8780 Federal Register / Vol. 62, No. 38 / Wednesday, February 26, 1997 / Notices

Reason for Closing: The proposals being

reviewed include information of a propriety

or confidential nature, including technical

information; financial data, such as salaries;

and personal information under 5 U.S.C.

552b(c), (4) and (6) of the Government in the

Sunshine Act.

Dated: February 21, 1997.

Linda Allen-Benton,

Deputy Director, Division of Human Resource

Management, Acting Committee Management

Officer.

[FR Doc. 97–4707 Filed 2–25–97; 8:45 am]

BILLING CODE 7555–01–M

Special Emphasis Panel in Biological

Sciences; Notice of Meeting

In accordance with the Federal

Advisory Committee Act (Pub. L. 92–

463, as amended), the National Science

Foundation announces the following

meeting:

Name: Special Emphasis Panel in

Biological Sciences; (1754).

Date and Time: March 19, 1997 (7:30PM to

9:30PM) and March 20, 1997 (8:30AM to

5PM).

Place: (March 19) Holiday Inn, 3845

Veterans Memorial Highway, Ronkonkoma,

NY 11779; (March 20) Biology Building,

Brookhaven National Laboratory, Upton, NY,

11973.

Type of Meeting: Closed.

Contact Person: Dr. Gerald Selzer, Program

Director, Division of Biological Infrastructure

(DBI), Room 615, National Science

Foundation, 4201 Wilson Blvd. Arlington,

VA 22230, Tel: (703) 306–1469.

Purpose of Meeting: To provide advice and

recommendations concerning BIO Database

Activities projects.

Agenda: To review and evaluate progress

report and proposal for future activities

provided by the Protein Data Bank project.

Reason for Closing: The report being

reviewed includes information of a

proprietary or confidential nature, including

technical information, financial data, such as

salaries; and personal information

concerning individuals associated with the

report. These matters are within exemptions

(4) and (6) of 5 U.S.C. 552b(c), the

Government in the Sunshine Act.

Dated: February 21, 1997.

Linda Allen-Benton,

Deputy Director, Division of Human Resource

Management, Acting Committee Management

Officer.

[FR Doc. 97–4712 Filed 2–25–97; 8:45 am]

BILLING CODE 7555–01–M

Special Emphasis Panel in Human

Resource Development; Notice of

Meeting

In accordance with the Federal

Advisory Committee Act (Pub. L. 92–

463, as amended), the National Science

Foundation announces the following

meeting:

Name: Special Emphasis Panel in Human

Resource Development (#1199).

Date and Time: March 16, 1997: 7:00 to

9:00 p.m., March 17, 1997: 8:00 a.m. to 5:00

p.m., March 18, 1997: 8:00 a.m. to 4:00 p.m.

Place: National Science Foundation, 4201

Wilson Blvd., Rooms 310, 320, 380, 390,

Arlington, VA 22230.

Type of Meeting: Closed.

Contact Person: Lola Rogers, Program

Director, 4201 Wilson Blvd., Suite 815,

Arlington, VA 22230. Telephone: 703/306–

1637.

Purpose of Meeting: To provide advice and

recommendations concerning proposals

submitted to NSF for financial support.

Agenda: To review and evaluate

unsolicited proposals submitted to the

Implementation and Development Projects

for Women and Girls Program.

Reason for Closing: The proposals being

reviewed include information of a

proprietary or confidential nature, including

technical information, financial data, such as

salaries, and personal information

concerning individuals associated with the

proposals. These matters are exempt under 5

U.S.C. 552b(c)(4) and (6) of the Government

in the Sunshine Act.

Dated: February 21, 1997.

Linda Allen-Benton,

Deputy Director, Division of Human Resource

Management, Acting Committee Management

Officer.

[FR Doc. 97–4710 Filed 2–25–97; 8:45 am]

BILLING CODE 7555–01–M

NUCLEAR REGULATORY

COMMISSION

[Docket Nos. 50–317 and 50–318]

Baltimore Gas and Electric Company;

Notice of Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing

The U.S. Nuclear Regulatory

Commission (the Commission) is

considering issuance of an amendment

to Facility Operating License Nos. DPR–

53 and DPR–69 issued to Baltimore Gas

and Electric Company, for operation of

the Calvert Cliffs Nuclear Power Plant,

Unit Nos. 1 and 2, located in Calvert

County, Maryland.

