59218 Federal Register / Vol. 65, No. 193 / Wednesday, October 4, 2000 / Notices

present, may exchange preliminary

views regarding matters to be

considered during the balance of the

meeting.

The Subcommittee will then hear

presentations by and hold discussions

with representatives of the NRC staff, its

consultants, and other interested

persons regarding this review.

Further information regarding topics

to be discussed, whether the meeting

has been canceled or rescheduled, the

Chairman’s ruling on requests for the

opportunity to present oral statements

and the time allotted therefor can be

obtained by contacting the cognizant

ACRS staff engineer, Dr. Medhat ElZeftawy (telephone 301/415–6889)

between 7:30 a.m. and 4:15 p.m. (EDT).

Persons planning to attend this meeting

are urged to contact the above named

individual one or two working days

prior to the meeting to be advised of any

potential changes in the proposed

agenda, etc., that may have occurred.

Date: September 27, 2000.

James E. Lyons,

Associate Director for Technical Support

ACRS/ACNW.

[FR Doc. 00–25461 Filed 10–3–00; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Sunshine Act Meeting

AGENCY HOLDING THE MEETING: Nuclear

Regulatory Commission.

DATES: Weeks of October 2, 9, 16, 23, 30,

and November 6, 2000.

PLACE: Commissioners’ Conference

Room, 11555 Rockville Pike, Rockville,

Maryland.

STATUS: Public and Closed.

MATTERS TO BE CONSIDERED: 

Week of October 2

Friday, October 6

9:25 a.m.

Affirmation Session (Public Meeting)

(If needed)

9:30 a.m.

Meeting with ACRS (Public Meeting)

(Contact: John Larkins, 301–415–

7360)

This meeting will be webcast live at the

Web address—

www.nrc.gov/live.html

Week of October 9—Tentative

There are no meetings scheduled for

the Week of October 9.

Week of October 16—Tentative

Tuesday, October 17

9:25 a.m.

Affirmation Session (Public Meeting)

(If needed)

Week of October 23—Tentative

Monday, October 23

1:55 p.m.

Affirmation Session (Public Meeting)

(If needed)

Week of October 30—Tentative

There are no meetings scheduled for

the week of October 30.

Week of November 6—Tentative

There are no meetings scheduled for

the Week of November 6.

The schedule for Commission

meetings is subject to change on short

notice. To verify the status of meetings

call (Recording)—)(301) 415–1292.

Contact Person for More Information:

Bill Hill (301) 415–1661.

The NRC Commission Meeting

Schedule can be found on the Internet

at:

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

schedule.htm

This notice is distributed by mail to

several hundred subscribers; if you no

longer wish to receive it, or would like

to be added to it, please contact the

Office of the Secretary, Attn: Operations

Branch, Washington, D.C. 20555 (301–

415–1661). In addition, distribution of

this meeting notice over the Internet

system is available. If you are interested

in receiving this Commission meeting

schedule electronically, please send an

electronic message to wmh@nrc.gov or

dkw@nrc.gov.

Dated: September 29, 2000.

William M. Hill, Jr.,

SECY Tracking Officer, Office of the

Secretary.

[FR Doc. 00–25564 Filed 10–2–00; 11:36 am]

BILLING CODE 7590–01–M

NUCLEAR REGULATORY

COMMISSION

Applications and Amendments to

Facility Operating Licenses Involving

No Significant Hazards

Considerations; Biweekly Notice

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 September

11, 2000, through September 22, 2000.

The last biweekly notice was published

on September 20, 2000 (65 FR 56946, as

corrected at 65 FR 57484 and 65 FR

58113).

Notice of Consideration of Issuance of

Amendments to Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a

proposed determination that the

following amendment requests involve

no significant hazards consideration.

Under the Commission’s regulations in

10 CFR 50.92, this means that operation

of the facility in accordance with the

proposed amendment would not (1)

involve a significant increase in the

probability or consequences of an

accident previously evaluated; or (2)

create the possibility of a new or

different kind of accident from any

accident previously evaluated; or (3)

involve a significant reduction in a

margin of safety. The basis for this

proposed determination for each

amendment request is shown below.

The Commission is seeking public

comments on this proposed

determination. Any comments received

within 30 days after the date of

publication of this notice will be

considered in making any final

determination.

Normally, the Commission will not

issue the amendment until the

expiration of the 30-day notice period.

However, should circumstances change

during the notice period such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received before

action is taken. Should the Commission

Federal Register / Vol. 65, No. 193 / Wednesday, October 4, 2000 / Notices 59219

take this action, it will publish in the

Federal Register a notice of issuance

and provide for opportunity for a

hearing after issuance. The Commission

expects that the need to take this action

will occur very infrequently.

Written comments may be submitted

by mail to the Chief, Rules Review and

Directives Branch, Division of Freedom

of Information and Publications

Services, Office of Administration, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555–0001, and

should cite the publication date and

page number of this Federal Register

notice. Written comments may also be

delivered to Room 6D22, Two White

Flint North, 11545 Rockville Pike,

Rockville, Maryland from 7:30 a.m. to

4:15 p.m. Federal workdays. Copies of

written comments received may be

examined at the NRC’s Public Document

Room, located at One White Flint North,

11555 Rockville Pike (first floor),

Rockville, Maryland 20852. The filing of

requests for a hearing and petitions for

leave to intervene is discussed below.

By November 3, 2000, 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

ADAMS Public Library component on

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

(the Electronic Reading Room). If a

request for a hearing or petition for

leave to intervene is filed by the above

date, the Commission or an Atomic

Safety and Licensing Board, designated

by the Commission or by the Chairman

of the Atomic Safety and Licensing

Board Panel, will rule on the request

and/or petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) The nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555–0001, Attention:

Docketing and Services Branch, or may

be delivered to the NRC’s Public

Document Room, One White Flint

North, 11555 Rockville Pike (first floor),

Rockville, Maryland 20852, by the

above date. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555–0001, and to the attorney for

the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

factors specified in 10 CFR

2.714(a)(1)(i)–(v) and 2.714(d).

For further details with respect to this

action, see the application for

amendment which is available for

public inspection at the 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

ADAMS Public Library component on

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

(the Electronic Reading Room).

