Federal Register / Vol. 66, No. 104 / Wednesday, May 30, 2001 / Notices 29349

8065) between 7:30 a.m. and 4:30 p.m.

(EDT). Persons planning to attend this

meeting are urged to contact the above

named individual one or two working

days prior to the meeting to be advised

of any potential changes to the agenda,

etc., that may have occurred.

Dated: May 22, 2001.

James E. Lyons,

Associate Director for Technical Support.

[FR Doc. 01–13488 Filed 5–29–01; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Sunshine Act Meeting

DATES: Weeks of May 28, June 4, 11, 18,

25, July 2, 2001.

PLACE: Commissioners’ Conference

Room, 11555 Rockville Pike, Rockville

Maryland.

STATUS: Public and Closed.

Matters To Be Considered

Week of May 28, 2001

Wednesday, May 30, 2001

10:25 a.m.—Affirmation Session

(Public Meeting) (If needed)

Week of June 4, 2001—Tentative

Tuesday, June 5, 2001

9:25 a.m.—Affirmation Session

(Public Meeting) (If needed)

2:00 p.m.—Discussion of Management

Issues (Closed-Ex. 2)

Wednesday, June 6, 2001

10:30 a.m.—All Employees Meeting

(Public Meeting)

1:30 p.m.—All Employees Meeting

(Public Meeting)

Week of June 11, 2001—Tentative

Thursday, June 14, 2001

9:55 a.m.—Affirmation Session

(Public Meeting) (If needed)

10:00 a.m.—Meeting with Nuclear

Waste Technical Review Board

(Public Meeting)

1:30 p.m.—Briefing on License

Renewal Program (Public Meeting)

(Contact: David Solorio, 301–415–

1973)

Week of June 18, 2001—Tentative

There are no meetings scheduled for the

Week of June 18, 2001

Week of June 25, 2001—Tentative

Wednesday, June 27, 2001

9:25 a.m.—Affirmation Session

(Public Meeting) (If needed)

Week of July 2, 2001—Tentative

There are no meetings scheduled for the

Week of July 2, 2001

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.

The NRC Commission Meeting

Schedule can be found on the Internet

at

http://www.nrc.gov/SECY/smji/

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 the Commission meeting

schedule electronically, please send an

electronic message to dkw@nrc.gov.

Dated: May 24, 2001.

David Louis Gamberoni,

Technical Coordinator, Office of the

Secretary.

[FR Doc. 01–13605 Filed 5–25–01; 10:16 am]

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 May 7, 2001

through May 18, 2001. The last

biweekly notice was published on May

16, 2001 (66 FR 27174).

Notice of Consideration of Issuance of

Amendments to Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a

proposed determination that the

following amendment requests involve

no significant hazards consideration.

Under the Commission’s regulations in

10 CFR 50.92, this means that operation

of the facility in accordance with the

proposed amendment would not (1)

involve a significant increase in the

probability or consequences of an

accident previously evaluated; or (2)

create the possibility of a new or

different kind of accident from any

accident previously evaluated; or (3)

involve a significant reduction in a

margin of safety. The basis for this

proposed determination for each

amendment request is shown below.

The Commission is seeking public

comments on this proposed

determination. Any comments received

within 30 days after the date of

publication of this notice will be

considered in making any final

determination.

Normally, the Commission will not

issue the amendment until the

expiration of the 30-day notice period.

However, should circumstances change

during the notice period such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received before

action is taken. Should the Commission

take this action, it will publish in the

Federal Register a notice of issuance

and provide for opportunity for a

hearing after issuance. The Commission

expects that the need to take this action

will occur very infrequently.

Written comments may be submitted

by mail to the Chief, Rules and

Directives Branch, Division of

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

29350 Federal Register / Vol. 66, No. 104 / Wednesday, May 30, 2001 / Notices

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 June 29, 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

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).

Carolina Power & Light Company, et al.,

Docket Nos. 50–325 and 50–324,

Brunswick Steam Electric Plant, Units 1

and 2, Brunswick County, North

Carolina

Date of amendments request: May 1,

2001.

Description of amendments request:

The proposed amendments would

revise the pressure-temperature limits

curves contained in Technical

Specification 3.4.9, ‘‘RCS Pressure and

Temperature (P/T) Limits.’’

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The Proposed License Amendments Do

Not Involve a Significant Increase in the

Probability or Consequences of an Accident

Previously Evaluated

The changes to the calculation

methodology for the pressure-temperature

limits are based on American Society of

Mechanical Engineers (ASME) Code Case N–

640, ‘‘Alternative Reference Fracture

Toughness for Development of P–T Limit

Curves for ASME Section XI, Division 1,’’

and provide adequate margin in the

prevention of a non-ductile type fracture of

Federal Register / Vol. 66, No. 104 / Wednesday, May 30, 2001 / Notices 29351

the reactor pressure vessel. The code case

was developed based upon the knowledge

gained through years of industry experience.

The pressure-temperature limits developed

using the allowances of ASME Code Case N–

640 provide more operating margin.

However, experience gained in the areas of

fracture toughness of materials and preexisting undetected defects shows that some

of the existing assumptions used for the

calculation of pressure-temperature limits are

unnecessarily conservative and unrealistic.

Therefore, use of the allowances of ASME

Code Case N–640 in developing the pressuretemperature limits will provide adequate

protection against nonductile-type fractures

of the reactor pressure vessel.

Development of the revised BSEP

[Brunswick Steam Electric Plant], Unit 1 and

2 pressure-temperature limits was performed

using the approved methodologies of 10 CFR

50, Appendix G, and using the allowances of

ASME Code Case N–640. The pressuretemperature limits generated using these

methods ensure the pressure-temperature

limits will not be exceeded during any phase

of reactor operation. Therefore, the

probability of occurrence and the

consequences of a previously analyzed event

are not significantly increased. Finally, the

proposed changes will not affect any other

system or piece of equipment designed for

the prevention or mitigation of previously

analyzed events.

Thus, the probability of occurrence and the

consequences of any previously analyzed

event are not significantly increased as the

result of the proposed changes to the

pressure-temperature limits.

2. The Proposed License Amendments Will

Not Create the Possibility of a New or

Different Kind of Accident From Any

Accident Previously Evaluated

The proposed changes provide more

operating margin in the pressure-temperature

limits for hydrostatic pressure testing, nonnuclear heatup and cooldown, and criticality,

with the benefits being primarily realizable

during the pressure tests. The changes also

extend the pressure-temperature limits for

use up to 32 EFPY [effective full-power

years] of operation. However, operation in

the ‘‘new’’ regions of the pressuretemperature limits has been analyzed and

will provide adequate protection against a

nonductile-type fracture of the reactor

pressure vessel. Otherwise, the proposed

pressure-temperature limits do not result in

any new or unanalyzed operation of any

system or piece of equipment important to

safety and, as a result, the possibility of a

new type event is not created.

3. The Proposed License Amendments Do

Not Involve a Significant Reduction in a

Margin of Safety

The revised pressure-temperature limits

provide more operating margin and

operational flexibility than the existing

pressure-temperature limits. With the

increased operational margin, a reduction in

the safety margin results with respect to the

existing limits. However, the industry

experience since the inception of pressuretemperature limits confirms that some of the

existing methodologies used to develop

pressure-temperature limits are unrealistic

and unnecessarily conservative. Accordingly,

ASME Code Case N–640 takes advantage of

this acquired knowledge by establishing

more realistic methodologies for the

development of pressure-temperature limits.

Therefore, operational flexibility is gained

and an acceptable margin of safety to reactor

pressure vessel non-ductile type fracture is

maintained. Evaluation of the revised

pressure-temperature limits for use up to 32

EFPY was performed using 10 CFR 50 and

ASME Code Case N–640; thus, the margin of

safety is not significantly reduced as the

result of the proposed changes.

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: William D.

Johnson, Vice President and Corporate

Secretary, Carolina Power & Light

Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Section Chief: Patrick M.

Madden, Acting.

Dominion Nuclear Connecticut, Inc.,

Docket No. 50–336, Millstone Nuclear

Power Station, Unit No. 2, New London

County, Connecticut

Date of amendment request: April 26,

2001.

Description of amendment request:

The proposed amendment would add

Section 6.22, ‘‘Reactor Coolant Pump

Flywheel Inspection Program’’ to

Section 6, ‘‘Administrative Controls’’ of

the Technical Specifications (TSs) and

relocate the requirements of TS 3/4.4.10,

‘‘Reactor Coolant System, Structural

Integrity’’ to the Millstone Unit No. 2

Technical Requirements Manual (TRM).

The Bases of the affected TSs would

also be relocated to the TRM. The Index

pages would also be updated to reflect

these changes.

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. Involve a Significant Increase in the

Probability or Consequences of An Accident

Previously Evaluated

Missile generation from a Reactor Coolant

Pump (RCP) flywheel could damage the

Reactor Coolant System, the Containment, or

other equipment or systems important to

safety. The fracture mechanics analyses

conducted to support the change to Inservice

Inspection (ISI) requirements in accordance

with the proposed Section 6.22, ‘‘Reactor

Coolant Pump Flywheel Inspection Program’’

shows that a pre-existing crack sized just

below the detection level will not grow to the

flaw size necessary to create flywheel

missiles within the life of the plant. This

analysis conservatively assumes minimum

material properties, maximum flywheel

accident speed, location of the flaw in the

highest stress area, and a number of startup/

shutdown cycles eight times greater than

expected. Since an existing flaw in a

Millstone Unit No. 2 flywheel will not grow

to the allowable flaw size under Loss of

Coolant Accident (LOCA) conditions over the

life of the plant, reducing the ISI

requirements for the detection of such cracks

over the life of the plant will not significantly

increase the probability or consequences of

an accident previously evaluated

The proposed Technical Specification

changes to relocate the requirements for

Technical Specification 3/4.4.10, ‘‘Reactor

Coolant System, Structural Integrity’’ (with

the exception of the RCP inspection

requirements) to the TRM will have no

adverse effect on plant operation or the

availability or operation of any accident

mitigation equipment. Therefore, the Reactor

Coolant System structural integrity (with the

exception of the RCP flywheel which is

addressed above) will not adversely impact

an accident initiator and can not cause an

accident. Therefore these changes will not

increase the probability or consequences of

an accident previously evaluated.

The Index pages will be updated to reflect

the proposed changes. These changes are

administrative in nature. These changes will

not increase the probability or consequences

of an accident previously evaluated.

2. Create the Possibility of a New or Different

Kind of Accident From Any Accident

Previously Evaluated

The proposed changes will not alter the

plant configuration (no new or different type

of equipment will be installed) or require any

new or unusual operator actions. They do not

alter the way any structure, system, or

component functions and do not alter the

manner in which the plant is operated. These

changes do not introduce any new failure

modes. Therefore, the proposed changes will

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Involve a Significant Reduction in a

Margin of Safety

The fracture mechanics analyses

conducted to support the change to ISI

requirements in accordance with the

proposed Section 6.22, ‘‘Reactor Coolant

Pump Flywheel Inspection Program’’ shows

that significant conservatism has been used

for calculating the allowable flaw size,

critical flaw size, and crack growth rate in the

RCP flywheels. These include minimum

material properties, maximum flywheel

accident speed, location of the flaw in the

highest stress area and a number of startup/

shutdown cycles eight times greater than

expected. Since an existing flaw in a

Millstone Unit No. 2 flywheel will not grow

to the allowable flaw size under normal

operating conditions or to the critical flaw

size under LOCA conditions over the life of

the plant, reducing ISI requirements for the

29352 Federal Register / Vol. 66, No. 104 / Wednesday, May 30, 2001 / Notices

detection of such cracks over the life of the

plant will not involve a significant reduction

in the margin of safety. The proposed

changes have no impact on plant equipment

operation. Therefore, the proposed changes

will not result in a reduction in a margin of

safety.

