs) or other safety

related systems from responding to accidents

described in Chapter 15 of the Final Safety

Analysis Report (FSAR). In addition, the

change does not alter any assumptions

previously made in the radiological

consequences of an accident described in the

FSAR. Therefore, the consequences of an

accident previously evaluated in the FSAR

are not increased. Thus, the operation of

WBN Unit 1 in accordance with this

proposed amendment does 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 amendment would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The Appendix G P/T limitations were

prepared using methods derived from the

ASME Boiler and Pressure Vessel Code

Section III and the criteria set forth in NRC

Regulatory Standard Review Plan 5.3.2,

‘‘Pressure-Temperature Limits.’’ The use of

ASME Code Case N–514 and the theoretical

fluence value for 7 EFPY does not modify the

RCS pressure boundary, nor make any

physical changes to the LTOP setpoints or

system design. The proposed change was

prepared in accordance with regulatory

requirements and provides evaluation of

LTOP events based on 7 EFPY theoretical

fluence which is more limiting than actual

expected neutron exposure for that same

period.

This proposed change is an administrative

change which incorporates by reference the

use of an NRC approved methodology;

therefore, the change does not cause the

initiation of any accident nor create any new

creditable limiting failure for safety-related
Federal Register / Vol. 62, No. 175 / Wednesday, September 10, 1997 / Notices 47701

systems and components. The change does

not result in an event previously deemed

incredible being made credible. As such, it

does not create the possibility of an accident

different than any evaluated in the FSAR.

The change does not have any effect on the

ability of the safety-related systems to

perform their intended safety functions. The

change does not create failure modes that

could adversely impact safety-related

equipment. Therefore, it will not create the

possibility of a malfunction of equipment

important to safety different than previously

evaluated in the FSAR. Thus, the proposed

amendment does not create the possibility of

a new or different kind of accident from any

accident previously evaluated.

Therefore, the proposed change does not

create the possibility of a new or different

type 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 margin of

safety.

The 10 CFR 50, Appendix G P/T

limitations were prepared using methods

derived from ASME Section III and criteria

set forth in NRC Regulatory Standard Review

Plan 5.3.2. These documents along with the

calculational limitations specified in 10 CFR

50.61 are an acceptable method for

implementing the requirements of 10 CFR 50

Appendices G and H. Inherent conservatisms

in the P/T limits resulting from these

documents include:

a. An assumed defect in the reactor vessel

wall with a depth equal to 1⁄4 of the thickness

(T) of the vessel wall and a length equal to

11⁄2 times the thickness of the vessel wall.

b. Assumed reference flaw oriented in both

longitudinal and circumferential directions

and limiting material property. At WBN, the

only weld in the core region is oriented in

the circumferential direction.

c. A factor of safety of 2 is applied to the

membrane stress intensity factor.

d. The limiting toughness is based upon a

reference value (KIM) which is the lower

bound of the dynamic crack initiation and

arrest toughness.

e. A 2-sigma margin term is applied in

determining the adjusted reference

temperature (ART) that is used in calculating

the limiting toughness.

Beyond the conservatisms described above,

WBN has the following additional margin:

a. The value of fluence used in the

calculation of the WBN Unit 1 Appendix G

P/T limits is a theoretical value calculated by

NRC approved methodology.

b. The ART for 7 EFPY is based on the

theoretical value for fluence and therefore is

conservative. The LTOP enabling

temperature of [less than or equal to] 350

degrees F in accordance with TS 3.4.12 is

conservative with respect to (RTNDT + 90

degrees F) which based on an ART of 181.1

degrees F would equal 271.1 degrees F. An

enabling temperature of (RTNDT + 90 degrees

F) is based on NRC BTP–RSB 5.2.

The ASME Working Group for Operating

Plant Criteria developed Code Case N–514 as

an alternative methodology to the safety

margin requirements of Appendix G to 10

CFR 50. The Code Case provides criteria to

determine pressure limits during LTOP

events that avoid certain operational

restrictions, provide adequate margins

against failure of the reactor vessel, and

reduce the potential for unnecessary

activation of the relief valves used for LTOP.

Specifically, the N–514 Code Case allows

determination of the LTOP setpoints such

that for LTOP events the maximum pressure

in the reactor vessel would not exceed 110%

of the P/T limits of the existing ASME

Appendix G curves, and redefines the

enabling temperature as a coolant

temperature less than 200 degrees F or a

reactor vessel metal temperature less than

RTNDT + 50 degrees F. Code Case N–514 has

been approved by the ASME Code Committee

and its content has been incorporated in

Appendix G of ASME Section XI and

published in the 1993 Addenda and 1995

Edition. Code Case N–514 has not been

approved for use in Regulatory Guide 1.147,

‘‘Inservice Inspection Code Case

Acceptability, ASME Section XI;’’ however, it

has been included in the Draft Regulatory

Guide 1.147 (Task DG–1050) which is

currently out for public review and comment.