Arizona Public Service Company, et al.,

Docket Nos. STN 50–528, STN 50–529,

and STN 50–530, Palo Verde Nuclear

Generating Station, Units 1, 2, and 3,

Maricopa County, Arizona

Date of amendments request: June 8,

2000.

Description of amendments request:

The licensee proposes to amend

Technical Specification (TS) 5.6.5,

‘‘Core Operating Limits Report (COLR),’’

to add a methodology using the

59220 Federal Register / Vol. 65, No. 193 / Wednesday, October 4, 2000 / Notices

CASMO–4 and SIMULATE–3 codes to

the list of analytical methods used to

determine core operating limits

contained in TS 5.6.5.b. The change

would allow the use of the CASMO–4

and SIMULATE–3 methodology to

perform nuclear design calculations.

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:

Standard 1—Does the proposed change

involve a significant increase in the

probability or consequences of an accident

previously evaluated?

No. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

Arizona Public Service Company (APS)

intends to replace the DIT/ROCS/MC

methodology with CASMO–4/SIMULATE–3

code package. The proposed amendment

would add methodology using CASMO–4

and SIMULATE–3 codes to the list of

analytical methods used to determine core

operating limits contained in Technical

Specification 5.6.5.b. This will allow the use

of the CASMO–4 and SIMULATE–3

methodology to perform all steady-state PWR

[pressurized-water reactor] core physics

analyses.

The probability of occurrence of an

accident previously evaluated will not be

increased by the proposed change in the

particular codes used for physics calculations

for nuclear design analysis. The results of

nuclear design analyses are used as inputs to

the analysis of accidents that are evaluated in

the Updated Final Safety Analysis Report

(UFSAR). These inputs do not alter the

physical characteristics or modes of

operation of any system, structure, or

component involved in the initiation of an

accident. Thus, there is no significant

increase in the probability of an accident

previously evaluated as a result of this

change.

The consequences of an accident evaluated

in the UFSAR are affected by the value of

inputs to the transient safety analysis. An

extensive benchmark of CASMO–4/

SIMULATE–3 predictions with measured

data using a variety of fuel designs and

operating conditions in power reactors and

critical experiments, was performed. The

accuracy of CASMO–4/SIMULATE–3 is

similar to, and sometimes better than, the

accuracy of DIT/ROCS/MC. Furthermore,

there is always the potential for the value of

the nuclear design parameters to change

solely as a result of the new reload fuel core

loading pattern. Regardless of the source of

a change, an assessment is always made of

changes to the nuclear design parameters

with respect to their effects on the

consequences of accidents previously

evaluated in the UFSAR. Refueling is an

anticipated activity which is described in the

UFSAR. If increased consequences are

anticipated, compensatory actions are

implemented to neutralize any expected

increase in consequences. These

compensatory actions include, but are not

limited to, crediting any existing margins in

the analysis or redefining the operating

envelope to avoid increased consequences.

Thus, the nuclear design parameters are

intermediate results and by themselves will

not result in an increase in the consequence

of an accident evaluated in the UFSAR.

Therefore, the replacement of the DIT/

ROCS/MC codes with the CASMO–4/

SIMULATE–3 code package, which will

perform the same functions as the DIT/

ROCS/MC codes with similar accuracy, does

not significantly increase the consequences

of an accident previously evaluated.

Standard 2—Does the proposed change

create the possibility of a new or different

kind of accident from any accident

previously evaluated?

No. The proposed change does not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

Arizona Public Service Company (APS)

intends to replace the DIT/ROCS/MC

methodology with CASMO–4/SIMULATE–3

code package. The proposed amendment

would add methodology using CASMO–4

and SIMULATE–3 codes to the list of

analytical methods used to determine core

operating limits contained in Technical

Specification 5.6.5.b.

The possibility for a new or different kind

of accident evaluated previously in the

UFSAR will not be created by the proposed

change to the particular codes used for

physics calculations for nuclear design

analyses. The change involves replacing the

NRC approved ABB Combustion Engineering

Nuclear Power (ABB/CE) DIT and ROCS/MC

codes, with the Studsvik CASMO–4 and

SIMULATE–3 codes. The results of nuclear

design analyses are used as inputs to the

analysis of accidents that are evaluated in the

UFSAR. These inputs do not alter the

physical characteristics or modes of

operation of any system, structure or

component involved in the initiation of an

accident.

Therefore, the replacement of the DIT/

ROCS/MC codes with the CASMO–4/

SIMULATE–3 code package, which will

perform the same functions as the DIT/

ROCS/MC codes with similar accuracy, does

not increase the possibility of a new or

different kind of accident from any accident

previously evaluated.

Standard 3—Does the proposed change

involve a significant reduction in a margin of

safety?

No. The proposed change does not involve

a significant reduction in a margin of safety.

The margin of safety as defined in the basis

for any technical specification will not be

reduced nor increased by the proposed

change to the particular codes used for

physics calculations for nuclear design

analyses. The change involves replacing the

NRC approved ABB/CE DIT and ROCS/MC

codes, with the Studsvik CASMO–4 and

SIMULATE–3 codes. Extensive

benchmarking of the CASMO–4/SIMULATE–

3 computer codes has demonstrated that the

values of those parameters used in the safety

analysis are not significantly changed relative

to the values obtained using the DIT/ROCS/

MC computer codes. For any changes in the

calculated values that do occur, the

application of appropriate biases and

uncertainties ensures that the current margin

of safety is maintained. Specifically, use of

these code specific biases and uncertainties

in safety evaluations continues to provide the

same statistical assurance that the values of

the nuclear parameters used in the safety

analysis are conservative with respect to the

actual values on at least a 95/95 probability/

confidence basis.

The NRC staff has reviewed the

licensee’s analysis and, based on that

review, it appears that the three

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: Nancy C. Loftin,

Esq., Corporate Secretary and Counsel,

Arizona Public Service Company, P.O.

Box 53999, Mail Station 9068, Phoenix,

Arizona 85072–3999

NRC Section Chief: Stephen Dembek.

Arizona Public Service Company, et al.,

Docket Nos. STN 50–528, STN 50–529,

and STN 50–530, Palo Verde Nuclear

Generating Station, Units 1, 2, and 3,

Maricopa County, Arizona

Date of amendments request: June 16,

2000.

