34278 Federal Register / Vol. 66, No. 124 / Wednesday, June 27, 2001 / Notices

constitute a clearly unwarranted

invasion of personal privacy.

The agenda for the subject meeting

shall be as follows:

Tuesday, July 10, 2001, 3:00 p.m. until

5:00 p.m.

The Subcommittee will discuss

proposed ACRS activities and related

matters. The purpose of this meeting is

to gather information, analyze relevant

issues and facts, and to formulate

proposed positions and actions, as

appropriate, for deliberation by the full

Committee. A portion of this meeting

may be closed pursuant to 5 U.S.C.

552b(c)(2) and (6) to discuss

organizational and personnel matters

that relate solely to internal personnel

rules and practices of ACRS, and

information the release of which would

constitute a clearly unwarranted

invasion of personal privacy.

Oral statements may be presented by

members of the public with the

concurrence of the Subcommittee

Chairman; written statements will be

accepted and made available to the

Committee. Electronic recordings will

be permitted only during those portions

of the meeting that are open to the

public, and questions may be asked only

by members of the Subcommittee, its

consultants, and staff. Persons desiring

to make oral statements should notify

the cognizant ACRS staff person named

below five days prior to the meeting, if

possible, so that appropriate

arrangements can be made.

Further information regarding topics

to be discussed, the scheduling of

sessions open to the public, whether the

meeting has been canceled or

rescheduled, the Chairman’s ruling on

requests for the opportunity to present

oral statements, and the time allotted

therefor can be obtained by contacting

the cognizant ACRS staff person, Dr.

John T. Larkins (telephone: 301/415–

7360) between 7:30 a.m. and 4:15 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 changes in schedule, etc., that

may have occurred.

Dated: June 21, 2001.

James E. Lyons,

Associate Director for Technical Support,

ACRS/ACNW.

[FR Doc. 01–16095 Filed 6–26–01; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Advisory Committee on Reactor

Safeguards; Meeting of the ACRS

Subcommittee on Plant Operations;

Notice of Meeting

The ACRS Subcommittee on Plant

Operations will hold a meeting on July

9, 2001, Room T–2B3, 11545 Rockville

Pike, Rockville, Maryland.

The entire meeting will be open to

public attendance.

The agenda for the subject meeting

shall be as follows:

Monday, July 9, 2001—9:30 a.m. until

12:30 p.m.

The Subcommittee will continue its

discussion of the Reactor Oversight

Process. The purpose of this meeting is

to gather information, analyze relevant

issues and facts, and to formulate

proposed positions and actions, as

appropriate, for deliberation by the full

Committee.

Oral statements may be presented by

members of the public with the

concurrence of the Subcommittee

Chairman and written statements will

be accepted and made available to the

Committee. Electronic recordings will

be permitted only during those portions

of the meeting that are open to the

public, and questions may be asked only

by members of the Subcommittee, its

consultants, and staff. Persons desiring

to make oral statements should notify

the cognizant ACRS staff engineer

named below five days prior to the

meeting, if possible, so that appropriate

arrangements can be made.

During the initial portion of the

meeting, the Subcommittee, along with

any of its consultants who may be

present, may exchange preliminary

views regarding matters to be

considered during the balance of the

meeting.

The Subcommittee will then hear

presentations by and hold discussions

with representatives of the NRC staff,

and other interested persons regarding

this review.

Further information regarding topics

to be discussed, whether the meeting

has been canceled or rescheduled, and

the Chairman’s ruling on requests for

the opportunity to present oral

statements and the time allotted

therefore, can be obtained by contacting

the cognizant ACRS staff engineer, Ms.

Maggalean W. Weston (telephone: 301/

415–3151) between 8:00 a.m. and 5:30

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: June 21, 2001.

James E. Lyons,

Associate Director for Technical Support.

[FR Doc. 01–16096 Filed 6–26–01; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Biweekly Notice; Applications and

Amendments to Facility Operating

Licenses Involving No Significant

Hazards Considerations

I. Background

Pursuant to Public Law 97–415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 97–415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from June 4

through June 15, 2001. The last

biweekly notice was published on June

12, 2001 (66 FR 31700).

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

Federal Register / Vol. 66, No. 124 / Wednesday, June 27, 2001 / Notices 34279

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

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 July 27, 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

34280 Federal Register / Vol. 66, No. 124 / Wednesday, June 27, 2001 / Notices

Public Document Room, located at One

White Flint North, 11555 Rockville Pike

(first floor), Rockville, Maryland.

Publicly available records will be

accessible from the Agencywide

Documents Assess and Management

Systems (ADAMS) Public Electronic

Reading Room on the internet at the

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

ADAMS/index.html. If you do not have

access to ADAMS or if there are

problems in accessing the documents

located in ADAMS, contact the NRC

Public Document room (PDR) Reference

staff at 1–800–397–4209, 304–415–4737

or by email to pdr@nrc.gov.

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: June 4,

2001.

Description of amendments request:

The proposed license amendments

revise, from 2 hours to 6 hours, the time

period in Surveillance Requirement

3.6.1.6.1 for verifying that each

suppression chamber-to-drywell

vacuum breaker is closed after any

discharge of steam to the suppression

chamber from any source. In

conjunction with this change, the

Completion Time associated with

Required Action B.1 for closing an open

vacuum breaker is being revised from 8

hours to 4 hours.

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 proposed changes provide additional

time to verify that each vacuum breaker is

closed and reduce the time allowed for

closing an open vacuum breaker. The safety

functions of the suppression chamber-todrywell vacuum breaker valves are to relieve

vacuum in the drywell following a postulated

loss-of-coolant accident and to remain

closed, except when the vacuum breakers are

performing their intended design function, in

order to ensure that no excessive bypass

leakage occurs from the drywell to the

suppression chamber. With a vacuum breaker

not closed, communication between the

drywell and suppression chamber airspaces

could occur and, if a loss-of-coolant accident

were to occur, there would be the potential

for primary containment overpressurization

due [to] steam leakage from the drywell to

the suppression chamber without quenching.

The vacuum breakers do not perform a safety

function that initiates, or alters initiation of,

an accident previously evaluated. Rather, the

vacuum breakers function to mitigate the

consequences of certain design basis

accidents. Therefore, the proposed changes

do not involve an increase in the probability

of an accident previously evaluated or the

method of performing their safety functions.

As noted above, the vacuum breakers

function to mitigate the consequences of

certain design basis accidents. The proposed

changes to the Surveillance Requirement and

Completion Time provide additional time to

verify that each vacuum breaker is closed and

reduce the time allowed for closing an open

vacuum breaker; however, the proposed

changes do not alter the safety functions of

the vacuum breakers. When performing the

surveillance to verify each vacuum breaker is

closed, the expected result is the verification

that the component is indeed closed.

However, if this surveillance result is not

obtained, the Technical Specifications limit

the time allowed to close the vacuum

breaker. Additional time is being provided to

verify that each vacuum breaker is closed;

however, the overall time allowed for closing

and verifying closure of a vacuum breaker is

not being increased. Since the overall time to

take action for an open vacuum breaker has

not been increased, the proposed changes do

not involve an increase in the consequences

of an accident previously evaluated.

2. The proposed license amendments will

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The suppression chamber-to-drywell

vacuum breakers are not an initiator of any

design basis accident. Rather, the safety

functions of the vacuum breaker valves are to

relieve vacuum in the drywell following a

loss-of-coolant accident and to remain closed

when not relieving vacuum to ensure that no

excessive bypass leakage occurs from the

drywell to the suppression chamber. Neither

safety function of these vacuum breakers is

altered by the proposed changes. Therefore,

the proposed changes will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The proposed license amendments do

not involve a significant reduction in a

margin of safety.

The proposed changes will not affect the

ability of the suppression chamber-to-drywell

vacuum breakers to perform their safety

functions. Rather, as previously stated, the

proposed changes provide additional time to

verify that each vacuum breaker is closed and

reduce the time allowed for closing an open

or inoperable vacuum breaker. As a result,

the overall time for taking action for an open

vacuum breaker is unchanged. The vacuum

breakers will continue to be verified closed

every 14 days, as part of a required functional

test of the vacuum breaker every 31 days, and

following any activity involving the

discharge of steam to the suppression

chamber. If a vacuum breaker is found to be

open and cannot be closed as required, plant

shutdown will continue to be required

within the same time requirements as

currently specified in the Technical

Specifications. Current Technical

Specifications allow up to 10 hours to close

an open vacuum breaker (i.e., 2 hours to

perform the surveillance to verify vacuum

breaker closure and, if necessary, 8 hours to

close the vacuum breaker). The proposed

change maintains the 10 hour limit by

reducing the time to 4 hours to close an open

or inoperable vacuum breaker while

increasing the time to 6 hours to complete

the surveillance to verify vacuum breaker

closure. Thus, on this basis, the proposed

license amendments will not change overall

plant risk and do not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

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.

Carolina Power & Light Company, et al.,

Docket No. 50–400, Shearon Harris

Nuclear Power Plant, Unit 1, Wake and

Chatham Counties, North Carolina

Date of amendment request: May 18,

2001.

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) 3/4.9.4

‘‘Containment Building Penetrations’’

and the associated Bases to permit

containment building penetrations to

remain open, under administrative

controls, during core alterations or the

movement of irradiated fuel within the

containment. Specifically, the licensee

proposes: (1) Incorporating an alternate

source term methodology in the fuel

handling accident analysis; (2) revising

TS 3.9.4 to remove portions of a note

restricting the applicability of

administrative controls with respect to

containment penetrations; and (3)

including the use of administrative

controls on the equipment hatch and

other penetrations that provide access

from containment atmosphere to outside

atmosphere.

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 does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed changes modify TS

requirements previously reviewed and

Federal Register / Vol. 66, No. 124 / Wednesday, June 27, 2001 / Notices 34281

approved by the NRC in improved Technical

Specifications (ITS) and changes to ITS as

described in TSTF [Technical Specification

Task Force]-312. An alternate source term

calculation has been performed for the HNP

[Harris Nuclear Plant] that demonstrates that

dose consequences remain below limits

specified in NRC Regulatory Guide 1.183 and

10 CFR 50.67. The proposed change does not

modify the design or operation of equipment

used to move spent fuel or to perform core

alterations[.]

Therefore, 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.