As stated above, WBN Unit 1 uses Appendix

G for the P/T limits for plant operation and

an LTOP enabling temperature greater than

RTNDT + 90 degrees F which is more

conservative than the alternative

methodology contained in Code Case N–514.

The need for implementation of Code Case

N–514 at WBN involves the avoidance of

certain operational restrictions associated

with low temperature operation of the plant.

Use of Appendix G P/T limits to determine

the PORV setpoints would result in pressure

setpoints within the operating window;

consequently, no margin would be available

for normal operating pressure surges.

Therefore, operating with these limits could

result an unnecessary challenge to the PORVs

and cavitation of the reactor coolant pumps

(RCP) during normal operation. Additionally,

the need to raise the RCS inventory by

external heating methods to a temperature

high enough to avoid PORV activation when

starting a RCP from a RCS cold shutdown

condition could result in undesirable thermal

transients in the RCS.

Utilizing the methodology set forth in the

ASME Boiler and Pressure Vessel Code

Section XI, Appendix G, which includes the

provisions of Code Case N–514, NRC

Regulatory Standard Review Plan 5.3.2, 10

CFR 50.61, and 10 CFR 50, Appendices G

and H with the above additional margins

ensures that proper limits and conservative

safety factors are maintained. Thus the

proposed change does not significantly

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: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga,

TN 37402.

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 10H,

Knoxville, Tennessee 37902.

NRC Project Director: Frederick J.

Hebdon.

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.

Commonwealth Edison Company,

Docket Nos. 50–237 and 50–249,

Dresden Nuclear Power Station, Units 2

and 3, Grundy County, Illinois

Date of application for amendments:

March 18, 1997.

Brief description of amendments: The

amendments revise the Technical

Specifications (TS) to increase the High

Pressure Coolant Injection (HPCI)
47702 Federal Register / Vol. 62, No. 175 / Wednesday, September 10, 1997 / Notices

system low pressure isolation setpoint

from greater than 80 psig to greater than

100 psig.

Date of issuance: August 21, 1997.

Effective date: Immediately, to be

implemented within 30 days.

Amendment Nos.: 161, 156.

Facility Operating License Nos. DPR–

19 and DPR–25: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: April 9, 1997 (62 FR 17228).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated August 21,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Morris Area Public Library

District, 604 Liberty Street, Morris,

Illinois 60450.

Duke Power Company, 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:

June 12, 1997.

Brief description of amendments: The

amendments change the name ‘‘Duke

Power Company’’ to ‘‘Duke Energy

Corporation’’ in the Catawba operating

licenses and appendices as a result of

Duke Power Company’s recent name

change.

Date of issuance: August 22, 1997.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment Nos.: 161 and 153.

Facility Operating License Nos. NPF–

35 and NPF–52: Amendments revised

the Facility Operating Licenses.

Date of initial notice in Federal

Register: July 2, 1997 (62 FR 35848).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated August 22,

1997, and an Environmental Assessment

dated July 31, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: York County Library, 138 East

Black Street, Rock Hill, South Carolina

29730.

Duke Power Company, Docket Nos. 50–

369 and 50–370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg

County, North Carolina

Date of application for amendments:

June 12, 1997.

Brief description of amendments: The

amendments change the name ‘‘Duke

Power Company’’ to ‘‘Duke Energy

Corporation’’ in the McGuire operating

licenses and appendices as a result of

Duke Power Company’s recent name

change.

Date of issuance: August 26, 1997.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment Nos.: 176 and 158.

Facility Operating License Nos. NPF–

9 and NPF–17: Amendments revised the

Licenses.

Date of initial notice in Federal

Register: July 2, 1997 (62 FR 35848).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated August 26,

1997. An Environmental Assessment

was issued and dated August 15, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: J. Murrey Atkins Library,

University of North Carolina at

Charlotte, 9201 University City

Boulevard, Charlotte, North Carolina.

Entergy Gulf States, Inc., Cajun Electric

Power Cooperative, and Entergy

Operations, Inc., Docket No. 50–458,

River Bend Station, Unit 1, West

Feliciana Parish, Louisiana

Date of amendment request: October

26, 1995 and supplemented by letters

dated April 7 and July 30, 1997.