Description of amendments request:

The licensee proposes to amend

Technical Specification (TS) Table

3.3.10–1, ‘‘Post Accident Monitoring

Instrumentation,’’ to add the High

Pressure Safety Injection cold leg flow

and hot leg flow instrumentation to this

table. This change is required because

this instrumentation meets the criteria

for a Regulatory Guide 1.97,

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

Plant and Environs Conditions During

and Following an Accident,’’ Revision

2, Type A, Category 1 variable.

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:

Standard 1—Does the proposed change

involve a significant increase in the

probability or consequences of an accident

previously evaluated?

No. The proposed change to TS Table

3.3.10–1, by adding the High Pressure Safety

Injection (HPSI) hot and cold leg flow

instrumentation, does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated because it does not represent a

change to design configuration or operation

of the plant. The amendment does not affect

Federal Register / Vol. 65, No. 193 / Wednesday, October 4, 2000 / Notices 59221

the operability or availability of the HPSI

system or any other safety related equipment.

Additionally, there are no effects on the

failure modes associated with the probability

of a failure of a system important to safety.

Standard 2—Does the proposed change

create the possibility of a new or different

kind of accident from any accident

previously evaluated?

No. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated because the change

does not impact the response or operation of

the plant. The availability and operability of

the plant equipment is unchanged, as the

design requirements have not changed.

The proposed change revises only the

Regulatory Guide (RG) 1.97, Revision 2,

classification of the HPSI hot leg and cold leg

flow indication loops. Regardless of RG

classification the instruments remain

seismically, electrically, and otherwise

qualified for the application. Hence, the

revised classification will not subject these

components to new modes of operation that

could result in a new failure mode, thus

initiating an accident of a different type.

Standard 3—Does the proposed change

involve a significant reduction in a margin of

safety?

No. This proposed amendment does not

involve a significant reduction in the margin

of safety because neither of the following

PVNGS Technical Specification (TS) Bases (B

3.5.3 ECCS [emergency core cooling

system]—Operating, or 3.3.10 Post Accident

Monitoring (PAM) Instrumentation) is

changed by the proposed amendment.

TS Bases B 3.5.3 ECCS—Operating—states

that the function of the ECCS is to provide

core cooling and negative reactivity to ensure

that the reactor core is protected after any of

the following accidents:

a. Loss of Coolant Accident (LOCA);

b. Control Element Assembly (CEA)

ejection accident;

c. Loss of secondary coolant accident,

including uncontrolled steam release or loss

of feedwater; and

d. Steam Generator Tube Rupture (SGTR).

Changing the RG 1.97 Type and Category

of these instruments does not affect the

ability of the ECCS to provide core cooling

and negative reactivity during these

accidents.

TS Bases B 3.3.10—Post Accident

Monitoring (PAM) Instrumentation—states

that the primary purpose of PAM

instrumentation is to display plant variables

that provide information required by the

control room operators during accident

situations. This information provides the

necessary support for the operator to take the

manual actions, for which no automatic

control is provided, that are required for

safety systems to accomplish their safety

functions for Design Basis Events.

The OPERABILITY of PAM

instrumentation ensures that there is

sufficient information available on selected

plant parameters to monitor and assess plant

status and behavior following an accident.

These Type A variables are required to be

included in this LCO [Limiting Condition for

Operation] because they provide the primary

information required to permit the control

room operator to take specific manually

controlled actions, for which no automatic

control is provided, that are required for

safety systems to accomplish their safety

functions for Design Basis Accidents (DBAs).

The addition of these instruments supports

this TS Bases. 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 that

review, it appears that the three

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: Nancy C. Loftin,

Esq., Corporate Secretary and Counsel,

Arizona Public Service Company, P.O.

Box 53999, Mail Station 9068, Phoenix,

Arizona 85072–3999.

NRC Section Chief: Stephen Dembek.

Consolidated Edison Company of New

York, Docket No. 50–247, Indian Point

Nuclear Generating Unit No. 2,

Westchester County, New York

Date of amendment request: May 5,

1999, as supplemented on December 22,

1999, and September 18, 2000.

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) 3.5,

‘‘Instrumentation Systems,’’ for the

reactor protection system and

engineered safety features actuation

system instrumentation. Specifically,

the proposed amendment would (1)

change the allowed outage times for the

instrumentation and the analog channel

test bypass time and (2) allow on-line

testing and maintenance of

instrumentation. The proposed

amendment also includes several

editorial changes to TS Tables 3.5–2 and

3.5–3. The proposed amendment was

originally noticed in the Federal

Register on September 8, 1999 (64 FR

48861). It is now being noticed to

correct errors made in the original

notice description of amendment

request.

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 reactor protection and engineered

safety features functions are not initiators of

any design basis accident or event and

therefore do not increase the probability of

any accident previously evaluated. The

proposed changes to the AOTs [allowed

outage times], bypass times, and allowing online testing and maintenance have an

insignificant impact on plant safety based on

the calculated CDF [core damage frequency]

increase being less than 1.0E–06. Therefore,

the proposed changes do not result in 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 changes do not result in a

change in the manner in which the RPS

[reactor protection system] and ESFAS

[engineered safety features actuation system]

provide plant protection. No change is being

made which alters the functioning of the RPS

and ESFAS. Rather, the likelihood or

probability of the RPS or ESF functioning

properly is affected as described above.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident nor involve a reduction in

the margin of safety as defined in the Safety

Analysis Report.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed changes do not alter the

manner in which safety limits, limiting safety

system setpoints or limiting conditions for

operations are determined. The impact of

increased AOTs, testing times, and allowing

on-line testing and maintenance are expected

to result in an overall improvement in safety

because:

The longer AOTs for the master relays,

logic cabinets, and analog channels will

promote improved maintenance practices

that will provide improved component

performance, improved availability of the

protection system, and a reduced number of

spurious reactor trips and spurious actuation

of safety equipment.

The longer AOTs and bypass times for the

analog channels will provide additional time

before being required to place the channel in

trip. With the channel in trip, the logic

required to cause a reactor trip or a safety

system actuation is reduced to 1 of 2 (for 2

of 3 logic) and to 1 of 3 (for 2 of 4 logic). With

the reduced logic requirement, the potential

for a spurious actuation is increased. Leaving

the channel in the bypass state for additional

time does reduce the availability of signals to

initiate component actuation for event

mitigation when required, but as shown in

this analysis, the impact on plant safety is

small due to the availability of other signals

or operator action to trip the reactor or cause

component actuation.