Containment penetrations are designed to

form part of the containment pressure

boundary. The proposed change provides for

administrative controls and operating

restrictions for containment penetrations

consistent with guidance approved by the

NRC staff. Containment penetrations are not

an accident initiating system as described in

the Final Safety Analysis Report [FSAR]. The

proposed change does not affect other

Structures, Systems, or Components. The

operation and design of containment

penetrations in operational modes 1–4 will

not be affected by this proposed change.

Therefore, the proposed change does 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 the margin

of safety.

The proposed changes modify required

Actions and Surveillance Requirements

previously reviewed and approved by the

NRC in improved Technical Specifications

(ITS) and changes to ITS, TSTF–312.

Additionally, the implementation of the

alternate source term methodology is

consistent with NRC Regulatory Guide 1.183.

The proposed change to containment

penetrations does not significantly affect any

of the parameters that relate to the margin of

safety as described in the Bases of the TS or

the FSAR. Accordingly, NRC Acceptance

Limits are not significantly affected by this

change.

Therefore, the proposed 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: 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.

Duke Energy Corporation, et al.,

Docket No. 50–414, Catawba Nuclear

Station, Unit 2, York County, South

Carolina

Date of amendment request: March 9,

2001.

Description of amendment request:

The amendment will revise the cold leg

elbow tap flow coefficients used in the

determination of Reactor Coolant

System (RCS) flow rate at Catawba

Nuclear Station, Unit 2. No changes in

Technical Specification are necessary

for this amendment.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The following discussion is a summary of

the evaluation of the changes contained in

this proposed amendment against the 10 CFR

50.92(c) requirements to demonstrate that all

three standards are satisfied. A no significant

hazards consideration is indicated if

operation of the facility in accordance with

the proposed amendment would not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated, or

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated, or

3. Involve a significant reduction in a

margin of safety.

First Standard

The proposed amendment will not involve

a significant increase in the probability or

consequences of an accident previously

evaluated. No component modification,

system realignment, or change in operating

procedure will occur which could affect the

probability of any accident or transient. The

revised cold leg elbow tap flow coefficients

will not change the probability of actuation

of any Engineered Safeguards Feature or

other device. The actual Unit 2 RCS flow rate

will not change. Therefore, the consequences

of previously analyzed accidents will not

change as a result of the revised flow

coefficients.

Second Standard

The proposed amendment will not create

the possibility of a new or different kind of

accident from any accident previously

evaluated. No component modification or

system realignment will occur which could

create the possibility of a new event not

previously considered. No change to any

methods of plant operation will be required.

The elbow taps are already in place, and are

presently being used to monitor flow for

Reactor Protection System purposes. They

will not initiate any new events.

Third Standard

The proposed amendment will not involve

a significant reduction in a margin of safety.

The removal of some of the excess flow

margin, which was introduced by the hot leg

streaming flow penalties in later

calorimetrics, will allow additional operating

margin between the indicated flow and the

Technical Specification minimum measured

flow limit. The proposed changes in the cold

leg elbow tap flow coefficients will continue

to be conservative with respect to the

analytical model flow predictions, since the

proposed coefficients will continue to

contain some hot leg streaming penalties

from the calorimetric determined coefficients

used in the average.

An increase in the RCS flow indication of

approximately 1.0% will increase the margin

to a reactor trip on low flow but will not

adversely affect the plant response to low

flow transients. Current UFSAR Chapter 15

transients that would be expected to cause a

reactor trip on the RCS low flow trip setpoint

are Partial Loss of Reactor Coolant Flow,

Reactor Coolant Pump Shaft Seizure and

Reactor Coolant Pump Shaft break transients.

Three reactor trip functions provide

protection for these transients, RCS low flow

reactor trip, RCP undervoltage reactor trip

and RCP underfrequency reactor trip. The

transient analyses of these events assume the

reactor is tripped on the low flow reactor trip

setpoint. This is conservative and produces

a more severe transient response since a

reactor trip on undervoltage or

underfrequency would normally be expected

to trip the reactor sooner and therefore

reduce the severity of these transients.

The RCS low flow reactor trip is currently

set at 91% of the Technical Specification

minimum measured flow of 390,000 gpm.

The setpoint will not be revised as a result

of this change, which means the transients

relying on this function will behave in the

same manner with the reactor trips occurring

at essentially the same conditions as

previously analyzed. Therefore, any small

increase in the reactor trip margin gained by

the small increase in the indicated RCS flow

will not adversely affect the plant response

during these low flow events.

Based upon the preceding discussion,

Duke Energy has concluded that the

proposed amendment does not involve a

significant hazards consideration.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 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: Ms. Lisa F.

Vaughn , Legal Department (PB05E),

Duke Energy Corporation, 422 South

Church Street, Charlotte, North Carolina

28201–1006.

NRC Section Chief: Richard L. Emch,

Jr.

Entergy Gulf States, Inc., and Entergy

Operations, Inc., Docket No. 50–458,

River Bend Station, Unit 1, West

Feliciana Parish, Louisiana

Date of amendment request: May 23,

2001.

34282 Federal Register / Vol. 66, No. 124 / Wednesday, June 27, 2001 / Notices

Description of amendment request:

The amendment request proposes a

change to the minimum critical power

ratio safety limit (SLMCPR) and changes

to the references for the analytical

methods used to determine the core

operating limits.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

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 Minimum Critical Power Ratio (MCPR)

safety limit is defined in the Bases to

Technical Specification [TS] 2.1.1 as that

limit which ‘‘ensures that during normal

operation and during AOOs [Anticipated

Operational Occurrences], at least 99.9% of

the fuel rods in the core do not experience

transition boiling.’’ The MCPR safety limit

satisfies the requirements of General Design

Criterion 10 of Appendix A to 10 CFR [Part]

50 regarding acceptable fuel design limits.

The MCPR safety limit is re-evaluated for

each reload using NRC [Nuclear Regulatory

Commission]-approved methodologies. The

analyses for RBS [River Bend Station] Cycle

11 have concluded that a two-loop MCPR

safety limit of 1.08, based on the application

of Framatome ANP Richland, Inc.’s [FRA–

ANP] [(proprietary)] NRC-approved MCPR

safety limit methodology, will ensure that

this acceptance criterion is met. For singleloop operation, a MCPR safety limit of 1.10,

also ensures that this acceptance criterion is

met.

In addition to the MCPR safety limit, core

operating limits are established to support

the Technical Specification 3.2 requirements

which ensure that the fuel design limits are

not exceeded during any conditions of

normal operation or in the event of any

anticipated operational occurences (AOO).

The methods used to determine the core

operating limits for each operating cycle are

based on methods previously found

acceptable by the NRC and listed in TS

section 5.6.5. A change to TS section 5.6.5 is

requested to include the FRA–ANP methods

in the list of NRC approved methods

applicable to RBS. These NRC approved

methods will continue to ensure that

acceptable operating limits are established to

protect the fuel cladding integrity during

normal operation and in the event of an

AOO.

The requested Technical Specification

changes do not involve any plant

modifications or operational changes that

could affect system reliability or performance

or that could affect the probability of operator

error. The requested changes do not affect

any postulated accident precursors, do not

affect any accident mitigating systems, and

do not introduce any new accident initiation

mechanisms.

Therefore, these changes to the Minimum

Critical Power Ration (MCPR) safety limit

and to the list of methods used to determine

the core operating limits do not involve a

significant increase in the probability or

consequences of any accident previously

evaluated.

2. Will operation of the facility in

accordance with this proposed change create

the possibility of a new or different kind of

accident from any accident previously

evaluated?

The ATRIUM–10 fuel to be used in Cycle

11 is of a design compatible with the coresident GE–11. Therefore, the introduction

of ATRIUM–10 fuel into the Cycle 11 core

will not create the possibility of a new or

different kind of accident. The proposed

changes do not involve any new modes of

operation, any changes to setpoints, or any

plant modifications. The proposed revised

MCPR safety limits have accounted for the

mixed fuel core and have been shown to be

acceptable for Cycle 11 operation.

Compliance with the criterion for incipient

boiling transition continues to be ensured.

The core operating limits will continue to be

developed using NRC approved methods

which also account for the mixed fuel core

design. The proposed MCPR safety limits or

methods for establishing the core operating

limits do not result in the creation of any

new precursors to an 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?

The MCPR safety limits have been

evaluated in accordance with Framatome

ANP Richland, Inc.’s NRC-approved cyclespecific safety limit methodology to ensure

that during normal operation and during

Anticipated Operational Occurrences

(AOO’s) at least 99.9% of the fuel rods in the

core are not expected to experience transition

boiling. On this basis, the implementation of

this Framatome ANP Richland, Inc.

methodology does not involve a significant

reduction in a 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: Mark

Wetterhahn, Esq., Winston & Strawn,

1400 L Street, NW., Washington, DC

20005.

NRC Section Chief: Robert A. Gramm.

Entergy Nuclear Operations, Inc.,

Docket No. 50–333, James A. FitzPatrick

Nuclear Power Plant, Oswego County,

New York

Date of amendment request: May 11,

2001.

Description of amendment request:

This amendment revised the Technical

Specifications to allow, on a one-time

basis only, Entergy Nuclear Operations,

Inc. to extend the allowed out-of-service

time for the Residual Heat Removal

Service Water (RHRSW) System from 7

days to 11 days. This amendment is

only applicable during installation of

the modification 00–12 to the ‘‘B’’

RHRSW Strainer.

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:

Involve an increase in the probability or

consequences of an accident previously

evaluated.

The CCDP [Conditional Core Damage

Probability] due to this proposed change is

calculated to be 4.33 E–8 (assuming no-risk

significant SSC maintenance), which falls

below the threshold probability of 1 E–6 for

risk significance of temporary changes to the

plant configuration in the EPRI PSA

Applications Guide (Reference 2). The

ICLERP [incremental conditional large early

release probability] is calculated to be 8.85

E–8, which falls below the threshold

probability of 1 E–7 for risk significance per

Reference 2 [see application dated May 11,

2001].

This proposed change does not increase

the consequences of an accident previously

evaluated because all relevant accidents

(LOCA) [loss-of-coolant accident] would

result in the transfer of decay heat to the

suppression pool. For this scenario, the same

compliment of equipment will be available to

achieve and maintain cold shutdown as is

required by the current TS LCO [limiting

condition for operation].