Relocation of Technical Specification 3/

4.4.10 (whole specification except the

portion specifying surveillance requirement

for the RCP flywheel) to the TRM does not

imply any reduction in its importance in

ensuring that the structural integrity and

operational readiness of ASME Code Class 1,

2, and 3 components will be maintained at

an acceptable level throughout the life of the

plant. The proposed change has no impact on

plant equipment operation. Therefore, the

proposed change will not result in a

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Lillian M.

Cuoco, Senior Nuclear Counsel,

Dominion Nuclear Connecticut, Inc.,

Rope Ferry Road, Waterford, CT 06385.

NRC Section Chief: James W. Clifford.

Entergy Operations, Inc., Docket Nos.

50–313 and 50–368, Arkansas Nuclear

One, Units 1 and 2 (ANO–1&2), Pope

County, Arkansas

Date of amendment request: January

27, 2000, as supplemented by letter

dated March 1, 2001.

Description of amendment request:

The proposed changes to Arkansas

Nuclear One, Unit 1 (ANO–1) and Unit

2 (ANO–2), Technical Specifications

(TSs) allow for the qualified condensate

storage tank (QCST) to be used for both

units as the preferred source of water for

emergency feedwater (EFW). Currently,

the QCST is aligned to the ANO–1 EFW

system while ANO–2 relies on nonsafety related tanks and an automatic

switchover to the Service Water System

as the source of EFW coolant water.

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

staff’s analysis is presented below.

Criterion 1—Does Not Involve a

Significant Increase in the Probability or

Consequences of an Accident Previously

Evaluated

The condensate storage tanks provide

a source of condensate grade water for

the EFW System. The tanks, one for

ANO–1 and two for ANO–2, are already

included in the plant TSs. The proposed

change allows for both units to operate

while aligned to the QCST, but does not

affect the physical design, construction,

or operation of the condensate storage

tanks. These tanks are not associated

with the precursors of any accident.

This change does not increase the

probability of any accident previously

evaluated.

As a source of EFW, the tanks serve

an accident mitigation function. The

proposed change does not alter this

function. In addition to the tanks, the

Service Water System is also available

as a long-term assured source of EFW.

The proposed change allows the use of

the QCST as the preferred source of

EFW for both units. The combination of

available sources of water for EFW

assures that both units are able to

respond to accidents previously

evaluated. Because this function

continues to be assured, the proposed

changes do not increase the

consequences of a previously evaluated

accident.

Therefore, this change does 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 Accident Previously

Evaluated

The condensate storage tanks serve an

accident mitigation function as a

temporary source of EFW. These tanks

have not been identified as a precursor

to any accident previously evaluated.

The design and operation of these tanks

have not changed. While the proposed

change does permit the qualified tank to

be used by ANO–2, the design has been

evaluated and it has been demonstrated

that the existing tank is capable of

meeting the intended design function of

both units.

Therefore, this change does not create

the possibility of a new or different kind

of accident from any accident

previously evaluated.

Criterion 3—Does Not Involve a

Significant Reduction in a Margin of

Safety

The existing sources of water for the

ANO–1 EFW system will continue to

ensure adequate EFW system

performance after the proposed change.

The QCST and, if necessary, the Service

Water System will ensure that the EFW

system performs to maintain margins of

safety. The Service Water System is the

assured long-term source of cooling

water for both units. The safety function

of decay heat removal and core cooling

continues to be met. There is no

reduction in the margin of safety for

ANO–1.

The proposed change to the ANO–2

specifications will provide a qualified

alternative source of EFW. The required

function of the tanks is the same as for

ANO–1; that is, to provide a source of

water until the unit can successfully

transfer to decay heat cooling or until

the Service Water System is aligned for

long-term cooling. The addition of this

QCST to the specification as an

alternative to the existing tanks does not

decrease the margin of safety.

Therefore, this change does not

involve a significant reduction in a

margin of safety.

Based on this analysis, 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.

Entergy Operations, Inc., Docket No. 50–

368, Arkansas Nuclear One, Unit No. 2,

Pope County, Arkansas

Date of amendment request: May 2,

2001.

Description of amendment request:

The proposed amendment would

relocate the requirements for the

containment recirculation system from

the technical specifications to the

Technical Requirements Manual (TRM).

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?

The containment recirculation fans, along

with the containment cooling units and

containment spray systems, provide a means

of circulating the containment atmosphere to

ensure adequate mixing of the containment

atmosphere. The containment cooling units

and containment spray systems are safetyrelated systems and required by TS 3.6.2.3

and 3.6.2.1, respectively. Adequate air

mixing is assured with the use of these two

systems. The containment recirculation fans

are not credited in the mitigation of any

accidents.

Based on an evaluation of the criteria listed

in 10 CFR 50.36(c)(2)(ii), the relocation of the

Federal Register / Vol. 66, No. 104 / Wednesday, May 30, 2001 / Notices 29353

containment recirculation fans to the TRM is

acceptable.

Therefore, this change does 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 containment recirculation fans are not

accident initiators. The function they fulfill

will continue to be maintained by the

containment cooling units and containment

spray pumps. Because the proposed

amendment will not change the design,

configuration or method of plant operation,

it will not create the possibility of a new or

different kind of accident.

Therefore, this change does 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?

Air mixing of the containment atmosphere

can be accomplished following a LOCA [lossof-coolant accident] by the containment

recirculation fans, the containment cooling

units, or the containment spray systems. Any

one of these systems is capable of providing

adequate air mixing. The proposed change

does not change the design function of the

containment recirculation fans. Additionally,

the containment recirculation fans are not

credited in any accident analysis. Since

adequate mixing of the containment

atmosphere is credited through the

containment cooling units and spray systems,

relocation of the containment recirculation

fan requirements to the TRM does not result

in any impact to the margin of safety.

Therefore, this change 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

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.

Entergy Operations Inc., Docket No. 50–

382, Waterford Steam Electric Station,

Unit 3, St. Charles Parish, Louisiana

Date of amendment request: May 3,

2001.

Description of amendment request:

The proposed amendment deletes

requirements from the Technical

Specifications (TSs) (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

Nuclear Station] 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

TSs 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 Nuclear Regulatory Commission

(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

May 3, 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 [Three Mile Island Nuclear

Station, Unit 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 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 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

29354 Federal Register / Vol. 66, No. 104 / Wednesday, May 30, 2001 / Notices

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: N. S. Reynolds,

Esquire, Winston & Strawn, 1400 L

Street NW., Washington, DC 20005–

3502.

NRC Section Chief: Robert A. Gramm.

Exelon Generation Company, LLC, PSEG

Nuclear LLC, and Atlantic City Electric

Company, Dockets Nos. 50–277 and 50–

278, Peach Bottom Atomic Power

Station, Units Nos. 2 and 3, York

County, Pennsylvania

Date of application for amendments:

February 8, 2001.

Description of amendment request:

The proposed amendment would revise

the Peach Bottom Atomic Power Station

(PBAPS), Units 2 and 3, technical

specifications (TSs) and the associated

TS Bases, to reflect changes to support

the activation of the trip outputs of the

oscillation power range monitor (OPRM)

portion of the power range neutron

monitoring (PRNM) system and delete

the interim corrective action

requirements from the TSs. The OPRM

trip function provides protection from

exceeding the fuel minimum critical

power ratio (MCPR) safety limit in the

event of thermal-hydraulic power

oscillations. PBAPS is currently

operating under interim corrective

actions that specify restrictions on plant

operations and actions by operators in

response to power oscillations. The

OPRM system provides an automatic

reactor trip which eases the burden on

the operators if power oscillations were

to occur.

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 review is presented below:

1. The Proposed Amendment Does Not

Involve a Significant Increase in the

Probability or Consequences of an

Accident Previously Evaluated

This modification has no impact on

any of the existing PRNM functions. It

connects the OPRM trip function to the

reactor protection system; connects the

associated trip alarm to the annunciator

circuitry; updates the TSs to add the

OPRM-related functions and to delete

Interim Corrective Actions (ICAs)

related requirements; and revises

affected procedures.

The ICAs include a restricted region

on the power-to-flow map where

thermal-hydraulic instabilities were

known to be more likely. Operation in

the restricted region requires more

frequent monitoring of the average

power range monitors (APRMs) and

local power range monitors (LPRMs),

which are part of the PRNM system.

This restricted region is less than 10

percent of the full power-to-flow map.

Plant operation in portions of the former

restricted region without the increased

monitoring of APRMs and LPRMs

previously required by the ICAs may

cause a slight increase in the probability

of occurrence of an instability. This

potential increase in probability is not

significant because operation in this

region will still result in a low

likelihood of core power oscillations.

Because of the more reliable detection of

an instability event, should it occur, the

automatic scram if preset limits are

exceeded, and the elimination of

dependence on the operator, the

consequences of an instability event are

not increased with this modification.

Based on the above discussion, the

proposed amendment does not involve

a significant increase in the probability

or consequences of an accident

previously evaluated.

2. The Proposed Amendment Does Not

Create the Possibility of a New or

Different Kind of Accident From Any

Accident Previously Evaluated

Enabling the OPRM reactor scram

function does not create any new system

interactions except for the reactor scram

function. The failure modes for the new

OPRM circuits would be to initiate a

reactor scram unnecessarily, or to fail to

initiate a reactor scram when

instabilities were present. These failures

would not create the possibility of a

new or different kind of accident. Since

the present system has no automatic

reactor scram for instabilities, the

operators insert a manual scram if

necessary, and the effect of core

instabilities has been analyzed. The use

of a manual scram is still available with

the OPRM scram function enabled.

Removing the ICAs from the TSs does

not create the possibility of a new or

different kind of accident, since the

effect of core instabilities has been

evaluated.

Based on the above discussion, the

proposed amendment will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The Proposed Amendment Does Not

Involve a Significant Reduction in a

Margin of Safety.

The current safety analyses assume

that the existing ICA related TS

requirements are adequate to prevent

exceeding the MCPR safety limit due to

an instability event. As a result, there is

currently no quantitative or qualitative

assessment of an instability event with

respect to its impact on MCPR.

The OPRM trip function is being

implemented to automate the detection

(via direct measurement of neutron flux)

and subsequent suppression (via reactor

scram) of an instability event prior to

exceeding the MCPR safety limit. The

OPRM trip provides a trip output of the

same type as currently used for the

APRMs. Its failure modes and types are

identical to those for the present APRM

output. Currently, the MCPR safety limit

is not impacted by an instability event

since the event is mitigated by manual

means via the ICAs. In both methods of

mitigation (manual and automated), the

margin of safety associated with the

MCPR safety limit is maintained.

Therefore, based on the fact that the

MCPR safety limit will not be exceeded

as a result of an instability event

following implementation of the OPRM

trip function in place of the existing

manual ICAs, it is concluded that the

proposed amendment does not reduce a

margin of safety.

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: Mr. Edward

Cullen, Vice President and General

Counsel, Exelon Generation Company,

LLC, 300 Exelon Way, Kennett Square,

PA 19348.

NRC Section Chief: James W. Clifford.

FirstEnergy Nuclear Operating

Company, et al., Docket No. 50–334,

Beaver Valley Power Station, Unit No. 1

(BVPS–1), Shippingport, Pennsylvania

Date of amendment request: March

28, 2001.

Federal Register / Vol. 66, No. 104 / Wednesday, May 30, 2001 / Notices 29355

Description of amendment request:

The requested amendment proposes

changes to the BVPS–1 Technical

Specifications (TSs) associated with the

reductions of the reactor coolant system

and secondary coolant system specific

activity limits. These TS changes

support a revised main steam line break

safety analysis with a higher assumed

primary-to-secondary leak rate in

accordance with the methodology

described in Nuclear Regulatory

Commission (NRC) Generic Letter (GL)

95–05, ‘‘Voltage-Based Repair Criteria

for Westinghouse Steam Generator

Tubes by Outside Diameter Stress

Corrosion Cracking.’’ TS Bases and

other administrative changes are

proposed for consistency.

