Federal Register / Vol. 66, No. 247 / Wednesday, December 26, 2001 / Notices 66461

collection request; Abstract: Instructors

will be asked to complete a

questionnaire before the course begins.

In most cases, it will be specified in the

instructor’s contract that they complete

the questionnaire. This survey

instrument contains six questions, most

requiring only a simple fill-in-the blank

response. Data and information from

this questionnaire will establish a

baseline of the instructor’s expectations

and intentions to be used in measuring

changes at the end of the course.

Affected Entities: Entities potentially

affected by this action are instructors

who lead training sessions sponsored by

the U.S. Institute. Burden Statement: It

is estimated that the annual national

public burden and associated costs will

be approximately 33.3 hours and

$1,200, respectively. These values were

calculated assuming that on average: (a)

instructors require 10 minutes to

complete the questionnaire; and (b) each

year there are 200 instructors who work

on training sessions sponsored by the

U.S. Institute. Cost burden estimates

assume: (a) there are no capital or startup costs for respondents, and (b)

respondents’ time is valued at $36 hr.

(17) Training—Instructor

Questionnaire, at the conclusion; New

collection request; Abstract: When the

course concludes, instructors will be

asked to complete a questionnaire. In

most cases, it will be specified in their

contract that they complete this

questionnaire. The survey instrument

contains five questions, most requiring

only a simple fill-in-the blank response.

Data and information from this

questionnaire will help establish a

contextual baseline for evaluating

survey data from the training

participants. As well, this instrument is

also intended to generate useful

feedback on ways to improve the U.S.

Institute’s training projects. Affected

Entities: Entities potentially affected by

this action are instructors who lead

training sessions sponsored by the U.S.

Institute. Burden Statement: It is

estimated that the annual national

public burden and associated costs will

be approximately 33.3 hours and

$1,200, respectively. These values were

calculated assuming that on average: (a)

instructors require 10 minutes to

complete the questionnaire; and (b) each

year there are 200 instructors who work

on training sessions sponsored by the

U.S. Institute. Cost burden estimates

assume: (a) there are no capital or startup costs for respondents, and (b)

respondents’ time is valued at $36 hr.

Meeting Facilitation

U.S. Institute staff and contractors

facilitate and provide leadership for

many public meetings, ranging from

small group meetings to large public

convenings of several hundred

attendees. In order to maximize the

probability that such meeting objectives

will be accomplished, the meeting

participants must both understand the

objectives for the meeting, and perceive

that the meeting was managed in a fair

and efficient manner. This requires that

the right facilitator run the meeting, and

the right people attend the meeting.

(18) Meeting Facilitation—Meeting

Attendees Questionnaire, at the

conclusion of the process; New

collection request; Abstract: Attendees

at public meetings run by U.S. Institute

staff or contractors will be asked to

complete a voluntary questionnaire at

the conclusion of the meeting. The

questionnaire used in this case contains

nine questions, two-thirds requiring

only a simple fill-in-the blank response.

Information from this questionnaire will

help evaluate the effectiveness of

individual facilitators and particular

meeting process designs. Affected

Entities: Entities potentially affected by

this action are individuals who

participate in these public meetings.

Burden Statement: It is estimated that

the annual national public burden and

associated costs will be approximately

833.3 hours and $22,500, respectively.

These values were calculated assuming

that on average: (a) meeting attendees

require 10 minutes to complete the

questionnaire; (b) the U.S. Institute

conducts 100 public meetings each year;

and (c) 50 people attend the average

meeting. Cost burden estimates assume:

(a) there are no capital or start-up costs

for respondents; and (b) respondents’

time is valued at $27 hr.

Dated: December 18, 2001.

Christopher L. Helms,

Executive Director, Morris K. Udall

Foundation.

[FR Doc. 01–31587 Filed 12–21–01; 8:45 am]

BILLING CODE 6820–FN–P

NATIONAL SCIENCE FOUNDATION

DOE/NSF Nuclear Science Advisory

Committee; Notice of Meeting

In accordance with the Federal

Advisory Committee Act (Pub. L. 92–

463, as amended), the National Science

Foundation announces the following

meeting.

Name: DOE/NSF Nuclear Science Advisory

Committee (1176).

Date and Time: Monday, Jan. 14, 2002 8

a.m.–6 p.m.; Tuesday, Jan. 15, 2002; 8 a.m.–

6 p.m.

Place: Rm 585–II 4201 Wilson Blvd.,

Arlington, VA 22230.

Contact Person: Dr. Bradley D. Keister,

Program Director for Nuclear Physics,

National Science Foundation, 4201 Wilson

Blvd., Arlington, VA 22230. Telephone (703)

292–7380.

Purpose of Meeting: To provide advice and

recommendations concerning the scientific

programs of the NSF and DOE in the area of

basic nuclear physics research.

Dated: December 19, 2001.

Susanne Bolton,

Committee Management Officer.

[FR Doc. 01–31640 Filed 12–21–01; 8:45 am]

BILLING CODE 7555–01–M

NUCLEAR REGULATORY

COMMISSION

Biweekly Notice; Applications and

Amendments to Facility Operating

Licenses Involving No Significant

Hazards Considerations

(Note: The publication date for this notice

will change from every other Wednesday to

every other Tuesday, effective January 8,

2002. The notice will contain the same

information and will continue to be

published biweekly.)

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 December 3,

2001 through December 14, 2001. The

last biweekly notice was published on

December 12, 2001 (66 FR 64284).

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

66462 Federal Register / Vol. 66, No. 247 / Wednesday, December 26, 2001 / Notices

of the facility in accordance with the

proposed amendment would not (1)

involve a significant increase in the

probability or consequences of an

accident previously evaluated; or (2)

create the possibility of a new or

different kind of accident from any

accident previously evaluated; or (3)

involve a significant reduction in a

margin of safety. The basis for this

proposed determination for each

amendment request is shown below.

The Commission is seeking public

comments on this proposed

determination. Any comments received

within 30 days after the date of

publication of this notice will be

considered in making any final

determination.

Normally, the Commission will not

issue the amendment until the

expiration of the 30-day notice period.

However, should circumstances change

during the notice period such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received before

action is taken. Should the Commission

take this action, it will publish in the

Federal Register a notice of issuance

and provide for opportunity for a

hearing after issuance. The Commission

expects that the need to take this action

will occur very infrequently.

Written comments may be submitted

by mail to the Chief, Rules and

Directives Branch, Division of

Administrative Services, Office of

Administration, U.S. Nuclear Regulatory

Commission, Washington, DC 20555–

0001, and should cite the publication

date and page number of this Federal

Register notice. Written comments may

also be delivered to Room 6D22, Two

White Flint North, 11545 Rockville

Pike, Rockville, Maryland from 7:30

a.m. to 4:15 p.m. Federal workdays.

Copies of written comments received

may be examined at the NRC Public

Document Room, located at One White

Flint North, 11555 Rockville Pike (first

floor), Rockville, Maryland. The filing of

requests for a hearing and petitions for

leave to intervene is discussed below.

By January 25, 2002, 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 NRC’s Public

Document Room, located at One White

Flint North, 11555 Rockville Pike (first

floor), Rockville, Maryland 20852.

Publicly available records will be

accessible electronically from the

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 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,

Federal Register / Vol. 66, No. 247 / Wednesday, December 26, 2001 / Notices 66463

Washington, DC 20555–0001, and to the

attorney for the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

factors specified in 10 CFR

2.714(a)(1)(i)–(v) and 2.714(d).

For further details with respect to this

action, see the application for

amendment which is available for

public inspection at the Commission’s

Public Document Room, located at One

White Flint North, 11555 Rockville Pike

(first floor), Rockville, Maryland.

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.

AmerGen Energy Company, LLC, Docket

No. 50–461, Clinton Power Station, Unit

1, DeWitt County, Illinois

Date of amendment request: April 17,

2001.

Description of amendment request:

The proposed amendment would make

editorial and administrative corrections

to Technical Specifications (TS) Section

3.3, ‘‘Instrumentation’’, and eliminate

minor discrepancies between TS

Section 3.3 and other plant licensing

basis documents.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration which is presented below:

Does the Change Involve a Significant

Increase in the Probability or Consequences

of an Accident Previously Evaluated?

The proposed changes involve correction

of editorial or administrative errors made

during the conversion of the Clinton Power

Station (CPS) Technical Specifications (TS)

to the improved TS (ITS). These proposed

changes are based upon current design and

licensing basis requirements. The proposed

changes involve correction or reformatting of

the TS and do not involve any physical

changes to plant systems, including those

that mitigate the consequences of accidents

or the manner in which these plant systems

are operated. As such, these changes do not

involve a significant increase in the

probability or consequences of any accident

previously evaluated.

Does the Change Create the Possibility of a

New or Different Kind of Accident From Any

Accident Previously Evaluated?

The proposed changes involve correcting

errors or reformatting existing TS

requirements that do not involve a physical

alteration of the plant (i.e., no new or

different type of equipment will be installed)

or a change in the methods governing normal

plant operation. These changes are consistent

with the assumptions in the safety analyses

and licensing basis. Thus, these changes do

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

Does the Change Involve a Significant

Reduction in a Margin of Safety?

The proposed changes involve correcting

editorial or administrative errors introduced

during the conversion of the CPS TS to the

ITS. The change to the Allowable Value for

the Control Room Ventilation System air

intake radiation monitors setpoint in TS

Table 3.3.7.1–1 is consistent with the

supporting analyses for the trip setpoint

value that was previously contained in the

TS. The changes involve reformatting or

correction of errors, and therefore will not

reduce any margin of safety because there is

no effect on any safety analysis assumptions.

These proposed changes maintain

requirements within the safety analyses and

licensing basis. Therefore, these 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: Robert Helfrich,

Mid-West Regional Operating Group,

Exelon Generation Company, LLC, 4300

Winfield Road, Warrenville, IL 60555.

NRC Section Chief: Anthony J.

Mendiola.

AmerGen Energy Company, LLC, Docket

No. 50–461, Clinton Power Station, Unit

1, DeWitt County, Illinois

Date of amendment request: May 21,

2001.

Description of amendment request:

The proposed amendment would revise

the actions required if the refueling

equipment interlocks become

inoperable. The proposed changes are

consistent with the changes submitted

to the Nuclear Regulatory Commission

by the Technical Specifications Task

Force, Issue number 225, Revision 1.

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:

Does the Change Involve a Significant

Increase in the Probability or Consequences

of an Accident Previously Evaluated?

The proposed addition of alternate actions

in the event that the refueling equipment

interlocks are determined to be in operable

ensures that the safety function provided by

the interlocks are enforced. This is

accomplished through manually inserting a

rod block to prevent the inadvertent

withdrawal of a control rod when fuel is

being moved over the core region.

The refueling equipment interlocks are

credited in the Control Rod Removal Error

During Refueling—Fuel Insertion with

Control Rod Withdrawn as described in

Updated Safety Analysis (USAR Section

15.4.1.1.2.2). The manual insertion of a

control rod withdrawal block provides

equivalent protection for the conditional rod

block provided by the refueling equipment

interlocks.