Create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed change does not physically

alter the plant. As such, no new or different

types of equipment will be installed. The

new design for the RHRSW strainer packing

gland will be evaluated under a separate 10

CFR 50.59 evaluation and is considered to be

functionally equivalent for the purposes of

this one-time-only proposed TS change.

The connection and use of a temporary

hose for achieving limited containment heat

removal in the event the ‘‘A’’ division of

RHRSW is rendered inoperable for some

reason is a contingency plan that is already

addressed by current plant procedures.

Involve a significant reduction in a margin

of safety.

The CCDP due to this proposed change is

calculated to be 4.33 E–8 (assuming no-risk

significant SSC maintenance). This value

falls below the threshold probability of 1 E–

6 for risk significance of temporary changes

to the plant configuration in the EPRI PSA

Applications Guide (Reference 2). The

CLERP is calculated to be 8.85 E–8, which

falls below the threshold probability of 1 E–

7 for risk significance per Reference 2.

The consequences of a postulated accident

occurring during the extended allowable outVerDate 11<MAY>2000 19:22 Jun 26, 2001 Jkt 194001 PO 00000 Frm 00142 Fmt 4703 Sfmt 4703 E:\FR\FM\27JNN1.SGM pfrm01 PsN: 27JNN1
Federal Register / Vol. 66, No. 124 / Wednesday, June 27, 2001 / Notices 34283

of-service time are bounded by existing

analyses, therefore, there is no 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 E.

Blabey, 1633 Broadway, New York, New

York 10019.

NRC Section Chief: Richard P.

Correia, Acting.

Entergy Operations Inc., Docket No. 50–

382, Waterford Steam Electric Station,

Unit 3, St. Charles Parish, Louisiana

Date of amendment request: May 22,

2001.

Description of amendment request:

The proposed change to Technical

Specification (TS) 3/4.7.1.2, Emergency

Feedwater (EFW) System expands and

clarifies the current TS.

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 administrative and more restrictive

changes will not affect the assumptions,

design parameters, or results of any accident

previously evaluated. The accident

mitigation features of the plant are not

affected by these proposed changes. The

proposed changes do not add or modify any

existing equipment. The administrative

change to test EFW pumps pursuant to the

Inservice Test Program will ensure the EFW

pumps are tested against the more restrictive

of the data points required by either the

safety analysis or the Inservice Test Program.

Therefore, the proposed administrative

changes do not involve a significant increase

in the probability or consequences of any

accident previously evaluated.

The less restrictive changes (allowing 7

days for an inoperable pump due to an

inoperable steam supply, allowing 24 hours

for an inoperable steam supply and one

inoperable motor driven EFW pump,

allowing 72 hours for two inoperable motor

driven EFW pumps, performing Surveillance

Requirements during other than shutdown

conditions, allowing the use of actual

actuation signals in addition to test signals,

and delaying the requirement to complete

Surveillance Requirement ‘‘d’’ to just prior to

Mode 2) will not affect the assumptions,

design parameters, or results of any accident

previously evaluated. The accident

mitigation features of the plant are not

affected by these proposed changes. The

proposed changes do not add or modify any

existing equipment. Therefore, the proposed

less restrictive changes do not involve a

significant increase in the probability or

consequences of any accident previously

evaluated.

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 proposed changes do not alter the

design or configuration of the plant. There

has been no physical change to plant

systems, structures, or components. The

proposed changes will not reduce the ability

of any of the safety-related equipment

required to mitigate Anticipated Operational

Occurrences or accidents.

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?

The proposed change to the LCO [Limiting

Conditions for Operation] requiring three

pumps and two flow paths be OPERABLE

maintains the functionality of the EFW such

that it is capable of performing its design

function as assumed in the Final Safety

Analysis Report. If the functionality of the

system is not maintained, Technical

Specifications require ACTIONs be taken,

within specified time limitations, to restore

EFW to OPERABLE status or shutdown the

reactor. This action is consistent with the

existing Technical Specifications and

NUREG–1432.

The allowed outage time for one inoperable

steam supply has been increased from 72

hours to 7 days in accordance with NUREG–

1432. This is acceptable due to the redundant

OPERABLE steam supply, the availability of

redundant OPERABLE motor-driven EFW

pumps, and the low probability of an event

requiring the inoperable steam supply. This

change is consistent with NUREG–1432 and

has therefore been previously approved by

the NRC [Nuclear Regulatory Commission].

The ACTION for an inoperable steam

supply to the turbine-driven EFW pump

steam turbine concurrent with one motordriven EFW pump being inoperable will

allow a 24 hour completion time. This

change is acceptable based on the ability of

the system to cool the reactor coolant system

to shutdown cooling entry conditions

following a loss of normal feedwater. The 24

hour completion time is reasonable based on

the redundant OPERABLE steam supply to

the turbine-driven EFW pump steam turbine,

the OPERABLE motor-driven EFW pump,

and the low probability of an event requiring

the inoperable steam supply to the turbinedriven EFW pump.

The ACTION for an inoperable steam

supply to the turbine-driven EFW pump

steam turbine concurrent with both motordriven EFW pumps being inoperable as

proposed requires a unit shutdown be

initiated immediately. This change is

appropriate due to the seriousness of the

condition and is acceptable due to the ability

of the EFW system to support the unit shut

down.

The ACTION for the EFW system

inoperable for reasons other than those

described in ACTION (a), (b), or (c) and able

to deliver at least 100% flow to either steam

generator as proposed will allow a 72 hour

completion time. This change is acceptable

based on the ability of the system to cool the

RCS [Reactor Coolant System] to SDC

[Shutdown Cooling] entry conditions

following a design basis accident assuming

no single active failure.

The ACTION for the EFW system

inoperable for reasons other than those

described in ACTION (a), (b), or (c) and able

to deliver at least 100% combined flow to the

steam generators as proposed requires a unit

shutdown be initiated immediately. This

change is appropriate due to the seriousness

of the condition and is acceptable due to the

ability of the EFW system to support the unit

shut down.

The ACTION for the EFW system

inoperable and unable to deliver at least

100% flow to the steam generators as

proposed requires immediate action be taken

to restore the ability to deliver at least 100%

flow to the steam generators. The unit is in

a seriously degraded condition in that the

EFW system is unable to support a unit

shutdown. This change is consistent with the

intent of the current EFW Technical

Specification and NUREG–1432.

Testing pursuant to Specification 4.0.5

(Inservice Testing Program) as proposed for

Surveillance Requirement ‘b’ will ensure the

EFW pumps are tested against the more

restrictive of the data points required by

either the safety analysis or ASME [American

Society of Mechanical Engineers] Section XI.

The remaining changes to the EFW

Technical Specification are consistent (other

than format) with NUREG–1432 and have

therefore been previously approved by the

NRC.

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: N. S. Reynolds,

Esquire, Winston & Strawn 1400 L

Street NW., Washington, DC 20005–

3502.

NRC Section Chief: Robert A. Gramm.

Exelon Generation Company, LLC,

Docket Nos. 50–295 and 50–304, Zion

Nuclear Power Station, Units 1 and 2,

Lake County, Illinois.

Date of amendment request: February

28, 2001

Description of amendment request:

The proposed amendments would

34284 Federal Register / Vol. 66, No. 124 / Wednesday, June 27, 2001 / Notices

revise the Technical Specifications to

eliminate the requirement for at least

one person qualified to stand watch to

be present in the control room when

nuclear fuel is stored in the spent fuel

pool.

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 Technical Specifications

(TS) change does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

The Defueled Safety Analysis Report

(DSAR) identifies three categories of events:

spent fuel pool events (i.e., operational

occurrences), fuel handling accidents in the

fuel building, and radioactive waste handling

accidents. There are no active controls in the

control room that affect spent fuel pool

equipment, or the handling of fuel or

radioactive waste. Actions to mitigate the

consequences of these events are taken

outside the control room. Emergency

response is not adversely affected by this

proposed change because the control room is

still available to the emergency response

team and communication capability and

timeliness will not be affected. Therefore, the

proposed change does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed TS change does not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The configuration, operation and accident

response of the systems, structures or

components that support safe storage of the

spent fuel are unchanged by the proposed TS

change. Current site surveillance

requirements ensure frequent and adequate

monitoring of system and component

functionality. Systems in the Spent Fuel

Nuclear Island will continue to be operated

in accordance with current design

requirements and no new components or

system interactions have been identified. No

new accident scenarios, failure mechanisms

or limiting single failures are introduced as

a result of the proposed change. The

proposed TS change does not have an

adverse affect on any system related to safe

storage of spent fuel. Therefore, the proposed

TS change does not create the possibility of

a new or different kind of accident from any

accident previously evaluated.

3. The proposed TS change does not

involve a significant reduction in a margin of

safety.

All design basis accident acceptance

criteria will continue to be met. The margin

of safety relative to the cooling of the spent

fuel is unaffected by the proposed change as

the SFP [spent fuel pool] parameters will

continue to be monitored at the same

frequency that they are monitored now. The

ability of the shift crew to respond to

abnormal or accident conditions is

unaffected by the proposed change since all

controls are located in the fuel building and

any necessary communication will be

handled by the DERO [Defueled Emergency

Response Organization]. Therefore, it is

concluded that the proposed TS 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

requested amendments involve no

significant hazards consideration.

Attorney for licensee: Mr. Robert

Helfrich, Senior Counsel, Nuclear, MidWest Regional Operating Group, Exelon

Generation Company, LLC, 1400 Opus

Place, Suite 900, Downers Grove,

Illinois 60515.

NRC Section Chief: Robert A. Gramm.

FirstEnergy Nuclear Operating

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

Ottawa County, Ohio

Date of amendment request: May 15,

2001.

Description of amendment request:

The proposed amendment would revise

refueling operation Technical

Specification (TS) requirements for

containment equipment hatch cover

closure during core alterations and

during movement of irradiated fuel both

inside containment and in the spent fuel

pool or cask pit. The proposed change

would allow the containment

equipment hatch cover to be off during

core alterations and movement of

irradiated fuel provided the Emergency

Ventilation System is operable with the

ability to filter any radioactive release.

The proposed changes involve TS 3/

4.9.4, Refueling Operations

-Containment Penetrations, and TS 3/

4.9.12, Refueling Operations—Storage

Pool Ventilation, and associated Bases.

