Federal Register / Vol. 66, No. 45 / Wednesday, March 7, 2001 / Notices 13797

Scoping Comments

Written comments should be mailed

to: Michael T. Lesar, Acting Chief, U.S.

Nuclear Regulatory Commission, Rules

& Directives Branch, Division of

Administrative Services, Office of

Administration, Mail Stop T6D59,

Washington, DC 20555.

Comments will also be accepted by email. Interested parties may e-mail their

comments to teh@nrc.gov. Comments

will be accepted by fax at 301–415–

5398, Attention: Tim Harris.

NRC will make the scoping

summaries and project-related materials

available for public review through our

electronic reading room: http://

www.nrc.gov/NRC/ADAMS/index.html.

The scoping meeting summaries and

project-related materials will also be

available on the NRC’s MOX web page:

http://www.nrc.gov/NRC/NMSS/MOX/

index.html (case sensitive).

The NEPA Process

The EIS for the MOX Facility will be

prepared according to the National

Environmental Policy Act of 1969, the

Council on Environmental Quality’s

Regulations for Implementing the

Procedural Provisions of NEPA (40 CFR

Parts 1500–1508), and NRC’s NEPA

Regulations (10 CFR Part 51).

The draft EIS is scheduled to be

published in February 2002. A 45-day

comment period on the draft EIS is

planned, and public meetings to receive

comments will be held approximately

three weeks after distribution of the

draft EIS. Availability of the draft EIS,

the dates of the public comment period,

and information about the public

meetings will be announced in the

Federal Register, on NRC’s MOX web

page, and in the local news media when

the draft EIS is distributed. The final

EIS, which will incorporate public

comments received on the draft EIS, is

expected in September 2002.

Signed in Rockville, MD, this 1st day of

March 2001.

For the Nuclear Regulatory Commission.

Charlotte E. Abrams,

Acting Chief, Environmental and

Performance Assessment Branch, Division of

Waste Management, Office of Nuclear

Material Safety and Safeguards.

[FR Doc. 01–5509 Filed 3–6–01; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Sunshine Act Meeting

AGENCY HOLDING THE MEETING: Nuclear

Regulatory Commission.

DATES: Weeks of March 5, 12, 19, 26,

April 2, 9, 2001.

PLACE: Commissioners’ Conference

Room, 11555 Rockville Pike, Rockville,

Maryland.

STATUS: Public and Closed.

MATTERS TO BE CONSIDERED: 

Weeks of March 5, 2001

There are no meetings scheduled for

the Week of March 5, 2001.

Week of March 12, 2001—Tentative

Monday, March 12, 2001

1:25 p.m.

Affirmation Session (Public Meeting)

(If needed).

1:30 p.m.

Discussion of Management Issues

(Closed-Ex. 2)

Week of March 19, 2001—Tentative

Thursday, March 22, 2001

10:25 a.m.

Affirmation Session (Public Meeting)

(If needed).

10:30 a.m.

Meeting with Advisory Committee on

Nuclear Waste (ACNW) (Public

Meeting) (Contact: John Larkins,

301–415–7360).

This meeting will be webcast live at

the Web address—www.nrc.gov/

live.html.

Week of March 26, 2001—Tentative

There are no meetings scheduled for

the Week of March 26, 2001.

Week of April 2, 2001—Tentative

There are no meetings scheduled for

the Week of April 2, 2001.

Week of April 9, 2001—Tentative

Monday, April 9, 2001.

1:30 p.m.

Briefing on 10 CFR Part 71

Rulemaking (Public Meeting)

(Contacts: Naiem Tanious, 301–

415–6103; David Pstrak, 301–415–

8486).

Tuesday, April 10, 2001

10:25 a.m.

Affirmation Session (Public Meeting)

(If needed).

10:30 a.m.

Meeting on Rulemaking and Guidance

Development for Uranium Recovery

Industry (Public Meeting) (Contact:

Michael Layton, 301–415–6676).

*The schedule for Commission

meetings is subject to change on short

notice. To verify the status of meetings

call (recording)—(301) 415–1292.

Contact person for more information:

David Louis Gamberoni (301) 415–1651.

Additional Information:

By a vote of 5–0 on February 23, the

Commission determined pursuant to

U.S.C. 552b(e) and § 9.107(a) of the

Commission’s rules that ‘‘Discussion of

Intragovernmental Issues (Closed-Ex.

9)’’ be held on February 26, and on less

than one week’s notice to the public.

The NRC Commission Meeting

Schedule can be found on the Internet

at: http://www.nrc.gov/SECY/smj/

schedule.htm.

This notice is distributed by mail to

several hundred subscribers; if you no

longer wish to receive it, or would like

to be added to the distribution, please

contact the Office of the Secretary,

Washington, D.C. 20555 (301–415–

1969). In addition, distribution of this

meeting notice over the Internet system

is available. If you are interested in

receiving this Commission meeting

schedule electronically, please send an

electronic message to dkw@nrc.gov.

Dated: March 1, 2001.

David Louis Gamberoni,

Technical Coordinator, Office of the

Secretary.

[FR Doc. 01–5723 Filed 3–5–01; 2:21 pm]

BILLING CODE 7590–01–M

NUCLEAR REGULATORY

COMMISSION

Biweekly Notice; Applications and

Amendments to Facility Operating

Licenses Involving No Significant

Hazards Considerations

I. Background

Pursuant to Public Law 97–415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 97–415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from February 12,

2001, through February 23, 2001. The

13798 Federal Register / Vol. 66, No. 45 / Wednesday, March 7, 2001 / Notices

last biweekly notice was published on

February 21, 2001 (66 FR 11050).

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 Review and

Directives Branch, Division of Freedom

of Administrative 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, located at One White

Flint North, 11555 Rockville Pike (first

floor), Rockville, Maryland 20852. The

filing of requests for a hearing and

petitions for leave to intervene is

discussed below.

By April 6, 2001, 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, located at One

White Flint North, 11555 Rockville Pike

(first floor), Rockville, Maryland 20852.

Publicly available records will be

accessible and electronically from the

ADAMS Public Library component on

the NRC Web site, http://www.nrc.gov

(the Electronic Reading Room). If a

request for a hearing or petition for

leave to intervene is filed by the above

date, the Commission or an Atomic

Safety and Licensing Board, designated

by the Commission or by the Chairman

of the Atomic Safety and Licensing

Board Panel, will rule on the request

and/or petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) The nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

Federal Register / Vol. 66, No. 45 / Wednesday, March 7, 2001 / Notices 13799

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555–0001, Attention:

Rulemaking and Adjudications Branch,

or may be delivered to the

Commission’s Public Document Room,

located at One White Flint North, 11555

Rockville Pike (first floor), Rockville,

Maryland 20852, 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, located at One

White Flint North, 11555 Rockville Pike

(first floor), Rockville, Maryland 20852.

Publicly available records will be

accessible and electronically from the

ADAMS Public Library component on

the NRC Web site, http://www.nrc.gov

(the Electronic Reading Room).

Calvert Cliffs Nuclear Power Plant, Inc.,

Docket Nos. 50–317 and 50–318, Calvert

Cliffs Nuclear Power Plant, Unit Nos. 1

and 2, Calvert County, Maryland

Date of amendments request:

December 20, 2000.

Description of amendments request:

The amendments would revise the

Technical Specifications to incorporate

changes required to support operation

with replacement steam generators. The

proposed changes will (1) accommodate

geometric differences between the

original and replacement steam

generators, (2) increase the reactor

coolant flow rate from the current value

which was recently established to

accommodate more tube plugging, and

(3) delete tube sleeving options

approved for the original steam

generators.

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. Would not involve a significant

increase in the probability or

consequences of an accident previously

evaluated

A. Technical Specification Table 3.3.1–

1, Item 7

Technical Specification Table 3.3.1–1,

‘‘Reactor Protective System

Instrumentation,’’ Item 7 sets the

allowable value for ‘‘Steam Generator

Level-Low’’ function to greater than or

equal to 10 inches below the top of the

feed ring. To accommodate the

geometric difference in the location of

the top of the feed ring with respect to

the pedestal between the original steam

generators (OSG) (510.8 inches) and the

replacement steam generators (RSG)

(484.8 inches), the proposed

amendment would change the allowable

value for ‘‘Steam Generator Level-Low’’

function to greater than or equal to 50

inches below normal water level. Since

normal water levels for RSG and OSG

with respect to the pedestal are identical

and the current steam generator levellow reactor trip setpoint ‘‘≥10 inches

below top of feed ring’’ is ‘‘≥ 50 inches

below normal water level’’ for both the

RSG and OSG, the functionality of the

steam generator level-low reactor trip

setpoint will be unchanged.

Furthermore, use of normal water level

as the point of reference instead of top

of the feed ring is more practical and

appropriate since it is the frame of

reference for steam generator water level

indication used in the Control Room by

the operators.

The design basis accident affected by

the proposed change is the Loss of

Feedwater Flow event. The Steam

Generator Level-Low Reactor Trip

Setpoint, in combination with the

Auxiliary Feedwater Actuation System,

ensures that adequate secondary side

water inventory exists in both RSGs to

remove decay heat following a Loss of

Feedwater Flow event. To ensure that

the acceptance criteria for the Loss of

Feedwater Flow event are met with the

RSGs, there must be at least as much

mass in RSG at the Safety Analysis

water level as in the OSG. The OSG

Safety Analysis water level is 116.4

inches below normal water level. Using

the same method to predict steam

generator inventory, at this water level,

OSG has 64,049 Ibm water mass and

RSG has 64,115 Ibm water mass.

Therefore, the RSG has more postreactor trip secondary side inventory

than the OSG which ensures the Loss of

Feedwater event acceptance criteria are

not challenged.

Therefore, the proposed revision to

change the reference setpoint for steam

generator low level reactor trip function

will not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

B. LCO [limiting condition for

operation] 3.4.1 and Surveillance

Requirement 3.4.1.3

The proposed amendment would

revise Technical Specification LCO

3.4.1 and Surveillance Requirement

3.4.1.3 to increase reactor coolant

minimum required total flow rate back

to the originally established value of

370,000 gpm [gallons per minute] from

the current value of 340,000 gpm, which

was recently established to

accommodate more tube plugging in the

OSG. The flow resistance of the RSG is

equivalent to that of the OSG with zero

plugged tubes. Therefore, the required

minimum RCS [reactor coolant system]

total flow rate can be increased to the

value previously established for the

original steam generators with zero

plugged tubes, 370,000 gpm.