The proposed amendment revises the

Technical Specifications (TSs) to reduce

the minimum Reactor Coolant System

(RCS) total flow rate from 370,000 gpm

to 340,000 gpm; reduce the Reactor

Protective Instrumentation trip setpoint

for Reactor Coolant Flow—Low from

greater than or equal to 95% to greater

than or equal to 92% of design reactor

coolant flow; adjust the reactor core

thermal margin safety limit lines to

reflect the reduced RCS flow rate; and

reduce the lift setting range for the eight

Main Steam Safety Valves (MSSVs) with

the highest allowable lift setting from

the current range of 935 to 1065 psig to

a more restrictive range of 935 to 1050

psig. In addition to the changes to the

TSs necessary to support an increased

number of plugged SG tubes, reanalysis

of the accident analyses affected by this

change identified an Unreviewed Safety

Question (USQ) associated with these

changes. The USQ results from the

determination that the Main Steam Line

Break (MSLB) and Seized Rotor Event

analyses involve an increased

percentage of failed fuel cladding.

Finally, three reanalyzed events (MSLB,

Loss of Coolant Flow, and Boron

Dilution) will require Nuclear

Regulatory Commission (NRC) approval

due to changes to the methodology or

assumptions used to analyze these

events.

Before issuance of the proposed

license amendment, the Commission

will have made findings required by the

Atomic Energy Act of 1954, as amended

(the Act) and the Commission’s

regulations.

The Commission has made a

proposed determination that the

amendment request involves no

significant hazards consideration. Under

the Commission’s regulations in 10 CFR

50.92, this means that operation of the

facility in accordance with the proposed

amendment would not (1) involve a

significant increase in the probability or

consequences of an accident previously

evaluated; or (2) create the possibility of

a new or different kind of accident from

any accident previously evaluated; or

(3) involve a significant reduction in a

margin of safety. As required by 10 CFR

50.91(a), the licensee has provided its

analysis of the issue of no significant

hazards consideration, which is

presented below:

1. Would not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

The proposed amendment defines changes

to the operating licenses for Calvert Cliffs

Nuclear Power Plant, Units 1 and 2,

necessary to support increased steam

generator tube plugging. The effects of

increased steam generator tube plugging

include reduced steam generator pressure

and RCS flow rate, and increased core outlet

(hot leg) temperature. The Technical

Specification changes necessary to account

for these effects are reducing the minimum

RCS total flow rate from 370,000 gpm to

340,000 gpm; reducing the Limiting Safety

System Setting for reactor coolant flow trip

function from greater than or equal to 95%

to greater than or equal to 92% of design

reactor coolant flow; revising the Reactor

Core Thermal Safety Limit lines to indicate

operation at the lower reactor coolant flow

rate; and decreasing the maximum allowable

lift settings for the eight highest set Main
Federal Register / Vol. 62, No. 38 / Wednesday, February 26, 1997 / Notices 8781

Steam Safety Valves from 1065 psig to 1050

psig. The Design Basis Events (DBEs) affected

by these changes were reanalyzed to

determine if the effects of increased steam

generator tube plugging, and the associated

changes to the Technical Specifications,

could result in exceeding the acceptance

criteria applicable to each of these events.

Although it was determined that the DBE

acceptance criteria would not be exceeded as

a result of increased steam generator tube

plugging, the analyses for the Main Steam

Line Break and Seized Rotor Events indicated

an increased percentage of fuel cladding

failure as a result of the lower RCS total flow

rate; therefore, it was determined that this

activity involves a USQ.

Technical Specification 2.1.1 will be

changed to establish more restrictive limits

on core thermal power and reflect a lower

minimum RCS flow of 340,000 gpm. Making

the core thermal power limits more

restrictive does not initiate a change to plant

conditions that would affect other plant

components. Therefore, the probability of a

previously evaluated accident is not

significantly increased. Additionally, the

Limiting Conditions for Operation and

Limiting Safety System Settings based on

these limits remain adequately conservative

or will be changed in the Core Operating

Limits Report, as appropriate. Therefore, the

consequences of a previously evaluated

accident are not significantly increased.

Technical Specification 2.2 will be

changed to reduce the Reactor Coolant

Flow—Low reactor trip setpoint from [greater

than or equal to] 95% to [greater than or

equal to] 92%, thereby providing additional

operating margin to this trip setpoint and the

associated pre-trip alarm. Reducing this

setpoint does not initiate a change to plant

conditions that would affect other plant

components. Therefore, the probability of a

previously evaluated accident is not

significantly increased.