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 review is presented below.

1. Does the Change Involve a Significant

Increase in the Probability or

Consequences of an Accident Previously

Evaluated?

The proposed change involves the

reduction of the Technical Specification

Dose Equivalent Iodine 131 (I–131)

activity limits for the reactor coolant

system (RCS) and the secondary system

which facilitates an increase in the

assumed accident-induced primary-tosecondary leak rate in the event of a

postulated main steam line break

(MSLB) accident. There are no proposed

changes to any facility structures,

systems, or components. The proposed

changes do not affect any initiators of

accidents previously evaluated nor does

the proposed change introduce any new

failure mechanisms that may initiate a

previously-evaluated accident.

Furthermore, the proposed change

would not affect the ability of any

accident mitigation system to perform

its design-basis function as defined in

the Updated Final Safety Analysis

Report (UFSAR). The dose consequence

analysis for a postulated MSLB accident

are being revised as part of this

amendment and the resulting calculated

dose consequences do not increase.

Therefore, the proposed 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 Evaluated?

The proposed change is only

associated with the reduction of the RCS

and secondary system I–131 activity

limits and does not involve changes to

any facility structures, systems, or

components. There are no proposed

changes to the facility or its operation.

Since there are no changes being made

to any structures, systems, or

components, no new failure

mechanisms are introduced by the

proposed changes that would result in

the occurrence of a new or different

kind of accident from any accident

previously evaluated. The accident

analyses contained in the UFSAR

continue to remain bounding with

regard to the spectrum of possible

accidents.

Therefore, the proposed 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?

There are no proposed changes to any

structure, systems, or components.

Changes proposed to the dose

consequence analysis for a postulated

MSLB accident are included in the

amendment request. The reduction in

the RCS and secondary system activity

limits are being made to offset the

effects of an increased accident-induced

primary-to-secondary leak rate resulting

from a postulated MSLB accident in

accordance with GL 95–05. The margins

to safety that could be affected are those

associated with the resulting calculated

doses to the public and facility

personnel. However, the dosedecreasing effect of lowering the activity

limits offsets the dose-increasing effect

of raising the assumed accident-induced

primary-to-secondary leak rate.

Consequently, the resulting calculated

doses do not increase.

Therefore, the proposed change does

not involve a significant reduction in a

margin to safety.

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: Mary O’Reilly,

FirstEnergy Nuclear Operating

Company, FirstEnergy Corporation, 76

South Main Street, Akron, OH 44308.

NRC Section Chief: Richard P. Correia

(Acting).

FirstEnergy Nuclear Operating

Company, Docket No. 50–346, DavisBesse Nuclear Power Station, Unit 1,

Ottawa County, Ohio

Date of amendment request: March

30, 2001.

Description of amendment request:

The proposed amendment would

involve changes to Technical

Specification (TS) 3/4.5.2, Emergency

Core Cooling—ECCS Subsystems—Tavg

≥ 280°F.

Technical Specification Limiting

Condition for Operation (LCO) 3.5.2

requires two independent Emergency

Core Cooling Systems (ECCS)

Subsystems to be operable. Surveillance

Requirement (SR) 4.5.2.f requires each

ECCS subsystem to be demonstrated

operable by performing a vacuum

leakage rate test of the watertight

enclosure for Decay Heat Removal

System valves DH–11 and DH–12 that

assures the motor operator on valves

DH–11 and DH–12 will not be flooded

for at least (7) days following a Loss-ofCoolant Accident (LOCA). The test is

required to be performed: (1) At least

once per 18 months, (2) After each

opening of the watertight enclosure, and

(3) After any maintenance on or

modification to the watertight enclosure

which could affect its integrity.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensees have provided their analysis of

the issue of no significant hazards

consideration, which is presented

below:

1a. Not involve a significant increase in the

probability of an accident previously

evaluated because no such accidents are

affected by the proposed changes. Initial

conditions and assumptions remain as

previously analyzed for accidents in the

Davis-Besse Nuclear Power Station Updated

Safety Analysis Report.

The proposed changes would increase the

surveillance test interval in Technical

Specification 4.5.2.f.1 from 18 to 24 months

for the vacuum leakage rate test of the

watertight enclosure for Decay Heat Removal

System valves DH–11 and DH–12. The

surveillance data and maintenance records

have been reviewed and support an increase

in the surveillance test interval from 18 to 24

months based on the low potential for a

significant increase in the failure rate of the

watertight enclosure due to an increased

surveillance interval, and based on the

introduction of no new failure modes. The

proposed change to the surveillance interval

has been evaluated consistent with the NRC

guidance on evaluating and proposing such

revisions as provided in Generic Letter 91–

04, ‘‘Changes in Technical Specification

Surveillance Intervals to Accommodate a 24-

Month Fuel Cycle,’’ dated April 2, 1991. The

watertight enclosure and its condition do not

contribute to the initiation of any accident.

29356 Federal Register / Vol. 66, No. 104 / Wednesday, May 30, 2001 / Notices

Therefore, the probability of any accident

previously evaluated is not increased.

1b. Not involve a significant increase in the

consequences of an accident previously

evaluated because the integrity of the

watertight enclosure sealing mechanisms has

been evaluated, and it has been determined

that the sealing mechanisms will remain

intact for the proposed increased surveillance

interval. Therefore, there is assurance that

the backup boric acid precipitation control

flow path will remain available, so that there

will be no impact on the source term,

containment isolation or radiological

releases.

2. Not create the possibility of a new or

different kind of accident from any accident

previously evaluated because the proposed

changes do not alter the manner in which the

watertight enclosure is sealed or tested, and

the operability requirements of Decay Heat

Removal System valves DH–11 and DH–12

will continue to be adequately addressed by

Surveillance Requirement 4.5.2.f.1.

No changes are being proposed to the type

of testing currently being performed, only to

the length of the surveillance test interval.

An increase in the surveillance test interval

from 18 to 24 months is justified based on

the low potential for a significant increase in

the failure rate of the watertight enclosure

due to an increased surveillance interval, and

based on the introduction of no new failure

modes.

No different accident initiators or failure

mechanisms are introduced by the proposed

change. Thus, it does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. Not involve a significant reduction in a

margin of safety.

An increase in the surveillance test interval

from 18 to 24 months is justified based on

the low potential for a significant increase in

the failure rate of the watertight enclosure

due to an increased surveillance interval, and

based on the introduction of no new failure

modes.

Since there are no new or significant

changes to the initial conditions contributing

to accident severity or consequences, there

are no significant reductions in a margin of

safety.

On the basis of the above, the Davis-Besse

Nuclear Power Station has determined that

the License Amendment Request does not

involve a significant hazards consideration.

As this License Amendment Request

concerns a proposed change to the Technical

Specifications that must be reviewed by the

Nuclear Regulatory Commission, this License

Amendment Request does not constitute an

unreviewed safety question.

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: Mary E.

O’Reilly, Attorney, FirstEnergy

Corporation, 76 South Main Street,

Akron, OH 44308

NRC Section Chief: Anthony J.

Mendiola.

FirstEnergy Nuclear Operating

Company, Docket No. 50–346, DavisBesse Nuclear Power Station, Unit 1,

Ottawa County, Ohio

Date of amendment request: April 1,

2001.

Description of amendment request:

The proposed amendment would add

new Technical Specification (TS)

Administrative Controls Section 6.17,

TS Bases Control Program, and make a

related change to the TS Index. The

proposed new TS Administrative

Control would provide requirements for

changing and updating the TS Bases.

This proposed new TS is similar to the

Specification 5.5.14 of NUREG–1430,

‘‘Standard Technical Specifications—

Babcock and Wilcox Plants,’’ Revision

1, April 1995.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensees have provided their analysis of

the issue of no significant hazards

consideration, which is presented

below:

1a. Not involve a significant increase in the

probability of an accident previously

evaluated because no such accidents are

affected by the proposed changes. The

amendment application proposes to add a

new Technical Specification (TS)

Administrative Controls Section 6.17,

‘‘Technical Specifications (TS) Bases Control

Program,’’ and to make a related change to

the TS Index. The proposed changes do not

involve a change to any structure, system, or

component or to the assumptions of any

accident analyses.

1b. Not involve a significant increase in the

consequences of an accident previously

evaluated because no equipment, accident

conditions, or assumptions are affected

which could lead to a significant increase in

radiological consequences.

2. Not create the possibility of a new or

different kind of accident from any accident

previously evaluated because no new or

different accident initiators are introduced by

these proposed changes.

3. Not involve a significant reduction in a

margin of safety because there are no new or

significant changes to the initial conditions

contributing to accident severity or

consequences. Consequently, there are no

significant reductions in a margin of safety.

On the basis of the above, the DBNPS has

determined that the License Amendment

Request does not involve a significant

hazards consideration. As this License

Amendment Request concerns a proposed

change to the Technical Specifications that

must be reviewed by the Nuclear Regulatory

Commission, this License Amendment

Request does not constitute an unreviewed

safety question.

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: Mary E.

O’Reilly, Attorney, FirstEnergy

Corporation, 76 South Main Street,

Akron, OH 44308.

NRC Section Chief: Anthony J.

Mendiola.

FirstEnergy Nuclear Operating

Company, Docket No. 50–346, DavisBesse Nuclear Power Station, Unit 1,

Ottawa County, Ohio

Date of amendment request: April 1,

2001.

Description of amendment request:

The proposed amendment would

involve changes to Technical

Specification (TS) 3/4.3.1, Reactor

Protection System Instrumentation; 3/

4.3.2.1, Safety Features Actuation

System Instrumentation; TS 3/4.3.2.2,

Steam and Feedwater Rupture Control

System Instrumentation; and Bases 3/

4.3.1 and 3/4.3.2, Reactor Protection

System and Safety System

Instrumentation.

The proposed changes would revise

TS Table 3.3–3, Safety Features

Actuation System (SFAS)

Instrumentation, TS Table 3.3–11,

Steam and Feedwater Rupture Control

System (SFRCS) Instrumentation, and

associated Bases to add a provision to

allow an eight-hour delay in entering an

Action statement when an SFAS or

SFRCS instrumentation channel is

undergoing Channel Functional Testing.

The proposed changes would provide a

reasonable time to perform the required

surveillance testing and relieve the

control room staff of the burden of

making multiple Action statement

entries and exits in order to complete

the testing. Additionally, Surveillance

Requirements 4.3.1.1.2, 4.3.2.1.2, and

4.3.2.2.2 would be revised to clarify the

term ‘‘total bypass function.’’

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensees have provided their analysis of

the issue of no significant hazards

consideration, which is presented

below:

1a. Not involve a significant increase in the

probability of an accident previously

evaluated because no such accidents are

affected by the proposed changes. The

amendment application proposes to add a

provision to TS Table 3.3–3, Safety Features

Actuation System (SFAS) Instrumentation,

and TS Table 3.3–11, Steam and Feedwater

Federal Register / Vol. 66, No. 104 / Wednesday, May 30, 2001 / Notices 29357

Rupture Control System (SFRCS)

Instrumentation, to permit certain SFAS and

SFRCS instrument channels to [be] placed in

an inoperable condition for up to 8 hours

during surveillance testing without declaring

the channel inoperable and entering the

Action statement. This proposed change

would reduce burden placed on the control

room operators and is essentially

administrative in nature. The proposed

change to the TS Bases 3/4.3.1 and 3/4.3.2,

Reactor Protection System and Safety System

instrumentation, is associated with the

changes to TS Tables 3.3–3 and 3.3–11.

These changes will not significantly change

testing methodology, system unavailability,

or system reliability. Initiating conditions

and assumptions remain as previously

analyzed for accidents in the DBNPS

Updated Safety Analysis Report (USAR).