Increasing the required minimum RCS

total flow rate has no adverse impact on

the safety analysis. Crediting more RCS

flow in the safety analysis allows for

greater flexibility in core design and

operation. The increase in RCS flow

associated with the RSG is within the

bounds previously analyzed for the

OSG. The hydraulic forces experienced

around the RCS loop, including the core

uplift force, are acceptable. The change

is more restrictive in nature in that more

RCS flow will be required to meet

Surveillance Requirement 3.4.1.3 and

more RCS flow ensures enhanced core

heat removal. The overall core thermal

margin in the safety analysis will

remain essentially the same.

Therefore, the proposed revision to

increase reactor coolant minimum

required total flow rate will not involve

a significant increase in the probability

or consequences of an accident

previously evaluated.

C. Technical Specification

Administrative Control 5.5.9

The proposed revision deletes three

sleeving options from Administrative

Technical Specification 5.5.9. The

sleeving options are: Westinghouse

Laser Welded sleeves, Asea Brown

Boveri, Inc. (ABB)-Combustion

Engineering Leak Tight sleeves, and the

ABB-Combustion Engineering Alloy 800

Leak Limiting sleeves. One of the

differences between the OSG and the

RSG design is the use of thermallytreated Alloy 690 tube material instead

of high temperature mill-annealed Alloy

600 used for the OSG. The three

sleeving tube repair options described

in Calvert Cliffs Nuclear Power Plant

(CCNPP) Technical Specification

13800 Federal Register / Vol. 66, No. 45 / Wednesday, March 7, 2001 / Notices

Administrative Control 5.5.9, are

designed specifically for the OSGs’ millannealed Alloy 600 tubes.

The three sleeving options were

acquired by CCNPP for economic

reasons to maintain OSG thermal output

by minimizing the number of tubes

plugged. Therefore, deletion of these

repair options from Administrative

Control 5.5.9 has no safety significance.

Therefore, the proposed revision will

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 [kind] of accident from

any accident previously evaluated

A. Technical Specification Table 3.3.1–

1, Item 7

The RSGs are equivalent in function

to the OSGs. Changing Technical

Specification Table 3.3.1–1, Item 7 is

required to provide a correct and

practical reference point from which to

measure the Reactor Trip Steam

Generator Level-Low Setpoint. As

described above in Item 1, the normal

water levels for RSG and OSG with

respect to the pedestal are identical and

the current steam generator level-low

reactor trip setpoint, ‘‘≥ 10 inches below

top of feed ring’’ is ‘‘≥ 50 inches below

normal water level’’ for both the RSG

and OSG. Hence, the functionality of the

reactor trip steam generator level-low

setpoint will be unchanged.

Furthermore, use of normal water level

as the point of reference instead of top

of the feed ring is more practical and

appropriate since it is the frame of

reference for steam generator water level

indication used in the Control Room by

the operators.

Therefore, the proposed revision to

change the reference setpoint for steam

generator low level reactor trip function

will not create the possibility of a new

or different [kind] of accident from any

accident previously evaluated.

B. LCO 3.4.1 and Surveillance

Requirement 3.4.1.3

As described above in Item 1,

increasing the required minimum RCS

total flow rate has no adverse impact on

the plant’s safety analyses. The increase

in RCS flow associated with the RSG is

within the bounds previously analyzed

for the OSG. The hydraulic forces

experienced around the RCS loop,

including the core uplift force, are

acceptable. The change is more

restrictive in nature in that more RCS

flow will be required to meet

Surveillance Requirement 3.4.1.3 and

more RCS flow ensures enhanced core

heat removal.

Therefore, the proposed revision to

increase reactor coolant minimum

required total flow rate will not create

the possibility of a new or different type

of accident from any accident

previously evaluated.

C. Technical Specification

Administrative Control 5.5.9

As described in Item I above, the three

sleeving options were acquired by

CCNPP for economic reasons to

maintain OSO thermal output by

minimizing the number of tubes

plugged. Therefore, deletion of these

repair options from Technical

Specification Administrative Control

5.5.9 has no safety significance.

Therefore, the proposed revision will

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 the margin of safety

A. Technical Specification Table 3.3.1–

1, Item 7

As described above in Item 1, the

design basis accident affected by the

proposed change is the Loss of

Feedwater Flow event. The Steam

Generator Level-Low Reactor Trip

Setpoint, in combination with the

Auxiliary Feedwater Actuation System,

ensures that adequate secondary side

water inventory exists in both RSGs to

remove decay heat following a Loss of

Feedwater Flow event. To ensure that

the acceptance criteria for the Loss of

Feedwater Flow event are met with the

RSGs, there must be at least as much

mass in RSG at the Safety Analysis

water level as in the OSG. The OSG

Safety Analysis water level is 116.4

inches below normal water level. Using

the same method to predict steam

generator inventory, at this water level,

OSO has 64,049 lbm water mass and

RSG has 64,115 Ibm water mass.

Therefore, the RSG has more postreactor trip secondary side inventory

than the OSG which ensures the Loss of

Feedwater event acceptance criteria are

not challenged.

Therefore, the proposed revision to

change the reference setpoint for steam

generator low level reactor trip function

does not involve a significant reduction

in the margin of safety.

B. LCO 3.4.1 and Surveillance

Requirement 3.4.1.3

As described above in Item 1,

increasing the required minimum RCS

total flow rate has no adverse impact on

the safety analysis. Crediting more RCS

flow in the safety analysis allows for

greater flexibility in core design and

operation. The increase in RCS flow

associated with the RSG is within the

bounds previously analyzed for the

OSG. The hydraulic forces experienced

around the RCS loop, including the core

uplift force, are acceptable. The change

is more restrictive in nature in that more

RCS flow will be required to meet

Surveillance Requirement 3.4.1.3 and

more RCS flow ensures enhanced core

heat removal. The overall core thermal

margin in the safety analysis will

remain essentially the same.

Therefore, the proposed revision to

increase reactor coolant minimum

required total flow rate does not involve

a significant reduction in the margin of

safety.

C. Technical Specification

Administrative Control 5.5.9

As described in Item 1C above, the

three sleeving options were acquired by

CCNPP for economic reasons to

maintain OSG thermal output by

minimizing the number of tubes

plugged. Therefore, deletion of these

repair options from Technical

Specification Administrative Control

5.5.9 has no safety significance.

Therefore, the proposed revision does

not involve a significant reduction in

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

amendments request involves no

significant hazards consideration.

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Section Chief: Marsha

Gamberoni.

Consumers Energy Company, Docket

No. 50–255, Palisades Plant, Van Buren

County, Michigan

Date of amendment request: January

26, 2001.

Description of amendment request:

The proposed amendment would

change Technical Specification (TS)

Surveillance Requirement (SR) 3.7.9.2,

‘‘Ultimate Heat Sink (UHS),’’ by

increasing the maximum allowable

temperature of Lake Michigan water

from 81.5 °F to 85 °F. The licensee also

proposes to reflect this change in the

associated TS Bases.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Federal Register / Vol. 66, No. 45 / Wednesday, March 7, 2001 / Notices 13801

The following evaluation supports the

finding that operation of the facility in

accordance with the proposed changes

would not:

a. Involve a significant increase in the

probability or consequences of an

accident previously evaluated.

The UHS is Lake Michigan which is

completely passive and is not an

accident initiator in any accident

previously evaluated. Therefore, this

change does not involve an increase in

the probability of an accident previously

evaluated.

The UHS, by design, mitigates the

consequences of accidents by supplying

a repository for the decay heat and other

excess energy removed in the process of

cooling the plant equipment. The safety

analysis has been revised to use a

maximum UHS water temperature of 85

°F. The results of these revised analyses

still meet all of the required acceptance

criteria. Therefore, the proposed

changes do not affect any of the results

of the FSAR [Final Safety Analysis

Report] Chapter 14 accident analyses.

Hence the consequences of accidents

previously evaluated do not change.

Therefore, operation of the facility in

accordance with the proposed changes

to the Technical Specifications would

not involve a significant increase in the

probability or consequences of an

accident previously evaluated.

b. Create the possibility of a new or

different kind of accident from any

previously evaluated.

The proposed change would not alter

the design, configuration, or method of

operation of the plant. The proposed

temperature limit has been verified to be

acceptable for UHS operability

determinations by its documented use

in plant equipment design

considerations, and in the FSAR

Chapter 14 accident analyses. Therefore,

operation of the facility in accordance

with the proposed change to the

Technical Specifications would not

create the possibility of a new or

different kind of accident from any

previously evaluated.

c. Involve a significant reduction in

the margin of safety.

The proposed change to the Technical

Specifications would impose

temperature limits already in use in

equipment designs and as an initial

assumption of the plant accident

analyses. The proposed SR limit has

been utilized in the accident analyses

since 1994. The results of these accident

analyses meet all of the required

acceptance criteria when using the 85 °F

UHS water temperature limit. Therefore,

the proposed change to the Technical

Specifications would not involve a

significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Arunas T.

Udrys, Esquire, Consumers Energy

Company, 212 West Michigan Avenue,

Jackson, Michigan 49201.

NRC Section Chief: Claudia M. Craig.

Consumers Energy Company, Docket

No. 50–255, Palisades Plant, Van Buren

County, Michigan

Date of amendment request: February

12, 2001.

Description of amendment request:

The proposed amendment would

change Technical Specification (TS)

Section 5.6.5b, ‘‘Reporting

Requirements—Core Operating Limits

Report (COLR),’’ by adding a reference

to the existing references of approved

analytical methods for determining core

operating 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:

The following evaluation supports the

finding that operation of the facility in

accordance with the proposed changes

would not:

(1) Involve a significant increase in

the probability or consequences of an

accident previously evaluated.

The proposed change to the list of

methodology documents in

Specification 5.6.5.b. would not

increase the probability or consequence

of an accident previously evaluated.

Accidents previously evaluated will be

unaffected by the addition of a

methodology reference because they

were analyzed using approved methods.

The results of these event analyses met

their respective acceptance criteria.

Therefore, operation of the facility in

accordance with the proposed change to

the Technical Specifications would not

involve a significant increase in the

probability or consequences of an

accident previously evaluated.

(2) Create the possibility of a new or

different kind of accident from any

previously evaluated.