The proposed change to the surveillance

frequency does not change the means in

which the refueling equipment operates. A

review of surveillance history was performed

for the past two refueling outages. In the last

seven performances of the refueling

equipment interlocks operability test, the

interlocks have operated successfully with no

corrective maintenance or corrective action

necessary. Therefore, since the proposed

changes do not result in any physical

changes to the facility, or involve any

modifications to plant systems or design

parameters or conditions that contribute to

the initiation of any accidents previously

evaluated, the proposed changes do not

increase the probability of any accident

previously evaluated.

Since the proposed changes maintain the

same level of protection provided by the

refueling equipment interlocks, the

conclusion of the accident scenario remain

valid. The probability of a criticality event

during refueling remains such that no

radioactive material would be released.

Therefore, the proposed changes do not

increase the consequences of an accident

previously evaluated.

In summary, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Does the Change Create the Possibility of a

New or Different Kind of Accident From Any

Accident Previously Evaluated?

The proposed changes do not involve a

change to the plant design or operation.

Inserting a manual rod block is not

considered an abnormal operation. The

change to the SR [surveillance requirement]

frequency does not increase the probability of

a malfunction of the refueling equipment

interlocks, since the interlocks are

considered reliable and their function can be

verified with each fuel move. As a result, the

proposed changes do not affect any of the

parameters or conditions that could

contribute to the initiation of any accidents.

No new accident modes or equipment failure

modes are created by these changes.

66464 Federal Register / Vol. 66, No. 247 / Wednesday, December 26, 2001 / Notices

Therefore, these proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

Does the Change Involve a Significant

Reduction in a Margin of Safety?

The major challenge to the margin of safety

would be a criticality event that would cause

a potential failure of the fuel cladding. The

proposed addition of alternative actions in

the event that the refueling equipment

interlocks are determined to be inoperable

ensure that equivalent protection is in place

during fuel loading movements. Given this

equivalent protection, a criticality event is

not credible. In addition, the increase in the

SR frequency for performing the channel

functional test of the refueling equipment

interlocks does not impact the ability of the

interlocks to perform their function, thereby

maintaining the refueling interlocks function.

Therefore, the proposed changes do not

involve a significant reduction in the margin

of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Robert Helfrich,

Mid-West Regional Operating Group,

Exelon Generation Company, LLC, 4300

Windfield Road, Warrenville, IL 60555.

NRC Section Chief: Anthony J.

Mendiola.

Duke Energy Corporation, Docket Nos.

50–369 and 50–370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg

County, North Carolina

Date of amendment request:

December 7, 2001.

Description of amendment request:

The proposed amendments would allow

implementation of 10 CFR Part 50,

Appendix J, Option B, which governs

performance-based containment leakage

testing requirements, for Type B and C

testing. In addition, the licensee also

proposes to (a) modify Technical

Specification (TS) 3.6.3 to delete the

requirement for conducting soap bubble

tests of welded penetrations during

Type A tests which are not individually

Type B or Type C testable, and (b) to

modify TS 3.6.3 to delete a separate

requirement for leak testing

containment purge lower and upper

compartment and instrument room

valves with resilient seals. These valves

will be covered by the overall

Containment Leakage Rate Testing

Program. Associated changes to the

Bases are also proposed.

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. Implementation of these changes

will provide continued assurance that

specified parameters associated with

containment integrity will remain within

acceptance limits as delineated in 10 CFR 50,

Appendix J, Option B. The changes are

consistent with current safety analyses.

Although some of the proposed changes

represent minor relaxation to existing TS

requirements, they are consistent with the

requirements specified by Option B of 10

CFR 50, Appendix J. The systems affecting

containment integrity related to this

proposed amendment request are not

assumed in any safety analyses to initiate any

accident sequence. Therefore, the probability

of any accident previously evaluated is not

increased by this proposed amendment. The

proposed changes maintain an equivalent

level of reliability and availability for all

affected systems. In addition, maintaining

leakage within analyzed limits assumed in

accident analyses does not adversely affect

either onsite or offsite dose consequences.

Therefore, the proposed amendment does not

increase the consequences of any accident

previously evaluated.

Second Standard

The proposed amendment will not create

the possibility of a new or different kind of

accident from any accident previously

evaluated. No changes are being proposed

which will introduce any physical changes to

the existing plant design. The proposed

changes are consistent with the current safety

analyses. Some of the changes may involve

revision in the testing of components;

however, these are in accordance with the

McGuire’s current safety analyses and

provide for appropriate testing or

surveillance that is consistent with 10 CFR

50, Appendix J, Option B. The proposed

changes will not introduce new failure

mechanisms beyond those already

considered in the current safety analyses. No

new modes of operation are introduced by

the proposed changes. The proposed changes

maintain, at minimum, the present level of

operability of any system that affects

containment integrity.

Third Standard

The proposed amendment will not involve

a significant reduction in a margin of safety.

The provisions specified in Option B of 10

CFR 50, Appendix J allow changes to Type

B and Type C test intervals based upon the

performance of past leak rate tests. 10 CFR

50, Appendix J, Option B allows longer

intervals between leakage tests based on

performance trends, but does not relax the

leakage acceptance criteria. Changing test

intervals from those currently provided in

the TS to those provided in 10 CFR 50,

Appendix J, Option B does not increase any

risks above and beyond those that the NRC

has deemed acceptable for the performance

based option. In addition, there are risk

reduction benefits associated with reduction

in component cycling, stress, and wear

associated with increased test intervals. The

proposed changes provide continued

assurance of leakage integrity of containment

without adversely affecting the public health

and safety and will not significantly reduce

existing safety margins. Similar proposed

changes have been previously reviewed and

approved by the NRC, and they are

applicable to McGuire.

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: Mr. Albert Carr,

Duke Energy Corporation, 422 South

Church Street, Charlotte, North

Carolina.

NRC Section Chief: Richard J. Laufer,

Acting.

Entergy Operations, Inc., System Energy

Resources, Inc., South Mississippi

Electric Power Association, and Entergy

Mississippi, Inc., Docket No. 50–416,

Grand Gulf Nuclear Station, Unit 1,

Claiborne County, Mississippi

Date of amendment request:

November 15, 2001.

Description of amendment request:

Entergy Operations, Inc. is proposing

that the Grand Gulf Nuclear Station

(GGNS) Operating License be amended

to revise the GGNS Technical

Specification Surveillance

Requirements (SRs) pertaining to testing

of the standby emergency diesel

generators (DGs) to allow DG testing

during reactor operation. The proposed

change would remove the restriction

associated with these SRs that prohibits

conducting the required testing of the

DGs during reactor operating Modes 1,

2, or 3.

Basis for proposed no significant

hazards consideration determination:

Federal Register / Vol. 66, No. 247 / Wednesday, December 26, 2001 / Notices 66465

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 any accident previously

evaluated.

The DGs and their associated emergency

loads are accident mitigating features, not

accident initiating equipment. Therefore,

there will be no impact on any accident

probabilities by the approval of the requested

amendment.

The design of plant equipment is not being

modified by these proposed changes. As

such, the ability of the DGs to respond to a

design basis accident will not be adversely

impacted by these proposed changes. The

capability of the DGs to supply power in a

timely manner will not be compromised by

permitting performance of DG testing during

periods of power operation. Additionally,

limiting testing to only one DG at a time

ensures that design basis requirements for

backup power is met, should a fault occur on

the tested DG. Therefore, there would be no

significant impact on any accident

consequences.

Based on the above, the proposed change

to permit certain DG surveillance tests to be

performed during plant operation will have

no effect on accident probabilities or

consequences.

(2) The proposed change does not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

No new accident causal mechanisms

would be created as a result of NRC [Nuclear

Regulatory Commission] approval of this

amendment request since no changes are

being made to the plant that would introduce

any new accident causal mechanisms.

Equipment will be operated in the same

configuration with the exception of the plant

mode in which the testing is conducted. This

amendment request does not impact any

plant systems that are accident initiators;

neither does it adversely impact any accident

mitigating systems.

Based on the above, implementation of the

proposed changes would 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 the margin of

safety.

Margin of safety is related to the

confidence in the ability of the fission

product barriers to perform their design

functions during and following an accident

situation. These barriers include the fuel

cladding, the reactor coolant system, and the

containment system. The proposed changes

to the testing requirements for the plant DGs

do not affect the operability requirements for

the DGs, as verification of such operability

will continue to be performed as required

(except during different allowed Modes).

Continued verification of operability

supports the capability of the DGs to perform

their required function of providing

emergency power to plant equipment that

supports or constitutes the fission product

barriers. Consequently, the performance of

these fission product barriers will not be

impacted by implementation of this proposed

amendment.

In addition, the proposed changes involve

no changes to setpoints or limits established

or assumed by the accident analysis. On this

and the above basis, no safety margins will

be impacted. Therefore, implementation of

the proposed changes would not involve a

significant reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, NW., 12th Floor,

Washington, DC 20005–3502.

NRC Section Chief: Robert A. Gramm.

FirstEnergy Nuclear Operating

Company, et al., Docket Nos. 50–334

and 50–412, Beaver Valley Power

Station, Unit Nos. 1 and 2, Beaver

County, Pennsylvania

Date of amendment request: October

29, 2001.

Description of amendment request:

The proposed amendments would

revise technical specification (TS) 3.9.3,

‘‘Refueling Operations—Decay Time,’’

by reducing the amount of time that the

reactor must be subcritical before the

licensee is allowed to move irradiated

fuel assemblies in the reactor pressure

vessel from 150 hours to 100 hours. The

amendment also makes various

editorial, format and administrative

changes.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

No. The proposed change does not alter the

manner in which fuel assemblies are handled

or core alterations are performed. The

proposed change does not alter the manner

in which heavy loads are controlled at BVPS.

The proposed change does not result in

changes being made to structures, systems, or

components (SSCs), or to event initiators or

precursors. Also, the proposed change does

not impact the design of plant systems such

that previously analyzed SSCs would now be

more likely to fail. The initiating conditions

and assumptions for accidents described in

the Updated Final Safety Analysis Report

(UFSAR) remain as previously analyzed.

Thus, the proposed change does not involve

a significant increase in the probability of an

accident previously evaluated.

The proposed revision of the decay time

from 150 hours to 100 hours is consistent

with the assumptions used in the NRC

approved fuel handling accident (FHA)

analyses for Beaver Valley Power Station

(BVPS) Unit Nos. 1 and 2. The BVPS

radiological analyses demonstrates that

should a FHA occur within the containment

or the fuel building that involves irradiated

fuel with at least 100 hours of decay, the

projected offsite doses for this event will be

well within the applicable regulatory limits.

Limiting Condition for Operation (LCO)

3.9.3, ‘‘Refueling Operations—Decay Time,’’

will continue to ensure that irradiated fuel is

not moved in the reactor pressure vessel until

at least 100 hours after shutdown which is

consistent with the FHA radiological

analysis. This LCO will continue to ensure

that key assumptions used in the radiological

safety analysis are met. The previously

analyzed SSCs are unaffected by the

proposed change and continue to provide

assurance that they are capable of performing

their intended design function in mitigating

the effects of design basis accidents (DBAs).