The proposed changes to Limiting

Condition for Operation 3.3.1.1, Surveillance

Requirement (SR) 4.3.1.1.2, SR 4.3.2.1.2, and

SR 4.3.2.2.2 to clarify the nomenclature of

the Reactor Protection System (RPS), SFAS,

and SFRCS bypass functions being tested are

administrative in nature. These changes will

not effect any plant hardware or the

performance of any test. Initiating conditions

and assumptions remain as previously

analyzed for accidents in the DBNPS USAR.

1b. Not involve a significant increase in the

consequences of an accident previously

evaluated because the source term,

containment isolation, or radiological

releases are not affected by the proposed

changes. Existing system and component

redundancy is not affected by the proposed

changes. The existing system and component

operation is not affected by the proposed

changes, and the assumptions used in

evaluating the radiological consequences in

the DBNPS USAR are not invalidated.

Therefore, for each postulated accident the

consequences remain bounded by the

consequences from the previously evaluated

accidents.

2. Not create the possibility of a new or

different kind of accident from any accident

previously evaluated because these proposed

changes do not involve any physical changes

to systems or components, nor do they alter

the manner in which the systems or

components are operated. No new or

different accident initiators or equipment

failure modes are introduced by the proposed

changes.

3. Not involve a significant reduction in a

margin of safety because, for the proposed

changes, there are no new or significant

changes to the initial conditions contributing

to accident severity or consequences.

Accordingly, there are no significant

reductions in a margin of safety.

On the basis of the above, the DBNPS has

determined that the License Amendment

Request does not involve a significant

hazards consideration. As this License

Amendment Request concerns a proposed

change to the Technical Specifications that

must be reviewed by the Nuclear Regulatory

Commission, this License Amendment

Request does not constitute an unreviewed

safety question.

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: Mary E.

O’Reilly, Attorney, FirstEnergy

Corporation, 76 South Main Street,

Akron, OH 44308.

NRC Section Chief: Anthony J.

Mendiola.

FirstEnergy Nuclear Operating

Company, Docket No. 50–346, DavisBesse Nuclear Power Station, Unit 1,

Ottawa County, Ohio

Date of amendment request: April 1,

2001.

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS)

Surveillance Requirement (SR) 4.0.5,

Applicability, TS Bases 4.0.5, and TS

Bases 3/4.4.2 and 3/4.4.3, Reactor

Coolant System—Safety Valves,

Regarding Inservice Testing

Requirements.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensees have provided their analysis of

the issue of no significant hazards

consideration, which is presented

below:

1a. Not involve a significant increase in the

probability of an accident previously

evaluated because no such accidents are

affected by the proposed changes. The

amendment application proposes to revise

DBNPS Technical Specification (TS)

Surveillance Requirement 4.0.5,

Applicability, and its associated Bases and

TS Bases 3/4.4.2 and 3/4.4.3, Reactor Coolant

System—Safety Valves. The proposed

changes would modify the Technical

Specifications to conform to the requirements

of Section 50.55a(f) of Title 10 of the Code

of Federal Regulations regarding the

inservice testing of pumps and valves for the

third and successive 120-month intervals.

The current DBNPS TS reference the

American Society of Mechanical Engineers

Boiler and Pressure Vessel Code (ASME

Code), Section XI requirements for the

inservice testing of ASME Code Class 1, 2,

and 3 pumps and valves. The proposed

changes would reference the ASME Code for

Operation and Maintenance of Nuclear

Power Plants (ASME OM Code) which is

consistent with Section 50.55a(f).

In addition, surveillance interval

definitions for ‘‘semi-quarterly,’’ ‘‘every 9

months,’’ and ‘‘biennially or every 2 years,’’

as used in the ASME Code would be added

to TS 4.0.5.b to ensure consistent

interpretation of the terms. The proposed

changes do not affect any plant hardware and

do not affect the probability of any

equipment malfunction or accident-initiating

event.

1b. Not involve a significant increase in the

consequences of an accident previously

evaluated because no equipment, accident

conditions, or assumptions are affected

which could lead to a significant increase in

radiological consequences.

2. Not create the possibility of a new or

different kind of accident from any accident

previously evaluated because no new or

different accident initiators are introduced by

these proposed changes.

3. Not involve a significant reduction in a

margin of safety because there are no changes

to the initial conditions contributing to

accident severity or consequences.

Consequently, there are no significant

reductions in a margin of safety.

On the basis of the above, the DBNPS has

determined that the License Amendment

Request does not involve a significant

hazards consideration. As this License

Amendment Request concerns a proposed

change to the Technical Specifications that

must be reviewed by the Nuclear Regulatory

Commission, this License Amendment

Request does not constitute an unreviewed

safety question.

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: Mary E.

O’Reilly, Attorney, FirstEnergy

Corporation, 76 South Main Street,

Akron, OH 44308.

NRC Section Chief: Anthony J.

Mendiola.

Florida Power and Light Company, et

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

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

County, Florida

Date of amendment request: April 17,

2001.

Description of amendment request:

The proposed amendments would

implement minor changes and

corrections to the Technical

Specifications (TS) to correct

administrative errors (e.g.,

typographical, amendment tracking

number, etc.), or to incorporate changes

that have been justified by previously

approved license amendments and

should have been made as part of those

submittals, or to correct logic errors

(e.g., TS operating mode breakpoints

based on pressurizer pressure and not

temperature). Also, the proposed

amendments would revise the Units 1

and 2 TS to delete obsolete terminology

and provide conforming changes to

reflect the recently implemented change

to 10 CFR 50.59.

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

29358 Federal Register / Vol. 66, No. 104 / Wednesday, May 30, 2001 / Notices

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

These proposed license amendments

require no plant hardware or operational

modifications. The proposed changes either

correct various administrative errors (e.g.,

typographical errors, amendment tracking

number errors), incorporate changes that

have been justified by previously approved

license amendments and should have been

made as part of those submittals, correct logic

errors, or are necessary to implement the 10

CFR 50.59 rule change.

Therefore, operation of the facility in

accordance with the proposed amendments

would not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

(2) Operation of the Facility in Accordance

with the Proposed Amendment Would Not

Create the Possibility of a New or Different

Kind of Accident From Any Accident

Previously Evaluated

No modifications to either plant hardware

or operational procedures are required to

support these proposed license amendments;

hence, no new failure modes are created. The

proposed changes either correct various

administrative errors (e.g., typographical

errors, amendment tracking number errors),

incorporate changes that have been justified

by previously approved license amendments

and should have been made as part of those

submittals, correct logic errors, or are

necessary to implement the 10 CFR 50.59

rule change.

Therefore, operation of the facility in

accordance with the proposed amendments

would not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

(3) Operation of the Facility in Accordance

With the Proposed Amendment Would Not

Involve a Significant Reduction in a Margin

of Safety

The majority of TS corrections proposed by

these license amendments are administrative

in nature in that they either correct

typographical errors (e.g., ODCM verses

OCDM), are justified by previous license

amendments (e.g., surveillance requirements

for Thot wide versus narrow range

instrumentation), or correct logic errors (e.g.,

ECCS subsystem TS headings based on

operating mode, with Mode 3 breakpoints

based on pressurizer pressure and not

temperature). The overly restrictive

emergency power requirements for non

critical single train quality related radiation

monitors are being removed, while critical

radiation monitor emergency power

requirements are unaffected by the change.

Therefore, operation of the facility in

accordance with the proposed amendment

would not involve a significant reduction in

a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Attorney for licensee: M.S. Ross,

Attorney, Florida Power & Light, P.O.

Box 14000, Juno Beach, Florida 33408–

0420.

NRC Section Chief: Patrick M.

Madden (Acting).

Florida Power and Light Company,

Docket No. 50–389, St. Lucie Plant, Unit

No. 2, St. Lucie County, Florida

Date of amendment request: April 18,

2001.

Description of amendment request:

The proposed amendment would

implement an improved heat flux

correlation (designated ABB–NV)

previously approved by the NRC for

Westinghouse-Combustion Engineering,

as documented in the topical

reportCENPD–387–P–A, Rev 000. The

proposed change updates Technical

Specification (TS) 6.9.1.11, ‘‘Core

Operating Limits Report (COLR),’’ to

include the topical report in the list of

analytical methods used. Additionally,

the Bases for TS 2.1.1, ‘‘Reactor Core,’’

would be modified to reflect use of the

improved heat flux correlation.

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 would allow the

implementation of ABB–NV critical heat flux

correlation to St. Lucie Unit 2 core. The

proposed changes have no adverse impact on

the operation of the plant and have no

relevance to the accident initiators. There are

no changes to the plant configuration, and

thus the frequency of occurrence of

previously analyzed accidents is not affected

by the proposed changes. With the

application of the added methodology (the

approved ABB–NV DNB correlation), the

safety analysis would continue to remain

consistent with the design basis

requirements. The proposed changes,

including changes to the TS Bases, have no

adverse effect on the safety analysis and thus

would not involve a significant increase in

the consequences of design basis accidents.

Changes to the COLR limits will continue to

be controlled per Generic Letter 88–16 under

the provisions of 10 CFR 50.59 and the

requirements of TS 6.9.1.11.c.

Therefore, 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.

(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 amendment updates the list

of approved methodology in TS 6.9.1.11 and

makes corresponding changes to the TS Bases

for TS 2.1.1. These changes would not create

the possibility of a new kind of accident

since there is no change to plant

configuration, systems, or components,

which would create new failure modes. The

modes of operation of the plant would

remain unchanged.

Therefore, operation of the facility in

accordance with the proposed amendment

would not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

(3) Use of the Modified Specification Would

Not Involve a Significant Reduction in a

Margin of Safety

The proposed changes have no significant

adverse impact on the safety analysis. As

such, these changes would continue to

provide margin to the acceptance criteria for

Specified Acceptable Fuel Design Limits

(SAFDL), 10 CFR 50.46(b) requirements,

primary and secondary overpressurization,

peak containment pressure, potential

radioactive releases, and existing limiting

conditions for operation. The future use of

updated approved methodology will follow

all design basis requirements to ensure that

a safety margin to the acceptance criteria

would continue to remain available at all

power levels for operation of St. Lucie Unit

2.

Therefore, operation of the facility in

accordance with the proposed amendment

would not involve a significant reduction in

a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Attorney for licensee: M.S. Ross,

Attorney, Florida Power & Light, P.O.

Box 14000, Juno Beach, Florida 33408–

0420.

NRC Section Chief: Patrick M.

Madden (Acting).

Niagara Mohawk Power Corporation,

Docket No. 50–410, Nine Mile Point

Nuclear Station Unit No. 2, Oswego

County, New York

Date of amendment request: March

29, 2001.

Description of amendment request:

The licensee proposed to amend the

Technical Specifications (TSs) in three

areas, adopting three NRC-approved

Technical Specification Task Force

(TSTF) issues. This notice is concerned

Federal Register / Vol. 66, No. 104 / Wednesday, May 30, 2001 / Notices 29359

with changes covered by one of the

three issues, identified as TSTF–51.

The licensee proposed to adopt

TSTF–51, reducing the operability

requirements for certain engineered

safeguard features (ESFs) such as

secondary containment, standby gas

treatment, control room envelop

filtration. The current requirements

specify that these ESFs be operable

during movement of irradiated fuel in

the secondary containment, and during

core operation. The proposed changes

would specify these ESFs be operable

during movement of recently irradiated

fuel in the secondary containment, and

would eliminate the applicability

during core alteration. The associated

licensee-controlled TS Basis document

would also be changed to reflect the

changes.

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 three standards of 10 CFR 50.92(c).

The NRC staff’s review is presented

below:

The first standard requires that

operation of the unit in accordance with

the proposed amendment will not

involve a significant increase in the

probability or consequences of an

accident previously evaluated. TSTF–51

involves no hardware design change,

thus there will be no adverse effect on

the functional performance of the ESFs

to mitigate accident consequences. The

ESFs are not initiators of any previously

analyzed accidents, thus the proposed

changes cannot increase the probability

of any previously analyzed accidents.

Therefore, the proposed changes will

not involve a significant increase in the

probability or consequences of an

accident previously evaluated.