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 TS 3/4.9.4, Refueling Operations—

Containment Penetrations, and its associated

Bases, and TS 3/4.9.12, Refueling

Operations—Storage Pool Ventilation, and its

associated Bases. The proposed changes

would provide for access to the containment

through the containment equipment hatch

during core alterations and movement of

irradiated fuel, provided that an Emergency

Ventilation System is operable with the

ability to filter any radioactivity release

through the containment equipment hatch.

The proposed changes would also permit

relying on the closing the containment

personnel air lock by a designated individual

to establish the negative pressure boundary

for the Emergency Ventilation System

servicing the storage pool. The use of a

designated individual to close the

containment personnel airlock is currently

permitted by TS 3.9.4 for meeting

containment closure requirements. Neither

the containment equipment hatch nor the

Emergency Ventilation System contributes to

the initiation of any accident described in the

DBNPS Updated Safety Analysis Report.

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. The approved

analysis for the fuel handling accident inside

containment does not take credit for

containment closure or Emergency

Ventilation System filtering. This analysis

results in a maximum calculated offsite does

well within the limits of 10 CFR 100.

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 means to mitigate the

consequences of an accident.

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 the resulting

consequences. Consequently, 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.

Florida Power and Light Company,

Docket Nos. 50–250 and 50–251, Turkey

Point Plant, Units 3 and 4, Dade County,

Florida

Date of amendment request: May 14,

2001.

Federal Register / Vol. 66, No. 124 / Wednesday, June 27, 2001 / Notices 34285

Description of amendment request:

The proposed amendments would

delete Technical Specifications (TS)

Figures 5.1–1, ‘‘Site Area Map,’’ and

5.1–2, ‘‘Plant Area Map,’’ and would

replace TS 5.1, ‘‘Site,’’ with a site

location description. Conforming

changes are requested to delete TS 5.1.1,

‘‘Exclusion Area,’’ TS 5.1.2, ‘‘Low

Population Zone,’’ and TS 5.1.3, ‘‘Map

Defining Unrestricted Areas and Site

Boundary for Radioactive Gaseous and

Liquid Effluents,’’ from TS 5.1 and the

TS Index. These changes conform to

NUREG–1431, Rev. 1, Improved

Standard TS for Westinghouse Plants,

and the requirements of 10 CFR 50.36

(c)(4).

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) Operation of the facility in

accordance with the proposed

amendments would not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed amendments are

administrative in nature, removing sections

and maps from the TS, which are located in

other documents previously approved by

NRC. These amendments will not involve a

significant increase in the probability or

consequences of an accident previously

evaluated because they do not affect

assumptions contained in plant safety

analyses, the physical design and/or

operation of the plant, nor do they affect TS

that preserve safety analysis assumptions.

Therefore, the proposed changes do not affect

the probability or consequences of accidents

previously analyzed.

(2) Operation of the facility in accordance

with the proposed amendments would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed changes to the TS are

administrative in nature and can not create

the possibility of a new or different kind of

accident from any previously evaluated since

the proposed amendments will not change

the physical plant or the modes of plant

operation defined in the facility operating

license. No new failure mode is introduced

due to the administrative changes since the

proposed changes do not involve the

addition or modification of equipment, nor

do they alter the design or operation of

affected plant systems, structures, or

components.

(3) Operation of the facility in accordance

with the proposed amendments would not

involve a significant reduction in a margin of

safety.

The proposed changes are administrative

in nature and do not affect operating limits

or functional capabilities of plant systems,

structures and components. The addition of

a site location description to the TS adds

geographical information to the TS.

Elimination of site and plant area maps from

the TS would have no effect on margin of

safety as they are located in other controlled

plant documents. Thus, the changes

proposed would not involve a significant

reduction in margin of safety of the facility.

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).

Nebraska Public Power District, Docket

No. 50–298, Cooper Nuclear Station,

Nemaha County, Nebraska

Date of amendment request: February

15, 2001.

Description of amendment request:

The proposed amendment to the Cooper

Nuclear Station (CNS) Operating

License (OL) DPR–46 would (1) delete

OL Condition 2.D, Additional

Conditions for Protection of the

Environment, and (2) remove the

depiction of railroad tracks in Technical

Specifications (TS) Figure 4.1–1, Site

and Exclusion Area Boundaries and

Low Population Zone.

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.

OL [Operating License] Condition 2.D has

become obsolete based upon it being satisfied

or superceded by amendments to the FSAR

[Final Safety Analysis Report] and OL. The

previous FSAR and OL amendments which

made it obsolete were reviewed and

approved based on their individual

Unreviewed Safety Question (USQ)

evaluations or no significant hazards

considerations. Since this proposed change

does not physically alter any plant

equipment or operating limitations, it

therefore does not impact any previously

evaluated accident initiator, nor change

mitigating systems or features or operating

limitations for accidents previously

evaluated in the Updated Safety Analysis

Report (USAR). Thus, it does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated. This is an administrative change.

2. Do the proposed changes create the

possibility of a new or different kind of

accident from any accident previously

evaluated?

Response: No.

This proposed change is administrative in

nature. It does not involve a physical

alteration of the plant. No new or different

equipment is being installed, and no

installed equipment is being operated in a

new or different manner. No setpoints for

parameters which initiate protective or

mitigative action are being changed. As a

result, no new failure modes are being

introduced. There are no changes in the

procedures or methods governing normal

plant operation, nor are the procedures

utilized to respond to plant transients altered

as a result of this administrative change. This

change does not impose any new or different

requirements or eliminate any existing

requirements. In addition, the change does

not alter assumptions made in the safety

analysis, nor does it impact the licensing

basis. Therefore, the 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 the margin of safety?

Response: No.

This proposed change is administrative in

nature. It does not alter any accident analysis

assumptions, conditions, or methodology.

Since this proposed change does not

physically alter plant systems, structures or

components (SSC’s), change mitigating

systems, features, operating limitations, nor

revise accident analysis assumptions,

conditions or methodology, it does not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Mr. John R.

McPhail, Nebraska Public Power

District, Post Office Box 499, Columbus,

NE 68602–0499.

NRC Section Chief: Robert A. Gramm.

Nuclear Management Company, LLC,

Docket No. 50–331, Duane Arnold

Energy Center, Linn County, Iowa

Date of amendment request: October

19, 2000, as supplemented March 23

and April 9, 2001.

Description of amendment request:

The proposed amendment would

authorize the licensee to change the

licensing basis to utilize the full scope

of an alternative radiological source

term for accidents as described in

NUREG–1465, ‘‘Accident Source Terms

for Light-Water Nuclear Power Plants,’’

and change the Technical Specifications

34286 Federal Register / Vol. 66, No. 124 / Wednesday, June 27, 2001 / Notices

to implement various assumptions in

the Alternative Source Term analyses.

The portion of this amendment request

regarding operability requirements

during core alterations and while

moving irradiated fuel assemblies

within the secondary containment, and

which provided for selective application

of the Alternative Source Term to the

design-basis fuel handling accident was

previously evaluated and issued as

Amendment No. 237 on April 16, 2001.

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 Alternative Source Term and those

plant systems affected by implementing the

setpoints and action levels specified in the

analyses are not assumed to initiate design

basis accidents. The Alternative Source Term

does not affect the design or operation of the

facility; rather, once the occurrence of an

accident has been postulated the new source

term is an input to evaluate the consequence.

The implementation of the Alternative

Source Term has been evaluated in revisions

to the analyses of the limiting design bases

accidents at DAEC [Duane Arnold Energy

Center]. Based on the results of these

analyses, it has been demonstrated that, with

the requested changes, the dose

consequences of these limiting events are

within the regulatory guidance provided by

the NRC for use with the Alternative Source

Term. This guidance is presented in NUREG

1465, 10 CFR 50.67, associated Regulatory

Guide 1.183, and Standard Review Plan

(SRP) Section, 15.0.1. Since secondary

containment operability is not assumed for

the fuel handling accident (FHA), the

consequences of eliminating the

requirements for secondary containment

operability, secondary containment isolation

valves/dampers, secondary containment

instrumentation and the Standby Gas

Treatment system during fuel movement or

core alterations will not increase the effects

of a FHA beyond those evaluated in the

Alternative Source Term analysis. Therefore,

the proposed changes do not significantly

increase the probability or consequences of

any previously evaluated accident.

2. The proposed amendment will not

create the possibility of a new or different

kind of accident from any previously

evaluated.

The Alternative Source Term and those

plant systems affected by implementing the

setpoints and action levels specified in the

analyses do not initiate design basis

accidents. Therefore, the proposed changes

do not create the possibility of a new or

different kind of accident from any

previously evaluated.

3. The proposed amendment will not

involve a significant reduction in a margin of

safety.

The changes proposed are associated with

the implementation of a new licensing basis

for DAEC. Approval of the basis change from

the original source term developed in

accordance with TID–14844 to a new

alternative source term as described in

NUREG–1465 is requested by this submittal.

The results of the accident analyses revised

in support of this submittal, and the

requested Technical Specification changes,

are subject to revised acceptance criteria.

These analyses have been performed using

conservative methodologies. Safety margins

and analytical conservatisms have been

evaluated and are satisfied. The analyzed

events have been carefully selected and

margin has been retained to ensure that the

analyses adequately bound all postulated

event scenarios. The dose consequences of

these limiting events are within the

acceptance criteria also found in the latest

regulatory guidance. This guidance is

presented in NUREG 1465, in the approved

rulemaking for 10 CFR 50.67, and in the

associated Regulatory Guide 1.183.

The proposed changes continue to ensure

that the doses at the exclusion area and low

population zone boundaries, as well as the

control room, are within the corresponding

regulatory limit. Specifically, the margin of

safety for these accidents is considered to be

that provided by meeting the applicable

regulatory limit, which, for most events, is

conservatively set below the 10 CFR 50.67

limit. With respect to the control room

personnel doses, the margin of safety (the

difference between the 10 CFR 50.67 limits

and the regulatory limit defined by 10

CFR50, Appendix A, Criterion 19 (GDC 19))

continues to be satisfied.

Therefore, because the proposed changes

continue to result in dose consequences

within the applicable regulatory limits, they

are considered to not result in a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Al Gutterman,

Morgan, Lewis & Bockius, 1800 M

Street, NW., Washington, DC 20036–

5869.

NRC Section Chief: Claudia M. Craig.

Nuclear Management Company, LLC,

Docket No. 50–263, Monticello Nuclear

Generating Plant, Wright County,

Minnesota

Date of amendment request: May 30,

2001.