The longer allowed outage times will

provide plant operators additional flexibility

in operating the plant. There will be

additional time available before an action

needs to be taken to shut down the plant or

place a channel in the tripped state. This

additional flexibility will facilitate

prioritizing component repairs.

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

59222 Federal Register / Vol. 65, No. 193 / Wednesday, October 4, 2000 / Notices

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Brent L.

Brandenburg, Esquire, 4 Irving Place,

New York, New York 10003.

NRC Section Chief: Marsha

Gamberoni.

Duke Energy Corporation, Docket Nos.

50–269, 50–270, and 50–287, Oconee

Nuclear Station, Units 1, 2, and 3,

Oconee County, South Carolina

Date of amendment request:

September 12, 2000.

Description of amendment request:

The proposed amendments would

revise the Technical Specifications

5.5.10, Item e.6, Steam Generator Tube

Surveillance Program, by (1) removing

the restriction on the lower tube sheet

area rolling, (2) removing the limitation

of only one reroll per steam generator

tube, (3) eliminating the requirement

that the reroll be one inch in depth, and

(4) changing the revision number

reference for Topical Report BAW–

2303P, August 2000, ‘‘OTSG Repair Roll

Qualification Report,’’ from Revision 3

to Revision 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:

Duke Energy Corporation (Duke) has made

the determination that this amendment

request involves a No Significant Hazards

Consideration by applying the standards

established by NRC regulations in

10CFR50.92. This ensures operation of the

facility in accordance with the proposed

amendment will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated?

No. The proposed change to the Technical

Specifications incorporates Revision 4 of

Topical Report BAW–2303P, OTSG Repair

Roll Qualification Report. This document is

also being submitted for NRC review and

approval. This revision addresses, and is

consistent with, the conclusions of all

applicable Oconee licensing basis analyses

and ensures that previously evaluated

accidents are bounding. All the established

acceptance criteria for the accidents analyzed

in the Oconee licensing basis continue to be

met. Therefore, no existing accident

probabilities or consequences will be

impacted.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated?

No. Revision 4 of BAW–2303P addresses

limiting events for steam generator tube reroll

repairs. These events include Main Steam

Line Break, the Small Break Loss of Coolant

Accident, and other transients on B&W OnceThrough Steam Generators. For Oconee, the

Main Steam Line Break is the limiting event.

This revised topical report confirms the

acceptability of the reroll repair techniques

previously used at Oconee. As a result, no

new failure modes are being created. BAW–

2303P, as submitted for NRC review and

approval, does not create the possibility of a

new or different kind of accident.

3. Involve a significant reduction in a

margin of safety?

No. 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. As part of the reactor

coolant system pressure boundary, the steam

generator tubes are unique in that they are

also relied upon as a heat transfer surface

between the primary and secondary systems,

such that residual heat can be removed from

the primary system. In addition, the steam

generator tubes also isolate the radioactive

fission products in the primary coolant from

the secondary system. Finally, the steam

generator tubes may be relied upon to

maintain their integrity under conditions

resulting from core damage severe accidents

consistent with the containment objectives of

preventing uncontrolled fission product

release. The functions of the steam generator

tubes will not be significantly affected by the

changes proposed in this license amendment

request. Implementation of BAW–2303P,

Revision 4, as submitted for NRC review and

approval, at Oconee will result in assurance

that parameters affecting the integrity of the

steam generator tubes continue to meet

applicable safety analyses and industry codes

and standards. Therefore, no safety margin

will be significantly reduced.

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: Anne W.

Cottington, Winston and Strawn, 1200

17th Street, NW., Washington, DC

20005.

NRC Section Chief: Richard L. Emch,

Jr.

Florida Power and Light Company, et

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

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

County, Florida

Date of amendment request: July 26,

2000.

Description of amendment request:

The amendments would revise the

Technical Specifications (TS) Index to

delete reference to the Bases since, in

accordance with 10 CFR 50.36(a), the

Bases are not a part of the TS. Future

changes to the TS Bases will be

evaluated per 10 CFR 50.59 and made

under administrative control and

reviews and in accordance with the

proposed TS Bases control program as

described in TS 5.5.14 of NUREG–1432,

Revision 1, ‘‘Standard Technical

Specifications Combustion Engineering

Plants.’’

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) Operation of the facility in accordance

with the proposed amendments would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed amendments are

administrative in nature and do not affect

assumptions contained in plant safety

analyses, the physical design and operation

of the plant, nor do they affect Technical

Specifications that preserve safety analysis

assumptions. The Technical Specification

BASES, per 10 CFR 50.36(a), are not part of

the Technical Specifications. Changes to the

TS BASES will be controlled by a plant

procedure under administrative controls and

reviews. Proposed changes to the TS BASES

will be evaluated in accordance with 10 CFR

50.59 and made under the programmatic

controls and requirements of the proposed

Technical Specifications (TS) Bases Control

Program. Therefore, the proposed changes do

not increase the probability or consequences

of accidents previously analyzed.

(2) Operation of the facility in accordance

with the proposed amendments would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed amendments are

administrative in nature. The proposed

amendments will not create the possibility of

a new or different kind of accident from any

accident previously evaluated since the

proposed amendments will not change the

physical plant or the modes of plant

operation defined in the facility operating

license. No new failure mode is introduced

due to the administrative change, since the

proposed change does not involve the

addition or modification of equipment nor

does it alter the design or operation of

affected plant systems, structures, or

components.

(3) Operation of the facility in accordance

with the proposed amendments would not

involve a significant reduction in a margin of

safety.

The operating limits and functional

capabilities of the affected systems,

structures, and components are unchanged

by the proposed amendments. The BASES

information, per 10 CFR 50.36(a), is not a

part of the Technical Specifications. Changes

to the TS BASES will be controlled by a plant

procedure under administrative controls and

reviews and made under the programmatic

controls and requirements of the proposed

Technical Specifications (TS) Bases Control

Program. Proposed changes to the TS BASES

will be evaluated in accordance with 10 CFR

50.59 and the TS BASES will be maintained

Federal Register / Vol. 65, No. 193 / Wednesday, October 4, 2000 / Notices 59223

in an FPL-controlled document. Therefore,

the proposed changes do not reduce any

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Attorney for licensee: M.S. Ross,

Attorney, Florida Power & Light, P.O.