The second standard requires that

operation of the unit in accordance with

the proposed amendment will not create

the possibility of a new or different kind

of accident from any accident

previously evaluated. TSTF–51 involves

no hardware design change or

procedural change; hence all

components, systems, and structures

will continue to perform as originally

designed by the licensee and previously

accepted by the NRC staff. Therefore,

the proposed changes covered by TSTF–

51 will not create the possibility of a

new or different kind of accident from

any previously evaluated.

The third standard requires that

operation of the unit in accordance with

the proposed amendment will not

involve a significant reduction in a

margin of safety. Since TSTF–51

involves no change to the design,

operational procedure, or analysis

methodology, TSTF–51 will not affect in

any way the performance characteristics

and original intended functions of any

system, structure or component.

Therefore, the proposed changes do not

involve a significant reduction in a

margin of safety.

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 part of

the amendment request identified as

TSTF–51 involves no significant

hazards consideration.

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston & Strawn,

1400 L Street, NW., Washington, DC

20005–3502.

NRC Section Chief: Richard P.

Correia, Acting.

Niagara Mohawk Power Corporation,

Docket No. 50–410, Nine Mile Point

Nuclear Station Unit No. 2, Oswego

County, New York

Date of amendment request: March

29, 2001.

Description of amendment request:

The licensee proposed to amend the

Technical Specifications (TSs) in three

areas, adopting three NRC-approved

Technical Specification Task Force

(TSTF) issues. This notice is concerned

with one of the three changes, identified

as TSTF–204.

The licensee proposed to adopt

TSTF–204, revising Limiting Condition

for Operation (LCO) 3.8.5. Currently,

LCO 3.8.5 requires that direct current

(DC) power subsystems shall be

OPERABLE to support the electrical

power distribution subsystems required

by LCO 3.8.9 (pertaining to shutdown

conditions). Adoption of TSTF–204

would change this to require either the

Division 1 or Division 2 DC electrical

power subsystems, in addition to the

Division 3 DC electrical power

subsystem, shall be OPERABLE. This

change would restore the TS to what it

was before the TS was converted to the

Improved TS format by Amendment No.

91 (February 15, 2000). The associated

licensee-controlled TS Basis document

would also be changed to reflect the

changes.

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 three standards of 10 CFR 50.92(c).

The NRC staff’s review is presented

below:

The first standard requires that

operation of the unit in accordance with

the proposed amendment will not

involve a significant increase in the

probability or consequences of an

accident previously evaluated. The

proposed changes to adopt TSTF–204

involve no hardware design change or

operational procedure change, thus

there will be no adverse effect on the

functional performance of any plant

SSC; the decreased operability

requirement pertains to times when

there is less demand on the electrical

subsystems (i.e., during shutdown

conditions). All structures, systems and

components (SSCs) will continue to

perform their design functions with no

decrease in their capabilities to mitigate

the consequences of postulated

accidents. Accordingly, the proposed

operability requirements will lead to no

significant increase in the consequences

of an accident previously evaluated, and

no increase of the probability of an

accident previously evaluated.

The second standard requires that

operation of the unit in accordance with

the proposed amendment will not create

the possibility of a new or different kind

of accident from any accident

previously evaluated. TSTF–204

involves no hardware design change or

procedural change; hence all SSCs will

continue to perform as originally

designed by the licensee and previously

accepted by the NRC staff. Therefore,

the proposed changes covered by TSTF–

204 will not create the possibility of a

new or different kind of accident from

any previously evaluated.

The third standard requires that

operation of the unit in accordance with

the proposed amendment will not

involve a significant reduction in a

margin of safety. Since TSTF–204

involves no change to the design,

operational procedure, or analysis

methodology, TSTF–204 will not affect

in any way the performance

characteristics and intended functions

of any SSC. Therefore, the proposed

changes do not involve a significant

reduction in a margin of safety.

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 part of

the amendment request identified as

TSTF–204 involves no significant

hazards consideration.

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston & Strawn,

1400 L Street, NW., Washington, DC

20005–3502.

NRC Section Chief: Richard Correia,

Acting.

29360 Federal Register / Vol. 66, No. 104 / Wednesday, May 30, 2001 / Notices

Niagara Mohawk Power Corporation,

Docket No. 50–410, Nine Mile Point

Nuclear Station Unit No. 2, Oswego

County, New York

Date of amendment request: March

29, 2001.

Description of amendment request:

The licensee proposed to amend the

Technical Specifications (TSs) in three

areas, adopting three NRC-approved

Technical Specification Task Force

(TSTF) issues. This notice is concerned

with one of the three changes, identified

as TSTF–287.

The licensee proposed to adopt

TSTF–287, adding to Section 3.7.2,

Control Room Envelope Filtration

System (CREFS), a note to permit the

control room envelope be opened

intermittently under administrative

control, and a new Condition B allowing

24 hours to restore operability of the

two CREFS subsystems if their

operability is lost due to inoperable

control room envelope boundary. These

proposed provisions would allow time

to diagnose, plan and possibly repair,

and test most problems with the control

room envelope boundary. The

associated licensee-controlled TS Basis

document would also be changed to

reflect the TS changes.

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 three standards of 10 CFR 50.92(c).

The NRC staff’s review is presented

below:

The first standard requires that

operation of the unit in accordance with

the proposed amendment will not

involve a significant increase in the

probability or consequences of an

accident previously evaluated. The

proposed change to adopt TSTF–287

involves no hardware design change or

operational procedure change, thus

there will be no adverse effect on the

functional performance of any plant

structures, systems or components

(SSCs). The allowance to open the

control room envelope intermittently

does not increase accident

consequences on control personnel

since the administrative controls would

rapidly restore integrity. Allowing 24

hours to restore the integrity of the

control room envelope could result in

an increase in consequences of a designbasis accident occurring during this

time to control room personnel, but the

administrative controls in place would

easily and quickly reverse the condition,

re-establishing control room envelope

integrity, and thus limiting increases in

consequences. Thus, all SSCs will

continue to perform their design

functions with no decrease in their

capabilities to mitigate the

consequences of postulated accidents.

Accordingly, the proposed operability

requirements will lead to no significant

increase in the consequences of an

accident previously evaluated, and no

increase of the probability of an

accident previously evaluated.

The second standard requires that

operation of the unit in accordance with

the proposed amendment will not create

the possibility of a new or different kind

of accident from any accident

previously evaluated. TSTF–287

involves no hardware design change or

procedural change; hence it does not

negatively affect the design or

performance of any SSC, and all SSCs

will continue to perform as originally

designed by the licensee and previously

accepted by the NRC staff. Therefore,

the proposed changes will not create the

possibility of a new or different kind of

accident from any previously evaluated.

The third standard requires that

operation of the unit in accordance with

the proposed amendment will not

involve a significant reduction in a

margin of safety. Since TSTF–287

involves no change to the design,

operational procedure, or analysis

methodology, TSTF–287 will not affect

in any way the performance

characteristics and intended functions

of any SSC. Therefore, the proposed

changes do not involve a significant

reduction in a margin of safety.

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 part of

the amendment request identified as

TSTF–287 involves no significant

hazards consideration.

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston & Strawn,

1400 L Street, NW., Washington, DC

20005–3502.

NRC Section Chief: Richard Correia,

Acting.

Nuclear Management Company, LLC,

Docket No. 50–263, Monticello Nuclear

Generating Plant, Wright County,

Minnesota

Date of amendment request: May 2,

2001.

Description of amendment request:

The proposed amendment would (1)

relocate requirements of the American

Society of Mechanical Engineers

(ASME) Boiler and Pressure Vessel Code

(the Code), Section XI, inservice testing

(IST) program currently contained in

technical specification surveillance

requirement (TSSR) 4.15.B to the TS

Administrative Control Section 6.8,

Programs and Manuals, (2) make

conforming changes to several

surveillance requirements to reflect the

change in reference from TSSR 4.15.B to

the licensee-controlled IST Program, (3)

reword TSSRs 4.5.A.3 and 4.5.D.1 to be

consistent with NUREG–1433, (4)

incorporate TS Task Force (TSTF)

initiative TSTF–279 into TS

Administrative Control Section 6.8, and

(5) revise TSSRs 4.6.H.1, 4.6.H.3, and

Table 4.6.1 to change the inspection and

functional testing interval extensions

reference from plus-or-minus 25 percent

to plus 25 percent.

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 an Accident

Previously Evaluated

The requested changes are administrative

in nature in that they relocate IST [inservice

testing] requirements from the Monticello TS

[Technical Specifications] to a licensee

controlled IST program, rewrite TS

Surveillance Requirements 4.5.A.3 and

4.5.D.1 for clarification using the wording

from NUREG–1433 and revise TS

surveillance requirements for inspection and

functional testing interval extensions. The

requested changes will not revise previous

commitments to 10 CFR 50.55a of [sic] ASME

[American Society of Mechanical Engineers]

Code [ASME Boiler and Pressure Vessel

Code], Section XI, IST requirements.

The proposed changes do not involve any

change to the configuration or method of

operation of any plant equipment that is used

to mitigate the consequences of an accident,

nor do they affect any assumptions or

conditions in any of the accident analyses.

Since the accident analyses remain

bounding, their radiological consequences

are not adversely affected.

Therefore, the probability or consequences

of an accident previously evaluated are not

affected.

2. The Proposed Amendment Will Not Create

the Possibility of a New or Different Kind of

Accident From Any Accident Previously

Analyzed

The requested changes are administrative

in nature in that they relocate IST

requirements from the Monticello TS to the

licensee controlled IST program, rewrite TS

Surveillance Requirements 4.5.A.3 and

4.5.D.1 for clarification using the wording

from NUREG–1433 and revise TS

surveillance requirements for inspection and

functional testing interval extensions. The

requested changes will not revise previous

commitments to 10 CFR 50.55a or ASME

Code, Section XI, IST requirements.

The proposed changes do not involve

changes to the configuration or method of

Federal Register / Vol. 66, No. 104 / Wednesday, May 30, 2001 / Notices 29361

operation of any plant equipment that is used

to mitigate the consequences of an accident,

nor do they affect any assumptions or

conditions in any of the accident analyses.

Accordingly, no new failure modes have

been defined for any plant system or

component important to safety nor has any

new limiting single failure been identified as

a result of the proposed changes.

Therefore, the possibility of a new or

different kind of accident from any accident

previously analyzed is not created.

3. The Proposed Amendment Will Not

Involve a Significant Reduction in the Margin

of Safety

The requested changes are administrative

in nature in that they relocate IST

requirements from the Monticello TS to the

licensee controlled IST program, rewrite TS

Surveillance Requirements 4.5.A.3 and

4.5.D.1 for clarification using the wording

from NUREG–1433 and revise TS

surveillance requirements for inspection and

functional testing interval extensions. The

requested changes will not revise previous

commitments to 10 CFR 50.55a or ASME

Code, Section XI, IST requirements. Program

requirements will remain to ensure that Code

requirements are met.

Therefore, a significant reduction in

the margin of safety is not involved.

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: Jay E. Silberg,

Esq., Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Section Chief: Claudia M. Craig.

PSEG Nuclear LLC, Docket No. 50–354,

Hope Creek Generating Station, Salem

County, New Jersey

Date of amendment request: April 11,

2001.

Description of amendment request:

The proposed amendment would revise

the Technical Specifications (TSs) to

relax the frequency for testing of excess

flow check valves (EFCVs). Specifically,

TS surveillance requirement 4.6.3.4

would be changed to revise required

testing of EFCVs from once per 18

months for all valves to a test of a

representative sample each 18 months

such that all valves are tested once in 10

years.

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’s review is

presented below:

(1) The proposed changes do not

involve a significant increase in the

probability or consequences of an

accident previously evaluated.