As such, the consequences of accidents

previously evaluated in the UFSAR will not

be increased and no additional radiological

source terms are generated. Therefore, there

will be no reduction in the capability of those

SSCs in limiting the radiological

consequences of previously evaluated

accidents and reasonable assurance that there

is no undue risk to the health and safety of

the public will continue to be provided.

Thus, the proposed change does not involve

a significant increase in the consequences of

an accident previously evaluated.

The proposed administrative, editorial, and

format changes do not affect the probability

or consequences of any accident.

Therefore, the proposed amendment does

not significantly increase the probability or

consequences of any accident previously

evaluated.

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

No. The proposed amendment does not

affect a previously evaluated accident; e.g.,

FHA. The proposed amendment takes credit

for the normal decay of irradiated fuel and

the existing radiological analyses for FHAs.

The proposed change does not involve

physical changes to analyzed SSCs or

changes to the modes of plant operation

defined in the technical specification. The

proposed change does not involve the

addition or modification of plant equipment

(no new or different type of equipment will

be installed) nor does it alter the design or

operation of any plant systems. No new

accident scenarios, accident or transient

initiators or precursors, failure mechanisms,

or limiting single failures are introduced as

a result of the proposed change.

The proposed change does not cause the

malfunction of safety-related equipment

assumed to be operable in accident analyses.

No new or different mode of failure has been

created and no new or different equipment

performance requirements are imposed for

66466 Federal Register / Vol. 66, No. 247 / Wednesday, December 26, 2001 / Notices

accident mitigation. As such, the proposed

change has no effect on previously evaluated

accidents.

The proposed administrative, editorial, and

format changes do not create the possibility

of a new or different kind of accident from

any accident previously evaluated.

Therefore, the proposed amendment does

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Does the change involve a significant

reduction in a margin of safety?

No. The proposed revision of the decay

time from 150 hours to 100 hours is

consistent with the assumptions used in the

NRC approved FHA accident analyses for

BVPS Unit Nos. 1 and 2 and thus does not

involve a significant reduction in a margin of

safety.

The proposed amendment does not alter

the manner in which fuel assemblies are

handled or core alterations are performed.

The proposed amendment does not alter the

manner in which heavy loads are controlled

at BVPS.

The proposed changes to the TS

requirements will continue to ensure that the

necessary plant equipment is operable in the

plant conditions where these systems are

required to operate to mitigate a DBA. The

proposed administrative, editorial, and

format changes do not affect plant safety.

Therefore, the proposed amendment 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: Mary O’Reilly,

FirstEnergy Nuclear Operating

Company, FirstEnergy Corporation, 76

South Main Street, Akron, OH 44308.

NRC Section Chief: L. Raghavan,

Acting.

FirstEnergy Nuclear Operating

Company, et al., Docket Nos. 50–334

and 50–412, Beaver Valley Power

Station, Unit Nos. 1 and 2, Beaver

County, Pennsylvania

Date of amendment request: October

31, 2001.

Description of amendment request:

The proposed amendments would

revise the Technical Specifications

(TSs) by relocating the pressure

temperature Limit Curves and Low

Temperature Overpressure Protection

(LTOP) and by creating a PressureTemperature Limits Report in

accordance with Generic Letter 96–03

(GL–96–03), ‘‘Relocation of the Pressure

Temperature Limit Curves and Low

Temperature Overpressure Protection

System Limits.’’

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

No. The proposed changes are a relocation

of the Reactor Coolant System (RCS)

pressure/temperature (P/T) limits,

overpressure protection system (OPPS)

setpoint, and the enable temperature from the

Technical Specifications to the proposed

Pressure and Temperature Limits Report

(PTLR). The PTLR is created in accordance

with the guidance provided by Generic Letter

(GL) 96–03 and is consistent with the content

of NUREG–1431. The RCS P/T limits, OPPS

setpoint, and enable temperature will

continue to meet the requirements of 10 CFR

50, Appendix G, and will be generated in

accordance with the NRC approved

methodology described in WCAP–14040–

NP–A, Rev. 2 with the exceptions noted in

Technical Specification Section 6.9.6.

Since the proposed changes are

administrative in nature and do not involve

any change to any values being relocated, the

proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

No. As stated above, the proposed changes

to relocate the RCS P/T limits, OPPS

setpoint, and the enable temperature from the

Technical Specifications to the PTLR are

administrative changes. The proposed

changes do not result in a physical change to

the plant or add any new or different

operating requirements on plant systems,

structures, or components.

Therefore, the proposed changes do not

result in a significant increase in the

possibility of a new or different accident

from any previously evaluated.

3. Does the change involve a significant

reduction in a margin of safety?

No. The margin of safety is not affected by

the creation of the proposed PTLR. Operation

of the plant in accordance with the limits

specified in the PTLR will continue to meet

the requirements of 10 CFR 50, Appendix G,

with the identified exceptions, and will

assure that a margin of safety is not

significantly decreased as the result of the

proposed changes.

Therefore, the proposed changes do not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Mary O’Reilly,

FirstEnergy Nuclear Operating

Company, FirstEnergy Corporation, 76

South Main Street, Akron, OH 44308.

NRC Section Chief: Lakshminaras

Raghavan (Acting).

FirstEnergy Nuclear Operating

Company, Docket No. 50–346, DavisBesse Nuclear Power Station (DBNPS),

Unit 1, Ottawa County, Ohio

Date of amendment request: October

9, 2001.

Description of amendment request:

The proposed amendment changes

affected Technical Specifications (TS) 3/

4.3.2.2, ‘‘Instrumentation—Steam and

Feedwater Rupture Control System

Instrumentation,’’ including Table 3.3–

11, ‘‘Steam and Feedwater Rupture

Control System Instrumentation,’’ Table

3.3–12, ‘‘Steam and Feedwater Rupture

Control System Instrumentation Trip

Setpoints,’’ and Table 4.3–11 ‘‘Steam

and Feedwater Rupture Control System

Instrumentation Surveillance

Requirements.’’ Related administrative

changes are proposed to TS 3/4.3.2.3,

‘‘Instrumentation—Anticipatory Reactor

Trip System Instrumentation,’’ Table

3.3–17, ‘‘Anticipatory Reactor Trip

System Instrumentation,’’ and TS 3/

4.3.3.1, ‘‘Instrumentation—Monitoring

Instrumentation—Radiation Monitoring

Instrumentation,’’ Table 3.3–6,

‘‘Radiation Monitoring

Instrumentation.’’ Related changes to

associated TS Bases 3/4.3.1 and 3/4.3.2,

‘‘Reactor Protection System and Safety

System Instrumentation,’’ are also

proposed.

The main purpose for this license

amendment request is to decrease the

channel functional test frequency from

monthly to quarterly for the Steam and

Feedwater Rupture Control System

(SFRCS) Instrumentation Channels.

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. These changes would:

1a. Not involve a significant increase in the

probability of an accident previously

evaluated because the proposed changes do

not change any accident initiator, initiating

condition, or assumption.

The proposed changes would revise

Technical Specification (TS) Table 3.3–11,

‘‘Steam and Feedwater Rupture Control

System Instrumentation,’’ and Table 4.3–11

‘‘Steam and Feedwater Rupture Control

System Instrumentation Surveillance

Requirements,’’ to identify the Steam and

Feedwater Rupture Control System (SFRCS)

output logic as a separate Functional Unit. In

addition, the proposed changes would revise

TS Table 3.3–12, ‘‘Steam and Feedwater

Federal Register / Vol. 66, No. 247 / Wednesday, December 26, 2001 / Notices 66467

Rupture Control System Instrumentation Trip

Setpoints,’’ to remove the ‘‘Trip Setpoint’’

values and also modify the ‘‘Allowable

Values’’ entry for Functional Unit 3, ‘‘Steam

Generator Feedwater Differential Pressure—

High,’’ consistent with updated calculations

and current setpoint methodology, and revise

the applicability of TS Allowable Values for

other SFRCS Functional Units in this table.

The proposed changes would also revise TS

Table 4.3–11 to change the Channel

Functional Test surveillance requirements for

the SFRCS instrument channels from

monthly to quarterly, consistent with current

methodology. The proposed changes would

also make related administrative changes to

TS Limiting Condition for Operation (LCO)

3.3.2.2, TS Table 3.3–17, ‘‘Anticipatory

Reactor Trip System Instrumentation,’’ TS

Table 3.3–6, ‘‘Radiation Monitoring

Instrumentation,’’ and the associated TS

Bases.

These proposed changes do not involve a

significant change to plant design or

operation.

1b. Not involve a significant increase in the

consequences of an accident previously

evaluated because the proposed changes do

not invalidate assumptions used in

evaluating the radiological consequences of

an accident, do not alter the source term or

containment isolation, and do not provide a

new radiation release path or alter

radiological consequences.

2. Not create the possibility of a new or

different kind of accident from any accident

previously evaluated because the proposed

changes do not introduce a new or different

accident initiator or introduce a new or

different equipment failure mode or

mechanism.

3. Not involve a significant reduction in a

margin of safety as defined in the basis for

any Technical Specification. The SFRCS

instrumentation setpoint analyses will

continue to adequately preserve the margin

of safety. In addition, there are no new or

significant changes to the initial conditions

contributing to accident severity or

consequences. Therefore, there are no

significant reductions in a margin of safety.

Conclusion:

On the basis of the above, the Davis-Besse

Nuclear Power Station has determined that

the License Amendment Request does not

involve a significant hazards consideration.

As this License Amendment Request

concerns a proposed change to the Technical

Specifications that must be reviewed by the

Nuclear Regulatory Commission, this License

Amendment Request does not constitute an

unreviewed safety question.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Mary E.

O’Reilly, Attorney, FirstEnergy

Corporation, 76 South Main Street,

Akron, OH 44308.

NRC Section Chief: Anthony J.

Mendiola.

FirstEnergy Nuclear Operating

Company, Docket No. 50–346, DavisBesse Nuclear Power Station (DBNPS),

Unit 1, Ottawa County, Ohio

Date of amendment request: October

12, 2001.

Description of amendment request:

The proposed amendment would

change the Operating License (OL)

paragraph 2.C(1), Maximum Power

Level; OL paragraph 2.C(3)(d),

Additional Conditions; Technical

Specification (TS) 1.3, Definitions—

Rated Thermal Power; TS 2.1.1, Safety

Limits—Reactor Core, and associated

Bases; TS 2.2.1, Limiting Safety System

Settings—Reactor Protection System

Setpoints, and associated Bases; TS 3/

4.1.1.3, Reactivity Control Systems—

Moderator Temperature Coefficient; TS

3/4.2.5, Power Distribution Limits—

DNB Parameters; TS 3/4.4.9.1, Reactor

Coolant System—Pressure/Temperature

Limits, and associated Bases; and TS

6.9.1.7, Core Operating Limits Report.

The purpose of this license amendment

application would make the necessary

revisions to the Davis-Besse Nuclear

Power Station (DBNPS) TS to reflect an

increase in the authorized rated thermal

power from 2772 MWt to 2817 MWt

(approximately 1.63 percent), based on

the use of Caldon Inc. Leading Edge

Flow Meter (LEFM) CheckPlusTM

System instrumentation to improve the

accuracy of the feedwater mass flow

input to the plant power calorimetric

measurement.