Box 14000, Juno Beach, Florida 33408–

0420.

NRC Section Chief: Richard P.

Correia.

Florida Power Corporation, et al.,

Docket No. 50–302, Crystal River

Nuclear Generating Plant, Unit No. 3,

Citrus County, Florida

Date of amendment request: August

31, 2000.

Description of amendment request:

This amendment would revise

Improved Technical Specification (ITS)

Table 3.3.18–1, ‘‘Remote Shutdown

System Instrumentation.’’ The table

would be updated to reflect plant

modifications and procedure changes

regarding placing and maintaining the

plant in a safe shutdown condition if

the control room becomes inaccessible.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The proposed changes do not involve a

significant hazards consideration. In support

of this conclusion, the following analysis is

provided:

(1) Does not involve a significant increase

in the probability or consequences of an

accident previously analyzed.

The instruments listed in Table 3.3.18–1

are used to provide information on selected

parameters to the operators that will allow

them to place and maintain the plant in a

safe shutdown condition in the event the

control room becomes inaccessible. The

proposed license amendment revises Table

3.3.18–1, Remote Shutdown System

Instrumentation, to more accurately reflect

the instruments that would be used by the

operators to perform abnormal operating

procedure AP–990, Shutdown from Outside

the Control Room. The proposed license

amendment also revises ITS Bases Section B

3.3.18 to add a table that identifies, by

equipment tag number, the specific

instruments used to satisfy the requirements

of ITS 3.3.18 and ITS Table 3.3.18–1. The

instruments identified in ITS Table 3.3.18–1

and ITS Bases Table B 3.3.18–1 are not

initiators of any design basis accidents. The

design functions of the Remote Shutdown

System Instrumentation and the initial

conditions for accidents that require the

Remote Shutdown System will not be

effected by the change. Therefore, the change

will not increase the probability or

consequences of an accident previously

evaluated.

(2) Does not create the possibility of a new

or different kind of accident from any

accident previously analyzed.

The proposed amendment involves no

changes to the CR–3 design or to the

functions or operation of the Remote

Shutdown System. The proposed amendment

will ensure that sufficient and appropriate

instrumentation is available to allow the

operators to place and maintain the plant in

a safe shutdown condition in the event the

control room becomes inaccessible. The

proposed amendment will also add

information to Bases Section B 3.3.18 that

will ensure timely and accurate operability

evaluations and entry into the appropriate

Conditions and Required Actions of ITS

3.3.18. The proposed amendment will not

create any new plant configurations different

from those already analyzed. Therefore, the

proposed change will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

(3) Does not involve a significant reduction

in the margin of safety.

The proposed amendment revises Table

3.3.18–1, Remote Shutdown System

Instrumentation, to more accurately reflect

the instruments that would be used by the

operators to perform a shutdown from

outside the control room. The proposed

amendment will revise ITS Bases Section B

3.3.18 to provide the operators with guidance

that will assist them in making timely and

accurate operability determinations and

entries into the appropriate Conditions and

Required Actions for ITS 3.3.18. The

proposed changes will not reduce the ability

of the Remote Shutdown System to monitor

and control reactivity, RCS [reactor coolant

system] pressure, core heat removal, or RCS

inventory. Thus, the proposed amendment

will not result in a 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: R. Alexander

Glenn, General Counsel, Florida Power

Corporation, MAC—A5A, P.O. Box

14042, St. Petersburg, Florida 33733–

4042.

NRC Section Chief: Richard P.

Correia.

Pacific Gas and Electric Company,

Docket No. 50–323, Diablo Canyon

Nuclear Power Plant, Unit No. 2, San

Luis Obispo County, California

Date of amendment requests: June 19,

2000.

Description of amendment requests:

The proposed license amendment

would revise Technical Specifications

5.5.9, ‘‘Steam Generator (SG) Tube

Surveillance Program,’’ and 5.6.10, ‘‘SG

Tube Inspection Report,’’ to add new

surveillance and reporting requirements

associated with a SG tube inspection

and repair. The new requirements

establish alternate repair criteria for

axial primary water stress corrosion

cracking at dented tube support plate

intersections.

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.

Examination of crack morphology for

primary water stress corrosion cracking

(PWSCC) at dented intersections has been

found to show one or two microcracks well

aligned with only a few uncorroded

ligaments and little or no other inside

diameter axial cracking at the intersection.

This relatively simple morphology is

conducive to obtaining good accuracy in

nondestructive examination (NDE) sizing of

these indications. Accordingly, alternate

repair criteria (ARC) is established based on

crack length and average and maximum

depth within the thickness of the tube

support plate (TSP).

The application of the ARC requires a

Monte Carlo condition monitoring

assessment to determine the as-found

condition of the tubing. The condition

monitoring analysis described in WCAP–

15128 Revision 3 is consistent with NRC

Generic Letter 95–05 requirements.

The application of the ARC requires a

Monte Carlo operational assessment to

determine the need for tube repair. The

repair bases are obtained by projecting the

crack profile to the end of the next operating

cycle and determining the burst pressure and

leakage for the projected profile using Monte

Carlo analysis techniques described in

WCAP–15128 Revision 3. The burst pressure

and leakage is compared to the requirements

in WCAP–15128 Revision 3. Separate

analyses are required for the total crack

length and the length outside the TSP due to

differences in requirements. If the projected

end of cycle (EOC) requirements are satisfied,

the tube will be left in service.

A steam generator (SG) tube rupture event

is one of a number of design basis accidents

that are analyzed as part of a plant’s licensing

basis. A single or multiple tube rupture event

would not be expected in a SG in which the

ARC has been applied. The ARC requires

repair of any indication having a maximum

crack depth greater than or equal to 40

percent outside the TSP, thus limiting the

potential length of a deep crack outside the

TSP at EOC conditions and providing margin

59224 Federal Register / Vol. 65, No. 193 / Wednesday, October 4, 2000 / Notices

against burst and leakage for free span

indications.