The proposed changes affect the

surveillance interval for the EFCV’s,

allowing a reduced number of valves to

be tested at each interval. There are no

physical plant modifications associated

with this change. The EFCV’s, which are

installed on instrument lines

penetrating containment, are designed

to close in order to isolate containment

upon a failure of the instrument line

downstream of the valve. Since the

EFCV’s are designed to provide an

accident mitigation function (i.e.,

minimize radiological effects due to an

instrument line break), their postulated

failure to close as a result of the

proposed reduced testing frequency is

not considered an initiator to any

previously evaluated accidents.

Therefore, there is no increase in the

probability of occurrence of an accident

as a result of the proposed changes.

The design basis analyses for an

instrument line break is evaluated in

Section 15.6.2 of the Hope Creek

Updated Final Safety Analysis Report

(UFSAR). These analyses do not take

credit for the closure of the EFCV’s. The

postulated failure of an EFCV to close as

a result of the proposed reduced testing

frequency is bounded by the existing

UFSAR analyses. Therefore, the

proposed changes do not involve a

significant increase in the consequences

of an accident previously evaluated.

(2) The proposed changes do not

create the possibility of a new or

different kind of accident from any

accident previously evaluated.

The proposed changes in TS

surveillance requirements allow a

reduced number of EFCV’s to be tested

each operating cycle. No other changes

are being requested. The proposed

changes do not introduce any new

modes of plant operation and do not

involve physical modifications to the

plant. These changes will not alter any

process variables, structures, systems, or

components as described in the safety

analyses. Therefore, the proposed

changes do not create the possibility of

a new or different kind of accident from

any previously evaluated.

(3) The proposed changes do not

involve a significant reduction in a

margin of safety.

The radiological consequences of an

unisolable break of an instrument line

have previously been evaluated in

Section 15.6.2 of the Hope Creek

UFSAR. The accident analyses assume

that the line break results in the release

of reactor coolant into the Reactor

Building until the reactor pressure

vessel is depressurized. The analyses do

not take credit for the closure of the

EFCV’s. The proposed reduced testing

frequency only changes the potential for

an undetected failure of an EFCV and

does not change the event sequence

upon which the current safety margin

related to radiological consequences is

based. Therefore, the proposed changes

do not involve a significant reduction in

a margin of safety.

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: Jeffrie J. Keenan,

Esquire, Nuclear Business Unit—N21,

P.O. Box 236, Hancocks Bridge, NJ

08038.

NRC Section Chief: James W. Clifford.

South Carolina Electric & Gas Company

(SCE&G), South Carolina Public Service

Authority, Docket No. 50–395, Virgil C.

Summer Nuclear Station, Unit No. 1,

Fairfield County, South Carolina

Date of amendment request: April 18,

2001.

Description of amendment request:

South Carolina Electric & Gas Company

(SCE&G) proposes a change to the Virgil

C. Summer Nuclear Station (VCSNS)

Technical Specifications (TS)

Surveillance Requirements to revise the

volumetric flow units for TS 4.7.6.c.1,

c.3, e.1, e.3, and f to identify standard

flow units expressed as standard cubic

feet per minute. Volumetric flow units

for TS 4.6.3.b.1, b.2, c.1, d, and g, and

TS 4.9.11.b.1, b.3, d.1, e, and f are being

revised to identify actual air flow units

and are expressed as actual cubic feet

per minute.

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:

South Carolina Electric & Gas Company

(SCE&G) has evaluated the proposed changes

to the VCSNS TS described above against the

significant Hazards Criteria of 10 CFR 50.92

and has determined that the changes do not

involve any significant hazard. The following

is provided in support of this conclusion.

1. Does the proposed change involve a

significant increase in the probability or

consequences of an accident previously

evaluated? Changes associated with the

identification of proper flow units are

editorial and have no impact.

2. Does the proposed change create the

possibility of a new or different kind of

accident from any accident previously

evaluated? Changes associated with the

identification of proper flow units are

editorial and have no impact.

3. Does this change involve a significant

reduction in margin of safety? The margin of

29362 Federal Register / Vol. 66, No. 104 / Wednesday, May 30, 2001 / Notices

safety for any of the ventilation systems

associated with the proposed change is not

compromised. Changes associated with the

identification of proper flow units are

editorial and have no impact.

There are no significant safety hazards

created by the change. There is no new or

different accident postulated since the

change is considered editorial. The design

requirements of Regulatory Guide 1.52

remain satisfied. Therefore, there is no

significant decrease 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: Thomas G.

Eppink, South Carolina Electric & Gas

Company, Post Office Box 764,

Columbia, South Carolina 29218.

NRC Section Chief: Richard L. Emch,

Jr.

Tennessee Valley Authority, Docket No.

50–327, Sequoyah Nuclear Plant, Unit 1,

Hamilton County, Tennessee

Date of application for amendments:

May 14, 2001 (TS 01–02).

Brief description of amendments: The

proposed amendment would change the

Sequoyah (SQN) Unit 1 Operating

License Condition 2.C.(9)(d) to clarify

the lower voltage threshold for eddy

current inspections.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a),

Tennessee Valley Authority (TVA), the

licensee, has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

A. The Proposed Amendment Does Not

Involve a Significant Increase in the

Probability or Consequences of an Accident

Previously Evaluated

The proposed change does not alter plant

equipment, system design, or operating

practices. The clarification of SQN’s Unit 1

[steam generator] SG inspection commitment

provides a conservative inspection strategy

that defines 1 volt as the lower threshold.

The 1-volt threshold is based on the

subjectivity uncertainties associated with

interpreting bobbin coil probe data to

distinguish a dent below 1 volt. Given the

current capability of eddy current

technology, TVA’s proposed change will

define a reasonable criteria for tube

inspection.

TVA’s proposed change continues to

ensure that structural and leakage integrity of

SQN’s Unit 1 SG tubes is maintained.

Accordingly, the proposed amendment does

not result in any increase in the probability

or consequences of an accident previously

evaluated within the SQN Final Safety

Analysis Report.

B. The Proposed Amendment Does Not

Create the Possibility of a New or Different

Kind of Accident From Any Accident

Previously Evaluated

SQN limits SG tube leakage between the

primary coolant system and the secondary

coolant system to 150 gallons per day per SG.

This leakage limit ensures that tube cracks

have an adequate margin of safety to

withstand the loads imposed during normal

operation and by postulated accidents. In

addition, inservice inspections are performed

in accordance with Regulatory Guide 1.83,

Revision 1, ‘‘Inservice Inspection of

Pressurized Water Reactor Steam Generator

Tubes,’’ to ensure that structural integrity of

SG tubes is maintained during the plant

operation cycle.

The proposed change does not modify

plant equipment, system design, or operating

practices. The clarification of SQN’s Unit 1

SG inspection commitment provides an

inspection strategy that defines a minimum

‘‘calling’’ threshold for dent inspection. The

1-volt threshold is an inspection strategy

based on the subjectivity associated with

interpreting bobbin coil probe data below 1

volt for dented intersections. TVA’s proposed

change will continue to provide conservative

inspection criteria that maintains structural

and leakage integrity of SQN’s Unit 1 SG

tubes.

Based on the above, the proposed

amendment does not create the possibility of

a new or different kind of accident from any

accident previously evaluated.

C. The Proposed Amendment Does Not

Involve a Significant Reduction in a Margin

of Safety

TVA’s proposed clarification of the 1-volt

threshold will continue to provide a

conservative inspection criteria that will

ensure that SG tube structural and leakage

integrity is maintained. Accordingly, the

margin of safety is not reduced.

The NRC 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: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 10H,

Knoxville, Tennessee 37902.

NRC Section Chief: Patrick M.

Madden (Acting).

TXU Electric, Docket Nos. 50–445 and

50–446, Comanche Peak Steam Electric

Station, Units 1 and 2, Somervell

County, Texas

Date of amendment request: April 25,

2001.

Brief description of amendments: The

proposed changes would revise

Technical Specification (TS) 3.8.1, ‘‘AC

[alternating current] Sources—

Operating,’’ to extend the allowable

Completion Times for the Required

Actions associated with restoration of

an inoperable Emergency Diesel

Generator (EDG) and an inoperable

offsite circuit (i.e., startup transformer).

In addition, the TS Surveillance

Requirement (SR) corresponding to the

24-hour EDG endurance run in SR

3.8.1.14 would be revised to allow the

SR to be performed during Modes 1 and

2. The proposed changes would also

revise TS 3.8.9, ‘‘Distribution Systems—

Operating,’’ to extend the allowable

Completion Times for the Required

Actions associated with restoration of

an inoperable AC electrical power

distribution subsystem (i.e., 6.9 kilovolt

(kV) AC safety bus).

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. Do the Proposed Changes Involve a

Significant Increase in the Probability or

Consequences of an Accident Previously

Evaluated?

Response: No.

The proposed Technical Specification

changes do not significantly increase the

probability of occurrence of a previously

evaluated accident because the 6.9 kV AC

components (i.e., Emergency Diesel

Generators (EDGs), startup transformers

(STs), and safety-related (Class 1E) busses)

are not initiators of previously evaluated

accidents involving a loss of offsite power.

The proposed changes to the Technical

Specification Action Completion Times do

not affect any of the assumptions used in the

deterministic or the Probabilistic Safety

Assessment (PSA) analysis.

The proposed Technical Specification

changes will continue to ensure the 6.9 kV

AC components perform their function when

called upon. Extending the Technical

Specification Completion Times to 14 days

and allowing the performance of the EDG 24-

hour run test in either MODES 1 or 2 does

not affect the design of the EDGs, the

operational characteristics of the EDGs, the

interfaces between the EDGs and other plant

systems, the function, or the reliability of the

EDGs. Thus, the EDGs will be capable of

performing either accident mitigation

function and there is no impact to the

radiological consequences of any accident

analysis.

To fully evaluate the effect of the changes

to the 6.9 kV AC components, Probabilistic

Safety Analysis (PSA) methods and

deterministic analysis were utilized. The

results of this analysis show no significant

increase in the Core Damage Frequency.

The Configuration Risk Management

Program (CRMP) in Technical Specification

5.5.18 is an administrative program that

assesses risk based on plant status. Adding

the requirement to implement the CRMP for

Technical Specification 3.8.1 and 3.8.9

requires the consideration of other measures

to mitigate consequences of an accident

occurring while a 6.9 kV AC component is

inoperable.

Federal Register / Vol. 66, No. 104 / Wednesday, May 30, 2001 / Notices 29363

The proposed changes do not alter the

operation of any plant equipment assumed to

function in response to an analyzed event or

otherwise increase its failure probability.

Therefore, these changes do not involve a

significant increase in the probability or

consequences of any accident previously

evaluated.

2. Do the Proposed Changes Create the

Possibility of a New or Different Kind of

Accident From Any Accident Previously

Evaluated?

Response: No.

These proposed changes do not change the

design, configuration, or method of operation

of the plant. The proposed activities involves

[sic] a change to the allowed plant mode for

the performance of specific Technical

Specification surveillance requirements. No

physical or operational change to the 6.9 kV

AC components or supporting systems are

made by this activity. Since the proposed

changes do not involve a change to the plant

design or operation, no new system

interactions are created by this change. The

proposed Technical Specification changes do

not produce any parameters or conditions

that could contribute to the initiation of

accidents different from those already

evaluated in the Final Safety Analysis

Report.

The proposed changes only address the

time allowed to restore the operability of the

6.9 kV AC components. Thus the proposed

Technical Specification changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Do the Proposed Changes Involve a

Significant Reduction in a Margin of Safety?

Response: No.

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 net changes are either risk neutral or

risk beneficial. The proposed activities

involves [sic] changes to certain Completion

Times and to the allowed plant mode for the

performance of specific Technical

Specification Requirements. The proposed

changes remain bounded by the existing

Surveillance Requirement Completion Times

and therefore have no impact to the margins

of safety.