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 based on the comprehensive

analytical efforts that were performed to

demonstrate the acceptability of the proposed

power uprate changes. The proposed changes

include: revision of the maximum power

level limit stated in Operating License (OL)

paragraph 2.C(1) and Technical Specification

(TS) Section 1.3, increasing the allowable

power level from 2772 MWt to 2817 MWt;

revision of the reactor core safety limits

specified in TS Section 2.1.1; revision of the

Reactor Protection System (RPS) high flux

and Reactor Coolant System (RCS) pressuretemperature setpoints provided in TS Section

2.2.1; revision of the RCS pressuretemperature limits in TS Section 3/4.4.9.1,

and a related change to OL paragraph

2.C(3)(d); and revision of administrative

controls associated with the Core Operating

Limits Report, as described in TS Section

6.9.1.7. In addition, related changes to the TS

Bases associated with these TS Sections are

proposed. An evaluation has been performed

that identified the systems and components

that could be affected by these proposed

changes. The evaluation determined that

these systems and components will function

as designed and that performance

requirements remain acceptable.

The primary loop components (reactor

vessel, reactor internals, control rod drive

mechanisms (CRDMs), loop piping and

supports, reactor coolant pumps, steam

generators and pressurizer) will continue to

comply with their applicable structural limits

and will continue to perform their intended

design functions. Thus, there is no increase

in the probability of a structural failure of

these components leading to an accident.

The Leak-Before-Break analysis

conclusions remain valid and the breaks

previously exempted from structural

consideration remain unchanged.

All of the Nuclear Steam Supply System

(NSSS) systems will continue to perform

their intended design functions during

normal and accident conditions. The

pressurizer spray flow remains above its

design value. Thus, the control system design

analyses, which credit the flow, do not

require any modification. The components

continue to comply with applicable

structural limits and will continue to perform

their intended design functions. Thus, there

is no increase in the probability of a

structural failure of these components.

All of the NSSS/Balance of Plant (BOP)

interface systems will continue to perform

their intended design functions. The main

steam safety valves will provide adequate

relief capacity to maintain the main steam

system within design limits.

The current loss of coolant accident

(LOCA) hydraulic forcing functions remain

bounding.

The reduction in power measurement

uncertainty through the use of the Caldon

Leading Edge Flow Meter (LEFM)

CheckPlusTM system, allows for certain safety

analyses to continue to be used, without

modification, at the 2827 MWt power level

(102% of 2772 MWt). Other safety analyses

performed at a nominal power level of 2772

MWt have been either re-performed or reevaluated at the 2817 MWt power level, and

continue to meet their applicable acceptance

criteria. Some existing safety analyses had

been previously performed at a power level

greater than 2827 MWt, and thus continue to

bound the 2817 MWt power level.

The proposed changes to the RCS pressuretemperature limit curves impose a

conservative projection of the increase in

neutron fluence associated with the power

uprate. This projection will ensure that the

requirements of 10 CFR 50 Appendix G,

‘‘Fracture Toughness Requirements,’’ will

continue to be met following the proposed

power uprate. The design basis events that

were protected against by these limits have

not changed, therefore, the probability of an

accident previously evaluated is not

increased.

In addition to the changes related to the

proposed power uprate, unrelated changes

are proposed to revise the moderator

temperature coefficient requirements listed

66468 Federal Register / Vol. 66, No. 247 / Wednesday, December 26, 2001 / Notices

in TS Section 3.1.1.3, and to revise

requirements relating to the Departure from

Nucleate Boiling (DNB) parameters listed in

TS Section 3.2.5. These proposed changes are

conservative changes and clarifications that

do not involve any physical change to

systems or components, nor do they alter the

typical manner in which the systems or

components are operated. Therefore, these

changes will not result in a significant

increase in the probability of an accident.

1b. Not involve a significant increase in the

consequences of an accident previously

evaluated because the proposed power uprate

changes do not alter any assumptions

previously made in the radiological

consequence evaluations, nor affect

mitigation of the radiological consequences

of an accident previously evaluated.

The accident radiation dose evaluation was

performed at 2827 MWt and is bounding

when operating at the proposed 2817 MWt

using the LEFM CheckPlusTM flow

instrumentation.

The proposed changes unrelated to the

power uprate also do not alter any

assumption previously made in the

radiological consequence evaluations, nor do

they affect mitigation of the radiological

consequences of an accident previously

evaluated. Therefore, these changes will not

involve a significant increase in the

consequences of an accident previously

evaluated.

2. Not create the possibility of a new or

different kind of accident from any accident

previously evaluated because no new

accident scenarios, failure mechanisms or

single failures are introduced as a result of

the proposed power uprate changes as well

as the proposed changes unrelated to the

power uprate. All systems, structures, and

components previously required for the

mitigation of an event remain capable of

fulfilling their intended design function. The

proposed changes have no adverse effects on

any safety-related system or component and

do not challenge the performance or integrity

of any safety-related system.

3. Not involve a significant reduction in a

margin of safety because extensive analyses

of the primary fission product barriers,

conducted in support of the proposed power

uprate, have concluded that all relevant

design criteria remain satisfied, both from the

standpoint of the integrity of at the primary

fission product barrier and from the

standpoint of compliance with the regulatory

acceptance criteria. As appropriate, all

evaluations have been performed using

methods that have either been reviewed and

approved by the Nuclear Regulatory

Commission (NRC) or that are in compliance

with applicable regulatory review guidance

and standards. The proposed changes

unrelated to the power uprate do not involve

a significant reduction in a margin of safety

because they do not involve the potential for

a significant increase in a failure rate of any

system or component, and existing system

and component redundancy is not affected.

Also, these changes do not involve any new

or significant changes to the initial

conditions contributing to accident severity

or consequences.

Conclusion:

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.

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.

Niagara Mohawk Power Corporation,

Docket No. 50–220, Nine Mile Point

Nuclear Station Unit No. 1, Oswego

County, New York 

Date of amendment request:

November 26, 2001.

Description of amendment request:

The licensee proposed to amend the

Technical Specifications (TSs) to delete

Section 3/4.2.6, ‘‘Inservice Inspection

and Testing,’’ and its associated bases,

revise Section 4.2.7, ‘‘Reactor Coolant

System Isolation Valves,’’ and its

associated bases, create a new Section

6.17, ‘‘Inservice Testing Program,’’ and

delete several reporting requirements in

Section 6.9.3, ‘‘Special Reports.’’ These

changes will improve the TSs, making it

consistent with current NRC guidance

and the improved Standard Technical

Specifications for General Electric (GE)

Boiling Water Reactor (BWR)/4 and

BWR/6 plants (NUREG–1433 and

NUREG–1434, respectively). Most of

these changes would also render the TSs

to be similar to the Nine Mile Point

Nuclear Station, Unit No. 2 TSs, which

is based on NUREG–1433 and NUREG–

1434.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The operation of Nine Mile Point Unit

1 in accordance with the proposed

amendment will not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

The proposed amendment deletes

duplicative and unnecessary inservice

inspection (ISI) and inservice testing (IST)

requirements from the Technical

Specifications; clarifies remaining IST

requirements; revises a requirement to

perform quarterly testing of the reactor

coolant isolation valves to conform to the

periodic testing requirements of the ASME

[American Society of Mechanical Engineers]

Boiler and Pressure Vessel Code (ASME

Code); and deletes unnecessary reporting

requirements relating to routine ISI, primary

containment leakage testing, and secondary

containment leakage testing. These changes

do not reduce the plant’s existing ISI/IST

commitments based on 10CFR50.55a, Section

XI of the ASME Code, and Generic Letter 88–

01. These changes also do not involve

hardware changes, changes in plant

setpoints, or changes in plant safety

parameters.

Based on the above, the operation of Nine

Mile Point Unit 1 (NMP1) in accordance with

the proposed amendment, will not involve a

significant increase in the probability or the

consequences of an accident previously

evaluated.

2. The operation of Nine Mile Point Unit

1 in accordance with the proposed

amendment will not create the possibility of

a new or different kind of accident from any

accident previously evaluated.

The proposed changes do not involve any

physical modifications to the plant nor alter

equipment configuration, setpoints, or safety

parameters. The ISI/IST related changes are

consistent with current NRC guidance and

industry standards and will continue to

ensure acceptable equipment operability and

availability.

Based on the above, the operation of NMP1

in accordance with the proposed amendment

cannot create the possibility of a new or

different kind of accident from any accident

previously evaluated.

3. The operation of Nine Mile Point Unit

1 in accordance with the proposed

amendment will not involve a significant

reduction in a margin of safety.

The proposed changes do not affect any of

the plant’s fission product barriers or safety/

operational limits. The ISI/IST related

changes will continue to ensure acceptable

equipment operability and availability.

Based on the above, the operation of NMP1

in accordance with the proposed amendment

will not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

requested amendment involves no

significant hazards consideration.

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston & Strawn,

1400 L Street, NW., Washington, DC

20005–3502.

NRC Section Chief: L. Raghavan,

Acting.

Niagara Mohawk Power Corporation,

Docket No. 50–410, Nine Mile Point

Nuclear Station Unit No. 2, Oswego

County, New York

Date of amendment request:

November 20, 2001.

Description of amendment request:

The licensee proposed to amend the

Technical Specifications (TSs) regarding

the safety limit minimum critical power

Federal Register / Vol. 66, No. 247 / Wednesday, December 26, 2001 / Notices 66469

ratio (SLMCPR) to reflect the results of

cycle-specific calculations performed

for the next fuel cycle (i.e., Cycle 9),

using NRC-approved methodology for

determining SLMCPR values. The

proposed amendment would also

editorially revise references to topical

reports which document the approved

methodology.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The operation of Nine Mile Point Unit

2, in accordance with the proposed

amendment, will not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

The derivation of the revised Safety Limit

Minimum Critical Power Ratio (SLMCPR)

values for Nine Mile Point Unit 2 (NMP2)

Cycle 9 for incorporation into the Technical

Specifications (TS) and their use to

determine cycle-specific thermal limits has

been performed using the NRC-approved

methods and procedures in [Topical Report]

NEDE–24011–P–A, ‘‘General Electric

Standard Application for Reactor Fuel’’

(GESTAR II). The analysis methodology

incorporates cycle-specific parameters and

reduced power distribution uncertainties in

the determination of the SLMCPR values.

These calculations do not change the method

of operating the plant and have no effect on

the probability of an accident initiating event

or transient.

The basis of the Minimum Critical Power

Ratio Safety Limit is to ensure no

mechanistic fuel damage is calculated to

occur if the limit is not violated. The new

SLMCPR values preserve the existing margin

to transition boiling and the probability of

fuel damage is not increased. The deletion of

listed documents that are already

incorporated by reference into GESTAR II is

administrative only. Therefore, the proposed

changes do not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

2. The operation of Nine Mile Point Unit

2, in accordance with the proposed

amendment, will not create the possibility of

a new or different kind of accident from any

accident previously evaluated.

The new SLMCPR values for the NMP2

Cycle 9 core reload have been calculated in

accordance with the methods and procedures

described in GESTAR II. These methods have

been reviewed and approved by the NRC.