Description of amendment request:

The proposed amendment would

eliminate local suppression pool

temperature limits from the Updated

Safety Analysis Report as the basis for

limiting suppression pool mechanical

loads due to unstable steam

condensation during safety relief valve

actuations.

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.

Eliminating the Local Suppression Pool

Temperature Limits (LSPTLs) will not

introduce new equipment or new equipment

methods of operation, and will not alter

existing system relationships. LSPTLs are not

an accident initiator and does [sic] not affect

other accident initiators. The integrity of

fission product barriers do not rely on

LSPTLs since mechanical loads on

containment will not be exceeded and ECCS

[emergency core cooling system] operation in

the event of an accident will not be adversely

affected as demonstrated and approved in

Reference 6 [letter from G. Holahan (NRC) to

R. Pinelli (Boiling Water Reactor Owners

Group), ‘‘Transmittal of the Safety Evaluation

of General Electric Co. Topical Reports;

NEDO–30832, Entitled ’Elimination of Limit

on BWR Suppression Pool Temperature for

SRV Discharge With Quenchers,’ and NEDO–

31695, Entitled ’BWR Suppression Pool

Temperature Technical Specification

Limits’,’’ dated August 29, 1994].

Therefore, the proposed amendment will

not significantly increase the probability or

the consequences of an accident previously

evaluated.

2. The proposed amendment will not

create the possibility of a new or different

kind of accident from any accident

previously analyzed.

Eliminating the LSPTLs will not introduce

new equipment or new equipment methods

of operation, and will not alter existing

system relationships. Since containment

integrity and ECCS operation will not be

challenged, new or different kinds of

accidents are not created.

Therefore, 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 will not

involve a significant reduction in the margin

of safety.

Since LSPTLs are not required to limit

mechanical loads on containment, the margin

of safety associated with containment

integrity is not significantly reduced. Since

LSPTLs are not required to prevent steam

binding of the ECCS pumps, the margin of

safety associated with ECCS operation is not

significantly reduced.

Therefore, the proposed amendment will

not involve a significant reduction in the

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

Federal Register / Vol. 66, No. 124 / Wednesday, June 27, 2001 / Notices 34287

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.

Omaha Public Power District, Docket

No. 50–285, Fort Calhoun Station, Unit

No. 1,

Washington County, Nebraska

Date of amendment request: May 15,

2001.

Description of amendment request:

The proposed changes would: (1)

Replace the titles of Manager—Fort

Calhoun Station and the Vice President

with generic titles, (2) relocate the

requirements for the Plant Review

Committee (PRC) and the Safety Audit

and Review Committee (SARC) to the

Fort Calhoun Station (FCS) Quality

Assurance Program, (3) relocate the

requirements for procedure controls and

records retention to the FCS Quality

Assurance Program, (4) enhance and

clarify the qualification and training

requirements for individuals who

perform licensed operator functions, (5)

incorporate the Westinghouse/CENP

definition of Azimuthal Power Tilt, and

(6) eliminate specific mailing address

and reporting requirements that are

redundant to Title 10 of the Code of

Federal Regulations (10 CFR).

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes: revise the FCS

definition of Azimuthal Power Tilt, remove

specific titles from the Technical

Specifications, provide minor clarifications

of the training requirements for plant staff,

and indicate the change in title of the

Licensed Senior Operator. This change also

relocates the requirements for the Plant

Review Committee (PRC) and the Safety

Audit and Review Committee (SARC),

procedure control, and records retention to

the Fort Calhoun Station Quality Assurance

Program as described in NRC Administrative

Letter 95–06.

The proposed change includes an update

to the definition of Azimuthal Power Tilt and

adds the bases for the definition of

Azimuthal Power Tilt to the bases section of

Section 2.10.4 as recommended in ABB

Combustion Engineering (CE) Infobulletin

Number 97–07, dated December 31, 1997. As

noted in the infobulletin, CE discovered a

discrepancy in the definition for CE analog

plants that use Combustion Engineering Core

Operating Report (CECOR) for monitoring

and surveillance purposes. Plants that use

CECOR should use the same definition as the

CE digital plants. This change will make the

FCS definition and bases agree with the

improved Standard Technical Specifications

for CE digital plants, which have previously

been approved by the NRC.

The proposed change would allow the use

of generic personnel titles as provided in

ANSI/ANS 3.1 and NUREG–1432, ‘‘Standard

Technical Specifications Combustion

Engineering Plants,’’ in lieu of plant-specific

personnel titles. This change does not

eliminate any of the qualifications,

responsibilities or requirements for these

positions, since the plant-specific personnel

titles are currently identified in licensee

controlled documents such as the Updated

Safety Analysis Report (USAR) or the Quality

Assurance Program. For example, Section 12

of the Updated Safety Analysis Report

describes the management structure and

reporting responsibilities of OPPD and

provides an organizational chart to determine

the corporate officer with responsibility for

overall plant nuclear safety from other

corporate officers within OPPD. Therefore,

changing the terminology within the

Technical Specifications, indicating this

reporting responsibility does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated. Changing the periodicity of review

for staff overtime is also considered an

administrative change. This includes a

change of the title of the Supervisor—

Operations to Manager—Shift Operations,

Licensed Senior Operator to Control Room

Supervisor, and crewman to crewmember.

The change to the number of Senior Operator

License present during Core Alterations and

the associated note is also considered

clarifying in nature and not a change of

intent.

The proposed change would update the

qualification requirements for the Manager—

Radiation Protection, the Shift Technical

Advisors, and those individuals that perform

the functions described in 10 CFR 50.54(m)

to Regulatory Guide 1.8, Revision 3, and

ANSI/ANS 3.1–1993. In the March 1987

revision to 10 CFR Part 55, the NRC included

the requirement that those facility licensees

that have made a commitment that is less

than that required by the new rules must

conform to the new rules automatically.

OPPD had previously considered that

commitments made to comply with the

requirements of NUREG–0737 and the

standards applied through the Institute of

Nuclear Power Operations (INPO)

accreditation process were equivalent to the

guidance provided in Regulatory Guide 1.8,

Revision 3. The proposed change provides

enhancement to the current requirements and

clarifies the qualifications and training

requirements for licensed personnel. This

provides additional assurance that these

personnel are properly trained and qualified

for their positions and conforms with the

guidance of NRC Regulatory Issues Summary

2001–01. Therefore, the proposed change

does not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change would relocate

specific requirements for SARC, PRC,

procedure control, and records retention to

the Fort Calhoun Station Quality Assurance

Program (Appendix A, of the FCS USAR).

This proposed revision does not change or

eliminate responsibilities or requirements for

these programs. The management level and

expertise of personnel who are PRC or SARC

members is not being changed. The review of

plant operations, procedures control, and

record retention is still required to be in

compliance with the Fort Calhoun Station

Quality Assurance (QA) Program. Any

changes in the QA Program which reduce the

effectiveness of the program must be

approved by the NRC in accordance with 10

CFR 50.54(a)(4). These changes meet the

criteria as described in NRC Administrative

Letter 95–06. Therefore, the proposed

relocation of these programs to the QA

Program does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

The proposed change would also remove

the requirements prescribing specific

submittal addresses, titles, and reporting

periods. For example, the requirement to

submit License[e] Event Reports within 30

days is replaced with a citation referencing

10 CFR 50.73. This is in agreement with 10

CFR 50.73 and 10 CFR 50.4(f). Additionally,

an administrative requirement prescribing

the submittal of a Special Maintenance

Report is being deleted, as it is redundant to

the requirements of 10 CFR 50.73. Therefore,

the proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes revise

organizational and administrative

requirements contained within the

Administrative Controls section of the TS.

The proposed change to the definition of

Azimuthal Power Tilt is as recommended in

CE Infobulletin 97–07 for CE analog plants

that use CECOR for monitoring and

surveillance purposes and will have no affect

on accidents previously evaluated. The

proposed changes do not revise any

equipment setpoints, change the manner in

which any plant equipment is operated, or

propose any new operating modes. Therefore,

the proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed changes revise

organizational and administrative

requirements contained within the

Administrative Controls section of the TS.

The proposed change to the definition of

Azimuthal Power Tilt has no affect on the

margin of safety. The proposed changes do

not revise any equipment setpoints, change

the manner in which any plant equipment is

34288 Federal Register / Vol. 66, No. 124 / Wednesday, June 27, 2001 / Notices

operated, or propose any new operating

modes. Therefore, the proposed changes do

not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: James R.

Curtiss, Esq., Winston & Strawn, 1400 L

Street, NW., Washington, DC 20005–

3502.

NRC Section Chief: Stephen Dembek.

PSEG Nuclear LLC, Docket No. 50–354,

Hope Creek Generating Station, Salem

County, New Jersey

Date of amendment request: May 17,

2001.

Description of amendment request:

The proposed amendment would revise

the Technical Specifications (TSs) to

permit an increase in the allowable leak

rate for the main steam isolation valves

(MSIVs) and to delete the MSIV Sealing

System (MSIVSS). These changes are

based on the use of an alternate source

term and the guidance provided in

Regulatory Guide 1.183, ‘‘Alternate

Radiological Source Terms for

Evaluating Design Basis Accidents at

Nuclear Power Reactors.’’

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. Does the proposed change involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

As described in Section 6.7 of the Hope

Creek Updated Final Safety Analysis Report

(UFSAR), the MSIVSS limits the leakage of

fission products through the MSIVs following

a design-basis accident large break Loss of

Coolant Accident (LOCA). The system is

manually actuated following a LOCA. The

licensee has proposed to remove the MSIVSS

from the plant and to delete the associated

requirements from the TSs. In addition, the

TSs would be revised to increase the

allowable MSIV leak rate. The MSIVSS lines

and main steamline drain valves that are

connected to the main steam piping will be

capped and welded closed to ensure primary

containment integrity is maintained. The

welding and post-weld examination

procedures will be in accordance with the

American Society of Mechanical Engineers

Code, Section III requirements. The welded

caps will be periodically tested as part of the

Containment Integrated Leak Rate Test. MSIV

leakage and operation of the MSIVSS do not

affect the precursors for accidents analyzed

in Chapter 15 of the Hope Creek UFSAR. In

addition, the proposed changes do not

adversely affect other structures, systems, or

components important to safety. Therefore,

there is no increase in the probability of

occurrence of an accident previously

evaluated as a result of the proposed changes.