The proposed change does [sic] not involve

a change to the plant design or operation and

thus does not affect the design of the 6.9 kV

AC components, the operation characteristics

of the 6.9 kV AC components, the interfaces

between the 6.9 kV AC components and

other plant systems, or the function or

reliability of the 6.9 kV AC components.

Because 6.9 kV AC components performance

and reliability will continue to be ensured by

the proposed Technical Specification

changes, the proposed changes do not result

in a reduction in the margin of safety.

Therefore the proposed change does not

involve a reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: George L. Edgar,

Esq., Morgan, Lewis and Bockius, 1800

M Street, NW., Washington, DC 20036.

NRC Section Chief: Robert A. Gramm.

Vermont Yankee Nuclear Power

Corporation, Docket No. 50–271,

Vermont Yankee Nuclear Power Station,

Vernon, Vermont

Date of amendment request: April 23,

2001.

Description of amendment request:

The amendment would update the

facility operating license (FOL) by

deleting obsolete information, correcting

errors, and making administrative

changes to enhance the context and

provide consistency.

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 Operation of Vermont Yankee Nuclear

Power Station in Accordance With the

Proposed Amendment Will Not Involve a

Significant Increase in the Probability or

Consequences of an Accident Previously

Evaluated

The proposed change makes editorial

changes and brings the FOL up to date with

the expectations of Massachusetts regulatory

agencies. Since reactor operation under the

proposed amendment is unchanged, no

design or analytical acceptance criteria will

be exceeded. As such, this change does not

impact initiators of analyzed events or

assumed mitigation of accident or transient

events. The structural and functional

integrity of plant systems is unaffected. Thus,

there is no significant increase in the

probability or consequences of accidents

previously evaluated.

2. The Operation of Vermont Yankee Nuclear

Power Station in Accordance With the

Proposed Amendment Will Not Create the

Possibility of a New or Different Kind of

Accident From Any Accident Previously

Evaluated

The proposed change does not affect any

parameters or conditions that could

contribute to the initiation of any accident.

No new accident modes are created. No

safety-related equipment or safety functions

are altered as a result of these changes.

Because it does not involve any change to the

plant or the manner in which it is operated,

the proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The Operation of Vermont Yankee Nuclear

Power Station in Accordance With the

Proposed Amendment Will Not Involve a

Significant Reduction in a Margin of Safety

The proposed change does not affect

design margins or assumptions used in

accident analyses, and has no effect on any

assumed analysis initial condition. The

capability of safety systems to function and

limiting safety system settings are similarly

unaffected as a result of this change. Thus,

the proposed change will 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: Mr. David R.

Lewis, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037–1128.

NRC Section Chief: James W. Clifford.

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)

29364 Federal Register / Vol. 66, No. 104 / Wednesday, May 30, 2001 / Notices

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

electronically from the ADAMS Public

Library component on the NRC Web

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

Public Reading Room).

AmerGen Energy Company, LLC, Docket

No. 50–289, Three Mile Island Nuclear

Station, Unit 1, Dauphin County,

Pennsylvania

Date of application for amendment:

December 20, 2000, as supplemented

March 14, 2001.

The March 14, 2001, letter 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. A March 23, 2001, letter

provided a camera-ready copy of the

revised technical specification pages.

Brief description of amendment: The

amendment allows the expanded use of

the Framatome Cogema Fuels M5 alloy

for fuel rod cladding and fuel assembly

spacer grids. A related Bases change is

included with the licensee’s

application.

Date of issuance: May 10, 2001.

Effective date: As of the date of

issuance and shall be implemented no

later than the startup of Cycle 14

operation, approximately October 1,

2001.

Amendment No.: 233.

Facility Operating License No. DPR–

50. Amendment revised the Technical

Specifications.Date of initial notice in

Federal Register: February 6, 2001 (66

FR 9379).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 10, 2001.

No significant hazards consideration

comments received: No.

Commonwealth Edison Company,

Docket Nos. STN 50–454 and STN 50–

455, Byron Station, Unit Nos. 1 and 2,

Ogle County, IllinoisDocket Nos. STN

50–456 and STN 50–457, Braidwood

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

Illinois

Date of application for amendments:

November 13, 2000.

Brief description of amendments: The

amendment revised the technical

specifications to delete the ‘‘Power

Range Neutron Flux High Negative

Rate,’’ Trip Function from Reactor Trip

System Instrumentation. The changes

allow elimination of this unnecessary

function and thereby reduces the

potential for a transient. The changes

are consistent with the Westinghouse

Topical report previously accepted by

the NRC.

Date of issuance: May 17, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 30 days.

Amendment Nos.: 114, 114, 120, 120.

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: February 21, 2001 (66 FR

11054).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated May 17, 2001.

No significant hazards consideration

comments received: No.

Consumers Energy Company, Docket

No. 50–255, Palisades Plant, Van Buren

County, Michigan

Date of application for amendment:

November 21, 2000, as supplemented

April 25, April 26, May 3 (two letters),

and May 8, 2001.

Brief description of amendment:

Amendment conforms the license to

reflect the transfer of operating authority

under Operating License No. DPR–20 to

Nuclear Management Company, LLC, as

approved by order of the Commission

dated April 19, 2001 (66 FR 21021 dated

April 26, 2001).

Date of issuance: May 15, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 45 days.

Amendment No.: 201.

Facility Operating License No. DPR–

20. Amendment revised the Operating

License.

Date of initial notice in Federal

Register: December 19, 2000 (65 FR

79431).

The supplemental letters dated April

25, April 26, May 3 (two letters), and

May 8, 2001, were within the scope of

the initial application as originally

noticed. The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

April 19, 2001.

No significant hazards consideration

comments received: No.

Consumers Energy Company, Docket

No. 50–255, Palisades Plant, Van Buren

County, Michigan

Date of application for amendment:

December 7, 2000.

Brief description of amendment: The

amendment changes the Technical

Specifications (TSs) in accordance with

changes to the ‘‘Standard Technical

Specifications, Combustion Engineering

Plants,’’ NUREG 1432, Revision 1, made

by the Nuclear Energy Institute

Technical Specifications Task Force

Change Number 258, Revision 4,

addressing changes to various

administrative controls in the TSs.

Date of issuance: May 3, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 90 days.

Amendment No.: 196.

Facility Operating License No. DPR–

20. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 24, 2001 (66 FR

7678).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 3, 2001.

No significant hazards consideration

comments received: No.

Consumers Energy Company, Docket

No. 50–255, Palisades Plant, Van Buren

County, Michigan

Date of application for amendment:

December 7, 2000.

Brief description of amendment: The

amendment changes the Technical

Specifications (TSs) in accordance with

changes to the ‘‘Standard Technical

Specifications, Combustion Engineering

Plants,’’ NUREG 1432, Revision 1, made

by the Nuclear Energy Institute

Technical Specifications Task Force

Change Number 287, Revision 5,

addressing allowances for breach of the

control room envelope. Also, the action

table for TS Limiting Condition for

Operation 3.7.10 is corrected by

restoring Required Action D.2 (now

renumbered to E.2), which was

inadvertently omitted in Amendment

No. 189, issued on November 30, 1999.

Date of issuance: May 3, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 90 days.

Amendment No.: 197.

Facility Operating License No. DPR–

20. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 24, 2001 (66 FR

7678).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 3, 2001.

No significant hazards consideration

comments received: No.

Consumers Energy Company, Docket

No. 50–255, Palisades Plant, Van Buren

County, Michigan

Date of application for amendment:

December 7, 2000.

Federal Register / Vol. 66, No. 104 / Wednesday, May 30, 2001 / Notices 29365

Brief description of amendment: The

amendment changes Technical

Specification (TS) 3.5.2 in accordance

with changes to the ‘‘Standard

Technical Specifications, Combustion

Engineering Plants,’’ NUREG 1432,

Revision 1, made by the Nuclear Energy

Institute Technical Specifications Task

Force change number 325, Revision 0,

addressing changes to the structure of

the TS Limiting Condition for Operation

(LCO) for the Emergency Core Cooling

System.

Date of issuance: May 3, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 90 days.

Amendment No.: 198.

Facility Operating License No. DPR–

20. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 24, 2001 (66 FR

7675).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 3, 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:

April 6, 2001, as supplemented by letter

dated May 3, 2001.

Brief description of amendment: The

amendment revises the surveillance

requirements pertaining to testing of the

emergency diesel generators. The

change removes the restrictions in plant

technical specifications that prohibit

performing the required testing during

plant operation (Modes 1, 2, and 3).

Additionally, the amendment modifies

plant technical specifications to allow

the endurance test to be performed in

lieu of the load-run test provided the

requirements of the load-run test, except

the upper limit, are met.

Date of issuance: May 18, 2001.

Effective date: May 18, 2001, to be

implemented within 30 days of the date

of issuance.

Amendment No.: 173.

Facility Operating License No. NPF–

21: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: April 17, 2001 (66 FR 19801).

The May 3, 2001, supplemental letter

provided 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.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 18, 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:

February 5, 2001, as supplemented on

April 13, 2001.

Brief description of amendment: This

amendment changes the Safety Limit

Minimum Critical Power Ratio in

Technical Specification (TS) 2.1.2 from

1.08 to 1.06. The amendment makes

administrative changes to TS 5.6.5,

‘‘Core Operating Limits Report,’’ section

a and b. The amendment makes

administrative changes to Bases section

2.1 to reflect this TS change and to

Bases section 3.11 to reflect an earlier

TS change.

Date of issuance: May 8, 2001.

Effective date: As of the date of

issuance, and shall be implemented

prior to startup from Refueling Outage

13.

Amendment No.: 191.

Facility Operating License No. DPR–

35: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 7, 2001 (66 FR 13802).

The April 13, 2001, letter provided

clarfying information that was within

the scope of the amendment request and

did not change the proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 8, 2001.

No significant hazards consideration

comments received: No.

FirstEnergy Nuclear Operating

Company, Docket No. 50–440, Perry

Nuclear Power Plant, Unit 1, Lake

County, Ohio

Date of application for amendment:

November 9, 2000.

Brief description of amendment: This

amendment revised the technical

specifications by approving thirteen of

the simpler, generic administrative/

editorial/consistency improvements

agreed upon between the Nuclear

Energy Institute (NEI) Technical

Specification Task Force (TSTF) and the

Nuclear Regulatory Commission (NRC),

subsequent to the conversion of the

PNPP Technical Specifications to the

improved Standard Technical

Specifications. The improvements

include TSTF–5, TSTF–32, TSTF–38,

TSTF–52, TSTF–65, TSTF–104, TSTF–

106, TSTF–118, TSTF–152, TSTF–166,

TSTF–258, TSTF–278, and TSTF–279.

Date of issuance: May 15, 2001

Effective date: As of the date of

issuance and shall be implemented

within 90 days.

Amendment No.: 120

Facility Operating License No.NPF–

58: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: December 13, 2000 (65 FR

77920).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 15, 2001.

No significant hazards consideration

comments received: No.

Florida Power and Light Company,

Docket Nos. 50–250 and 50–251, Turkey

Point Plant, Units 3 and 4, Dade County,

Florida

Date of application for amendments:

March 12, 2001.

Brief description of amendments: The

amendments reduced the requirement

for average reactor coolant temperature

during the rod cluster control assembly

drop test from greater than or equal to

541°F to greater than or equal to 500°F.

Date of issuance: May 7, 2001.

Effective date: May 7, 2001.

Amendment Nos: 214 and 208.

Facility Operating License Nos. DPR–

31 and DPR–41: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: April 4, 2001 (66 FR 17967).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated May 7, 2001.

No significant hazards consideration

comments received: No.

Nuclear Management Company, LLC,

Docket No. 50–331, Duane Arnold

Energy Center, Linn County, Iowa

Date of application for amendment:

June 9, 2000.

Brief description of amendment: The

amendment revised the Technical

Specifications (TS) 3.7.7 for the Main

Turbine Bypass Valve surveillance test

frequency, TS surveillance requirement

SR 3.7.7.1 frequency from 31 days to 92

days.