The deletion of listed documents that are

already incorporated by reference into

GESTAR II is administrative only. The

changes do not involve any new method for

operating the facility and do not involve any

facility modifications. No new initiating

events or transients result from these

changes. Therefore, the proposed TS changes

do not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

3. The operation of Nine Mile Point Unit

2, in accordance with the proposed

amendment, will not involve a significant

reduction in a margin of safety.

The margin of safety as defined in the TS

bases will remain the same. The new, cyclespecific SLMCPR values are calculated using

NRC-approved methods and procedures that

are in accordance with the current fuel

design and licensing criteria. The SLMCPR

values remain high enough to ensure that

greater than 99.9% of all fuel rods in the core

are expected to avoid transition boiling if the

limits are not violated, thereby preserving the

fuel cladding integrity. The deletion of listed

documents that are already incorporated by

reference into GESTAR II is administrative

only. Therefore, the proposed TS changes do

not involve a significant reduction in [a]

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

requested amendment involves no

significant hazards consideration.

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston & Strawn,

1400 L Street, NW., Washington, DC

20005–3502.

NRC Section Chief: L. Raghavan,

Acting.

Nuclear Management Company, LLC,

Docket Nos. 50–266 and 50–301, Point

Beach Nuclear Plant, Units 1 and 2,

Town of Two Creeks, Manitowoc

County, Wisconsin

Date of amendment request:

November 19, 2001.

Description of amendment request: A

change is proposed to Technical

Specification (TS) 3.0.3 to allow a

longer period of time to perform a

missed surveillance. The time is

extended from the current limit of

‘‘* * * up to 24 hours or up to the limit

of the specified Frequency, whichever is

less’’ to ‘‘* * * up to 24 hours or up to

the limit of the specified Frequency,

whichever is greater.’’ In addition, the

following requirement would be added

to the specification: ‘‘A risk evaluation

shall be performed for any Surveillance

delayed greater than 24 hours and the

risk impact shall be managed.’’

The Nuclear Regulatory Commission

(NRC) staff issued a notice of

opportunity for comment in the Federal

Register on June 14, 2001 (66 FR 32400),

on possible amendments concerning

missed surveillances, including a model

safety evaluation and model no

significant hazards consideration

(NSHC) determination, using the

consolidated line item improvement

process. The NRC staff subsequently

issued a notice of availability of the

models for referencing in license

amendment applications in the Federal

Register on September 28, 2001 (66 FR

49714). The licensee affirmed the

applicability of the following NSHC

determination in its application dated

November 7, 2001.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), an

analysis of the issue of no significant

hazards consideration is presented

below:

Criterion 1—The Proposed Change Does Not

Involve a Significant Increase in the

Probability or Consequences of an Accident

Previously Evaluated

The proposed change relaxes the time

allowed to perform a missed surveillance.

The time between surveillances is not an

initiator of any accident previously

evaluated. Consequently, the probability of

an accident previously evaluated is not

significantly increased. The equipment being

tested is still required to be operable and

capable of performing the accident mitigation

functions assumed in the accident analysis.

As a result, the consequences of any accident

previously evaluated are not significantly

affected. Any reduction in confidence that a

standby system might fail to perform its

safety function due to a missed surveillance

is small and would not, in the absence of

other unrelated failures, lead to an increase

in consequences beyond those estimated by

existing analyses. The addition of a

requirement to assess and manage the risk

introduced by the missed surveillance will

further minimize possible concerns.

Therefore, this change does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

Criterion 2—The Proposed Change Does Not

Create the Possibility of a New or Different

Kind of Accident From Any Previously

Evaluated

The proposed change does not involve a

physical alteration of the plant (no new or

different type of equipment will be installed)

or a change in the methods governing normal

plant operation. A missed surveillance will

not, in and of itself, introduce new failure

modes or effects and any increased chance

that a standby system might fail to perform

its safety function due to a missed

surveillance would not, in the absence of

other unrelated failures, lead to an accident

beyond those previously evaluated. The

addition of a requirement to assess and

manage the risk introduced by the missed

surveillance will further minimize possible

concerns. Thus, this change does not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

Criterion 3—The Proposed Change Does Not

Involve a Significant Reduction in the Margin

of Safety

The extended time allowed to perform a

missed surveillance does not result in a

significant reduction in the margin of safety.

As supported by the historical data, the likely

66470 Federal Register / Vol. 66, No. 247 / Wednesday, December 26, 2001 / Notices

outcome of any surveillance is verification

that the LCO [Limiting Condition for

Operation] is met. Failure to perform a

surveillance within the prescribed frequency

does not cause equipment to become

inoperable. The only effect of the additional

time allowed to perform a missed

surveillance on the margin of safety is the

extension of the time until inoperable

equipment is discovered to be inoperable by

the missed surveillance. However, given the

rare occurrence of inoperable equipment, and

the rare occurrence of a missed surveillance,

a missed surveillance on inoperable

equipment would be very unlikely. This

must be balanced against the real risk of

manipulating the plant equipment or

condition to perform the missed surveillance.

In addition, parallel trains and alternate

equipment are typically available to perform

the safety function of the equipment not

tested. Thus, there is confidence that the

equipment can perform its assumed safety

function.

Therefore, this change does not involve a

significant reduction in a margin of safety.

Based upon the reasoning presented above

and the previous discussion of the

amendment request, the requested change

does not involve a significant hazards

consideration.

The NRC staff proposes to determine

that the amendment request involves no

significant hazards consideration.

Attorney for licensee: John H. O’Neill,

Jr., Shaw, Pittman, Potts, and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Section Chief: William D.

Reckley, Acting.

Nuclear Management Company, LLC,

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

Island Nuclear Generating Plant, Units

1 and 2, Goodhue County, Minnesota

Date of amendment request: February

2, 2001, supplemented August 31, 2001.

Description of amendment request:

The proposed amendments would

revise the technical specifications (TSs)

to clarify the plant conditions under

which various specifications are

applicable. The licensee stated in its

amendment request that a literal reading

of the current technical specifications

wording may result in situations where

a routine plant shutdown would seem to

be prohibited by TSs and, thereby,

require entry into TS 3.0.C. This

amendment request also makes several

administrative changes to the TSs,

including revising references to the

Chief Nuclear Corporate Officer,

capitalizing defined terms, and updating

references to previously relocated TS

paragraphs and correcting the List of

Figures. The licensee’s supplement to

the amendment request, dated August

31, 2001, proposed a correction of a

typographical error in TS Table 3.5–2B,

Action 33.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does operation of the facility with the

proposed amendment involve a significant

increase in the probability or consequences

of an accident previously evaluated?

The proposed changes are administrative

in nature and clarify existing specifications

without reducing or altering the requirements

imposed by existing specifications. The

proposed changes do not significantly affect

any system that is a contributor to initiating

events for previously evaluated accidents.

Neither do the changes significantly affect

any system that is used to mitigate any

previously evaluated accidents. Therefore,

the proposed changes do not involve any

significant increase in the probability or

consequence of an accident previously

evaluated.

2. Does operation of the facility with the

proposed amendment create the possibility of

a new or different kind of accident from any

accident previously evaluated?

The proposed changes are administrative

in nature and clarify existing specifications

without reducing or altering the requirements

imposed by existing specifications. The

proposed changes do not alter the design,

function, or operation of any plant

component and do not install any new or

different equipment, therefore a possibility of

a new or different kind of accident from

those previously analyzed has not be[en]

created.

3. Does operation of the facility with the

proposed amendment involve a significant

reduction in a margin of safety?

The proposed changes are

administrative in nature and clarify

existing specifications without reducing

or altering the requirements imposed by

existing specifications. Thus, the

proposed change[s] do not involve a

significant reduction in the margin of

safety associated with the safety limits

inherent in either the principle barriers

to a radiation release (fuel cladding,

RCS [reactor coolant system] boundary,

and reactor containment), or the

maintenance of critical safety functions

(subcriticality, core cooling, ultimate

heat sink, RCS inventory, RCS boundary

integrity, and containment integrity).

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment requests involve no

significant hazards consideration.

Attorney for licensee: Peter

Marquardt, Legal Department, 688 WCB,

Detroit Edison Company, 2000 2nd

Avenue, Detroit, Michigan 48226–1279.

NRC Acting Section Chief: William D.

Reckley, Acting.

Omaha Public Power District, Docket

No. 50–285, Fort Calhoun Station, Unit

No. 1, Washington County, Nebraska

Date of amendment request:

November 21, 2001.

Description of amendment request:

The proposed amendment will revise

Technical Specifications 2.15(5) and

2.15(6) to identify: (1) all indication and

control functions required for the

alternate (remote) shutdown panels, (2)

panel locations of the functions, and (3)

the number of operable channels

required.

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 to Technical

Specifications Sections 2.15(5) and 2.15(6)

identify functions, instruments, and controls

along with their location and the number of

required channels. New Technical

Specifications Section 2.15(5) addresses the

regulatory requirements for equipment

required for Alternative and Dedicated

Shutdown Capability per 10 CFR part 50,

Appendix R. It will ensure that proper

Limiting Conditions for Operation are

entered for equipment or functional

inoperability. There are no physical

alterations being made to the Alternate

Shutdown Panels and the Auxiliary

Feedwater Panel or related systems.

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 will not result in

any physical alterations to the Alternate

Shutdown Panels or the Auxiliary Feedwater

Panel, or any plant configuration, systems,

equipment, or operational characteristics.

There will be no changes in operating modes,

or safety limits, or instrument limits. With

the proposed changes in place, Technical

Specifications retain requirements for the

Alternate Shutdown Panels and the Auxiliary

Feedwater Panel. Therefore, the proposed

changes do 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.

The proposed changes clarify the

regulatory requirements for the Alternative

and Dedicated Shutdown Capability as

defined by 10 CFR Part 50, Appendix R. The

proposed changes will not alter any physical

Federal Register / Vol. 66, No. 247 / Wednesday, December 26, 2001 / Notices 66471

or operational characteristics of the Alternate

Shutdown Panels and the Auxiliary

Feedwater Panel and their associated systems

and equipment. Therefore, the proposed

changes do not involve a reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: James R.

Curtiss, Esq., Winston & Strawn, 1400 L

Street, N.W., Washington, DC 20005–

3502.

NRC Section Chief: Stephen Dembek.

Omaha Public Power District, Docket

No. 50–285, Fort Calhoun Station, Unit

No. 1, Washington County, Nebraska

Date of amendment request:

November 21, 2001.

Description of amendment request:

The proposed amendment will add

three topical report references to

Technical Specification 5.9.5, ‘‘Core

Operating Limit Reports.’’

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 amendment incorporates

three additional Framatome ANP topical

reports for conducting core reload analyses.

Since the intent of the amendment request is

to add references to NRC-approved reload

analysis methods, 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.

No new or different modes of operation are

proposed as a result of these changes. The

proposed revision does not change any

equipment required to mitigate the

consequences of an accident. The proposed

addition of NRC-approved topical reports to

the Technical Specification does not modify

the manner in which the topical reports may

be implemented. The plant will continue to

operate within the limits specified by the

Core Operating Limits Report and corrective

actions will be taken in accordance with the

Technical Specifications should these limits

be exceeded. 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.