For other design basis accidents such as a

main steam line break, main feed line break,

control rod ejection, and locked reactor

coolant pump motor, the tubes are assumed

to retain their structural integrity.

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.

Implementation of the proposed SG tube

ARC does not introduce any significant

changes to the plant design basis. A single or

multiple tube rupture event would not be

expected in a SG in which the ARC has been

applied. Both condition monitoring and

operational assessments are completed as

part of the implementation of ARC to

determine that structural and leakage margin

exists prior to returning SGs to service

following inspections. If the condition

monitoring requirements are not satisfied for

burst or leakage, the causal factors for EOC

indications exceeding the expected values

will be evaluated. The methodology and

application of this ARC will continue to

ensure that tube integrity is maintained

during all plant conditions consistent with

the requirements of Regulatory Guide (RG)

1.121 and Revision 1 of RG 1.83. Therefore,

a permanent ARC is justified.

In the analysis of a SG tube rupture event,

a bounding primary-to-secondary leakage rate

equal to the operational leakage limits in the

Technical Specifications (TS), plus the leak

rate associated with the double ended

rupture of a single tube, is assumed. For

other design basis accidents, the tubes are

assumed to retain their structural integrity

and exhibit primary-to-secondary leakage

within the limits assumed in the current

licensing basis accident analyses. Steam line

break leakage rates from the proposed

PWSCC ARC are combined with leakage rates

from other approved ARC (i.e., voltage-based

ARC and W* ARC). The combined leakage

rates will not exceed the limits assumed in

the current licensing basis accident analyses.

The 40 percent maximum depth repair

limit for free span indications provides a very

low likelihood of free span leakage under

design basis or severe accident conditions.

Leakage from indications inside the TSP is

limited by the constraint of the TSP even

under severe accident conditions, and

leakage behavior in a severe accident would

be similar to that found acceptable by the

NRC under approved ARC for axial outside

diameter stress corrosion cracking (ODSCC)

at TSP intersections. Therefore, even under

severe accident conditions, it is concluded

that application of the proposed ARC for

PWSCC at dented TSP locations results in a

negligible difference in risk of a tube rupture

or large leakage event, when compared to

current 40 percent repair limits or previously

approved ARC.

DCPP continues to implement a maximum

operating condition leak rate limit of 150

gallons per day per SG to preclude the

potential for excessive leakage during all

plant conditions.

The possibility of a new or different kind

of accident from any previously evaluated is

not created because SG tube integrity is

maintained by inservice inspection,

condition monitoring, operational

assessment, tube repair, and primary-tosecondary leakage monitoring.

3. The proposed change does not involve

a significant reduction in a margin of safety.

Tube repair limits provide reasonable

assurance that tubes accepted for continued

service without plugging or repair will

exhibit adequate tube structural and leakage

integrity during subsequent plant operation.

The implementation of the proposed ARC is

demonstrated to maintain SG tube integrity

consistent with the criteria of draft NRC

Regulatory Guide 1.121. The guidelines of RG

1.121 describe a method acceptable to the

NRC staff for meeting General Design Criteria

(GDC) 2, 4, 14, 15, 31, and 32 by ensuring

the probability or the consequences of SG

tube rupture remain within acceptable limits.

This is accomplished by determining the

limiting conditions of degradation of SG

tubing, for which tubes with unacceptable

cracking should be removed from service.

Upon implementation of the proposed

ARC, even under the worst-case conditions,

the occurrence of PWSCC at the tube support

plate elevations is not expected to lead to a

SG tube rupture event during normal or

faulted plant conditions. The ARC involves

a computational assessment to be completed

for each indication left in service ensuring

that performance criteria for tube integrity

and leak tightness are met until the next

scheduled outage. Therefore, a permanent

ARC is justified.

As discussed below, certain tubes are

excluded from application of ARC. Existing

tube integrity requirements apply to these

tubes, and the margin of safety is not

reduced.

In addressing the combined loading effects

of a loss-of-coolant (LOCA) and safe

shutdown earthquake (SSE) on the SGs (as

required by GDC 2), the potential exists for

yielding of the TSP in the vicinity of the

wedge groups, accompanied by deformation

of tubes and a subsequent postulated inleakage. Tube deformation could lead to

opening of pre-existing tight through wall

cracks, resulting in secondary to primary inleakage following the event, which could

have an adverse affect on the Final Safety

Analysis Report (FSAR) results. Based on a

DCPP analysis of LOCA and SSE, SG tubes

located in wedge region exclusion zones are

susceptible to deformation, and are excluded

from application of ARC.

A DCPP tube stress analysis for feed line

break (FLB)/steam line break (SLB) plus SSE

loading determined that high bending

stresses occur in certain SG tubes at the

seventh TSP, because the stresses exceed the

maximum imposed bending stress for

existing test data (equal to approximately the

lower tolerance limit yield stress). These

tubes are located in rows 11 to 15 and 36 to

46, and are excluded from application of

ARC.

Tube intersections that contain TSP

ligament cracking are also excluded from

application of ARC.

Based on the above, it is concluded that the

proposed license amendment requires does

not result in a significant reduction in margin

[of safety] with respect to the plant safety

analyses as defined in the FSAR or TS.

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 requests involve no

significant hazards consideration.

Attorney for licensee: Christopher J.

Warner, Esq., Pacific Gas and Electric

Company, P.O. Box 7442, San

Francisco, California 94120.

NRC Section Chief: Stephen Dembek.

Tennessee Valley Authority, Docket

Nos. 50–259, 50–260 and 50–296,

Browns Ferry Nuclear Plant (BFN),

Units 1, 2 and 3, Limestone County,

Alabama

Date of amendment request: August

28, 2000.