As demonstrated by the revised Loss of

Coolant Flow analysis, the proposed Reactor

Coolant Flow—Low reactor trip setpoint will

continue to provide adequate core protection.

A trip setpoint of [greater than or equal to]

92% ensures fuel is not damaged, and the

site boundary dose remains a small fraction

of the 10 CFR Part 100 guidelines. Therefore,

the consequences of a previously evaluated

accident are not significantly increased.

Technical Specification 3.2.5.c will be

changed to reduce the minimum RCS total

flow rate from 370,000 gpm to 340,000 gpm.

This change reduces the core heat removal

rate and slightly increases the core outlet and

average coolant temperatures. This change

involves a USQ, as the Main Steam Line

Break and Seized Rotor Event analyses have

indicated an increase in the number of failed

fuel pins during these events as a result of

reducing the initial RCS flow rate. The

probability of malfunction of equipment

important to safety (i.e., fuel pin cladding)

during these accidents increases. However,

this malfunction is not an accident initiator.

Rather, it is a consequence of an accident.

Therefore, the probability of a previously

evaluated accident is not significantly

increased. The consequences of the Main

Steam Line Break and Seized Rotor Events

are not significantly increased, as the results

of the analyses of these events are within the

current acceptance criteria established by the

NRC.

Analyses and evaluations have been

performed to demonstrate that the new

flow and temperature conditions are

acceptable:

Fuel and core performance remain within

acceptable limits. Analysis and evaluation of

fuel mechanical design, core physics,

parameters, fuel pin performance, fuel

assembly thermal/hydraulic performance,

and fuel pin corrosion all demonstrate

acceptable results.

The effect of the slightly elevated core

outlet and average coolant temperature on

the structural integrity of the RCS is

acceptable. The RCS penetration inspection

program and the steam generator tube

inspection program will continue to identify

and repair or isolate Alloy 600 cracks prior

to inservice failure of these components. The

stress analysis for the reactor vessel and

piping remain bounding.

The performance of control systems (i.e.,

feedwater, pressurizer level, and pressurizer

pressure) will maintain RCS and steam

generator parameters within appropriate

limits by periodic adjustment, as necessary.

Reactor coolant pump operation will be

maintained within acceptable limits by

periodic adjustment of the operating curves.

Therefore, the probability of a

previously evaluated accident is not

significantly increased.

Analyses and evaluations of the DBEs

have been performed demonstrating that

the NRC acceptance criteria for these

events are met. The revised analyses

and evaluations consider reduced RCS

flow, increased reactor coolant

temperature, and increased steam

generator tube plugging conditions.

The results of analyses and evaluations of

the Postulated Accidents demonstrate that

the site boundary dose is within 10 CFR Part

100 guidelines and the core geometry

remains coolable. Loss-of-Coolant Accident

analysis results meet the acceptance criteria

stipulated in 10 CFR 50.46(b).

The results of analyses and evaluations of

Anticipated Operational Occurrences

demonstrate that fuel parameters do not

exceed the specified acceptable fuel design

limits and site boundary dose is a small

fraction of 10 CFR Part 100 guidelines.

Primary and secondary system pressure

remain below the pressure upset limits for

the RCS and steam generators, respectively.

Therefore, the consequences of a

previously evaluated accident are not

significantly increased.

Technical Specification 4.7.1.1. will be

changed to reduce the maximum allowable

lift setting for the eight Main Steam Safety

Valves with the highest lift setpoint. This

change will place more restrictive limits on

the allowable range of lift settings for these

eight valves. The allowable range of lift

settings for the proposed change is also

allowed by current Technical Specification.

Therefore, the probability of a previously

evaluated accident occurring is not

significantly increased.

The revised safety analyses will credit the

highest lift setting for these eight valves as

being 1050 psig. The more restrictive limit on

the maximum lift setting is required in order

to make this Technical Specification

consistent with the revised safety analyses.

Analyses performed assuming the proposed

maximum lift setting for these valves

demonstrates that secondary system pressure

does not exceed 110% of the system design

pressure. Therefore, the consequences of a

previously evaluated accident are not

significantly increased.

Therefore, operation of the facility in

accordance with this amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Would not create the possibility of a new

or different type of accident from any

accident previously evaluated.