The proposed change to the list of

methodology documents in

Specification 5.6.5.b. would not create

the possibility of a new or different

accident than previously analyzed. The

proposed change only adds an approved

methodology document. All accidents

remain analyzed using applicable NRC

approved methodologies.

Therefore, operation of the facility in

accordance with the proposed change to

the Technical Specifications would not

create the possibility of a new or

different kind of accident from any

previously evaluated.

(3) Involve a significant reduction in

the margin of safety.

The proposed change to the list of

methodology documents in

Specification 5.6.5.b. would not reduce

the margin of safety. Because all

analyses use approved methodologies

and their results satisfy their respective

acceptance criteria, the margin of safety

is not reduced.

Therefore, the proposed change to the

Technical Specifications would not

involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Arunas T.

Udrys, Esquire, Consumers Energy

Company, 212 West Michigan Avenue,

Jackson, Michigan 49201

NRC Section Chief: Claudia M. Craig.

Entergy Gulf States, Inc., and Entergy

Operations, Inc., Docket No. 50–458,

River Bend Station, Unit 1, West

Feliciana Parish, Louisiana

Date of amendment request: January

24, 2001.

Description of amendment request:

The proposed amendment would revise

the Technical Specifications (TSs) to

incorporate the provisions to perform

routine diesel generator (DG) monthly

testing by gradually accelerating the DG

to operating speed, as opposed to

requiring the DG to attain rated voltage

and frequency within 10 seconds for DG

1A and DG 1B, and within 13 seconds

for DG 1C. In addition, a new TS would

be added to require fast start tests of the

DGs on a 184-day frequency.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Will operation of the facility in

accordance with this proposed change

involve a significant increase in the

probability or consequences of an

accident previously evaluated?

13802 Federal Register / Vol. 66, No. 45 / Wednesday, March 7, 2001 / Notices

The proposed changes affect the

surveillance requirements for the

emergency diesel generators. The

emergency diesel generators are onsite

standby power sources intended to

provide redundant and reliable power to

ESF [engineered safety feature] systems

credited as accident mitigating features

in design basis analyses. As discussed

in Regulatory Guide (RG) 1.9, Revision

3, the proposed changes are intended to

allow slower starts of the diesel

generators during testing in order to

reduce diesel generator aging effects due

to excessive testing conditions. As such,

the proposed changes should result in

improved diesel generator reliability

and availability, thereby providing

additional assurance that the diesel

generators will be capable of performing

their safety function. The method of

starting the emergency diesel generators

for testing purposes does not affect the

probability of any previously evaluated

accident. Although the changes allow

slower starts for the monthly tests, the

more rapid start function assumed in

the accident analysis is unchanged and

will be verified on a 184 day frequency.

Therefore the accident analysis

consequences are not affected.

Therefore, these changes do not

involve a significant increase in the

probability or consequences of any

accident previously evaluated.

2. Will operation of the facility in

accordance with this proposed change

create the possibility of a new or

different kind of accident from any

accident previously evaluated?

The proposed changes affect the

surveillance requirements for the onsite

ac [alternating current] sources, i.e. the

diesel generators. Accordingly, the

proposed changes do not involve any

change to the configuration or method

of operation of any plant equipment that

could cause an accident. In addition, no

new failure modes have been created

nor has any new limiting failure been

introduced as a result of the proposed

surveillance changes.

Therefore, these changes do not create

the possibility of a new or different kind

of accident from any previously

evaluated.

3. Will operation of the facility in

accordance with this proposed change

involve a significant reduction in a

margin of safety?

The proposed changes are intended to

bring the existing RBS [River Bend

Station] TS requirements for the onsite

ac sources in line with regulatory

guidance. Under the proposed changes,

the emergency diesel generators will

remain capable of performing their

safety function, and the effects of aging

on the diesel generators will be reduced

by eliminating unnecessary testing. The

diesel generator start times assumed in

the current accident analyses are

unchanged and will be verified on a 6-

month frequency.

Therefore, these changes do not

involve a significant reduction in 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.

Attorney for licensee: Mark

Wetterhahn, Esq., Winston & Strawn,

1400 L Street, NW., Washington, DC

20005.

NRC Section Chief: Robert A. Gramm.

Entergy Nuclear Generation Company,

Docket No. 50–293, Pilgrim Nuclear

Power Station, Plymouth County,

Massachusetts

Date of amendment request: February

5, 2001.

Description of amendment request:

The proposed amendment would

change the Safety Limit Minimum

Critical Power Ratio (SLMCPR) in

Technical Specification (TS) 2.1.2 from

1.08 to 1.06. The proposed amendment

would also change the parenthetical

statements after certain references listed

in TS 5.6.5.b to clarify that the

analytical methods described in General

Electric Nuclear Energy documents

inclusive of the latest amendment or

revision are used to determine core

operating limits. Also, the proposed

amendment would add a new reference

to TS 5.6.5.b.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration. The NRC staff has

reviewed the licensee’s analysis against

the standards of 10 CFR 50.92(c). The

NRC staff’s analysis is presented below:

1. The proposed changes to technical

specification do not involve a

significant increase in the probability of

an accident previously evaluated.

The proposed Safety Limit MCPR

(SLMCPR), and its use to determine the

Cycle 14 thermal limits, have been

derived using NRC approved methods

[See application dated February 5,

2001]. These methods do not change the

method of operating the plant and have

no effect on the probability of an

accident initiating event or transient.

The basis of the SLMCPR is to ensure

no mechanistic fuel damage is

calculated to occur if the limit is not

violated. The new SLMCPR preserves

the margin to transition boiling, and the

probability of fuel damage is not

increased.

Therefore, the proposed changes to

technical specifications do not involve

an increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed changes to technical

specifications do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes result only

from revised methods of analysis for the

Cycle 14 core reload. These methods

have been reviewed and approved by

the NRC, do not involve any new or

unapproved method for operating the

facility, and do not involve any facility

modifications. No new initiating events

or transients result from these changes.

Therefore, the proposed changes to

technical specifications do not create

the possibility of a new or different kind

of accident from any accident

previously evaluated.

3. The proposed changes to technical

specifications do not involve a

significant reduction in a margin of

safety.

The margin of safety will remain the

same. The new SLMCPR was derived

using NRC approved methods which are

in accordance with the current fuel

design and licensing criteria. The

SLMCPR remains high enough to ensure

that greater than 99.9% of all fuel rods

in the core will avoid transition boiling

if the limit is not violated, which is the

current margin of safety used to

preserve the fuel cladding integrity.

Therefore, the proposed changes to

technical specifications do not involve a

significant reduction in the margin of

safety.

Based on this review, it appears that

the three standards of 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: J. M. Fulton,

Esquire, Assistant General Counsel,

Pilgrim Nuclear Power Station, 600

Rocky Hill Road, Plymouth,

Massachusetts, 02360–5599

NRC Section Chief: James W. Clifford.

Entergy Nuclear Generation Company,

Docket No. 50–293, Pilgrim Nuclear

Power Station, Plymouth County,

Massachusetts

Date of amendment request: February

16, 2001.

Description of amendment request:

This amendment would substitute a

surveillance interval of ‘‘Once/

Federal Register / Vol. 66, No. 45 / Wednesday, March 7, 2001 / Notices 13803

Operating Cycle’’ for the current

surveillance interval of ‘‘Each Refueling

Outage,’’ for the following instruments

in Technical Specification Table 4.2.F:

Containment High Radiation Monitor,

Reactor Building Vent Radiation

Monitor, Main Stack Vent Radiation

Monitor, and Turbine Building Vent

Radiation Monitor.

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 changes do not involve

a significant increase in the probability

or consequences of an accident

previously evaluated.

There are no physical changes to

Pilgrim being introduced by the

proposed changes to the specified

instruments. The proposed changes do

not modify Pilgrim, i.e., there are no

changes in operating pressure, materials

or seismic loading. No plant safety

limits, setpoints, or design parameters

are adversely affected by the proposed

changes. The proposed changes do not

adversely affect the integrity of the

reactor coolant pressure boundary such

that its function in the control of

radiological consequences is affected.

The proposed changes do enlarge the

opportunity-period for performing the

subject calibrations by substituting one

established Technical Specification

definition for another; hence, the

proposed changes are administrative in

nature because they do not change any

methodology, interval, configuration or

equipment at Pilgrim.

Thus, the proposed changes do not

affect any significant parameter

associated with the instruments or

calibration interval; therefore, the ability

of the instruments to perform their

designed safety function is maintained.

The change does not impact plant

operation. Consequently, operating

Pilgrim in conformance with the

proposed changes does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes do not create

the possibility of a new or different kind

of accident from any accident

previously evaluated.

The proposed change substitutes one

Technical Specification definition for

another concerning certain radiationmonitoring instruments. The ability of

these instruments to perform their

designed-function is not affected by this

change, and the surveillance interval

remains nominally 24 months. No new

modes of operation are introduced by

the proposed changes. No plant safety

limits, setpoints, or design parameters

are herein proposed, nor is any adverse

consequence introduced by the

proposed changes. The proposed

changes will not create any failure mode

not bounded by previously evaluated

accidents. Therefore, the proposed

changes do not create the possibility of

a new or different kind of accident from

any accident previously evaluated.

The proposed changes do not involve

a significant reduction in a margin of

safety.

The proposed changes entail the

substitution of one Technical

Specification definition for another

concerning radiation-monitoring

instruments. This is an administrative

change because such substitution does

not modify the operation, configuration,

or processes of Pilgrim, nor does the

change modify the nominal 24-month

surveillance/calibration interval

currently in force for these instruments.

The substitution of one Technical

Specification definition for another

concerning radiation monitoring

instruments potentially reduces

personnel exposure from calibrationsource radiation because site population

is less during non-refueling periods. No

plant safety limits, setpoints, or design

parameters are changed, nor is any

adverse consequence introduced by the

proposed changes. Therefore, the

proposed changes do not involve a

significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Attorney for licensee: J. M. Fulton,

Esquire, Assistant General Counsel,

Pilgrim Nuclear Power Station, 600

Rocky Hill Road, Plymouth,

Massachusetts, 02360–5599

NRC Section Chief: James W. Clifford.