As required by Technical Specification

5.9.5, the analytical methods used to

determine the core operating limits shall be

those previously reviewed and approved by

the NRC. The proposed change incorporates

methodologies applicable for use with fuel

supplied by Framatome ANP that have been

approved by the NRC as documented by

Safety Evaluation Reports (References 10.1,

10.2, and 10.3 [of the November 21, 2001,

amendment request]). Because Technical

Specification 5.9.5 also requires that the core

operating limits shall be determined and

requires that all applicable limits of the

safety analysis are met, the proposed change

does not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: James R.

Curtiss, Esq., Winston & Strawn, 1400 L

Street, N.W., Washington, DC 20005–

3502.

NRC Section Chief: Stephen Dembek.

Pacific Gas and Electric Company,

Docket No. 50–133, Humboldt Bay

Power Plant, Unit 3, Humboldt County,

California

Date of amendment request:

December 28, 2000, as supplemented by

letters dated March 29 and October 31,

2001.

Description of amendment request:

The proposed amendment would

convert the Humboldt Bay Power Plant

Unit 3 Current Technical Specifications

to a set of Permanently Defueled

Technical Specifications with a more

standardized format and content based

on a revision to 10 CFR 50.36 (Technical

Specifications) and technical

specifications approved for other

permanently shutdown nuclear power

plants (Millstone Unit 1 and Trojan

Nuclear Plant).

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analyses of the

issue of no significant hazards

consideration, which are presented

below.

The conversion of the Humboldt Bay

Power Plant (HBPP) Current Technical

Specifications (CTS) to Permanently

Defueled Technical Specifications (PDTS)

involves the following four types of

dispositions:

A Administrative reformatting and

rewording

D Item deleted from the Technical

Specifications (TS)

LG Relocating items from CTS to the

Defueled Safety Analysis Report (DSAR),

PDTS, or other Licensee-Controlled

Document

N Addition of new requirements of new

sections to the PDTS

Administrative Reformatting and Rewording

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed change involves reformatting

and editorially rewording of the CTS. As

such, this change is administrative in nature

and does not impact initiators of analyzed

events or assumed mitigation of accidents or

transient events. Therefore, this 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.

The proposed change does not necessitate

a physical alteration of the plant (no new or

different type of equipment will be installed)

or changes in parameters governing normal

plant operation. The proposed change will

not impose any different operational

requirements and any administrative

additions are non-operational in nature and

have not been identified and justified. Thus,

this change does 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 change will not reduce a

margin of safety because it has no impact on

any safety analysis assumptions. This change

is administrative in nature. As such, no

question of safety is involved.

Items Deleted from the Technical

Specifications that are Duplicative in Nature

to Other Regulatory Requirements

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed change involves deleting

information from the CTS. The information

being deleted is still required to be performed

and is being performed by the licensee

because the information is contained in

regulatory requirements contained in the

Code of Federal Regulations. Therefore, this

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.

The proposed change does not necessitate

a physical alteration of the plant (no new or

different type of equipment will be installed)

or changes in parameters governing normal

plant operation. The proposed change will

not impose any different operational

requirements. Thus, this change does 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 change will not reduce a

margin of safety because it has no impact on

66472 Federal Register / Vol. 66, No. 247 / Wednesday, December 26, 2001 / Notices

any safety analysis assumptions. This change

is administrative in nature. The requirements

being deleted from the CTS are still required

to be met and are being met by the licensee

because these requirements exist in the Code

of Federal Regulations. As such, no question

of safety is involved.

Items Deleted from the Technical

Specifications That Have No Application in

the Proposed HBPP PDTS

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed change involves deleting

information from the CTS. The deletion

process involves no technical changes to the

CTS. As such, this change is administrative

in nature and does not impact initiators of

analyzed events or assumed mitigation of

accidents or transient events. Therefore, this

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.

The proposed change does not necessitate

a physical alteration of the plant (no new or

different type of equipment will be installed)

or changes in parameters governing normal

plant operation. The proposed change will

not impose any different operational

requirements. Thus, this change does 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 change will not reduce a

margin of safety because it has no impact on

any safety analysis assumptions. This change

is administrative in nature. As such, no

question of safety is involved.

Relocating Information from CTS to the

DSAR, PDTS Bases or Other LicenseeControlled Documents

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed change relocates

requirements and descriptive information

from the CTS to the PDTS Bases, DSAR, or

other licensee-controlled documents. The

PDTS Bases, DSAR, or other licenseecontrolled documents containing the

relocated requirements and information will

be maintained using provisions of 10 CFR

50.59 or other appropriate regulatory

controls. Since any future changes to the

PDTS Bases, DSAR, or other licenseecontrolled documents will be evaluated per

the requirements of 10 CFR 50.59 or other

appropriate regulatory controls, proper

controls are in place to adequately limit the

probability or consequences of an accident

previously evaluated. Therefore, this 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.

The proposed change does not necessitate

a physical alteration of the plant (no new or

different type of equipment will be installed)

or changes in parameters governing normal

plant operation. The proposed change will

not impose any different requirements and

adequate control of the information will be

maintained. Thus, this change does 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 change will not reduce a

margin of safety because it has no impact on

any safety analysis assumptions. In addition,

the requirements and information to be

relocated from the CTS to the PDTS Bases,

DSAR, or other licensee-controlled

documents are not being revised; they are

being relocated verbatim. Since any future

changes to these requirements in the PDTS

Bases, DSAR, or other licensee-controlled

documents will be evaluated per the

requirements of 10 CFR 50.59 or other

appropriate regulatory controls, proper

controls are in place to maintain an

appropriate margin of safety. Therefore this

change does not involve a significant

reduction in a margin of safety.

Addition of New Requirements or New

Sections to the PDTS

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed change involves the addition

of requirements or sections to the proposed

PDTS. Each addition either provides

equivalent or potentially more restrictive

controls than previously provided. The

additional requirements or controls do not

impact initiators of analyzed events or

assumed mitigation of accidents or transient

events. Therefore, this 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.

The proposed change does not necessitate

a physical alteration of the plant (no new or

different type of equipment will be installed)

or changes in parameters governing normal

plant operation. The proposed change will

not impose any different operational

requirements and any addition is nonoperational in nature and has been identified

and justified. Thus, this change does 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 change will not reduce a

margin of safety because it has no impact on

any safety analysis assumptions. This change

provides the equivalent or more restrictive

requirements on the surveillance and control

of TS parameters. As such, no question of

safety is involved.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Christopher J.

Warner, Esquire, Pacific Gas and

Electric Company, P.O. Box 7442, San

Francisco, California 94120.

NRC Section Chief: Stephen Dembek.

STP Nuclear Operating Company,

Docket Nos. 50–498 and 50–499, South

Texas Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request: August

22, 2001.

Description of amendment request:

Approve reactor core power uprate, and

revise the Technical Specifications to

reflect the power uprate.

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:

STPNOC [South Texas Project Nuclear

Operating Company] has evaluated whether

or not a significant hazards consideration is

involved with the proposed amendment by

focusing on the three standards set forth in

10 CFR 50.92, ‘‘Issuance of amendment,’’ as

discussed below.

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The comprehensive analytical efforts

performed to support the proposed uprate

conditions include a review and evaluation

of all components and systems (including

interface systems and control systems) that

could be affected by this change. The revised

power uprate value was input to applicable

safety analyses. The proposed change is not

an initiator of any design-basis accident. All

of the Nuclear Steam Supply System or

Balance of Plant interface systems will

continue to perform their intended design

functions and meet all performance

requirements. The primary loop components

(reactor vessel, reactor internals, control rod

drive mechanisms, loop piping and supports,

reactor coolant pump, steam generator, and

pressurizer) continue to comply with their

applicable structural limits and will continue

to perform their intended design functions.

Therefore, there is no increase in the

probability of a structural failure of these

components.

The auxiliary systems and components

continue to comply with applicable

structural limits and will continue to perform

their intended design functions. Therefore,

there is no increase in the probability of a

structural failure of these components. The

steam generator safety valves will provide

adequate relief capacity to maintain the

steam generators within design limits. The

steam dump system will still relieve 40

Federal Register / Vol. 66, No. 247 / Wednesday, December 26, 2001 / Notices 66473

percent of the maximum full-load steam

flow.

Therefore, the proposed change does not

involve a significant increase in the

probability of an accident previously

evaluated.

The applicable analyses have been

evaluated with respect to the increase in core

power associated with this change. All

applicable radiological acceptance criteria

continue to be met. Therefore, the proposed

change does not involve a significant

increase in the 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 change neither causes the

initiation of any accident nor creates any new

limiting single failures. All of the affected

systems and components continue to perform

their intended design functions. The

proposed change has no adverse effects on

any safety-related system or component and

does not challenge the performance or

integrity of any safety-related system.

The proposed change does 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 WRB–2M DNB methodology is used to

demonstrate that core thermal-hydraulic

limits are maintained without any significant

reduction in margin of safety for the uprated

power level of 3853 MWt (1.4-percent uprate)

assuming core designs composed of Robust

Fuel Assemblies. The WRB–1 DNB

correlation demonstrates that there is no

significant reduction in the margin of safety

for core designs composed of standard or

Vantage 5 Hybrid (V5H) fuel types. Extensive

analyses of the primary fission product

barriers have concluded that all relevant

design criteria remain satisfied, both from the

standpoint of the integrity of the primary

fission product barrier and from the

standpoint of compliance with the regulatory

acceptance criteria. As appropriate, all

evaluations have been performed using

methods that either have been reviewed and

approved by the Nuclear Regulatory

Commission or are in compliance with all

applicable regulatory review guidance and

standards.

Therefore, the proposed change does not

involve a significant reduction in a margin of

safety.

Based on the above, STPNOC concludes

that the proposed amendment presents no

significant hazards consideration under the

standards set forth in 10 CFR 50.92(c), and

accordingly, a finding of ‘‘no significant

hazards consideration’’ is justified.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the standards of

10 CFR 50.92(c) are satisfied. Therefore,

the NRC staff proposes to determine that

the request for amendments involves no

significant hazards consideration.

Attorney for licensee: Jack R.

Newman, Esq., Morgan, Lewis &

Bockius, 1800 M Street, NW.,

Washington, DC 20036–5869.

NRC Section Chief: Robert A. Gramm.

TXU Electric, Docket Nos. 50–445 and

50–446, Comanche Peak Steam Electric

Station, Units 1 and 2, Somervell

County, Texas

Date of amendment request: October

23, 2001.

Brief description of amendments: The

proposed amendment would revise

Technical Specification (TS) 5.5.9,

‘‘Steam Generator Tube Surveillance

Program,’’ to permit tube sleeving repair

techniques, developed by Westinghouse

Electric Company (Westinghouse) and

referred to as ‘‘Westinghouse Leak Tight

Sleeves.’’

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Do the proposed changes involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

Response: No.

The Westinghouse Leak Tight Sleeves are

designed using the applicable American

Society of Mechanical Engineers (ASME)

Boiler and Pressure Vessel Code and,

therefore, meet the design objectives of the

original steam generator tubing. The

applicable design criteria for the sleeves

conforms to the stress limits and margins of

safety of Section III of the ASME code.