The licensee’s submittal states that the

radiological consequences associated with

the proposed changes have been analyzed

based on the results of revised offsite and

control room operator dose calculations for a

LOCA, which is the most limiting Hope

Creek design-basis accident. The current

design-basis analysis for the radiological

consequences associated with a LOCA is

shown in Hope Creek UFSAR Sections 6.4.7

and 15.6.5.5. The revised analysis was

performed using an alternate source term in

accordance with the requirements in 10 CFR

50.67 and the guidance in Regulatory Guide

1.183. The dose calculations assess the

effects of the proposed increase in allowable

MSIV leak rate and take no credit for the

MSIVSS. In addition, the calculations assume

an unfiltered control room inleakage designbasis value that is higher than the current

design basis value to address control room

habitability issues associated with NEI 99–

03. The revised analysis was performed in

accordance with the current accepted

methodology discussed in Regulatory Guide

1.183 and the radiological consequences

were evaluated in terms of Total Effective

Dose Equivalent (TEDE) dose as per the

acceptance criteria specified in 10 CFR 50.67.

The Regulatory Guide 1.183 methodology is

not exactly comparable to the current Hope

Creek design basis analysis which is in terms

of whole body and thyroid doses. The results

of the licensee’s analysis associated with the

proposed changes indicate that the postLOCA doses will result in an increase in the

dose exposures for the control room, the

Exclusion Area Boundary (EAB), and the

Low Population Zone (LPZ), compared to the

current design basis analysis. However, the

revised post-LOCA doses will remain below

the TEDE dose acceptance criteria for the

control room, EAB, and LPZ, as specified in

10 CFR 50.67. The methodology and

guidance provided in Regulatory Guide 1.183

has been developed for the purpose of

performing design basis radiological

consequence analyses using an alternate

source term such that meeting the 10 CFR

50.67 acceptance criteria demonstrates

adequate protection of public health and

safety. Therefore, the proposed changes do

not involve a significant increase in the

consequences of any accident previously

evaluated.

2. Does the proposed change create the

possibility of a new or different kind of

accident from any accident previously

evaluated?

The proposed change to increase the

allowed MSIV leakage rate does not affect the

operability the MSIVs and will not inhibit

the capability of the MSIVs to perform their

function of isolating the primary containment

as assumed in the Hope Creek accident

analyses in UFSAR Chapter 15. The proposed

change to delete the MSIVSS does not

introduce any new modes of plant operation

and, as previously discussed, the designbasis LOCA analysis was reanalyzed without

taking credit for the operation of MSIVSS.

The affected main steam piping will be

welded and/or capped closed to assure that

the primary containment integrity, isolation,

and leak testing capability are not

compromised. Based on the above

considerations, the proposed changes do not

create the possibility of a new or different

kind of accident from any previously

evaluated.

3. Does the proposed change involve a

significant reduction in a margin of safety?

As previously discussed, the results of the

licensee’s analysis associated with the

proposed changes indicate that the postLOCA doses will result in an increase in the

dose exposures for the control room, the

EAB, and the LPZ, compared to the current

design basis analysis. Since there will be an

increase in dose exposure, the margin of

safety will be decreased. However, the

revised post-LOCA doses will remain below

the TEDE dose acceptance criteria for the

control room, EAB, and LPZ, as specified in

10 CFR 50.67. Meeting the 10 CFR 50.67

acceptance criteria demonstrates adequate

protection of public health and safety. An

acceptable margin of safety is inherent in

these acceptance criteria. Therefore, there is

no significant reduction in the margin of

safety as a result of the proposed changes.

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.

PSEG Nuclear LLC, Docket Nos. 50–272

and 50–311, Salem Nuclear Generating

Station, Unit Nos. 1 and 2, Salem

County, New Jersey

Date of amendment request: March 5,

2001.

Description of amendment request:

The proposed amendment would (1)

change the Security Plan provision that

a member of the security force escort all

vehicles, other than designated licensee

vehicles, and to delete the related

Security Training and Qualification

Plan task, (2) change the requirement of

the Security Plan that all areas of the

protected area be illuminated to a

minimum of 0.2 footcandle, and (3)

change the frequency of protected area

patrols in the Security Plan.

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

Federal Register / Vol. 66, No. 124 / Wednesday, June 27, 2001 / Notices 34289

consequences of an accident previously

evaluated?

The proposed changes involving security

activities do not reduce the ability for the

security organization to prevent radiological

sabotage and therefore do not increase the

probability or consequences of a radiological

release previously evaluated.

2. Do the proposed changes create the

possibility of a new or different kind of

accident from any accident previously

evaluated?

The proposed changes involve functions of

the security organization concerning vehicle

control, protected area illumination, and

protected area patrol frequency. Analysis of

the proposed changes has not indicated nor

identified a new or different kind of accident

from any previously evaluated.

3. Do the proposed changes involve a

significant reduction in a margin of safety?

Analysis of the proposed changes show

that they affect only the functions of the

Security organization and have no impact

upon nor cause a significant reduction in

margin of safety for plant operation. The

failure points of key safety parameters are not

affected.

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: Jeffrie J. Keenan,

Esquire, Nuclear Business Unit—N21,

P.O. Box 236, Hancocks Bridge, NJ

08038.

NRC Section Chief: James W. Clifford.

Union Electric Company, Docket No.

50–483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of amendment request: May 30,

2001 (ULNRC–04481).

Description of amendment request:

The proposed amendment changes the

technical specifications to remove the

phrase ‘‘and the charging flow control

valve full open’’ from Limiting

Condition for Operation 3.5.5, Required

Action A.1, and Surveillance

Requirement 3.5.5.1 for the reactor

coolant pump seal injection flow.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The emergency core cooling system (ECCS)

analysis models the reactor coolant pump

(RCP) seal injection flow path as a hydraulic

flow resistance. The proposed change

clarifies that RCP seal injection flow is a

function of system conditions. The seal

injection flow rate can vary during operation,

but the hydraulic flow resistance is fixed by

positioning the manual seal injection throttle

valves. The resistance does not change if the

valve adjustments are not changed. Thus,

RCP seal injection flow variation due to

changing reactor coolant system (RCS)

backpressure following a loss of coolant

accident (LOCA) is explicitly accounted for

as a result of modeling the RCP seal injection

flow path resistance.

The proposed change does not impact the

way the RCP seal injection flow should be

established per the safety analysis and does

not affect RCP seal integrity. The seal

injection flow resistance only affects ECCS

flow. Since ECCS flow occurs after an

accident, the proposed change cannot impact

the probability of an accident.

Overall ECCS performance will remain

within the bounds of the previously

performed accident analyses since there are

no hardware changes. The ECCS will

continue to function in a manner consistent

with the plant design basis. All design,

material, and construction standards that

were applicable prior to the proposed change

are [still] maintained.

The proposed change will not affect the

probability of any event initiators. There will

be no degradation in the performance of, or

an increase in the number of challenges

imposed on, safety-related equipment

assumed to function during an accident

situation. There will be no change to normal

plant operating parameters or accident

mitigation performance.

The proposed change will not alter any

assumptions or change any mitigation actions

in the radiological consequence evaluations

in the FSAR [Final Safety Analysis Report].

Therefore, the proposed change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

There are no hardware changes nor are

there any changes in the method by which

any safety-related plant system performs its

safety function. The proposed change will

not affect the normal method of plant

operation. No performance requirements will

be affected.

Since the proposed change continues to

assure that the assumed ECCS flow is

available after a large break LOCA, no new

accident scenarios, transient precursors,

failure mechanisms, or limiting single

failures are introduced as a result [of the

proposed change]. There will be no adverse

effect or challenges imposed on any safetyrelated system as a result of this request.

The proposed change does not alter the

design or performance characteristics of the

ECCS. It simply corrects the description of

how to properly set the position of the RCP

seal injection throttle valves in support of the

ECCS flow balance assumptions.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any previously

evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

There will be no effect on the manner in

which safety limits or limiting safety system

settings are determined nor will there be any

effect on those plant systems necessary to

assure the accomplishment of protection

functions. There will be no impact on the

overpower limit, departure from nucleate

boiling ratio limits, heat flux hot channel

factor (FQ) nuclear enthalpy rise hot channel

factor (FN/DH), loss of coolant accident peak

cladding temperature (LOCA PCT), peak

local power density, or any other margin of

safety. The radiological dose consequence

acceptance criteria listed in the Standard

Review Plan will continue to be met.

Therefore, the proposed change does not

involve a significant reduction in any 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: John O’Neill,

Esq., Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Section Chief: Stephen Dembek.

Virginia Electric and Power Company,

Docket Nos. 50–280 and 50–281, Surry

Power Station, Unit Nos. 1 and 2, Surry

County, Virginia

Date of amendment request: April 11,

2000, as supplemented by letters dated

August 28, 2000, November 20, 2000,

and April 11, 2001.

Description of amendment request:

The proposed amendments would

revise Technical Specifications (TS) 3.7,

3.10, and 3.22, as well as the Bases of

TS 3.4, 3.8, 3.10, 3.19, and 3.22. The

proposed changes would implement an

alternate accident source term

methodology previously approved by

NRC. Implementation of the alternate

source term could permit a number of

plant changes that have been proposed,

including: Permitting a slight

atmospheric pressure in containment for

a short time following a loss-of-coolant

accident (LOCA), deletion of automatic

function requirements and setpoints for

containment particulate and gas

monitors, deletion of the requirement to

filter fuel building and containment

purge exhaust during refueling, and a

number of other related operational and

configuration requirements.

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:

34290 Federal Register / Vol. 66, No. 124 / Wednesday, June 27, 2001 / Notices

The proposed TS changes allow relaxation

of containment integrity requirements during

refueling operations by allowing the

personnel airlock, equipment access hatch

and certain penetrations to remain open

during fuel movement in containment. The

changes also eliminate the requirement to

filter the exhaust from containment or the

fuel building during refueling operations.

Also proposed is a relaxation of the current

containment design basis acceptance criteria

to allow an interval of four hours following

the design basis LOCA until containment is

depressurized to subatmospheric conditions.