Entergy Operations, Inc., Docket No. 50–

368, Arkansas Nuclear One, Unit No. 2,

Pope County, Arkansas

Date of amendment request: February

6, 2001

Description of amendment request:

The proposed amendment would revise

the Technical Specifications (TSs)

associated with the reactor coolant

system (RCS) leakage detection systems,

to make them consistent with the

requirements in NUREG–1432,

‘‘Standard Technical Specifications,

Combustion Engineering Plants.’’

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:

Criterion 1—Does Not Involve a

Significant Increase in the Probability or

Consequences of an Accident Previously

Evaluated

The aforementioned revisions do not

involve any physical change to plant

design. Relocating the requirements

associated with the RCS Leak Detection

System from various TSs to ANO–2

[Arkansas Nuclear One, Unit 2]

Specification 3.4.6.1 is administrative in

nature and does not affect the accident

analyses. The RCS water inventory

balance is more accurate than normal

leak detection methods in regard to

actual RCS leak rates, and therefore is

an excellent alternative when other leak

detection components may become

inoperable. Since the proposed changes

only affect the requirements for the

detection of RCS leakage, the probability

that an accident previously evaluated

will occur remains unchanged. The

proposed changes do not prevent nor

limit the diversity of acceptable

detection of RCS leakage and, therefore,

do not significantly affect the

consequences of an accident previously

evaluated since leak rate information

will remain available to station

personnel. Although the nonadministrative revisions result in less

restrictive requirements, the proposed

changes remain within the acceptability

of General Design Criteria (GDC) 30 of

Appendix A to 10 CFR [Part] 50 and

Regulatory Guide (RG) 1.45, and are

consistent with the philosophies of the

RSTS [Revised Standard Technical

Specifications].

Therefore, the proposed changes do

not involve a significant increase in the

probability or consequences of any

accident previously evaluated.

Criterion 2—Does Not Create the

Possibility of a New or Different Kind of

Accident From Any Previously

Evaluated

The aforementioned revisions do not

involve any physical change to plant

design. Relocating the requirements

associated with the RCS Leak Detection

System from various TSs to ANO–2

Specification 3.4.6.1 is administrative in

nature and does not affect the accident

analyses. The RCS water inventory

balance is more accurate than normal

leak detection methods in regard to

actual RCS leak rates, and therefore is

an excellent alternative when other leak

13804 Federal Register / Vol. 66, No. 45 / Wednesday, March 7, 2001 / Notices

detection components may become

inoperable. The proposed changes do

not prevent acceptable detection of RCS

leakage by diverse methods. The

detection of a RCS leak does not cause

an accident or prevent an accident from

occurring. Likewise, detecting a RCS

leak while in its initial stages does not

create the possibility of a new or

different kind of accident than any

previously analyzed. Therefore, a new

or different kind of accident than that

previously analyzed is not expected to

result due to the proposed changes of

this submittal. Although the nonadministrative revisions result in less

restrictive requirements, the proposed

changes remain within the acceptability

of General Design Criteria (GDC) 30 of

Appendix A to 10 CFR [Part] 50,

Regulatory Guide (RG) 1.45, and are

consistent with the philosophies of the

RSTS.

Therefore, the proposed changes do

not create the possibility of a new or

different kind of accident from any

previously evaluated.

Criterion 3—Does Not Involve a

Significant Reduction in the Margin of

Safety

The aforementioned revisions do not

involve any physical change to plant

design. Relocating the requirements

associated with the RCS Leak Detection

System from various TSs to the ANO–

2 Specification 3.4.6.1 is administrative

in nature and does not affect the margin

of safety. The RCS water inventory

balance is more accurate than normal

leak detection methods in regard to

actual RCS leak rates, and therefore is

an excellent alternative when other leak

detection components may become

inoperable. Maintaining diverse and

accurate RCS leak detection methods

available helps to ensure RCS leaks will

be detected within an acceptable period

of time and, therefore, the proposed

changes do not significantly reduce the

margin to safety. Although the nonadministrative revisions result in less

restrictive requirements, the proposed

changes remain within the acceptability

of General Design Criteria (GDC) 30 of

Appendix A to 10 CFR [Part] 50 and

Regulatory Guide (RG) 1.45, and are

consistent with the philosophies of the

RSTS.

Therefore, the proposed changes do

not involve a significant reduction in

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.

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, NW., Washington, DC

20005–3502.

NRC Section Chief: Robert A. Gramm.

Florida Power and Light Company,

Docket No. 50–335, St. Lucie Plant, Unit

No. 1, St. Lucie County, Florida

Date of amendment request: January

17, 2001

Description of amendment request:

The licensee proposes to revise the

Technical Specifications (TS)

requirements for the Emergency Diesel

Generator (EDG) 24-hour surveillance

test run. Currently, the TS restrict

performance of this test to shutdown

periods due to historical concerns

regarding the effects of a potential

failure while the EDGs are paralleled to

the off-site power system. The proposed

amendment would allow the

surveillance test to be conducted with

the plant on-line. The licensee has

performed an analysis, which shows

that conducting the 24-hour EDG test

run with the plant on-line results in a

very small change in core damage

frequency, and is acceptable under the

guidelines of Regulatory Guide 1.174.

The risks incurred by performing the

test on-line will be substantially offset

by plant benefits associated with

avoiding unnecessary plant transitions

and/or reducing risks during shutdown

operations.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) Operation of the facility in

accordance with the proposed

amendment would not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed amendment does not

involve a significant increase in the

probability or consequences of an

accident previously evaluated for the

following reasons:

The change relocating the ‘‘during

shutdown’’ requirement from TS

4.8.1.1.2.e to the individual surveillance

requirements under TS 4.8.1.1.2.e is

strictly administrative in nature.

Therefore, it does not involve any

increase in the probability or

consequences of an accident previously

evaluated.

For the change that revises Unit 1 TS

4.8.1.1.2.e.6 to remove the restriction to

perform the EDG 24-hour endurance test

during shutdown, the emergency diesel

generators (EDG) and their associated

emergency busses are not accident

initiating equipment. Therefore, there

will be no impact on any accident

probabilities by the approval of this

amendment. The design of this

equipment is not being modified by

these proposed changes. In addition, the

ability of the EDGs to respond to a

design basis accident will not be

significantly impacted by these

proposed changes. Consequences are no

different than presently when an EDG is

out-of-service in the current TS allowed

outage time during operation in Modes

1 and 2.

Therefore, performing the EDG 24-

hour endurance test in Modes 1 and 2

does not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

(2) Use of the modified specification

would not create the possibility of a

new or different kind of accident from

any previously evaluated.

The proposed change does not create

the possibility of a new or different kind

of accident from any accident

previously evaluated for the following

reasons:

No new accident causal mechanisms

are created as a result of this

amendment request. Equipment will be

operated in the same configuration with

the exception of the plant Mode in

which testing is conducted. No changes

are being made to the plant which

introduce any new accident causal

mechanisms. This amendment request

does not impact any plant systems that

are accident initiators; neither does it

adversely impact accident mitigating

systems.

The changes removing the restriction

to perform the tests during shutdown for

Unit 1 TS 4.8.1.1.2.e.6, in its simplest

form, is just a request to extend the

amount of time the EDG is synchronized

to the grid in Modes 1 and 2 from

approximately 18 hours (one hour per

month) to approximately 42 hours per

cycle. The existing surveillance

requirement TS 4.8.1.1.2.a.5 requires, in

part, that every 31 days each EDG be

demonstrated operable by

synchronizing to the grid for at least an

hour. It is simply a time extension of the

existing surveillance requirement.

Therefore, performing the EDG 24-hour

endurance test in Modes 1 and 2 does

not create the possibility of a new or

different kind of accident from any

previously evaluated.

(3) Use of the modified specification

would not involve a significant

reduction in a margin safety.

The AC electrical distribution system

has been designed to provide sufficient

Federal Register / Vol. 66, No. 45 / Wednesday, March 7, 2001 / Notices 13805

redundancy and reliability to ensure the

availability of the EDGs to provide the

required safety function under design

basis events to protect the power plant,

the public, and plant personnel.

The proposed changes do not affect

the limiting conditions for operation or

their bases that are used in the

deterministic analysis to establish any

margin of safety. PSA evaluations were

used to evaluate these changes, and

these evaluations determined that the

changes are not risk significant. The

proposed activity involves changes to

the allowed plant mode for the

performance specific Technical

Specification surveillance requirements.

During the performance of the EDG

endurance surveillance test for a 24-

hour period, at least one EDG will be

available and will adequately respond

within the time necessary to mitigate

anticipated operational occurrences or

postulated design basis accidents.

The calculated total change in CDF,

including the conservatively estimated

fire risk contribution, is less than 1E–06

per reactor year and the calculated total

change in the LERF, including the

conservatively estimated fire risk

contribution, is less than 1E–07 per

reactor year. The change in CDF and

LERF is, therefore, within Region III of

Regulatory Guide 1.174 Figures 3 and 4,

and is considered very small. When the

full scope of plant risk is considered,

the risks incurred by performing the

EDG 24-hour surveillance test during

power operation will be substantially

offset by plant benefits associated with

avoiding unnecessary plant transitions

and/or reducing risks during shutdown

operations.

The proposed change does not

involve a change to the plant design or

operation, and thus, does not affect the

design of the EDGs, the operational

characteristics of the EDGs, the

interfaces between the EDGs and other

plant systems, or the function or

reliability of the EDGs. Because EDG

performance and reliability will

continue to be ensured by the proposed

Technical Specification changes, the

proposed changes do not result in a

significant reduction of the margin of

safety.

Based on the above, FPL has

determined that the proposed

amendment does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated; or 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; and therefore, does 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 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Attorney for licensee: M.S. Ross,

Attorney, Florida Power & Light, P.O.

Box 14000, Juno Beach, Florida 33408–

0420.

NRC Section Chief: Richard P.

Correia.

Maine Yankee Atomic Power Company,

Docket No. 50–309, Maine Yankee

Atomic Power Station, Lincoln County,

Maine

Date of amendment request: January

4, 2001.

Description of amendment request:

The proposed amendment requests

NRC’s approval of the Maine Yankee

Atomic Power Company’s (MYAPC)

Security Plan, Training and

Qualification Plan, and Contingency

Plan. These plans reflect the addition of

provisions related to the loading and

storage of spent fuel into the

independent spent fuel storage

installation (ISFSI) under construction

on owner-controlled property adjacent

to the plant site.