Brief description of amendment: The

amendment revised the technical

specifications for 16 editorial changes

and deletes the reuirement for a

program to prevent and detect Asiatic

Clams (Corbicula) in the service water

system (SWS). The Corbicula program is

no longer needed because the facility

has been modified and SWS no longer

takes water from the Mississippi River;

source of the larvae and infestation.

Date of issuance: August 26, 1997.

Effective date: August 26, 1997.

Amendment No.: 95.

Facility Operating License No. NPF–

47: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: December 6, 1995 (60 FR

62492).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated August 26,

1997.

No significant hazards consideration

comments received. No.

Local Public Document Room

location: Government Documents

Department, Louisiana State University,

Baton Rouge, LA 70803.

Entergy Gulf States, Inc., Cajun Electric

Power Cooperative, and Entergy

Operations, Inc., Docket No. 50–458,

River Bend Station, Unit 1, West

Feliciana Parish, Louisiana

Date of amendment request:

November 15, 1996, as supplemented

May 9 and August 15, 1997.

Brief description of amendment: The

amendment revises the technical

specifications to increase the two

recirculation loop Minimum Critical

Power Ratio (MCPR) from 1.07 to 1.10

and the single recirculation loop MCPR

limit from 1.08 to 1.12.

Date of issuance: August 26, 1997.

Effective date: August 26, 1997.

Amendment No.: 96.

Facility Operating License No. NPF–

47: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: January 2, 1997 (62 FR 127).

The May 9 and August 15, 1997,

submittal provided clarifying

information that did not change the

initial no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated August 26,

1997.

No significant hazards consideration

comments received. No.

Local Public Document Room

location: Government Documents

Department, Louisiana State University,

Baton Rouge, LA 70803.

Entergy Gulf States, Inc., Cajun Electric

Power Cooperative, and Entergy

Operations, Inc., Docket No. 50–458,

River Bend Station, Unit 1, West

Feliciana Parish, Louisiana

Date of amendment request: January

20, 1997 as supplemented by letter

dated July 7, 1997.

Brief description of amendment: The

amendment revises the technical

specifications to allow the use of flow

control spectral shift strategies to

increase cycle energy. The revision is

based on a Maximum Extended Load

Line Limit (MELLL) analysis for the

River Bend Station.

Date of issuance: August 26, 1997.

Effective date: August 26, 1997.

Amendment No.: 97.

Facility Operating License No. NPF–

47: The amendment revised the

Technical Specifications/operating

license.

Date of initial notice in Federal

Register: February 26, 1997 (62 CFR

8799).

The July 7, 1997 submittal provided

clarifying information and did not

change the initial no significant hazards

consideration determination.
Federal Register / Vol. 62, No. 175 / Wednesday, September 10, 1997 / Notices 47703

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated August 26,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Government Documents

Department, Louisiana State University,

Baton Rouge, LA 70803.

Entergy Gulf States, Inc., Cajun Electric

Power Cooperative, and Entergy

Operations, Inc., Docket No. 50–458,

River Bend Station, Unit 1, West

Feliciana Parish, Louisiana

Date of amendment request:

November 6, 1996, as supplemented by

letter dated July 31, 1997.

Brief description of amendment: The

amendment revises the Technical

Specifications to delete the requirement

for the Penetration Valve Leakage

Control System. The licensee requested

deferal of the proposal to increase the

allowed leakage by main steam isolation

valves and to delete the requirement for

the Main Steam Positive Leakage

Control System.

Date of issuance: August 26, 1997.

Effective date: August 26, 1997.

Amendment No.: 98.

Facility Operating License No. NPF–

47: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: January 2, 1997 (62 FR 125).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated August 26,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Government Documents

Department, Louisiana State University,

Baton Rouge, LA 70803.

Entergy Operations, Inc., System Energy

Resources, Inc., South Mississippi

Electric Power Association, and

Mississippi Power & Light Company,

Docket No. 50–416, Grand Gulf Nuclear

Station, Unit 1, Claiborne County,

Mississippi

Date of application for amendment:

October 22, 1996, as supplemented by

letter dated June 26, 1997.

Brief description of amendment: The

amendment revises Figure 3.4.11–1,

‘‘Minimum Reactor Vessel Metal

Temperature vs. Reactor Vessel

Pressure,’’ in Limiting Condition for

Operation 3.4.11, ‘‘RCS [Reactor Coolant

System] Pressure and Temperature (P/T)

Limits,’’ of the Technical Specifications.