The proposed amendment revises limiting

parameters to assure safe operation

commensurate with the effects of steam

generator tube plugging, and will not change

the modes of operation defined in the facility

license. The analysis of transients associated

with steam generator malfunctions are part of

the design and licensing bases. This change

does not add any new equipment, modify

any interfaces with any existing equipment,

or change the equipments’s function, or the

method of operating the equipment. The

proposed change does not change plant

conditions in a manner which could affect

other plant components. Reactor core, RCS,

and steam generator parameters remain

within appropriate design limits during

normal operation.

Therefore, the proposed change could not

cause any existing equipment to become an

accident initiator.

Therefore, the proposed change does not

create the possibility of a new or different

type of accident from any accident

previously evaluated.

3. Would not involve a significant

reduction in a margin of safety.

The margins of safety associated with this

change are defined in the fuel and corerelated analyses, the Alloy 600 stress

corrosion cracking evaluation, the RCS

structural evaluation, the operational

evaluation, and in each of the transient and

accident analyses affected by the increased

steam generator tube plugging.

Reanalysis of the fuel and core-related

analyses for fuel mechanical design, core

physics, fuel performance, thermal

hydraulics, and fuel rod corrosion verified

that the fuel and core performance will

remain within acceptable limits and will be

bounded by the current assumptions for fuel

performance in the transient and accident

analyses. The Alloy 600 RCS penetration

inspection program and the steam generator

tube inspection program will continue to find

and repair Alloy 600 cracks at the slightly

elevated core exit temperature prior to any

postulated inservice failure of these

components. The stress analyses performed

for the reactor vessel and piping remain

bounding for the slightly elevated core exit
8782 Federal Register / Vol. 62, No. 38 / Wednesday, February 26, 1997 / Notices

temperature. Additionally, the performance

of non-safety-related control systems remains

adequate to maintain RCS and steam

generator parameters within appropriate

operating limits. Therefore, the margins of

safety associated with the physical and

operational effects of this change will not be

significantly reduced.

An evaluation of the affected DBEs

confirmed that the established acceptance

criteria for specified acceptable fuel design

limits, primary and secondary system overpressurization, 10 CFR 50.46(b), Acceptance

Criteria for Emergency Core Cooling Systems

for Light-Water Nuclear Power Reactors, and

potential radiation dose during accidents

have been completed in support of this

license amendment request. The evaluation

concludes that, when considering the

proposed Limiting Safety System Setting for

the Reactor Coolant Flow—Low trip,

Limiting Conditions for Operation for RCS

total flow rate, and reduced lift settings for

eight Main Steam Safety Valves per unit, all

applicable acceptance limits are met.

Furthermore, the USQ resulting from the

reduced RCS total flow rate does not

represent a reduction in the margin of safety,

as the site boundary dose calculated in the

affected DBE analyses is within the current

established radiation dose limits and the core

geometry remains coolable. Therefore, the

margins of safety associated with the

transient and accident analyses affected by

this change will not be significantly reduced.

Therefore, the proposed change does not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

The Commission is seeking public

comments on this proposed

determination. Any comments received

within 30 days after the date of

publication of this notice will be

considered in making any final

determination.

Normally, the Commission will not

issue the amendment until the

expiration of the 30-day notice period.

However, should circumstances change

during the notice period such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received. Should

the Commission take this action, it will

publish in the Federal Register a notice

of issuance and provide for opportunity

for a hearing after issuance. The

Commission expects that the need to

take this action will occur very

infrequently.

Written comments may be submitted

by mail to the Chief, Rules Review 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 hearing and

petitions for leave to intervene is

discussed below.

By March 28, 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 Calvert

County Library, Prince Frederick,

Maryland 20678. 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.
Federal Register / Vol. 62, No. 38 / Wednesday, February 26, 1997 / Notices 8783

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:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington, DC, by

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

Union at 1–(800) 248–5100 (in Missouri

1–(800) 342–6700). The Western Union

operator should be given Datagram

Identification Number N1023 and the

following message addressed to S. Singh

Bajwa: petitioner’s name and telephone

number, date petition was mailed, plant

name, and publication date and page

number of this Federal Register notice.

A copy of the petition should also be

sent to the Office of the General

Counsel, U.S. Nuclear Regulatory

Commission, Washington, DC 20555–

0001, and to Jay E. Silbert, Esquire,

Shaw, Pittman, Potts and Trowbridge,

2300 N Street, NW., Washington, DC,

20037 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 January 31, 1997, as

supplemented February 13, 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

Calvert County Library, Prince

Frederick, Maryland 20678.

Dated at Rockville, Maryland, this 20th day

of February 1997.

For the Nuclear Regulatory Commission.