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 approved Security Plan, or

Defueled Security Program, currently

implemented is not being changed. The

FIT [Fuel in Transit] Security Program

and the ISFSI Security Program are

being added to the scope of the overall

security scheme at the Maine Yankee

site. The additions to the overall plan

have been evaluated in accordance with

10 CFR 50.54(p) and 10 CFR

72.212(b)(4) and it has been determined

that the implementation of the ISFSI

and FIT Security Programs would not

decrease the effectiveness of the

Defueled Security Program, the

Defueled Security Guard Training and

Qualification Program, or the first four

categories of the Defueled Safeguards

Contingency Program.

The Defueled Security Program

Staffing will be augmented as and if

necessary to support Fuel in Transit

evolutions. The ISFSI Security Program

staffing will be separate from and

parallel to the staffing requirements of

the Defueled Security Program.

The operational and physical venues

of the Defueled Security Program, the

FIT Security Program, and the ISFSI

Security Program are separate and

distinct. The line of demarcation

between the three programs is clearly

defined and not overlapping. The

implementation of any of the programs

therefore does not degrade or inhibit the

implementation of the other two

programs.

The Defueled Program Guard Training

and Qualification Plan and the Defueled

Safeguards Contingency plan also have

not been changed. A separate and

parallel ISFSI Training and

Qualification Plan and Contingency

Plan is included in the ISFSI Security

Program. The FIT program uses the

Defueled Program, Training and

Qualification Plan and Contingency

Plan. The physical protection systems

described in the ISFSI and FIT Programs

are designed to protect against the loss

of control of the facility that could be

sufficient to cause a radiation exposure

exceeding the dose as described in 10

CFR 72.106.

Therefore, the ISFSI Program

revisions of the Security Plan, Guard

Training and Qualification Plan and the

Safeguards Contingency Plan will not

increase the probability or the

consequences of an accident previously

evaluated since the previously approved

Defueled Training and Qualification

Plan and Contingency Plan remain

unchanged.

2. The proposed change does not

create the possibility of a new or

different kind of accident from any

accident previously evaluated.

The FIT and ISFSI Security Programs

have no impact on the existing Defueled

Security Program since they operate in

different physical and licensing venues.

The accidents considered for the Spent

Fuel Pool, the venue of the Defueled

Security Program, are described in the

Maine Yankee Defueled Safety Analysis

Report. The accidents considered for the

FIT and ISFSI are contained in the NAC

International, Inc. Final Safety Analysis

Report for the UMS Universal Storage

System Docket No. 72–1015.

The FIT and ISFSI Security Programs

have been crafted to meet or exceed all

of the assumptions of the NAC

International FSAR concerning accident

analyses and the programs meet or

exceed all of the applicable

requirements of 10 CFR 73.55 with

approved exceptions or approved

alternative measures. The physical

protection systems described in the

ISFSI and FIT Programs are designed to

protect against the loss of control of the

13806 Federal Register / Vol. 66, No. 45 / Wednesday, March 7, 2001 / Notices

facility that could be sufficient to cause

a radiation exposure exceeding the dose

as described in 10 CFR 72.106.

The proposed action does not affect

plant systems, structures or components

within the venue of the existing

Security Plan. The ISFSI and FIT

program additions to the Security Plan,

Guard Training and Qualification Plan

and the Safeguards Contingency Plan do

not create the possibility of a new or

different kind of accident from any

accident previously evaluated since the

previously approved Defueled Security

Plan, Training and Qualification Plan

and Contingency plan remain as is,

unaltered.

3. The proposed change does not

involve a significant reduction in a

margin of safety.

The addition of a separate, parallel

ISFSI and FIT Safeguards Program,

Training and Qualification Plan, and

Contingency Plan does not alter or

reduce the effectiveness of the

previously approved Defueled Program.

The physical protection systems

described in the ISFSI and FIT Programs

are designed to protect against the loss

of control of the facility that could be

sufficient to cause a radiation exposure

exceeding the dose as described in 10

CFR 72.106. Therefore, the margin of

safety will not be reduced as a result of

the ISFSI and FIT additions to the

Security Plan, or an ISFSI specific

addition of a Guard Training and

Qualification Plan or an ISFSI specific

addition of a Safeguards Contingency

Plan

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

requested amendment involves no

significant hazards consideration.

Attorney for licensee: Joseph Fay,

Esquire, Maine Yankee Atomic Power

Company, 321 Old Ferry Road,

Wiscasset, Maine 04578.

NRC Section Chief: Michael T.

Masnik.

Northern States Power Company,

Docket Nos. 50–282 and 50–306, Prairie

Island Nuclear Generating Plant, Units

1 and 2, Goodhue County, Minnesota

Date of amendment requests: April

17, 2000, as supplemented February 2,

2001.

Description of amendment requests:

The proposed amendments would

change the Technical Specifications

(TSs) for the removal of boric acid

storage tanks (BASTs) from the safety

injection (SI) system. These changes

would accomplish two objectives: (1)

Eliminate high concentration boric acid

from the SI system and (2) align this

specific Prairie Island TS section with

the Standard TSs.

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 amendment will not

involve a significant increase in the

probability or consequences of accidents

previously evaluated.

The proposed change to the CVCS

[chemical volume control system] and

SI system (increasing the concentration

of boric acid in the RWST [refueling

water storage tank] and eliminating the

BAST as a suction source, respectively)

and elimination of or change to

associated Technical Specifications do

not affect accident initiation. None of

the equipment being removed from

Sections 3.2 or 3.5 of Technical

Specifications are accident initiators.

Thus, the proposed changes will not

significantly increase the probability of

an accident previously evaluated.

Consequences are evaluated in terms

of off-site and on-site (control room

personnel) dose. Loss of coolant

accident (LOCA) dose is unaffected by

the proposed changes because the LOCA

analysis input assumptions are not

changed by the changes proposed in this

amendment request. The approved

steam line break (SLB) methodology

(approved by the NRC in letter dated

January 19, 2000) and the expected dose

are unaffected by the proposed change.

Therefore, the proposed changes will

not involve a significant increase in the

probability or consequences of an

accident previously evaluated.

(2) The proposed amendment will not

create the possibility of a new or

different kind of accident from any

accident previously evaluated.

The proposed changes to the plant

and its Technical Specifications do not

introduce any new accident initiators.

The proposed changes reduce the

number of automatic component

actuations needed to support Safety

Injection accident mitigation functions.

The proposed changes also remove the

Technical Specification requirements

for the balance of the CVCS

components. These requirements were

in Technical Specifications to support

the boration function of CVCS; however,

all boration functions can be met by the

safety-related SI system. All the other

functions of the CVCS are either backed

up by a safety related system or are not

required to preclude an accident

(reference NSP [Northern States Power]

letter of June 14, 1995 and NRC letter of

January 8, 1996).

Therefore, the proposed changes will

not create the possibility of a new or

different kind of accident.

(3) The proposed amendment will not

involve a significant reduction in the

margin of safety.

The proposed changes do not

significantly impact the plant response

to an accident with respect to the ability

to protect fission product barriers. The

proposed changes will not result in any

significant increase in fuel cladding

damage in the event of a postulated

accident (accident analyses show the

proposed changes meet all acceptance

criteria related to maintaining cladding

integrity). The proposed changes will

not reduce the integrity of the RCS

[reactor coolant system] (reduction of

boric acid concentrations in the SI

systems will not promote any

degradation of the components that

make up the RCS pressure boundary).

The proposed changes will not result in

a reduction in containment integrity in

the event of a postulated accident (the

changes proposed by this amendment

do not change the results of the accident

analyses with respect to containment

response.)

Therefore, the proposed changes will

not involve a significant reduction in

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 requests involve no

significant hazards consideration.

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts, and Trowbridge,

2300 N Street, NW., Washington, DC

20037.

NRC Section Chief: Claudia M. Craig.

Sacramento Municipal Utility District

(SMUD), Docket No. 50–312, Rancho

Seco Nuclear Station, Sacramento

County, California

Date of amendment request: October

23, 2000.

Description of amendment request:

The proposed amendment (PA–194) as

supplemented by SMUD letter to the

USNRC dated January 11, 2001, would

change the Permanently Defueled

Technical Specification (PDTS) by

deleting the definitions for ‘‘site

boundary’’ and ‘‘unrestricted area;’’

revising the definition of the ‘‘site;’’

deleting figures D5.1–1, ‘‘Emergency

Planning Zone,’’ D5.1–2, ‘‘Site

Boundary for Gaseous Effluent,’’ and

D5.1–3, ‘‘Site Boundary for Liquid

Effluent;’’ and making editorial changes

Federal Register / Vol. 66, No. 45 / Wednesday, March 7, 2001 / Notices 13807

to the other PDTSs because of the above

proposed changes. The information

proposed for removal from the PDTS is

contained in or will be relocated to

other licensee-controlled documents.

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:

SMUD has reviewed the proposed

PDTS change against each of the criteria

in 10 CFR 50.92 and has concluded that

the amendment request involves no

significant hazards consideration. The

following provides SMUD’s analysis of

the issue of no significant hazards

consideration:

1. Does the proposed license

amendment involve a significant

increase in the probability or

consequences of an accident previously

evaluated?

No. The proposed changes are

administrative and involve deleting the

definitions of SITE BOUNDARY and

UNRESTRICTED AREA from the

DEFINITIONS section, revising the

definition of the site in Section 5.1

‘‘SITE,’’ deleting all three figures from

the DESIGN FEATURES section [SMUD

proposes, as described in its January 11,

2001, letter, that these or equivalent

figures will be relocated to either the

Emergency Plan or the Offsite Dose

Calculation Manual, as appropriate],

revising Sections D6.8.3.a(2) and

D6.8.3.a(4) so that the term

‘‘unrestricted area’’ is lower case, and

revising Sections D6.8.3.a(8),

D6.8.3.a(9), D6.8.3.a(10), and D6.8.3.b(2)

so that the term ‘‘site boundary’’ is

lower case.

These changes do not affect possible

initiating events for accidents

previously evaluated or alter the

configuration or operation of the

facility. Safety limits, limiting safety

system settings, and limiting control

systems are no longer applicable to

Rancho Seco Technical Specifications

in the permanently defueled mode, and

are therefore not relevant.

The proposed changes do not affect

the emergency planning zone, the

boundaries used to evaluate compliance

with liquid or gaseous effluent limits,

and have no impact on plant operations.