Description of amendment request:

The proposed amendment would revise

the Units 1, 2 and 3 Technical

Specifications (TS) to incorporate

Technical Specifications Task Force

(TSTF) Nos.TSTF–71, TSTF–208,

TSTF–222, TSTF–284, TSTF–258 and

TSTF–364. TSTFs are changes that were

submitted to the staff by the nuclear

power industry TSTF that have generic

applicability. A description of each of

the six TSTFs follows: (1) TSTF–71,

Revision 2, adds an example of the

application of the Safety Function

Determination Program (SFDP) to the

Bases for Limiting Conditions for

Operation (LCO) 3.0.6. (2) TSTF–208,

Revision 0, extends the allowed time to

reach MODE 2 in LCO 3.0.3 from 7

hours to 10 hours. The change is based

on plant experience regarding the time

needed to perform a controlled

shutdown in an orderly manner. (3)

TSTF–222, Revision 1, clarifies

Improved Technical Specification (ITS)

Section 3.1.4, Control Rod Scram Times,

Surveillance Requirements (SRs) to

better delineate the requirements for

testing control rods following refueling

outages and for control rods requiring

testing due to work activities. (4) TSTF–

258, Revision 4, revises TS Section 5.0,

Administrative Controls, to delete

specific TS staffing requirement

provisions for Reactor Operators (ROs),

eliminates TS details for working hour

limits, clarifies requirements for the

Shift Technical Advisor (STA) position,

adds regulatory definitions for Senior

ROs and ROs, revises the Radioactive

Effluent Controls Program to be

consistent with the intent of 10 CFR Part

Federal Register / Vol. 65, No. 193 / Wednesday, October 4, 2000 / Notices 59225

20, deletes periodic reporting

requirements for mainsteam relief valve

openings, and revises radiological area

control requirements for radiation areas

to be consistant with those specified in

10 CFR 20.1601(c). (5) TSTF–284,

Revision 3, modifies Improved ITS

Section 1.4, Frequency, to clarify the

usage of the terms ‘‘met’’ and

‘‘performed’’ to facilitate the application

of SR Notes. Two new SR Examples,

1.4–5 and 1.4–6, are added to illustrate

the application of the terms. (6) TSTF–

364, Revision 0, revises Section 5.5.10,

TS Bases Control Program, to reference

10 CFR 50.59 rather than ‘‘unreviewed

safety question.’’ Also, editorial change

WOG–ED–24, which substitutes

‘‘require’’ for ‘‘involve’’ in 5.5.10.b is

made for consistency in usage.

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:

(1) TSTF–71, Revision 2.

A. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The change adds an example of SFDP use

to facilitate the application of the TS, which

serves to improve TS usefulness. The

proposed change is an administrative

clarification of existing requirements, and

does not change TS requirements. Therefore,

the proposed amendment does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

B. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed change does not involve a

physical alteration of the plant, add any new

equipment, or require any existing

equipment to be operated in a manner

different from the present design. The

proposed change will not impose any new or

eliminate any existing requirements.

Therefore, the proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

C. The proposed amendment does not

involve a significant reduction in a margin of

safety.

The proposed change will not reduce a

margin of safety because it has no effect on

any safety analyses assumptions. This change

is administrative in nature. For these reasons,

the proposed amendment does not involve a

significant reduction in the margin of safety.

(2) TSTF–208, Revision 0.

A. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated. The proposed change

relaxes the Action time for LCO 3.0.3. The

subject Action time is not an initiating

condition for any accident previously

evaluated and the accident analyses do not

assume that equipment is out of service

(requiring entry into LCO 3.0.3) prior to

postulated events. Consequently, the

extended action time does not significantly

increase the probability of an accident

previously evaluated. The consequences of

an analyzed accident during the extended

action time are the same as the consequences

during the existing action time. As a result,

the consequences of an accident previously

evaluated are not significantly increased.

Therefore, the proposed amendment does

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

B. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed change does not involve a

physical alteration of the plant, add any new

equipment, or require any existing

equipment to be operated in a manner

different from the present design. Therefore,

the proposed amendment does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

C. The proposed amendment does not

involve a significant reduction in a margin of

safety.

The proposed change will not reduce a

margin of safety because it has no effect on

any safety analyses assumptions. The TS

defines specific time limits during which

operation with degraded condition is

permitted. In this case, actual plant

experience indicates that the Action time in

existing TS is too short to accomplish the

specified action to be in MODE 2 in an

orderly manner. Extension of the time would

allow the reactor to be shutdown in a

controlled manner while minimizing risks

associated with the initiation of inadvertent

transients. This maximizes reactor safety. For

these reasons, the proposed amendment does

not involve a significant reduction in the

margin of safety.

(3) TSTF–222, Revision 1.

A. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change is an administrative

clarification of existing TS requirements

which clarifies scram time testing

requirements for control rods. The rewording

and reformatting involves no technical

changes to the existing TS. As such, there is

no effect on initiators of analyzed events or

assumed mitigation of accidents or

transients. Therefore, the proposed

amendment does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

B. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed change does not involve a

physical alteration of the plant, add any new

equipment, or require any existing

equipment to be operated in a manner

different from the present design. Therefore,

the proposed amendment does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

C. The proposed amendment does not

involve a significant reduction in a margin of

safety.

The proposed change will not reduce a

margin of safety because it has no effect on

any safety analyses assumptions. This change

is administrative in nature. For these reasons,

the proposed amendment does not involve a

significant reduction in the margin of safety.

(4) TSTF–258, Revision 4.

A. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change is an administrative

clarification of existing TS requirements

which clarifies and modifies administrative

controls in the areas of operator staffing

requirements, working hour limits, STA

position, Radioactive Effluent Controls

Program, periodic reporting requirements for

relief valve openings, and radiological

control requirements. These TS revisions do

not affect analysis inputs for analyzed

accidents and transients. Therefore, the

proposed amendment does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

B. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed change does not involve a

physical alteration of the plant, add any new

equipment, or require any existing

equipment to be operated in a manner

different from the present design. Therefore,

the proposed amendment does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

C. The proposed amendment does not

involve a significant reduction in a margin of

safety.

The proposed changes are administrative

type revisions and do not reduce a margin of

safety because they have no effect on any

safety analyses assumptions. For these

reasons, the proposed amendment does not

involve a significant reduction in the margin

of safety.

(5) TSTF–284, Revision 3.

A. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change is an administrative

clarification of existing requirements. The

change clarifies the TS terminology to

facilitate the use and application of

Surveillance Requirement Notes to improve

TS use. Also, two additional examples of the

application of Surveillance Requirement

Notes are incorporated. Therefore, the

proposed amendment does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

B. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

59226 Federal Register / Vol. 65, No. 193 / Wednesday, October 4, 2000 / Notices

The proposed change does not involve a

physical alteration of the plant, add any new

equipment, or require any existing

equipment to be operated in a manner

different from the present design. The

proposed change will not impose any new or

eliminate any existing requirements.