Mechanical testing has shown that the

structural strength of repair sleeves under

normal, upset, and faulted conditions

provides margin to the acceptance limits.

These acceptance limits bound the most

limiting (three times normal operating

pressure differential) burst margin

recommended by Draft Regulatory Guide

1.121. Burst testing of sleeved tubes has

demonstrated that no unacceptable levels of

primary-to-secondary leakage are expected

during any plant condition.

Evaluation of the repaired steam generator

tubes indicates no detrimental effects on the

sleeve or sleeve-tube assembly from reactor

coolant system flow, primary or secondary

coolant chemistries, thermal conditions or

transients, or pressure conditions as may be

experienced at CPSES [Comanche Peak

Steam Electric Station]. Corrosion testing of

sleeve-tube assemblies indicates no evidence

of sleeve or tube corrosion considered

detrimental under anticipated service

conditions.

The installation of the proposed sleeves is

controlled via the sleeving vendor’s

proprietary processes and equipment. The

Westinghouse process has been in use since

1984 and has been implemented more than

24 times for the installation of over 4,200

sleeves. The CPSES steam generator design

was reviewed and found to be compatible

with the installation processes and

equipment.

The implementation of the proposed

amendment has no significant effect on either

the configuration of the plant or the manner

in which it is operated. The consequences of

a hypothetical failure of the sleeved tube is

bounded by the current steam generator tube

rupture (SGTR) analysis described in the

CPSES FSAR [Final Safety Analysis Report].

Due to the slight reduction in diameter

caused by the sleeve wall thickness, primary

coolant release rates would be slightly less

than assumed for the steam generator tube

rupture analysis, depending on the break

location, and therefore, would result in lower

total primary fluid mass release to the

secondary system. A main steam line break

or feed line break will not cause a SGTR

since the sleeves are analyzed for a maximum

accident differential pressure greater than

that predicted in the CPSES safety analysis.

The proposed reduction of the steam

generator primary to secondary operational

leakage limit provides added assurance that

leaking flaws will not propagate to burst

prior to commencement of plant shutdown.

In conclusion, based on the discussion

above, these changes will not significantly

increase the probability or consequences of

an accident previously evaluated.

2. Do the proposed changes create the

possibility of a new or different kind of

accident from any accident previously

evaluated?

Response: No.

The Westinghouse Leak Tight Sleeves are

designed using the applicable ASME Code as

guidance; therefore, they meet the objectives

of the original steam generator tubing. As a

result, the functions of the steam generators

will not be significantly affected by the

installation of the proposed sleeves. The

proposed repair sleeves do not interact with

any other plant systems. Any accident as a

result of potential tube or sleeve degradation

in the repaired portion of the tube is bounded

by the existing tube rupture accident

analysis. The continued integrity of the

installed sleeve is periodically verified by the

Technical Specification requirements.

The implementation of the proposed

amendment has no significant effect on either

the configuration of the plant or the manner

in which it is operated. As discussed above,

the reduced primary to secondary leakage

limit is considered a conservative change in

the plant limiting conditions for operation.

Therefore, TXU Electric concludes that this

proposed change does not create the

possibility of 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?

Response: No.

The repair of degraded steam generator

tubes with Westinghouse Leak Tight Sleeves

restores the structural integrity of the

degraded tube under normal operating and

postulated accident conditions. The design

safety factors utilized for the repair sleeves

are consistent with the safety factors in the

ASME Code used in the original steam

generator design. The portions of the

installed sleeve assembly that represents the

reactor coolant pressure boundary can be

monitored for the initiation and progression

of sleeve/tube wall degradation. Use of the

66474 Federal Register / Vol. 66, No. 247 / Wednesday, December 26, 2001 / Notices

previously identified design criteria and

design verification testing assures that the

margin of safety is not significantly different

from the original steam generator tubes. The

proposed sleeve inspection requirements are

more stringent than existing requirements for

inspection of the steam generator tubes, and

the reduction in the operational limit for

primary to secondary leakage through the

steam generator tubes is more conservative

than current requirements. Therefore, TXU

Electric concludes that the proposed change

does not involve a significant reduction in a

margin of safety.

EPRI [Electric Power Research Institute]

qualified eddy current techniques will be

used for the detection of tube degradation in

3/4 inch welded sleeved tubes. Alternate

inspection techniques, may be used as they

become available, as long as it can be

demonstrated that the technique used

provides the same degree or greater degree of

inspection rigor.

The effect of sleeving on the design

transients and accident analyses were

reviewed and found to remain valid up to the

level of steam generator tube plugging

consistent with the minimum reactor flow

rate as specified in Technical Specification

3.4.1. Continued compliance with the RCS

[Reactor Coolant System] flow limits of

Technical Specification 3.4.1 is assured

through precision flow measurements.

Because all relevant safety analyses were

reviewed and found to remain valid, and

because the appropriate design margins are

maintained through compliance with the

relevant ASME Code requirements, it is

concluded that the proposed change does not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: George L. Edgar,

Esq., Morgan, Lewis and Bockius, 1800

M Street, NW., Washington, DC 20036.

NRC Section Chief: Robert A. Gramm.

Vermont Yankee Nuclear Power

Corporation, Docket No. 50–271,

Vermont Yankee Nuclear Power Station,

Vernon, Vermont

Date of amendment request:

November 20, 2001.

Description of amendment request:

The proposed amendment would: (1)

Move Table 4.7.2, ‘‘Primary

Containment Isolation Valves’’ and

references to the Table from the

Vermont Yankee Nuclear Power Station

(VY) Technical Specifications (TSs) to

the Technical Requirements Manual; (2)

change Surveillance Requirement

4.7.B.1.b to reflect that the Standby Gas

Treatment system (SBGT) duct heater

needs to meet relative humidity design

basis; (3) add section 3.7.E, ‘‘Reactor

Building Automatic Ventilation System

Isolation Valves,’’ to the Table of

Contents; (4) remove wording in

3.5.A.4.a and b referencing a one-time

30-day Limiting Condition for

Operation; and (5) make administrative

changes to Sections 5.3 and 6.4.

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. The operation of the Vermont

Yankee Nuclear Power Station in

accordance with the proposed

amendment will not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes consist of

removal of the primary containment

isolation valve component list from the

VY TS, revision of the SBGT inlet heater

surveillance minimum power rating and

other administrative changes. The

probability of occurrence of a previously

evaluated accident is not increased

because neither containment isolation

nor the SBGT heater are accident

initiators, and the proposed changes do

not impact any accident initiating

conditions. The consequences of an

accident previously evaluated are not

increased because the proposed changes

do not impact the ability of containment

to restrict, or SBGT to filter, the release

of any fission product radioactivity to

the environment. The proposed changes

to remove the primary containment

isolation valve component list from TS,

relocate the information to a licensee

controlled document, and to change the

SBGT inlet heater power input

surveillance requirement, will have no

significant impact on any safety related

structures, systems or components. The

TS requirements for the primary

containment isolation valves and SBGT

operability and surveillance will not be

changed. Additionally, the

administrative changes do not affect any

system operation or function.

Therefore, the proposed changes do

not involve a significant increase in the

probability or consequences of an

accident previously evaluated.

2. The operation of Vermont Yankee

Nuclear Power Station in accordance

with the proposed amendment will not

create the possibility of a new or

different kind of accident from any

accident previously evaluated.

The proposed changes do not involve

any physical alteration of plant

equipment and do not change the

method by which any safety-related

system performs its function. No new or

different types of equipment will be

installed. The proposed changes do not

create any new accident initiators or

involve an activity that could be an

initiator of an accident of a different

type.

Therefore, the proposed changes will

not create the possibility of a new or

different kind of accident from any

accident previously evaluated.

3. The operation of Vermont Yankee

Nuclear Power Station in accordance

with the proposed amendment will not

involve a significant reduction in a

margin of safety.

The proposed administrative changes,

the removal of the primary containment

isolation valve component list from TS

and the change to the SBGT inlet heater

power input surveillance requirement,

do not alter the TS requirements for

containment integrity, containment

isolation, SBGT operability, or adversely

affect their capability. The changes will

not alter the basic operation of process

variables, systems, or components as

described in the safety analysis. No new

equipment is introduced.

The proposed changes do no impact

design margins of the primary

containment isolation system, SBGT or

any other system to perform their safety

functions. The essential safety functions

of providing primary containment

integrity and providing filtration of

airborne radioactive releases, are

maintained. There is no physical or

operational change being made which

would alter the sequence of events,

plant response, or margins in existing

safety analyses. The proposed changes

result in no impact on analyzed

accident event precursors or effects.

These proposed changes do not alter

the physical design of the plant. There

is no change in methods of operation.

The proposed changes do not alter the

means by which primary containment

isolation capability is maintained and

SBGT is operated.

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: Mr. David R.

Lewis, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037–1128.

NRC Section Chief: James W. Clifford.

Federal Register / Vol. 66, No. 247 / Wednesday, December 26, 2001 / Notices 66475

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.

Connecticut Yankee Atomic Power

Company, Docket No. 50–213, Haddam

Neck Plant, Middlesex County,

Connecticut

Date of amendment request: May 30,

2001, as supplemented on November 6,

2001.

Brief description of amendment:

Connecticut Yankee Atomic Power

Company (the licensee) requested

changes to the Technical Specifications

(TSs) for the Haddam Neck Plant. The

changes to Sections 5 and 6 of the TSs

correct terminology, clarify the

specifications for consistency with

established programs and Standard TSs,

and reflect current plant conditions. The

changes also reflect the licensee’s

current organization titles. For

information only, the licensee also

included proposed changes to the TS

Bases for spent fuel pool water level and

cooling. The NRC staff did not review

the proposed changes to the TS Bases.

Date of issuance: December 4, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment No.: 196.

Facility Operating License No. DPR–

61: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: August 22, 2001 (66 FR

44164).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated December 4,

2001.

No significant hazards consideration

comments received: No.

Dominion Nuclear Connecticut Inc., et

al., Docket Nos. 50–336 and 50–423,

Millstone Nuclear Power Station, Unit

Nos. 2 and 3, New London County,

Connecticut Date of application for

amendment: August 9, 2001

Brief Description of amendment: The

amendments modify the Millstone

Nuclear Power Station, Unit Nos. 2 and

3 Technical Specifications to clarify the

licensed operator qualification

standards.

Date of issuance: December 5, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment No.: 258 and 199.

Facility Operating License Nos. NPD–

69 and NPF–49: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: October 17, 2001 (66 FR

52798).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated December 5,

2001.

No significant hazards consideration

comments received: No.

Dominion Nuclear Connecticut, Inc.,

Docket No. 50–336, Millstone Nuclear

Power Station, Unit No. 2, New London

County, Connecticut

Date of application for amendment:

April 23, 2001.

Brief description of amendment: The

amendment approves a change to

Technical Specification (TS) 3.8.1.1,

‘‘Electrical Power System—A.C.

Sources.’’ The change removes

Surveillance Requirement 4.8.1.1.2.c.1

regarding Emergency Diesel Generator

inspection at least once per 18 months

during shutdown condition.

Date of issuance: December 7, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 60 days from the date of

issuance.

Amendment No.: 259.