Therefore, the proposed license

amendment does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. Does the proposed license

amendment create the possibility of a

new or different kind of accident from

any accident previously evaluated?

No. As described above, the proposed

changes are administrative. The safety

analysis for the facility remains

complete and accurate. There are no

physical changes to the facility and the

plant conditions for which the design

basis accidents have been evaluated are

still valid.

The operating procedures and

emergency procedures are not affected.

The proposed changes do not affect the

emergency planning zone, the

boundaries used to evaluate compliance

with liquid or gaseous effluent limits,

and have no impact on plant operations.

Consequently, no new failure modes are

introduced as the result of the proposed

changes. Therefore, the proposed

changes will not create the possibility of

a new or different kind of accident from

any accident previously evaluated.

3. Does the proposed license

amendment involve a significant

reduction in a margin of safety?

No. As described above, the proposed

changes are administrative. There are no

changes to the design or operation of the

facility. The proposed changes do not

affect the emergency planning zone, the

boundaries used to evaluate compliance

with liquid or gaseous effluent release

limits, and have no impact on plant

operations. Accordingly, neither the

design basis nor the accident

assumptions in the Defueled Safety

Analysis Report (DSAR), nor the

Technical Specification Bases are

affected. Therefore, the proposed

changes do not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

requested amendment involves no

significant hazards consideration.

Attorney for licensee: Dana Appling,

Esq., Sacramento Municipal Utility

District, P.O. Box 15830, Sacramento,

California 95852–1830.

NRC Section Chief: Michael T.

Masnik.

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 amendment request: August

25, 2000.

Description of amendment request:

The proposed amendments would

revise the Updated Final Safety

Analysis Report (UFSAR) described

offsite dose analyses based on changes

to the letdown flow rate and iodine

spike postulated concurrent with a Main

Steam Line Break or a Steam Generator

Tube Rupture.

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. Does the change involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

The proposed changes do not

significantly increase the probability or

consequences of an accident previously

evaluated in the UFSAR. The

comprehensive engineering review

included evaluations or re-analysis of

all accident analyses. Calculations for

letdown flow measurement and

indication have verified the

acceptability of the analyzed letdown

flow rate. The letdown flow rate does

not initiate any accident; therefore, the

probability of an accident has not been

increased. All dose consequences have

been analyzed or evaluated with respect

to the proposed changes, and all

acceptance criteria continue to be met.

Therefore, these changes do not involve

a significant increase in the probability

or consequences of an accident

previously evaluated.

2. Does the change create the

possibility of a new or different kind of

accident from any accident previously

analyzed?

The proposed changes do not create

the possibility of a new or different kind

of accident than any accident already

evaluated in the UFSAR. No new

accident scenarios, failure mechanisms

or limiting single failures are introduced

as a result of the proposed changes. The

changes have no adverse effects on any

safety-related system and do not

challenge the performance or integrity

of any safety-related system. Therefore,

all accident analyses criteria continue to

be met and these changes do not create

the possibility of a new or different kind

of accident from any accident

previously evaluated.

3. Does the change involve a

significant reduction in a margin of

safety?

The proposed changes do not involve

a significant reduction in a margin of

safety. All analyses and evaluations

using letdown flow rate as an input

have been revised to reflect the

proposed value. The calculations are

based on FNP instrumentation and test

methods and include uncertainty

allowances. The evaluations and

analyses results [a small change]

demonstrate applicable acceptance

criteria are met. Therefore, the proposed

13808 Federal Register / Vol. 66, No. 45 / Wednesday, March 7, 2001 / Notices

changes do not involve a significant

reduction in 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.

Attorney for licensee: M. Stanford

Blanton, Esq., Balch and Bingham, Post

Office Box 306, 1710 Sixth Avenue

North, Birmingham, Alabama 35201.

NRC Section Chief (Acting): Maitri

Banerjee.

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 amendment request:

December 8, 2000.

Description of amendment request:

The proposed amendments would either

delete or modify existing license

conditions from the Unit 1 and Unit 2

Operating Licenses, which have been

completed or are otherwise no longer in

effect. These activities have now been

completed, and the license conditions

are either obsolete or no longer needed.

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. Does the change involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

The proposed amendment deletes

license conditions which are completed

or are otherwise obsolete. As such, the

change is strictly administrative.

Therefore, this change does not involve

a significant increase in the probability

or consequences of an accident

previously evaluated.

2. Does the change create the

possibility of a new or different kind of

accident from any accident previously

analyzed?

The proposed amendment deals with

operating license reporting conditions

and has no effect on the type of

accidents that have been considered at

Plant Farley. Therefore, this change

does not create the possibility of a new

or different kind of accident from any

accident previously evaluated.

3. Does the change involve a

significant reduction in a margin of

safety?

The requirements associated with the

deleted license conditions have been

completed; the conditions are therefore

obsolete. Removing these conditions

from the license is an administrative

and editorial activity. 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.

Attorney for licensee: M. Stanford

Blanton, Esq., Balch and Bingham, Post

Office Box 306, 1710 Sixth Avenue

North, Birmingham, Alabama 35201.

NRC Section Chief (Acting): M.

Banerjee.

Union Electric Company, Docket No.

50–483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application request: January

18, 2001 (ULNRC–04371).

Description of amendment request:

The proposed amendment deletes

Section 5.5.3, ‘‘Post Accident

Sampling,’’ from the administrative

controls section of the Technical

Specifications (TS). The proposed

amendment deletes requirements from

the TS (and, as applicable, other

elements of the licensing bases) to

maintain a Post Accident Sampling

System (PASS). Licensees were

generally required to implement PASS

upgrades as described in NUREG–0737,

‘‘Clarification of TMI [Three Mile

Island] Action Plan Requirements,’’ and

Regulatory Guide 1.97,

‘‘Instrumentation for Light-WaterCooled Nuclear Power Plants to Assess

Plant and Environs Conditions During

and Following an Accident.’’

Implementation of these upgrades was

an outcome of the lessons learned from

the accident that occurred at TMI Unit

2. Requirements related to PASS were

imposed by Order for many facilities

and were added to or included in the TS

for nuclear power reactors currently

licensed to operate. Lessons learned and

improvements implemented over the

last 20 years have shown that the

information obtained from PASS can be

readily obtained through other means or

is of little use in the assessment and

mitigation of accident conditions.

The NRC staff issued a notice of

opportunity for comment in the Federal

Register on August 11, 2000 (65 FR

49271) on possible amendments to

eliminate PASS, including a model

safety evaluation and model no

significant hazards consideration

(NSHC) determination, using the

consolidated line item improvement

process. The NRC staff subsequently

issued a notice of availability of the

models for referencing in license

amendment applications in the Federal

Register on October 31, 2000 (65 FR

65018). The licensee affirmed the

applicability of the following NSHC

determination in its application dated

January 18, 2001.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), an

analysis of the issue of no significant

hazards consideration is presented

below:

Criterion 1—The Proposed Change Does

Not Involve a Significant Increase in the

Probability or Consequences of an

Accident Previously Evaluated.

The PASS was originally designed to

perform many sampling and analysis

functions. These functions were

designed and intended to be used in

post accident situations and were put

into place as a result of the TMI–2

accident. The specific intent of the

PASS was to provide a system that has

the capability to obtain and analyze

samples of plant fluids containing

potentially high levels of radioactivity,

without exceeding plant personnel

radiation exposure limits. Analytical

results of these samples would be used

largely for verification purposes in

aiding the plant staff in assessing the

extent of core damage and subsequent

offsite radiological dose projections. The

system was not intended to and does

not serve as a function for preventing

accidents and its elimination would not

affect the probability of accidents

previously evaluated.

In the 20 years since the TMI–2

accident and the consequential

promulgation of post accident sampling

requirements, operating experience has

demonstrated that a PASS provides

little actual benefit to post accident

mitigation. Past experience has

indicated that there exists in-plant

instrumentation and methodologies

available in lieu of a PASS for collecting

and assimilating information needed to

assess core damage following an

accident. Furthermore, the

implementation of Severe Accident

Management Guidance (SAMG)

emphasizes accident management

strategies based on in-plant instruments.

These strategies provide guidance to the

plant staff for mitigation and recovery

from a severe accident. Based on current

severe accident management strategies

and guidelines, it is determined that the

PASS provides little benefit to the plant

staff in coping with an accident.

The regulatory requirements for the

PASS can be eliminated without

Federal Register / Vol. 66, No. 45 / Wednesday, March 7, 2001 / Notices 13809

degrading the plant emergency

response. The emergency response, in

this sense, refers to the methodologies

used in ascertaining the condition of the

reactor core, mitigating the

consequences of an accident, assessing

and projecting offsite releases of

radioactivity, and establishing

protective action recommendations to

be communicated to offsite authorities.

The elimination of the PASS will not

prevent an accident management

strategy that meets the initial intent of

the post-TMI–2 accident guidance

through the use of the SAMGs, the

emergency plan (EP), the emergency

operating procedures (EOP), and site

survey monitoring that support

modification of emergency plan

protective action recommendations

(PARs).

Therefore, the elimination of PASS

requirements from Technical

Specifications (TS) (and other elements

of the licensing bases) does not involve

a significant increase in the

consequences of any accident

previously evaluated.

Criterion 2—The Proposed Change Does

Not Create the Possibility of a New or

Different Kind of Accident from any

Previously Evaluated

The elimination of PASS related

requirements will not result in any

failure mode not previously analyzed.

The PASS was intended to allow for

verification of the extent of reactor core

damage and also to provide an input to

offsite dose projection calculations. The

PASS is not considered an accident

precursor, nor does its existence or

elimination have any adverse impact on

the pre-accident state of the reactor core

or post accident confinement of

radionuclides within the containment

building.

Therefore, this change does not create

the possibility of a new or different kind

of accident from any previously

evaluated.

Criterion 3—The Proposed Change Does

Not Involve a Significant Reduction in

the Margin of Safety.

The elimination of the PASS, in light

of existing plant equipment,

instrumentation, procedures, and

programs that provide effective

mitigation of and recovery from reactor

accidents, results in a neutral impact to

the margin of safety. Methodologies that

are not reliant on PASS are designed to

provide rapid assessment of current

reactor core conditions and the

direction of degradation while

effectively responding to the event in

order to mitigate the consequences of

the accident. The use of a PASS is

redundant and does not provide quick

recognition of core events or rapid

response to events in progress. The

intent of the requirements established as

a result of the TMI–2 accident can be

adequately met without reliance on a

PASS.