The previous figure was only up to 10

Effective Full Power Years (EFPYs) and

this amendment revises the figure up to

32 EFPYs. There are now five curves of

Figure 3.4.11–1 for five different EFPY

periods: up to 16, 16 to 20, 20 to 24, 24

to 28, and 28 to 32. The licensee

submitted two sets of curves. The first

set replaced TS Figure 3.4.11–1. The

second set were duplicates of the first

set except the second set also contained

detailed information used in

development of the curves and would

be included in the next update of the

Updated Final Safety Analysis Report.

There were also minor additions to

Surveillance Requirements (SRs)

3.4.11.1 and 3.4.11.2 to have the SRs

reference the ‘‘applicable Figure 3.4.11–

1 based on the current effective full

power year (EFPY).’’

Date of issuance: August 27, 1997.

Effective date: August 27, 1997.

Amendment No: 132.

Facility Operating License No. NPF–

29: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: February 26, 1997 (62 FR

8797).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated August 27,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Judge George W. Armstrong

Library, 220 S. Commerce Street,

Natchez, MS 39120.

Entergy Operations, Inc., Docket No. 50–

382, Waterford Steam Electric Station,

Unit 3, St. Charles Parish, Louisiana

Date of amendment request: April 11,

1997.

Brief description of amendment: The

amendment modifies Technical

Specifications 3.3.3.7.3, and

Surveillance Requirements (SR)

4.3.3.7.3 for the broad range gas

detection system. Also it includes some

changes to the Bases in Section 3/4.3.3.7

to incorporate information associated

with the proposed modifications. The

licensee is planning to replace the

existing toxic gas monitors in the system

with a new, more advanced gas

monitors.

Date of issuance: August 19, 1997.

Effective date: August 19, 1997, to be

implemented within 90 days.

Amendment No.: 133.

Facility Operating License No. NPF–

38: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: May 7, 1997 (62 FR 24987)

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated August 19,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, LA 70122.

Georgia Power Company, Oglethorpe

Power Corporation, Municipal Electric

Authority of Georgia, City of Dalton,

Georgia, Docket No. 50–321, Edwin I.

Hatch Nuclear Plant, Unit 1, Appling

County, Georgia

Date of application for amendment:

April 29, 1997, as supplemented by

letter dated May 28, 1997.

Brief description of amendment: The

amendment revises Hatch Unit 1 reactor

vessel pressure and temperature limits

to reflect data collected from the

material sample recovered during the

March 1996 Unit 1 outage.

Date of issuance: August 19, 1997.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 207.

Facility Operating License No. DPR–

57: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 16, 1997 (62 FR 38138).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated August 19,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Appling County Public

Library, 301 City Hall Drive, Baxley,

Georgia 31513.

GPU Nuclear Corporation, et al., Docket

No. 50–219, Oyster Creek Nuclear

Generating Station, Ocean County, New

Jersey

Date of application for amendment:

October 4, 1996, as supplemented June

10 and August 15, 1997 (TSCR 250).

Brief description of amendment: The

amendment changes the Safety Limit

Minimum Critical Power Ratio and as a

result, the operating Minimum Critical

Power Ratio. The amendment also

capitalized certain definitions and

provided a uniform type font for

Sections 2.1 and 3.10.

Date of Issuance: August 26, 1997.

Effective date: August 26, 1997, with

full implementation within 30 days.

Amendment No.: 192.

Facility Operating License No. DPR–

16. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: November 6, 1996 (61 FR

57484).

The Commission’s related evaluation

of this amendment is contained in a
47704 Federal Register / Vol. 62, No. 175 / Wednesday, September 10, 1997 / Notices

Safety Evaluation dated August 26,

1997.

The June 10 and August 15, 1997,

submittals provided clarifying

information that did not alter the staff’s

initial proposed no significant hazards

considerations determination.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Ocean County Library,

Reference Department, 101 Washington

Street, Toms River, NJ 08753.

GPU Nuclear Corporation, et al., Docket

No. 50–289, Three Mile Island Nuclear

Station, Unit No. 1, Dauphin County,

Pennsylvania

Date of application for amendment:

April 21, 1997, as supplemented July

17, 1997.

Brief description of amendment: The

amendment reduces the required

volume of borated water in each core

flood tank from 1040 ft 3 to 940 ft 3,

reduces the required high pressure

injection pump flowrate from 500

gallons per minute (gpm) to 431 gpm,

and deletes the local manual valve

operability option for decay heat system

valves DH–V–6A and DH–V–6B.