We have reviewed the proposed TS changes

relative to the requirements of 10 CFR 50.92

and determined that a significant hazards

consideration is not involved. Specifically,

operation of Surry Power Station with the

proposed changes will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The probability remains unaffected since

the accident analyses involve no change to a

system, component or structure that affects

initiating events for any of the accidents

evaluated. The consequences of the

reanalyzed events is expressed in terms of

the TEDE [total effective dose equivalent]

dose, which is not directly comparable to

either the thyroid or whole body doses

reported in existing analyses. However, even

taking this comparison into consideration,

any dose increase is not significant.

Furthermore, the revised analysis results

meet the applicable TEDE dose acceptance

criteria for alternative source term

implementation.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The implementation of the proposed

changes does not create the possibility of an

accident of a different type than was

previously evaluated in the SAR [Safety

Analysis Report]. The proposed Technical

Specifications changes allow relaxation of

these current requirements: (1) maintaining

subatmospheric containment conditions

following a LOCA; (2) filtration of

containment & fuel building exhaust during

fuel movement; (3) maintaining the

personnel airlock, equipment access hatch &

penetrations closed during fuel movement

and (4) operability of containment purge

isolation during refueling. These changes do

not alter the nature of events postulated in

the UFSAR [Updated Final SAR] nor do they

introduce any unique precursor mechanisms.

Therefore, there is no possibility for

accidents of a different type than previously

evaluated.

3. Involve a significant reduction in the

margin of safety.

The implementation of the proposed

changes does not reduce the margin of safety.

The radiological analysis results, even

though compared with the revised TEDE

acceptance criteria, meet the applicable

limits. These criteria have been developed for

application to analyses performed with

alternative source terms. These acceptance

criteria have been developed for the purpose

of use in design basis accident analyses such

that meeting the stated limits demonstrates

adequate protection of public health and

safety. It is thus concluded that the margin

of safety will not be reduced by the

implementation of the changes.

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: Donald P. Irwin,

Esq., Hunton and Williams, Riverfront

Plaza, East Tower, 951 E. Byrd Street,

Richmond, Virginia 23219.

NRC Section Chief: Richard L. Emch.

Notice of Issuance of Amendments to

Facility Operating Licenses

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application

complies with the standards and

requirements of the Atomic Energy Act

of 1954, as amended (the Act), and the

Commission’s rules and regulations.

The Commission has made appropriate

findings as required by the Act and the

Commission’s rules and regulations in

10 CFR Chapter I, which are set forth in

the license amendment.

Notice of Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for A Hearing in

connection with these actions was

published in the Federal Register as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the applications for

amendment, (2) the amendment, and (3)

the Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, located at One White Flint North,

11555 Rockville Pike (first floor),

Rockville, Maryland. Publicly available

records will be accessible from the

Agencywide Documents Access and

Management Systems (ADAMS) Public

Electronic Reading Room on the internet

at the NRC web site, http://

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

If you do not have access to ADAMS or

if there are problems in accessing the

documents located in ADAMS, contact

the NRC Public Document Room (PDR)

Reference staff at 1–800–397–4209, 301–

415–4737 or by email to pdr@nrc.gov.

Exelon Generation Company, LLC,

Docket Nos. STN 50–454 and STN 50–

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

Ogle County, Illinois; Docket Nos. STN

50–456 and STN 50–457, Braidwood

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

Illinois

Date of application for amendments:

November 30, 2000.

Brief description of amendments: The

amendments revise TS 5.5.13, ‘‘Diesel

Fuel Oil Testing Program,’’ to relocate

the specific American Society for

Testing and Materials (ASTM) Standard

reference from the Administrative

Controls Section of TS to a licenseecontrolled document, i.e., the Diesel

Fuel Oil Program in the Technical

Requirements Manual (TRM). In

addition, the ‘‘clear and bright’’ test

used to establish the acceptability of

new fuel oil for use prior to addition to

storage tanks has been expanded to

allow a water and sediment content test

to be performed to establish the

acceptability of new fuel oil in lieu of

the ‘‘clear and bright’’ test.

Date of issuance: June 13, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 30 days.

Amendment Nos.: 122, 122, 116, and

116.

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: March 21, 2001.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated June 13, 2001.

No significant hazards consideration

comments received: No.

Exelon Generation Company, Docket

Nos. 50–352 and 50–353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of application for amendments:

July 31, 2000.

Brief description of amendments:

Revised Technical Specification (TS)

Surveillance Requirement (SR) 4.5.1.d.1,

concerning the operability of the

Automatic Depressurization System,

and relocated the existing requirements

Federal Register / Vol. 66, No. 124 / Wednesday, June 27, 2001 / Notices 34291

in TS SR 4.5.1.d.1 and TS SR 4.5.1.d.2.c

to the Technical Requirements Manual.

Date of issuance: June 12, 2001.

Effective date: As of date of issuance

and shall be implemented within 30

days.

Amendment Nos.: 152 and 116.

Facility Operating License Nos. NPF–

39 and NPF–85. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: October 18, 2000 (65 FR

62389).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated June 12, 2001.

No significant hazards consideration

comments received: No.

FirstEnergy Nuclear Operating

Company, et al., Docket No. 50–334,

Beaver Valley Power Station, Unit No. 1,

Beaver County, Pennsylvania

Date of application for amendment:

November 8, 2000, as supplemented on

February 6, and May 7, 2001.

Brief description of amendment: The

amendment changed the technical

specifications associated with the

deletion of TS 3/4.4.1.6, ‘‘Reactor

Coolant Pump—Startup.’’

Date of issuance: June 13, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 60 days.

Amendment No: 238.

Facility Operating License No. DPR–

66: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: December 27, 2000 (65 FR

81917).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated June 13, 2001.

No significant hazards consideration

comments received: No.

Indiana Michigan Power Company,

Docket No. 50–316, Donald C. Cook

Nuclear Plant, Unit 2, Berrien County,

Michigan

Date of application for amendment:

January 19, 2001, as supplemented

April 20 and May 9, 2001.

Brief description of amendment: The

amendment would change the TSs to

extend surveillance intervals associated

with the emergency diesel generator

(EDG) engines and station batteries that

are currently required to be completed

beginning June 27, 2001. The license

amendment would allow these

requirements to be performed during the

next refueling outage, but no later than

December 31, 2001. This would

preclude the need for a mid-cycle

shutdown of the Unit.

Date of issuance: June 11, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 30 days.

Amendment No.: 234.

Facility Operating License No. DPR–

74: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: March 21, 2001 (66 FR 15926).

The April 20 and May 9, 2001,

supplemental letters, did not change the

scope of the proposed action and did

not change the Nuclear Regulatory

Commission’s (NRC’s) preliminary no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated June 11, 2001.

No significant hazards consideration

comments received: No.

Nuclear Management Company, LLC,

Docket No. 50–305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of application for amendment:

January 13, 2000, as supplemented

March 7, March 30, and May 4, 2001.

Brief description of amendment: The

amendment revises the Kewaunee

Nuclear Power Plant (KNPP) Technical

Specifications (TSs) 3.6, ‘‘Containment’’

to add Limiting Condition for Operation

(LCO) and Allowed Outage Times

(AOT) for containment isolation

devices. In addition, the amendment

provides additional information,

clarification, and uniformity to the bases

of the associated TSs.

Date of issuance: June 8, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 30 days.

Amendment No.: 155.

Facility Operating License No. DPR–

43: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 21, 2001 (66 FR

11061). The March 7, March 30, and

May 4, 2001, letters, provided clarifying

information that was within the scope of

the original application, did not change

the NRC staff’s initial proposed no

significant hazards consideration

determination, and did not expand the

amendment beyond the scope of the

original notice (66 FR 11061).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated June 8, 2001.

No significant hazards consideration

comments received: No.

Nuclear Management Company, LLC,

Docket No. 50–255, Palisades Plant, Van

Buren County, Michigan

Date of application for amendment:

January 26, 2001, as supplemented by

letter dated March 13, 2001.

Brief description of amendment: The

amendment changes Technical

Specification Surveillance Requirement

3.7.9.2, ‘‘Ultimate Heat Sink (UHS),’’ by

increasing the maximum allowable

temperature of Lake Michigan water

from 81.5 °F to 85 °F.

Date of issuance: June 4, 2001.

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment No.: 202.

Facility Operating License No. DPR–

20. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 7, 2001 (66 FR 13800).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated June 4, 2001.

No significant hazards consideration

comments received: No.

Omaha Public Power District, Docket

No. 50–285, Fort Calhoun Station, Unit

No. 1, Washington County, Nebraska

Date of amendment request: August 3,

2000, as supplemented by letters dated

November 17, 2000, and February 14,

2001.

Brief description of amendment: The

amendment deletes Section 3.D,

‘‘License Term,’’ from the Fort Calhoun

Station, Unit No. 1 operating license.

Date of issuance: June 6, 2001.

Effective date: June 6, 2001, to be

implemented within 30 days from the

date of issuance.

Amendment No.: 199.

Facility Operating License No. DPR–

40: The amendment revised the

operating license.

Date of initial notice in Federal

Register: January 10, 2001 (66 FR 2019).

The November 17, 2000, and February

14, 2001, supplemental letters 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 determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated June 6, 2001.

No significant hazards consideration

comments received: No.

34292 Federal Register / Vol. 66, No. 124 / Wednesday, June 27, 2001 / Notices

Southern California EdisonCompany, et

al., Docket Nos. 50–361 and 50–362,

San Onofre Nuclear Generating Station,

Units 2 and 3, San Diego County,

California

Date of application for amendments:

April 6, 2001 and supplemented by

letter dated April 20, 2001.

Brief description of amendments: The

amendments proposed to revise the San

Onofre Nuclear Generating Station,

Units 2 and 3 Technical Specification

Surveillance Requirements 3.8.1.2,

3.8.1.3, 3.8.1.9, 3.8.1.10, and 3.8.1.19 to

assure that an emergency diesel

generator automatic voltage regulator

(AVR) is operable and regularly tested.

AVR operability would be demonstrated

by conducting SR 3.8.1.2 and 3.8.1.3

within the past 60 days, and any one of

SR 3.8.1.9, 3.8.1.10, or 3.8.1.19 within

the past 24 months.

Date of issuance: June 8, 2001.

Effective date: June 8, 2001, to be

implemented within 30 days of

issuance.

Amendment Nos.: Unit 2–179; Unit

3–170.

Facility Operating License Nos. NPF–

10 and NPF–15: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: May 2, 2001 (66 FR 22032).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated June 8, 2001.