Facility Operating License No. DPR–

65: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 12, 2001 (66 FR 31705).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated December 7,

2001.

No significant hazards consideration

comments received: No.

Duke Energy Corporation, Docket Nos.

50–369 and 50–370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg

County, North Carolina Date of

application for amendments: June 13,

2000, as supplemented August 30 and

September 10, 2001.

Brief description of amendments: The

amendments revised the Facility

Operating License of each unit to (1)

delete license conditions that have been

fulfilled; and (2) make other corrections

and editorial changes.

Date of issuance: December 5, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment Nos.: 200 and 181.

Facility Operating License Nos. NPF–

9 and NPF–17: Amendments revised the

Facility Operating License.

Date of initial notice in Federal

Register: November 1, 2000 (65 FR

65341).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated December 5,

2001.

No significant hazards consideration

comments received: No.

66476 Federal Register / Vol. 66, No. 247 / Wednesday, December 26, 2001 / Notices

Exelon Generation Company, LLC,

Docket Nos. 50–373 and 50–374, LaSalle

County Station, Units 1 and 2, LaSalle

County, Illinois

Date of application for amendments:

June 15, 2001.

Brief description of amendments:

Eliminate the Technical Specifications

(TS) requirement that the Automatic

Depressurization System (ADS)

designated Safety/Relief Valves (S/RVs)

open during the manual actuation of the

ADS and rewords the Surveillance

Requirement (SR) 3.5.1.8 frequency to

require the testing of all required ADS

manual actuation solenoids during the

performance of SR 3.5.1.8 in place of

testing on a staggered basis.

Date of issuance: December 13, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 30 days.

Amendment Nos.: 151 and 137.

Facility Operating License Nos. NPF–

11 and NPF–18: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: August 8, 2001 (66 FR 41618).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated December 13,

2001.

No significant hazards consideration

comments received: No.

FirstEnergy Nuclear Operating

Company, et al., Docket Nos. 50–334

and 50–412, Beaver Valley Power

Station, Unit Nos. 1 and 2, Beaver

County, Pennsylvania

Date of application for amendments:

August 13, 2001.

Brief description of amendments:

These amendments delete Technical

Specifications (TS) Section 6.8.4, which

required a Post-Accident monitoring

program, for Beaver Valley Power

Station, Unit Nos. 1 and 2, and thereby

eliminate the requirements to have and

maintain the post-accident sampling

system (PASS) for those units.

Date of Issuance: December 6, 2001.

Effective date: Upon issuance and

shall be implemented within 180 days.

Amendment Nos.: 245, 123.

Facility Operating License Nos. DPR–

66 and NPF–73: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: September 19, 2001 (66 FR

48286).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated December 6,

2001.

No significant hazards consideration

comments received: No.

Indiana Michigan Power Company,

Docket Nos. 50–315 and 50–316, Donald

C. Cook Nuclear Plant, Units 1 and 2,

Berrien County, Michigan

Date of application for amendments:

May 17, 2001, as supplemented by letter

dated September 5, 2001.

Brief description of amendments: The

amendments revised Technical

Specification (TS) 3/4.9.3, ‘‘Decay

Time,’’ to allow the start of core offload

at 100 hours after reactor subcriticality

between September 15 and June 15,

when the lake temperature is assumed

to be not higher than 77.8°F, and 148

hours after reactor subcriticality

between June 16 and September 14,

when the lake temperature is assumed

to be not higher than 85°F. TS 3/4.9.3

currently prohibits fuel movement in

the reactor pressure vessel until the

reactor has been subcritical for at least

168 hours. The 168-hour decay time was

placed in the CNP TS with Amendment

Nos. 169 and 152 to DPR–58 and DPR–

74, respectively, on January 14, 1993.

Date of issuance: November 30, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 45 days.

Amendment Nos.: 260 and 243.

Facility Operating License Nos. DPR–

58 and DPR–74: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: August 22, 2001 (66 FR

44174)

The supplemental letter contained

clarifying information and did not

change the initial no significant hazards

consideration determination and did not

expand the scope of the original Federal

Register notice.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated November 30,

2001.

No significant hazards consideration

comments received: No.

Nuclear Management Company, LLC,

Docket No. 50–263, Monticello Nuclear

Generating Plant, Wright County,

Minnesota

Date of application for amendment:

August 30, 2001, as supplemented

October 10 and November 16, 2001.

Brief description of amendment: The

amendment revises the Technical

Specification safety limit minimum

critical power ratio for two recirculation

pump operation for Cycle 21.

Date of issuance: December 6, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 60 days.

Amendment No.: 125.

Facility Operating License No. DPR–

22: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 3, 2001 (66 FR 50470)

The October 10 and November 16,

2001, supplements provided clarifying

information that was within the scope of

the original Federal Register notice and

did not change the staff’s initial

proposed no significant hazards

considerations determination. The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated December 6, 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: February

7, 2001, as supplemented by letters

dated October 17 and November 2,

2001.

Brief description of amendment: The

requested changes replaced the current

accident source term used in the design

basis radiological analyses for control

room habitability with an alternative

source term (AST) pursuant to 10 CFR

50.67, ‘‘Accident Source Term.’’ OPPD

requested a full implementation of the

AST. Changes were also made to the Ft.

Calhoun Technical Specifications to

make them consistent with the revised

associated accident analysis.

Date of issuance: December 5, 2001.

Effective date: December 5, 2001, to

be implemented within 60 days from

the date of issuance.

Amendment No.: 201.

Facility Operating License No. DPR–

40: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: May 2, 2001 (66 FR 22031).

The October 17 and November 2,

2001, supplemental letters provided

additional information that clarified the

application, 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 December 5,

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: May 3,

2001.

Brief Description of amendments: The

amendments relocate cycle-specific

Federal Register / Vol. 66, No. 247 / Wednesday, December 26, 2001 / Notices 66477

reactor coolant system parameter limits

from the Technical Specifications (TS)

and associated Bases, to the Core

Operating Limits Report. The

amendments also, add a reference to the

Refueling Boron Concentration to TS

5.6.5 to correct an omission.

Date of issuance: December 4, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment Nos.: 151 and 143.

Facility Operating License Nos. NPF–

2 and NPF–8: Amendments revise the

Technical Specifications. Date of initial notice in Federal

Register: October 31, 2001 (66 FR

55024).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated December 4,

2001.

No significant hazards consideration

comments received: No.

STP Nuclear Operating Company,

Docket Nos. 50–498 and 50–499, South

Texas Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request: February

12, 2001.

Brief description of amendments: The

amendments consist of deleting

Surveillance Requirement 4.4.6.2.2.e of

South Texas Project Technical

Specifications Section 3/4.4.6.2. Date of issuance: December 11, 2001.

Effective date: As of the date of

issuance, and shall be implemented

within 30 days from the date of

issuance.

Amendment Nos.: Unit 1—134; Unit

2—123.

Facility Operating License Nos. NPF–

76 and NPF–80: The amendments

revised the Technical Specifications. Date of initial notice in Federal

Register: June 12, 2001 (66 FR 31715). The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated December 11,

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:

December 14, 2000.

Brief description of amendment:

These amendments revise Technical

Specifications Sections 4.7.7.1.d.1 and

4.7.7.2.a. These changes increase the

specified minimum number of

compressed bottles of air from 84 to 102,

and revise the differential pressure limit

across the Control Room Emergency

Ventilation System HEPA Filter,

demister filter, and charcoal adsorber.

Date of issuance: December 12, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment Nos.: 228 and 209.

Facility Operating License Nos. NPF–

4 and NPF–7: Amendments change the

Technical Specifications.

Date of initial notice in Federal

Register: January 24, 2001 (66 FR

7687).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated December 12,

2001.

No significant hazards consideration

comments received: No.

(Note: The publication date for this notice

will change from every other Wednesday to

every other Tuesday, effective January 8,

2002. The notice will contain the same

information and will continue to be

published biweekly.

Dated at Rockville, Maryland, this 17th day

of December, 2001.

For the Nuclear Regulatory Commission.

Ledyard B. Marsh,

Acting Director, Division of Licensing Project

Management, Office of Nuclear Reactor

Regulation.

[FR Doc. 01–31473 Filed 12–21–01; 8:45 am]

BILLING CODE 7590–01–P

OFFICE OF MANAGEMENT AND

BUDGET

Cost of Hospital and Medical Care

Treatment Furnished by the United

States; Certain Rates Regarding

Recovery From Tortiously Liable Third

Persons

By virtue of the authority vested in

the President by section 2(a) of Public

Law 87–693 (76 Stat. 593; 42 U.S.C.

2652), and delegated to the Director of

the Office of Management and Budget

by Executive Order No. 11541 of July 1,

1970 (35 FR 10737), the two sets of rates

outlined below are hereby established.

These rates are for use in connection

with the recovery, from tortiously liable

third persons, of the cost of hospital and

medical care and treatment furnished by

the United States (Part 43, Chapter I,

Title 28, Code of Federal Regulations)

through three separate Federal agencies.

The rates have been established in

accordance with the requirements of

OMB Circular A–25, requiring

reimbursement of the full cost of all

services provided and will remain in

effect until further notice. The rates for

VA that were published in the Federal

Register on October 31, 2000 remain in

effect until further notice. The rates are

as follows:

1. Department of Defense

The Department of Defense (DoD)

reimbursement rates for inpatient,

outpatient, and other services are

provided in accordance with Title 10,

United States Code, section 1095. Due to

size, the sections containing the Drug

Reimbursement Rates (section III.D.)

and the rates for Ancillary Services

Requested by Outside Providers (section

III.E.) are not included in this package.

Those rates are available from the

TRICARE Management Activity’s

Uniform Business Office web site:

http://www.tricare.osd.mil/ebc/

rm_home/imcp/ubo/ubo_01.htm. The

medical and dental service rates in this

package (including the rates for

ancillary services and other procedures

requested by outside providers) are

effective October 1, 2001. Pharmacy

rates are updated on an as needed basis.

2. Health and Human Services

The tortiously liable rates for Indian

Health Service health facilities are based

on Medicare cost reports. The

obligations for the Indian Health Service

hospitals participating in the cost report

project were identified and combined

with applicable obligations for area

offices costs and headquarters costs. The

hospital obligations were summarized

for each major cost center providing

medical services and distributed

between inpatient and outpatient. Total

inpatient costs and outpatient costs

were then divided by the relevant

workload statistic (inpatient day,

outpatient visit) to produce the

inpatient and outpatient rates. In

calculation of the rates, the

Department’s unfunded retirement

liability cost and capital and equipment

depreciation costs were incorporated to

conform to requirements set forth in

OMB Circular A–25.

In addition, the obligations for each

cost center include obligations from

certain other accounts, such as Medicare

and Medicaid collections and the

Contract Health fund, that were used to

support the inpatient and outpatient

workload. Obligations were excluded

for certain cost centers that primarily

support workloads outside of the

directly operated hospitals or clinics

(public health nursing, public health

nutrition, health education). These

obligations are not a part of the

traditional cost of hospital operations

and do not contribute directly to the

inpatient and outpatient visit workload.

Separate rates per inpatient day and

outpatient visit were computed for

Alaska and the rest of the United States.