Therefore, this change does not

involve a significant reduction in the

margin of safety.

Based upon the reasoning presented

above and the previous discussion of

the amendment request, the requested

change does not involve a significant

hazards consideration.

The NRC staff proposes to determine

that the amendment request involves no

significant hazards consideration.

Attorney for licensee: John O’Neill,

Esq., Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Section Chief: Stephen Dembek.

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, located at One White Flint North,

11555 Rockville Pike (first floor),

Rockville, Maryland 20852. Publicly

available records will be accessible and

electronically from the ADAMS Public

Library component on the NRC Web

site, http://www.nrc.gov (the Electronic

Reading Room).

Calvert Cliffs Nuclear Power Plant, Inc.,

Docket Nos. 50–317 and 50–318, Calvert

Cliffs Nuclear Power Plant, Unit Nos. 1

and 2, Calvert County, Maryland

Date of application for amendments:

September 14, 2000.

Brief description of amendments: The

amendments add two analytical

methods to the list of approved core

operating limit analytical methods in

Technical Specification 5.6.5.b.

Date of issuance: February 8, 2001.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment Nos.: 241 and 215.

Renewed Facility Operating License

Nos. DPR–53 and DPR–69: Amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: October 18, 2000 (65 FR

62383).

The Commission’s related evaluation

of these amendments is contained in a

Safety Evaluation dated February 8,

2001.

No significant hazards consideration

comments received: No.

Consolidated Edison Company of New

York, Docket No. 50–247, Indian Point

Nuclear Generating Unit No. 2,

Westchester County, New York

Date of application for amendment:

November 22, 1999, as supplemented on

September 11, 2000.

Brief description of amendment: The

amendment revises Technical

Specification Sections 4.5.D,

‘‘Containment Air Filtration System,’’

4.5.E, ‘‘Control Room Air Filtration

System,’’ 4.5.F, ‘‘Fuel Storage Building

Air Filtration System,’’ and 4.5.G, ‘‘PostAccident Containment Venting

System,’’ to address the testing

requirements in Generic Letter 99–02,

‘‘Laboratory Testing of Nuclear-Grade

Activated Charcoal.’’ The laboratory

testing of the engineered safeguards

features ventilation system charcoal

samples will meet the requirements of

the American Society for Testing and

Materials Standard D3803–1989.

Date of issuance: February 21, 2001.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 215.

13810 Federal Register / Vol. 66, No. 45 / Wednesday, March 7, 2001 / Notices

Facility Operating License No. DPR–

26: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: November 15, 2000 (65 FR

69059).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated February 21,

2001.

No significant hazards consideration

comments received: No.

Energy Northwest, Docket No. 50–397,

Columbia Generating Station, Benton

County, Washington

Date of application for amendment:

October 30, 2000.

Brief description of amendment: The

amendment revises Surveillance

Requirement 3.6.1.3.8 to allow a

representative sample of reactor

instrument line excess flow check

valves (EFCVs) to be tested every 24

months such that each reactor

instrument EFCV will be tested at least

once every 10 years. The amendment

also limits the surveillance requirement

to only the reactor instrument line

EFCVs.

Date of issuance: February 20, 2001.

Effective date: February 20, 2001, and

shall be implemented within 30 days

from the date of issuance.

Amendment No.: 170.

Facility Operating License No. NPF–21:

The amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: November 29, 2000 (65 FR

71135).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated February 20,

2001.

No significant hazards consideration

comments received: No.

Entergy Nuclear Generation Company,

Docket No. 50–293, Pilgrim Nuclear

Power Station, Plymouth County,

Massachusetts

Date of application for amendment:

November 22, 1999, as supplemented on

November 21, 2000.

Brief description of amendment: This

amendment approves changes related to

Technical Specification (TS) Sections

3.7.B.1 and 3.7.B.2, ‘‘Containment

Systems.’’ TS Section 5.0,

‘‘Administrative Controls,’’ was also

modified to reflect the addition of an

omitted page from a previous

amendment.

Date of issuance: February 13, 2001.

Effective date: As of the date of

issuance, and shall be implemented

within 60 days.

Amendment No.: 187.

Facility Operating License No. DPR–35:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: April 5, 2000 (65 FR 17913).

The November 21, 2000, letter

provided clarifying information that did

not change the initial proposed no

significant hazards consideration

determination. The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

February 13, 2001.

No significant hazards consideration

comments received: No.

Exelon Generation Company, Docket

Nos. STN 50–454 and STN 50–455,

Byron Station, Unit Nos. 1 and 2, Ogle

County, Illinois; Docket Nos. STN 50–

456 and STN 50–457, Braidwood

Station, Unit Nos. 1 and 2, Will County,

Illinois

Date of application for amendments:

February 15, 2000, as supplemented on

July 26, 2000. The July 26, 2000, letter

provided clarifying information that did

not change the scope of the February 15,

2000, application or the initial proposed

no significant hazards consideration

determination.

Brief description of amendments: The

amendments allow the use of the

Westinghouse core monitoring system

know as Best Estimate Analyzer for Core

Operations Nuclear.

Date of issuance: February 13, 2001.

Effective date: February 13, 2001.

Amendment Nos.: 116, 116, 110, and

110.

Facility Operating License Nos. NPF–

37, NPF–66, NPF–72 and NPF–77: The

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: April 5, 2000 (65 FR 17909).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated February 13,

2001.

No significant hazards consideration

comments received: No.

Niagara Mohawk Power Corporation,

Docket No. 50–220, Nine Mile Point

Nuclear Station, Unit 1, Oswego County,

New York

Date of application for amendment:

September 26, 2000.

Brief description of amendment: The

amendment changes the Technical

Specifications to (1) allow reactor vessel

hydrostatic tests, leakage tests, scram

time tests and excess flow check valve

tests be performed; (2) require

containment building integrity be

maintained; and (3) establish a limit and

a surveillance requirement on reactor

coolant radioactive iodine activity,

when coolant temperature is above 215

°F, the reactor is not critical, and

primary containment integrity has not

been established.

Date of issuance: February 20, 2001.

Effective date: As of the date of

issuance to be implemented within 30

days of issuance.

Amendment No.: 170.

Facility Operating License No. DPR–63:

Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: November 1, 2000 (65 FR

65344).

The staff’s related evaluation of the

amendment is contained in a Safety

Evaluation dated February 20, 2001.

No significant hazards consideration

comments received: No.

Nuclear Management Company, LLC,

Docket No. 50–305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of application for amendment:

November 10, 2000.

Brief description of amendment: The

amendment revised several sections of

the Kewaunee Nuclear Power Plant

(KNPP) Technical Specifications (TSs).

These sections include administrative

changes, Table 4.1–1, and Sections 1.0,

6.4, and 6.10.

Date of issuance: February 12, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 30 days.

Amendment No.: 151.

Facility Operating License No. DPR–43:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: December 13, 2000 (65 FR

77923).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated February 12,

2001.

No significant hazards consideration

comments received: No.

Portland General Electric Company, et

al., Docket No. 50–344, Trojan Nuclear

Plant, Columbia County, Oregon

Date of application for amendment:

August 5, 1999, as supplemented by

letters dated November 23, 1999,

December 27, 1999, May 4, 2000,

October 19, 2000, and November 22,

2000.

Brief description of amendment: The

amendment revised the Facility

Operating (Possession Only) License to

annotate approval of the Trojan Nuclear

Plant License Termination Plan.

Date of issuance: February 12, 2001.

Effective date: February 12, 2001, and

shall be implemented within 30 days of

the effective date.

Federal Register / Vol. 66, No. 45 / Wednesday, March 7, 2001 / Notices 13811

Amendment No.: 206.

Facility Operating License No. NPF–1:

The amendment changes the Facility

Operating (Possession Only) License.

Date of initial notice in Federal

Register: December 29, 1999 (64 FR

73083). The November 23, 1999,

December 27, 1999, May 4, 2000,

October 19, 2000, and November 22,

2000, supplemental letters provided

additional clarifying information, did

not expand the scope of the application

as originally noticed, and did not

change the staff’s original proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated February 12,

2001.

No significant hazards consideration

comments received: No.

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of application for amendments:

August 4, 2000 (TS 99–20).

Brief description of amendments:

Deletes Sequoyah License Condition for

Shift Technical Advisor and revises

Technical Specifications (TSs) that

specify shift manning requirements.

Date of issuance: February 16, 2001.

Effective date: February 16, 2001.

Amendment Nos.: 266 and 257.

Facility Operating License Nos. DPR–

77 and DPR–79: Amendments revise the

Operating Licenses and TSs.

Date of initial notice in Federal

Register: September 6, 2000 (65 FR

54088).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated February 16,

2001.

No significant hazards consideration

comments received: No.

Union Electric Company, Docket No.

50–483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application for amendment:

November 21, 2000 (ULNRC–04346).

Brief description of amendment: The

amendment changes Table 3.3.2–1,

‘‘Engineered Safety Feature Actuation

System [ESFAS] Instrumentation,’’ of

the Technical Specifications. The

change adds Surveillance Requirement

(SR) 3.3.2.10 for the following two

ESFAS instrumentation in the table:

item 6.f, loss of offsite power, and item

6.h, auxiliary feedwater pump suction

transfer on suction pressure—low.

Date of issuance: February 12, 2001.

Effective date: February 12, 2001, and

shall be implemented prior to entering

Mode 3 from Mode 4 during the startup

from Refuel Outage 11, including the

revision of the FSAR to reflect the

ESFAS response times in accordance

with the application.

Amendment No.: 141.

Facility Operating License No. NPF–

30: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: December 27, 2000 (65 FR

81931).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated February 12,

2001.

No significant hazards consideration

comments received: No.

Previously Published Notices of

Consideration of Issuance of

Amendments to Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing

The following notices were previously

published as separate individual

notices. The notice content was the

same as above. They were published as

individual notices either because time

did not allow the Commission to wait

for this biweekly notice or because the

action involved exigent circumstances.

They are repeated here because the

biweekly notice lists all amendments

issued or proposed to be issued

involving no significant hazards

consideration.

For details, see the individual notice

in the Federal Register on the day and

page cited. This notice does not extend

the notice period of the original notice.