Date of issuance: May 16, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 30 days.

Amendment No.: 239.

Facility Operating License No. DPR–

49: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: July 26, 2000 (65 FR 46009).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 16, 2001.

No significant hazards consideration

comments received: No.

29366 Federal Register / Vol. 66, No. 104 / Wednesday, May 30, 2001 / Notices

Nuclear Management Company, LLC,

Docket Nos. 50–266 and 50–301, Point

Beach Nuclear Plant, Units 1 and 2,

Town of Two Creeks, Manitowoc

County, Wisconsin

Date of application for amendments:

November 20, 2000, as supplemented

February 6 and May 3, 2001.

Brief description of amendments:

These amendments incorporate changes

to the Technical Specifications to

increase the allowable deviation in

individual rod position indication. By

the February 6, 2001, supplemental

letter, the licensee withdrew portions of

the original application that dealt with

operation at greater than 85-percent

power. The licensee plans to submit

those portions that deal with operation

at greater than 85-percent power as a

separate amendment request at a later

time.

Date of issuance: May 8, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 45 days.

Amendment Nos.: 200 and 205.

Facility Operating License Nos. DPR–

24 and DPR–27: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: February 7, 2001 (66 FR

9386).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated May 8, 2001.

No significant hazards consideration

comments received: No.

Pacific Gas and Electric Company,

Docket Nos. 50–275 and 50–323, Diablo

Canyon Nuclear Power Plant, Unit Nos.

1 and 2, San Luis Obispo County,

California

Date of application for amendments:

June 8, 2000, as supplemented by letter

dated January 4, 2001.

Brief description of amendments: The

amendments revised Technical

Specification (TS) Section 3.5.5,

‘‘Emergency Core Cooling Systems—

Seal Injection Flow,’’ to replace the

description of the seal injection flow

with a description consistent with the

method used to establish and verify

reactor coolant pump seal injection flow

limits and the method used to calculate

the seal injection flow in the safety

analyses for the Diablo Canyon Nuclear

Power Plant.

Date of issuance: May 7, 2001.

Effective date: May 7, 2001, and shall

be implemented within 30 days from

the date of issuance.

Amendment Nos.: Unit 1—148; Unit

2—148.

Facility Operating License Nos. DPR–

80 and DPR–82: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: April 4, 2001 (66 FR 17968).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated May 7, 2001.

No significant hazards consideration

comments received: No.

Southern Nuclear Operating Company,

Inc., et al., Docket Nos. 50–424 and 50–

425, Vogtle Electric Generating Plant,

Units 1 and 2, Burke County, Georgia

Date of application for amendments:

August 28, 2000.

Brief description of amendments: The

amendments revised the Technical

Specifications (TS) Section 5.5.2.b,

‘‘Primary Coolant Sources Outside

Containment’’ by changing the system

leak test frequency from ‘‘at refueling

cycle intervals or less’’ to ‘‘at least once

every 18 months.’’ The proposed change

will also allow the provisions of

Surveillance Requirement (SR) 3.0.2 to

apply to TS Section 5.5.2.b.

Date of issuance: May 11, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment Nos.: 119 and 97.

Facility Operating License Nos. NPF–

68 and NPF–81: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: September 20, 2000 (65 FR

56955).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated May 11, 2001.

No significant hazards consideration

comments received: No.

Southern Nuclear Operating Company,

Inc., et al., Docket Nos. 50–424 and 50–

425, Vogtle Electric Generating Plant,

Units 1 and 2, Burke County, Georgia

Date of application for amendments:

October 5, 2000.

Brief description of amendments: The

amendments revise the licenses to

reflect changes to the Updated Final

Safety Analysis Report due to revisions

to the dose equivalent iodine analysis.

Date of issuance: May 11, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment Nos.: 120 and 98.

Facility Operating License Nos. NPF–

68 and NPF–81: Amendments revised

the license.

Date of initial notice in Federal

Register: December 13, 2000 (65 FR

77925).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated May 11, 2001.

No significant hazards consideration

comments received: No.

Southern Nuclear Operating Company,

Inc., et al., Docket Nos. 50–424 and 50–

425, Vogtle Electric Generating Plant,

Units 1 and 2, Burke County, Georgia

Date of application for amendments:

November 6, 2000, as supplemented by

letter dated February 9, 2001.

Brief description of amendments: The

amendments revised Technical

Specifications (TS) 3.7.10, ‘‘Control

Room Emergency Filtration System

(CREFS)—Both Units Operating,’’ TS

3.7.11, ‘‘Control Room Emergency

Filtration System (CREFS)—One Unit

Operating,’’ and TS 3.7.13, ‘‘Piping

Penetration Area Filtration and Exhaust

System (PPAFES),’’ to establish actions

to be taken for inoperable ventilation

systems due to a degraded control room

pressure boundary or piping penetration

area pressure boundary, respectively.

Specifically, the changes allow the

pressure boundaries of ventilation

systems such as CREFS and PPAEFS to

be opened intermittently under

administrative control. A new condition

is also added that allows 24 hours to

restore inoperable CREFS and PPAFES

pressure boundaries before requiring the

units to perform an orderly shutdown.

The applicable TS Bases have been

revised to document these TS changes

and to provide supporting information.

These changes are based on Technical

Specifications Task Force (TSTF) –287,

Revision 5, to the Standard Technical

Specifications.

Date of issuance: May 14, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment Nos.: 121 and 99.

Facility Operating License Nos. NPF–

68 and NPF–81: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: December 13, 2000 (65 FR

77926).

The supplemental letter dated

February 9, 2001, provided clarifying

information that did not change the

scope of the November 6, 2000,

application nor the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated May 14, 2001.

No significant hazards consideration

comments received: No.

Federal Register / Vol. 66, No. 104 / Wednesday, May 30, 2001 / Notices 29367

1 15 U.S.C. 78l(d).

2 17 CFR 240.12d2–2(d).

3 15 U.S.C. 78l(b).

4 17 CFR 200.30–3(a)(1).

Tennessee Valley Authority, Docket No.

50–327, Sequoyah Nuclear Plant, Unit 1,

Hamilton County, Tennessee

Date of application for amendment:

March 9, 2001.

Brief description of amendment: The

amendment revised the Technical

Specifications (TSs) by allowing

insertion of up to four lead test

assemblies containing downblended

uranium, in accordance with

Framatome Cogema Fuels Topical

Report BAW 2328, into the Sequoyah

Unit 1 core for up to two fuel cycles.

Date of issuance: May 9, 2001.

Effective date: May 9, 2001.

Amendment No.: 268.

Facility Operating License No. DPR–

77: Amendment revised the TSs.

Date of initial notice in Federal

Register: April 4, 2001 (66 FR 17970).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 9, 2001.

No significant hazards consideration

comments received: No.

Virginia Electric and Power Company, et

al., Docket Nos. 50–280 and 50–281,

Surry Power Station, Units 1 and 2,

Surry County, Virginia

Date of application for amendments:

December 7, 2000.

Brief Description of amendments:

These amendments revise the Technical

Specifications (TS) in Section 3.23 for

the Main Control Room and Emergency

Switchgear Room Ventilation and Air

Conditioning Systems; TS Surveillance

Requirement (SR) Section 4.20 for the

Control Room Air Filtration System; and

TS SR Section 4.12 for the Auxiliary

Ventilation Exhaust Filter Trains. The

proposed changes will revise the above

SRs for the laboratory testing of the

carbon samples for methyl iodide

removal efficiency to be consistent with

American Society for Testing and

Materials Standard D3803–1989,

‘‘Standard Test Method for NuclearGrade Activated Carbon,’’ with

qualification, as the laboratory testing

standard for both new and used

charcoal adsorbent used in the

ventilation system.

Date of issuance: May 14, 2001.

Effective date: May 14, 2001.

Amendment Nos.: 225 and 225.

Facility Operating License Nos. DPR–

32 and DPR–37: Amendments change

the Technical Specifications.

Date of initial notice in Federal

Register: March 21, 2001, (66 FR

15931), supersedes March 20, 2000 (65

FR 15388).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated May 14, 2001.

No significant hazards consideration

comments received: No.

Yankee Atomic Electric Co., Docket No.

50–29, Yankee Nuclear Power Station

(YNPS) Franklin County, Massachusetts

Date of application for amendment:

November 22, 2000.

Brief description of amendment: The

amendment relocated certain

administrative requirements from the

Yankee Nuclear Power Station (YNPS)

Defueled Technical Specifications to the

YNPS Decommissioning Quality

Assurance Program. Additional editorial

changes to titles and designations were

also made.

Date of issuance: May 15, 2001.

Effective date: May 15, 2001.

Amendment No.: 155.

Facility Operating License No. DPR–3.

The amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: April 4, 2001 (66 FR 17972).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 15, 2001.

No significant hazards consideration

comments received: No.

Dated at Rockville, Maryland this 22nd day

of May 2001.

For the Nuclear Regulatory Commission.

John A. Zwolinski,

Director, Division of Licensing Project

Management, Office of Nuclear Reactor

Regulation.

[FR Doc. 01–13400 Filed 5–29–01; 8:45 am]

BILLING CODE 7590–01–P

SECURITIES AND EXCHANGE

COMMISSION

[File No. 1–12514]

Issuer Delisting; Notice of Application

To Withdraw From Listing and

Registration; (Keystone Property

Trust, Common Stock, Par Value $.01

Per Share)

May 23, 2001.

Keystone Property Trust, a Maryland

real estate investment trust (‘‘Issuer’’),

has filed an application with the

Securities and Exchange Commission

(‘‘Commission’’), pursuant to Section

12(d) of the Securities Exchange Act of

1934 (‘‘Act’’) 1 and Rule 12d2–2(d)

hereunder,2 to withdraw its Common

Stock, par value $.01 per share

(‘‘Security’’), from listing and

registration on the American Stock

Exchange (‘‘Amex’’).

The Issuer stated in its application

that it has met the requirements of

Amex Rule 18 by complying with all

applicable laws in effect in the State of

Maryland, in which it is incorporated,

and with the Amex’s rules governing an

issuer’s voluntary withdrawal of a

security from listing and registration.

The Amex has in turn informed the

Issuer that its does not object to the

proposed withdrawal of the Issuer’s

Security from listing and registration on

the Exchange.

The Board of Trustees (‘‘Board’’)

approved a resolution on April 17, 2001

to withdraw the Issuer’s Security from

listing on the Amex and to list such

Security on the New York Stock

Exchange, effective May 9, 2001. The

Issuer stated that the Board took such

action in order to increase the profile

and visibility of the Issuer in the public

markets and to attract more interest in

the Issuer from individuals and

institutional investors.

The Issuer’s application relates solely

to the withdrawal of the Security from

listing and registration on the Amex and

shall have no effect upon the Security’s

continued listing and registration on the

NYSE under section 12(b) of the Act.3

Any interested person may, on or

before June 13, 2001, submit by letter to

the Secretary of the Securities and

Exchange Commission, 450 Fifth Street,

NW., Washington, DC 20549–0609, facts

bearing upon whether the application

has been made in accordance with the

rules of the Amex and what terms, if

any, should be imposed by the

Commission for the protection of

investors. The Commission, based on

the information submitted to it, will

issue an order granting the application

after the date mentioned above, unless

the Commission determines to order a

hearing on the matter.

For the Commission, by the Division of

Market Regulation, pursuant to delegated

authority.4

Jonathan G. Katz,

Secretary.

[FR Doc. 01–13527 Filed 5–29–01; 8:45 am]

BILLING CODE 8010–01–M

SECURITIES AND EXCHANGE

COMMISSION

[File No. 1–15237]

Issuer Delisting; Notice of Application

To Withdraw From Listing and

Registration; (OTR Express, Inc.,

Common Stock, $.01 Par Value)

May 23, 2001.

OTR Express, Inc., a Kansas

corporation (‘‘Issuer’’), has filed an