Date of issuance: August 27, 1997.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 203.

Facility Operating License No. DPR–

50. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: May 21, 1997 (62 FR 27795).

The July 17, 1997, submittal provided

clarifying information that did not alter

the initial no significant hazards

determination. The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

August 27, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Law/Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Walnut

Street and Commonwealth Avenue, Box

1601, Harrisburg, PA 17105.

Houston Lighting & Power Company,

City Public Service Board of San

Antonio, Central Power and Light

Company, City of Austin, Texas, Docket

Nos. 50–498 and 50–499, South Texas

Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request: April 22,

1997.

Brief description of amendments: The

proposed amendment revised Technical

Specifications 5.3.1, Fuel Assemblies,

and 6.9.1.6, Core Operating Limits

Report, to allow use of an alternate

zirconium-based fuel cladding, ZIRLO,

and limited substitution of fuel rods by

ZIRLO filler rods.

Date of issuance: August 19, 1997.

Effective date: August 19, 1997.

Amendment Nos.: Unit 1—

Amendment No. 89; Unit 2—

Amendment No. 76.

Facility Operating License Nos. NPF–

76 and NPF–80: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: May 21, 1997 (62 FR 27795).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated August 19,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Wharton County Junior

College, J. M. Hodges Learning Center,

911 Boling Highway, Wharton, TX

77488.

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:

November 20, 1995.

Brief description of amendment: The

amendment changes the Technical

Specifications (TSs) by providing

clarifications to the applicability and

action statements in TS Table 3.3–12

relating to the Steam Generator

Blowdown Monitor and the Condensate

Polishing Facility Waste Neutralizing

Sump radiation monitor.

Date of issuance: August 26, 1997.

Effective date: As of the date of

issuance to be implemented within 60

days.

Amendment No.: 207.

Facility Operating License No. DPR–

65: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: December 20, 1995 (60 FR

65683).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated August 26,

1997.

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, CT 06360, and the Waterford

Library, ATTN: Vince Juliano, 49 Rope

Ferry Road, Waterford, CT 06385.

Northeast Nuclear Energy Company, et

al., Docket No. 50–423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of application for amendment:

May 1, 1997.

Brief description of amendment:

Technical Specifications 3/4.8.2.2 and

3/4.8.3.2 specify which electrical power

systems are required to be operable in

Modes 5 and 6. The amendment

clarifies the requirements by identifying

the specific equipment required and

their alignments in Modes 5 and 6.

Date of issuance: August 21, 1997.

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment No.: 146.

Facility Operating License No. NPF–

49: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 4, 1997 (62 FR 30637).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated August 21,

1997.

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 06360, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut 06385.

Northeast Nuclear Energy Company, et

al., Docket No. 50–423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of application for amendment:

May 5, 1997.

Brief description of amendment:

Technical Specification Surveillance

4.5.2.b.1 requires that the emergency

core cooling system piping be verified

full of water at least once per 31 days.

The amendment revises the surveillance

to exempt the operating charging

pump(s) and associated piping from the

requirement to be verified full of water

and moves the description of the

verification method from the

surveillance to the Bases section.

Date of issuance: August 28, 1997.

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment No.: 147.

Facility Operating License No. NPF–

49: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 4, 1997 (62 FR 30638).

The Commission’s related evaluation

of the amendment is contained in a
Federal Register / Vol. 62, No. 175 / Wednesday, September 10, 1997 / Notices 47705

Safety Evaluation dated August 28,

1997.

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 06360, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope.

PECO Energy Company, Public Service

Electric and Gas Company Delmarva

Power and Light Company, and Atlantic

City Electric Company, Docket Nos. 50–

277 and 50–278, Peach Bottom Atomic

Power Station, Unit Nos. 2 and 3, York

County, Pennsylvania

Date of application for amendments:

March 31, 1997, as supplemented by

letter dated June 25, 1997.

Brief description of amendments:

These amendments extend the APRM

flow bias instrumentation surveillance

interval from 18 months to 24 months.

This will eliminate the need to perform

on-line APRM surveillance testing,

which requires plant operators to place

an operating unit in a half scram

configuration.

Date of issuance: August 19, 1997.

Effective date: Units 2 and 3 effective

as of date of issuance.

Amendments Nos.: 219 and 222.

Facility Operating License Nos. DPR–

44 and DPR–56: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: May 7, 1997 (62 FR 24988).