Nuclear Management Company, LLC,

Docket No. 50–263, Monticello Nuclear

Generating Plant, Wright County,

Minnesota

Date of amendment request: February

1, 2001.

Brief description of amendment

request: The amendment would remove

the inservice inspection requirements of

Section XI of the American Society of

Mechanical Engineers Boiler and

Pressure Vessel Code from the

Monticello Technical Specifications and

relocates them to a licensee-controlled

program.

Date of publication of individual

notice in Federal Register: February 15,

2001 (66 FR 10535).

Expiration date of individual notice:

March 1, 2001.

Notice of Issuance of Amendments to

Facility Operating Licenses and Final

Determination of No Significant

Hazards Consideration and

Opportunity for a Hearing (Exigent

Public Announcement or Emergency

Circumstances)

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 for the

amendment 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.

Because of exigent or emergency

circumstances associated with the date

the amendment was needed, there was

not time for the Commission to publish,

for public comment before issuance, its

usual 30-day Notice of Consideration of

Issuance of Amendment, Proposed No

Significant Hazards Consideration

Determination, and Opportunity for a

Hearing.

For exigent circumstances, the

Commission has either issued a Federal

Register notice providing opportunity

for public comment or has used local

media to provide notice to the public in

the area surrounding a licensee’s facility

of the licensee’s application and of the

Commission’s proposed determination

of no significant hazards consideration.

The Commission has provided a

reasonable opportunity for the public to

comment, using its best efforts to make

available to the public means of

communication for the public to

respond quickly, and in the case of

telephone comments, the comments

have been recorded or transcribed as

appropriate and the licensee has been

informed of the public comments.

In circumstances where failure to act

in a timely way would have resulted, for

example, in derating or shutdown of a

nuclear power plant or in prevention of

either resumption of operation or of

increase in power output up to the

plant’s licensed power level, the

Commission may not have had an

opportunity to provide for public

comment on its no significant hazards

consideration determination. In such

case, the license amendment has been

issued without opportunity for

comment. If there has been some time

for public comment but less than 30

days, the Commission may provide an

13812 Federal Register / Vol. 66, No. 45 / Wednesday, March 7, 2001 / Notices

opportunity for public comment. If

comments have been requested, it is so

stated. In either event, the State has

been consulted by telephone whenever

possible.

Under its regulations, the Commission

may issue and make an amendment

immediately effective, notwithstanding

the pendency before it of a request for

a hearing from any person, in advance

of the holding and completion of any

required hearing, where it has

determined that no significant hazards

consideration is involved.

The Commission has applied the

standards of 10 CFR 50.92 and has made

a final determination that the

amendment involves no significant

hazards consideration. The basis for this

determination is contained in the

documents related to this action.

Accordingly, the amendments have

been issued and made effective 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 application for

amendment, (2) the amendment to

Facility Operating License, 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, located at One White Flint North,

11555 Rockville Pike (first floor),

Rockville, Maryland 20852, and

electronically from the ADAMS Public

Library component on the NRC Web

site, http://www.nrc.gov (the Electronic

Reading Room).

The Commission is also offering an

opportunity for a hearing with respect to

the issuance of the amendment. By

April 6, 2001, 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, located at One

White Flint North, 11555 Rockville Pike

(first floor), Rockville, Maryland 20852,

and electronically from the ADAMS

Public Library component on the NRC

Web site, http://www.nrc.gov (the

Electronic Reading Room). 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. Since the Commission has

made a final determination that the

amendment involves no significant

hazards consideration, if a hearing is

requested, it will not stay the

effectiveness of the amendment. Any

hearing held would take place while the

amendment is in effect.

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 20555–001, Attention:

Rulemakings and Adjudications Staff, or

may be delivered to the Commission’s

Public Document Room, located at One

White Flint North, 11555 Rockville Pike

(first floor), Rockville, Maryland 20852,

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 20555–001, 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 the

factors specified in 10 CFR

2.714(a)(1)(i)–(v) and 2.714(d).

AmerGen Energy Company, LLC, Docket

No. 50–289, Three Mile Island Nuclear

Station, Unit 1, Dauphin County,

Pennsylvania

Date of application for amendment:

February 14, 2001, as supplemented

February 16 and 19, 2001. The February

Federal Register / Vol. 66, No. 45 / Wednesday, March 7, 2001 / Notices 13813

16 and 19, 2001, letters provided

additional clarifying information which

did not change the initial proposed no

significant hazards consideration

determination or expand the

amendment beyond the scope of the

original notice (Harrisburg, PA, Patriot

News, February 18–20, 2001).

Brief description of amendment: The

amendment allows a one-time exception

to the system configuration and

maintenance requirements in Technical

Specification (TS) 3.3.2 related to the

nuclear service river water (NR) system

at TMI–1, in order to allow an up to 14-

day repair of a leaking underground

concrete pipe. The requirements of TS

3.3.1.4 to have two NR pumps

OPERABLE are unchanged. During the

14-day repair period, the NR pumps

flow will be realigned to pass through

a portion of the nonseismic secondary

services river water system.

Date of issuance: February 23, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 30 days.

Amendment No.: 229.

Facility Operating License No. DPR–

50. Amendment revised the Technical

Specifications.

Public comments requested as to

proposed no significant hazards

consideration: Yes.

The NRC published a public notice of

the proposed amendment, issued a

proposed finding of no significant

hazards consideration and requested

that any comments on the proposed no

significant hazards consideration be

provided to the staff by the close of

business on February 23, 2001. The

notice was published in the Harrisburg,

PA, Patriot News, from February 18

through February 20, 2001.

The Commission’s related evaluation

of the amendment, finding of exigent

circumstances, consultation with the

State of Pennsylvania, and final no

significant hazards consideration

determination are contained in a Safety

Evaluation dated February 23, 2001.

Attorney for licensee: Edward J.

Cullen, Jr., Esquire, PECO Energy

Company, 2301 Market Street (S23–1),

Philadelphia, PA 19103.

NRC Section Chief: Marsha

Gamberoni.

Dated at Rockville, Maryland this 27th day

of February 2001.

For the Nuclear Regulatory Commission.

John A. Zwolinski,

Director, Division of Licensing Project

Management, Office of Nuclear Reactor

Regulation.

[FR Doc. 01–5216 Filed 3–6–01; 8:45 am]

BILLING CODE 7590–01–P

OFFICE OF PERSONNEL

MANAGEMENT

Proposed Collection; Comment

Request for Review of a Revised

Information Collection: RI 94–7

AGENCY: Office of Personnel

Management.

ACTION: Notice.

SUMMARY: In accordance with the

Paperwork Reduction Act of 1995

(Public Law 104–13, May 22, 1995), this

notice announces that the Office of

Personnel Management (OPM) intends

to submit to the Office of Management

and Budget a request for review of a

revised information collection. RI 94–7,

Death Benefit Payment Rollover

Election for Federal Employees

Retirement System (FERS), provides

FERS surviving spouses and former

spouses with the means to elect

payment of the FERS rollover-eligible

benefits directly or to an Individual

Retirement Account.

Comments are particularly invited on:

whether this information is necessary

for the proper performance of functions

of OPM, and whether it will have

practical utility; whether our estimate of

the public burden of this collection of

information is accurate, and based on

valid assumptions and methodology;

and ways in which we can minimize the

burden of the collection of information

on those who are to respond, through

the use of appropriate technological

collection techniques or other forms of

information technology.

Approximately 700 RI 94–7 forms will

be completed annually. We estimate it

takes approximately 60 minutes to

complete the form. The annual

estimated burden is 700 hours.

For copies of this proposal, contact

Mary Beth Smith-Toomey on (202) 606–

8358, or E-mail to mbtoomey@opm.gov

DATES: Comments on this proposal

should be received on or before May 7,

2001.

ADDRESSES: Send or deliver comments

to: John C. Crawford, Chief, FERS

Division, Retirement and Insurance

Service, U.S. Office of Personnel

Management, 1900 E Street, NW, Room

3313, Washington, DC 20415.

FOR INFORMATION REGARDING

ADMINISTRATIVE COORDINATION CONTACT:

Donna G. Lease, Team Leader, Forms

Analysis and Design, Budget and

Administrative Services Division, (202)

606–0623.

U.S. Office of Personnel Management.

Steven R. Cohen,

Acting Director.

[FR Doc. 01–5517 Filed 3–6–01; 8:45 am]

BILLING CODE 6325–50–U

SECURITIES AND EXCHANGE

COMMISSION

[Release No. IC–24881; 812–12266]

ING Pilgrim Investments, LLC, et al.;

Notice of Application

February 28, 2001.

AGENCY: Securities and Exchange

Commission (‘‘Commission’’).

ACTION: Notice of application for an

order under section 6(c) of the

Investment Company Act of 1940

(‘‘Act’’) for an exemption from sections

18(c) and 18(i) of the Act, under

sections 6(c) and 23(c)(3) of the Act for

an exemption from rule 23c-3 under the

Act, and pursuant to section 17(d) of the

Act and rule 17d-1 under the Act.

SUMMARY OF APPLICATION: Applicants

request on order to permit certain

registered closed-end management

investment companies to issue multiple

classes of shares and to impose assetbased distribution fees and early

withdrawal charges.

APPLICANTS: Pilgrim Senior Income

Fund (‘‘Fund’’), ING Pilgrim

Investments, LLC (‘‘Investment

Adviser’’), and ING Pilgrim Securities,

Inc. (‘‘ING Pilgrim Securities’’).

FILING DATES: The application was filed

on September 25, 2000 and amended on

February 28, 2001.

HEARING OR NOTIFICATION OF HEARING: An

order granting the application will be

issued unless the Commission orders a

hearing. Interested persons may request

a hearing by writing to the

Commission’s Secretary and serving

applicants with a copy of the request,

personally or by mail. Hearing requests

should be received by the Commission

by 5:30 p.m. on March 26, 2001, and

should be accompanied by proof of

service on applicants, in the form of an

affidavit, or for lawyers, a certificate of

service. Hearing requests should state

the nature of the writer’s interest, the

reason for the request, and the issues

contested. Persons who wish to be

notified of a hearing may request

notification by writing to the

Commission’s Secretary.

ADDRESSES: Secretary, Commission, 450

Fifth Street, NW, Washington, DC

20549–0609. Applicants, 7337 East

Doubletree Ranch Road, Scottsdale,

Arizona, 85258.