Alexander W. Dromerick,

Senior Project Manager, Project Directorate

I–1, Division of Reactor Projects—I/II, Office

of Nuclear Reactor Regulation.

[FR Doc. 97–4701 Filed 2–25–97; 8:45 am]

BILLING CODE 7590–01–P

[Docket No. 50–331]

IES Utilities Inc., Central Iowa Power

Cooperative, Corn Belt Power

Cooperative, Duane Arnold Energy

Center; Notice of Consideration of

Approval of Application Regarding

Merger

Notice is hereby given that the United

States Nuclear Regulatory Commission

(the Commission) is considering the

issuance of an order approving under 10

CFR 50.80 an application regarding the

proposed merger involving IES

Industries (IESI), the parent company of

IES Utilities Inc. (IESU). IESU is the

licensee for the Duane Arnold Energy

Center (DAEC). By letter dated

September 27, 1996, IESU informed the

Commission that under a merger

agreement among IESI, WPL Holdings,

and Interstate Power Corporation, IESI

will merge with and into a WPL

Holdings (WPLH), and Interstate Power

Corporation will become a subsidiary of

WPLH. WPLH will be renamed

Interstate Energy Corporation (IEC) of

which IESU would become a whollyowned subsidiary. IESU will remain the

holder of its license for DAEC. Under

the merger agreement, current

stockholders of IESI, WPLH, and IPC

will become stockholders of IEC

pursuant to a formula stipulated in the

merger agreement. IESU requested the

Commission’s approval regarding the

proposed transactions to the extent they

effect an indirect transfer of control of

the DAEC license, pursuant to 10 CFR

50.80. IESU would remain an electric

utility as defined in 10 CFR 50.2,

engaged in the generation, transmission,

and distribution of electric energy for

wholesale and retail sale, subject to the

rate regulation of the Iowa Utilities

Board and the Federal Energy

Regulatory Commission.

Pursuant to 10 CFR 50.80, the

Commission may approve the transfer of

control of a license after notice to

interested persons. Such approval is

contingent upon the Commission’s

determination that the holder of the

license following the transfer is

qualified to hold the license and that the

transfer is otherwise consistent with

applicable provisions of law,

regulations, and orders of the

Commission.

For further details with respect to this

proposed action, see the licensee’s letter

dated September 27, 1996, with the

following exhibits: (A) Information to

support the request for the

Commission’s consent. (B) A copy of the

merger agreement executed among IES

Industries Inc., WPL Holdings, Inc., and

Interstate Power Corporation. These

documents are available for public

inspection at the Commission’s Public

Document Room, 2120 L Street, NW.,

Washington, DC, and at the local public

document room located in the Cedar

Rapids Public Library, 500 First Street,

SE., Cedar Rapids, IA 52401.

Dated at Rockville, Maryland, this 19th day

of February 1997.

Glenn B. Kelly,

Project Manager, Project Directorate III–3,

Division of Reactor Projects III/IV, Office of

Nuclear Reactor Regulation.

[FR Doc. 97–4700 Filed 2–25–97; 8:45 am]

BILLING CODE 7590–01–P

Toledo Edison Company Centerior

Service Company; and the Cleveland

Electric Illuminating Company; DavisBesse Nuclear Power Station, Unit No.

1 Notice of Consideration of Issuance

of Amendment to Facility Operating

License, Proposed no Significant

Hazards Consideration Determination,

and Opportunity for a Hearing

[Docket No. 50–346]

The U.S. Nuclear Regulatory

Commission (the Commission) is

considering issuance of an amendment

to Facility Operating License No. NPF–

3, issued to the Toledo Edison

Company, Centerior Service Company,

and The Cleveland Electric Illuminating

Company (the licensee), for operation of

the Davis-Besse Nuclear Power Station,

Unit No. 1 located in Ottawa County,

Ohio.

The proposed amendment would

revise Technical Specification (TS)

Section 3/4.5.2, ‘‘Emergency Core

Cooling Systems, ECCS Subsystems—

Tavg ≥ 280 °F.’’ Surveillance requirement

(SR) 4.5.2.f would be modified to state

that opening and closing of the

inspection port on the watertight

enclosure for the decay heat valve pit

would not require this surveillance

procedure to be performed. The

applicable TS bases would also be

changed.

The licensee’s submittal is being

processed as an exigent TS amendment

request pursuant to 10 CFR 50.91(a)(6),

as a followup to the Notice of

Enforcement Discretion (NOED) issued

by the Commission on February 12,

1997.