No significant hazards consideration

comments received: No.

Southern Nuclear Operating Company,

Inc., Docket Nos. 50–348 and 50–364,

Joseph M. Farley Nuclear Plant, Units 1

and 2, Houston County, Alabama

Date of amendments request: August

17, 2000, as supplemented by letter

dated April 2, 2001. The April 2, 2001,

letter requested a new implementation

date, but did not change the August 17,

2000, application and the initial

proposed no significant hazards

consideration determination.

Brief description of amendments: The

amendments eliminate the need for the

licensee to perform periodic response

time testing of selected reactor trip

system and engineered safety feature

actuation system equipment as defined

in Westinghouse report WCAP–14036–

P–A, Revision 1, ‘‘Elimination of

Periodic Protection Channel Response

Time Tests.’’

Date of issuance: June 7, 2001.

Effective date: As of the date of

issuance and shall be implemented on

Unit 1 entry in Mode 3 for Cycle 18

following the 2001 fall refueling.

Amendment Nos.: 149 and 141.

Facility Operating License Nos. NPF–

2 and NPF–8: Amendments revise the

Technical Specifications.

Date of initial notice in Federal

Register: January 10, 2001 (66 FR 2023).

The supplement dated April 2, 2001,

provided clarifying information that did

not change the scope of the August 17,

2001, 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 June 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:

January 11, 2001.

Brief description of amendments: The

amendments revise TS 5.5.17,

‘‘Containment Leakage Rate Testing

Program,’’ to add an exception to

Regulatory Guide 1.163 related to visual

examination of containment concrete

surfaces.

Date of issuance: June 6, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment Nos.: 122 and 100.

Facility Operating License Nos. NPF–

68 and NPF–81: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: May 2, 2001 (66 FR 22033).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated June 6, 2001.

No significant hazards consideration

comments received: No.

Tennessee Valley Authority, Docket

Nos. 50–259, 50–260, and 50–296,

Browns Ferry Nuclear Plant , Units 1, 2,

and 3, Limestone County, Alabama

Date of application for amendments:

November 6, 2000.

Description of amendment request:

These amendments revised the

Technical Specifications (TS) to allow

four residual heat removal suppression

pool cooling subsystems to be

inoperable for 8 hours.

Date of issuance: June 8, 2001.

Effective date: June 8, 2001.

Amendment Nos.: 241, 272, and 230.

Facility Operating License Nos. DPR–

33, DPR–52, and DPR–68. Amendments

revised the TS.

Date of initial notice in Federal

Register: November 29, 2000 (65 FR

71139).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated June 8, 2001.

No significant hazards consideration

comments received: No.

TXU Electric, Docket Nos. 50–445 and

50–446, Comanche Peak Steam Electric

Station, Unit Nos. 1 and 2, Somervell

County, Texas

Date of amendment request: April 3,

2001.

Brief description of amendments: The

amendments revise Technical

Specification (TS) 3.3.6, ‘‘Containment

Ventilation Isolation Instrumentation,’’

to modify the Note for Required Action

B.1 such that it applies only to * * *

Required Action and associated

Completion Time of Condition A not

met * * * This change is the result of

the discovery of an error which

occurred when the TSs were converted

to the improved TS with issuance of

License Amendment Nos. 64 and 64, for

Comanche Peak Steam Electric Station,

Units 1 and 2, on February 26, 1999.

Date of issuance: June 4, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 60 days from the date of

issuance.

Amendment Nos.: 86 and 86.

Facility Operating License Nos. NPF–

87 and NPF–89: The amendments revise

the Technical Specifications.

Date of initial notice in Federal

Register: May 2, 2001 (66 FR 22034).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated June 4, 2001.

No significant hazards consideration

comments received: No.

Virginia Electric and Power Company,

Docket Nos. 50–338 and 50–339, North

Anna Power Station, Units 1 and 2,

Louisa County, Virginia

Date of application for amendment:

September 27, 2000, as supplemented

November 21 and December 18, 2000,

and February 2, March 2, and May 21,

2001.

Brief description of amendment:

These amendments add Technical

Specification (TS) 3.7.14, TS 4.7.14, TS

3.7.15, TS 4.7.15, Figure 3.7.15–1, and

Figure 3.7.15–2; and revise TS 5.3.1 and

TS 5.6.1.1. The purpose of these

amendments is to increase the limit on

the fuel enrichment from the current

limit of 4.3 weight percent U235 to a

maximum of 4.6 weight percent U235,

establish TS Limiting Conditions for

Operations for the Spent Fuel Pool

(SFP) boron concentration and fuel

storage restrictions, and eliminate the

value of uncertainties in the calculation

for Keff in the SFP criticality calculation.

Federal Register / Vol. 66, No. 124 / Wednesday, June 27, 2001 / Notices 34293

Date of issuance: June 15, 2001.

Effective date: As of the date of

issuance and shall be implemented by

December 21, 2001.

Amendment Nos.: 227 and 208.

Facility Operating License Nos. NPF–

4 and NPF–7: Amendments change the

Technical Specifications.

Date of initial notice in Federal

Register: December 13, 2000 (65 FR

77929). The December 18, 2000,

February 2, March 2, and May 21, 2001,

supplements contained clarifying

information only, and did not change

the initial no significant hazards

consideration determination, or expand

the scope of the initial application.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated June 15, 2001.

No significant hazards consideration

comments received: No.

Dated at Rockville, Maryland, this 19th day

of June 2001.

For the Nuclear Regulatory Commission.

John A. Zwolinski,

Director, Division of Licensing Project

Management, Office of Nuclear Reactor

Regulation.

[FR Doc. 01–15818 Filed 6–26–01; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Preliminary Impact Assessment of

Nuclear Industry Consolidation onNRC

Oversight: Request for Comments

AGENCY: Nuclear Regulatory

Commission (NRC).

ACTION: Request for comments.

SUMMARY: Economic deregulation of the

electric utility industry has resulted in

consolidation and restructuring of the

nuclear power industry. The

transformation of the once strictly

regulated industry has led to separation

of the generation, transmission and

distribution sectors, corporate mergers

and asset transfers, acquisitions by

outright purchase, and a general

transition to a nationwide competitive

market. There have also been numerous

nuclear power plant license transfer

applications, which the NRC staff must

review and approve before a license can

be transferred to a new entity.

The NRC staff has identified and

performed a preliminary assessment of

the impacts of nuclear industry

consolidation on the NRC and whether

the NRC needs to change its regulations,

policies, processes, guidance, or

organizational structure to continue to

meet its strategic public health and

safety goals. The initial object of this

effort is to identify impacts that need to

be considered further.

The NRC staff has identified a number

of consolidation and a few deregulationrelated impacts on NRC oversight of the

nuclear industry, grouped them by

category, and performed preliminary

impact assessments. The individual

assessments follow this notice.

The NRC staff requests comments and

suggestions from stakeholders on the

identified issues and the preliminary

impact assessments. The NRC staff will

consider all comments received. A

public workshop will be held at NRC

Headquarters in the October/November

2001 timeframe to discuss the regulatory

oversight issues attendant to industry

consolidation, the staff’s preliminary

impact assessments, and the comments

received from the stakeholders. Notice

of this workshop will be published at a

later date. Commenters should indicate

their interest in attending and

participating in this workshop.

The product of this effort will be staff

recommendations of impacts that the

Commission needs to consider further.

DATES: The comment period ends

August 27, 2001. Comments received

after this date will be considered if it is

practical to do so, but the staff

guarantees consideration only of

comments received on or before this

date.

ADDRESSES: Mail written comments to

Chief, Rules and Directives Branch,

Division of Administrative Services,

Office of Administration, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555–0001. Comments may also be

sent by completing the online comment

form at http://www.nrc.gov/NRC/

REACTOR/CONSOLIMPACT/

index.html.

Deliver comments to Room 6D59,

Two White Flint North, 11545 Rockville

Pike, Rockville, Maryland, between 7:30

a.m. and 4:15 p.m. on Federal workdays.

For further information contact

Herbert N. Berkow, Mail Stop O 8 H–12,

Office of Nuclear Reactor Regulation,

U.S. Nuclear Regulatory Commission,

Washington, DC 20555; telephone (301)

415–1485 and e-mail at

HNB@NRC.GOV.

Dated at Rockville, Maryland, this 20th day

of June 2001.

For the Nuclear Regulatory Commission.

Herbert N. Berkow,

Director, Project Directorate II, Division of

Licensing Project Management, Office of

Nuclear Reactor Regulation.

Industry Consolidation Preliminary

Impact Assessments

Categorization of Industry Consolidation

Issues

Category 1 Plant Operational Safety

Issue 1.a Possible Cost-cutting Initiatives

Issue 1.b Technology-related Issues

Issue 1.c Spent Fuel Storage and

Transportation

Issue 1.d Low-Level Radioactive Waste

Management

Issue 1.e Emergency Preparedness

Issue 1.f Reliable Off-site Power

Category 2 Licensing

Issue 2.a License Transfer Process

Issue 2.b New License Applications, Site

Approvals, and Reactivations of Deferred

Plants

Issue 2.c License Renewal

Issue 2.d NRC Organizational Structure

Category 3 Inspection, Enforcement, and

Assessment

Issue 3.a NRC Reactor Oversight Process

Issue 3.b Other NRC Inspection Programs

Issue 3.c NRC Enforcement Program

Issue 3.d NRC Allegation Program

Category 4 Decommissioning

Category 5 External Regulatory Interfaces

Category 6 Fuel Cycle Facilities

Category 7 Financial

Issue 7.a Foreign Ownership

Issue 7.b License Fee Structure

Issue 7.c Insurance

Issue 7.d Joint and Several Regulatory

Responsibility

Issue 7.e Bankruptcy Protection

Issue 7.f Financial Qualifications

Category 8 Non-NRC Regulatory

Considerations

Issue 8.a Grid Stability/Reliability

Issue 8.b Antitrust Considerations

Issue Category: 1. Plant Operational

Safety

Issue: 1.a Possible Cost-Cutting

Initiatives

Discussion

In a more consolidated, economically

deregulated market, the nuclear power

industry will be faced with new

pressures to operate more efficiently.

Cost controls could result in shorter

outages (and thus longer run times),

increased use of on-line maintenance,

power uprate amendments, increased

use of risk-informed technology and

decisions and other changes that would

result in lower costs and increased

productivity.