The supplemental letter provided

clarifying information that did not

change the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated August 19,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Education

Building, Walnut Street and

Commonwealth Avenue, Box 1601,

Harrisburg, PA 17105.

Southern Nuclear Power Company, Inc.,

Georgia Power Company, Oglethorpe

Power Corporation, Municipal Electric

Authority of Georgia, City of Dalton,

Georgia, Docket Nos. 50–424 and 50–

425, Vogtle Electric Generating Plant,

Units 1 and 2, Burke County, Georgia

Date of application for amendments:

June 13, 1997, as supplemented by letter

dated July 18, 1997.

Brief description of amendments: The

amendments revise the pressurizer

safety relief valve setpoint specified in

Technical Specification 3.4.10.

Date of issuance: August 26, 1997.

Effective date: As of the date of

issuance to be implemented for Unit 1

prior to or after initial entry into Mode

3 (in accordance with the provisions of

the note to the Applicability for LCO

3.4.10) following the fall 1997 refueling

outage; for Unit 2 prior to or after initial

entry into Mode 3 (in accordance with

the provisions of the note to the

Applicability for LCO 3.4.10) following

the spring 1998 refueling outage.

Amendment Nos.: 98 and 76.

Facility Operating License Nos. NPF–

68 and NPF–81: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: July 16, 1997 (62 FR 38139).

The supplemental material did not

change the no significant hazards

finding or expand the scope of the

Federal Register notice.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated August 26,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Burke County Library, 412

Fourth Street, Waynesboro, Georgia

30830.

TU Electric Company, Docket Nos. 50–

445 and 50–446, Comanche Peak Steam

Electric Station, Unit Nos. 1 and 2,

Somervell County, Texas

Date of amendment request:

December 7, 1994 (TXX–94326), as

supplemented by letter dated June 21,

1996 (TXX–96384).

Brief description of amendments:

These changes revised Section 3.7.1.5 of

the Technical Specification to increase

the Allowed Outage Time for one

inoperable Main Steam Isolation Valve

(MSIV) while in Mode 1, and to clarify

requirements related to inoperable

MSIVs while in Modes 2 and 3.

Date of issuance: August 18, 1997.

Effective date: August 18, 1997, to be

implemented within 60 days.

Amendment Nos.: 54 and 40.

Facility Operating License Nos. NPF–

87 and NPF–89: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: February 1, 1995 (60 FR 6312).

The additional information contained

in the supplemental letter dated June

21, 1996, was clarifying in nature and

thus, within the scope of the initial

notice and did not affect the staff’s

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated August 18,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of Texas at

Arlington Library, Government

Publications/Maps, 702 College, P.O.

Box 19497, Arlington, TX 76019.

Wisconsin Electric Power Company,

Docket Nos. 50–266 and 50–301, Point

Beach Nuclear Plant, Unit Nos. 1 and 2,

Town of Two Creeks, Manitowoc

County, Wisconsin

Date of application for amendments:

April 14, 1997 (TSCR 197), as

supplemented on August 11, 1997.

Brief description of amendments:

These amendments revise Technical

Specifications (TS) Sections 15.6.2,

‘‘Organization,’’ TS 15.6.5.1, ‘‘Manager’s

Supervisory Staff,’’ TS 15.6.6,

‘‘Reportable Event Action,’’ TS 15.6.7,

‘‘Actions To Be Taken If A Safety Limit

Is Exceeded,’’ and TS 15.7.8,

‘‘Administrative Controls,’’ by changing

the title of the corporate officer

responsible for nuclear operations from

the ‘‘Vice President-Nuclear Power,’’ to

the ‘‘Chief Nuclear Officer.’’

Date of issuance: August 25, 1997.

Effective date: August 25, 1997, with

full implementation within 45 days.

Amendment Nos.: 177 and 181.

Facility Operating License Nos. DPR–

24 and DPR–27: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: May 21, 1997 (62 FR 27802),

as corrected May 29, 1997 (62 FR 29163)

The August 11, 1997, submittal

provided a corrected TS page. This

information was within the scope of the

action noticed and did not change the

staff’s initial proposed no significant

hazards considerations determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated August 25,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: The Lester Public Library,

1001 Adams Street, Two Rivers,

Wisconsin 54241.

Dated at Rockville, Maryland, this 3rd day

of September 1997.

For The Nuclear Regulatory Commission.

Bruce E. Boger,

Director, Division of Reactor Projects—I/II

Office of Nuclear Reactor Regulation.

[FR Doc. 97–23820 Filed 9–9–97; 8:45 am]

BILLING CODE 7590–01–P