Therefore, the proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

C. The proposed amendment does not

involve a significant reduction in a margin of

safety.

The proposed change will not reduce a

margin of safety because it has no effect on

any safety analyses assumptions. This change

is administrative in nature.

For these reasons, the proposed

amendment does not involve a significant

reduction in the margin of safety.

(6) TSTF–364, Revision 0.

A. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change is an administrative

modification of existing TS requirements for

the TS Bases change program to simply

reference changes pursuant to 10 CFR 50.59

rather than ‘‘unreviewed safety question’’.

This change is administrative and has no

affect on the current review and approval

process for Final Safety Analyses Report and

Bases changes. As such, there is no effect on

initiators of analyzed events or assumed

mitigation of accidents or transients.

Therefore, the proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

B. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated. The proposed change

does not involve a physical alteration of the

plant, add any new equipment, or require

any existing equipment to be operated in a

manner different from the present design.

Therefore, the proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

C. The proposed amendment does not

involve a significant reduction in a margin of

safety.

The proposed change modifies [sic] is an

administrative modification of existing TS

requirements for the TS FSAR and Bases

change program to simply reference changes

pursuant to 10 CFR 50.59 rather than

‘‘unreviewed safety question.’’ This change is

administrative and has no affect on the

current review process for FSAR and Bases

changes, and will not reduce a margin of

safety because it has no effect on any safety

analyses assumptions.

For these reasons, the proposed

amendment 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: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 10H,

Knoxville, Tennessee 37902.

NRC Section Chief: Richard P.

Correia.

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of application for amendments:

August 31, 2000 (TS 00–05).

Brief description of amendments: The

proposed amendment would revise the

Sequoyah Nuclear Plant (SQN)

Technical Specifications (TSs). The

revision would relocate certain

specifications related to reactivity

control that are not required to be

contained in the TSs by NRC

regulations. These specifications

include TSs 3.1.2.1 and 3.1.2.2 for

boration flow paths, TSs 3.1.2.3 and

3.1.2.4 for boration charging pumps, TSs

3.1.2.5 and 3.1.2.6 for borated water

sources, TS 3.1.3.3 for position

indication systems during shutdown,

and TS 3.10.5 for special test exceptions

for the position indication system.

These specifications will be relocated in

their entirety to the SQN Technical

Requirements Manual without changing

the requirements currently contained in

the TSs.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a),

Tennessee Valley Authority (TVA), the

licensee, has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

A. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed revision relocates the

boration specifications and one rod position

indication system specification without a

change to the requirements and deletes a

special test exception for rod position

indication that is no longer applicable to

SQN. Relocation to the TRM continues to

provide an acceptable level of applicability to

plant operation and requires revisions to be

processed in accordance with the provisions

in 10 CFR 50.59. Evaluations of revisions in

accordance with 10 CFR 50.59 will continue

to ensure that these specifications adequately

control the functions for boration and rod

position indication systems to maintain safe

operation of the plant. The boration systems

and the rod position indication system is not

postulated to be the initiator of a design basis

accident. Since there are no changes to these

functions and their operation will remain the

same, the probability of an accident is not

increased by relocating these requirements to

the TRM. Additionally, the accident

mitigation capability and offsite dose

consequences associated with accidents will

not change because these functions will not

be altered by the proposed relocation.

Therefore the consequences of an accident

are not increased by this relocation to the

TRM and the control of revisions to these

specifications in accordance with 10 CFR

50.59.

B. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed revision will not alter the

functions for the boration or the rod position

indication systems such that accident

potential would be changed. The location of

these specifications in the TRM and the

performance of revisions in accordance with

10 CFR 50.59 will continue to maintain

acceptable operability requirements.

Therefore, the possibility of an accident of a

new or different kind is not created by the

proposed relocation and deletion.

C. The proposed amendment does not

involve a significant reduction in a margin of

safety.

The proposed specification relocation and

special test deletion will not affect plant

setpoints or functions that maintain the

margin of safety. This is based on the

relocation to the TRM continuing to maintain

the same level of operability requirements

and surveillance testing to adequately ensure

functionality of the boration and rod position

indication systems. Control of TRM

requirements in accordance with 10 CFR

50.59 will ensure that revisions to these

functions will not inappropriately impact the

health and safety of the public without prior

review and approval by NRC. Therefore, the

proposed relocation and deletion is

acceptable and will not reduce the margin of

safety.

The NRC has reviewed the licensee’s

analysis and, based on this review, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 10H,

Knoxville, Tennessee 37902.

NRC Section Chief: Richard P.

Correia.

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: May 2,

2000, as supplemented by letter dated

August 30, 2000.

Brief description of amendment: The

proposed license amendment would

obtain approval from the Nuclear

Regulatory Commission (NRC) of

changes to the Comanche Peak Steam

Electric Station, Units 1 and 2 (CPSES)

Federal Register / Vol. 65, No. 193 / Wednesday, October 4, 2000 / Notices 59227

Security Plans prior to their

implementation. Prior approval is being

requested from the NRC, in accordance

with the requirements of 10 CFR

50.54(p)(1), because some of the changes

could have the potential of reducing the

effectiveness of the security plans.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Do the proposed changes involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

Response: No

The proposed changes involving Security

activities, do not reduce the ability for the

Security organization to prevent radiological

sabotage and therefore does not increase the

probability or consequences of a radiological

release previously evaluated.

2. Do the proposed changes create the

possibility of a new or different kind of

accident from any accident previously

evaluated?

Response: No

The proposed changes involve functions of

the Security organization concerning

intrusion detection, material search

requirements, alarm response and

compensation, and vehicle control. Analysis

of the proposed changes has not indicated

nor identified a new or different kind of

accident previously evaluated.

3. Do the proposed changes involve a

significant reduction in a margin of safety?

Response: No

Analysis of the proposed changes show

that the proposed changes affect only the

functions of the Security organization and

have no impact upon nor cause a significant

reduction in margin of safety for plant

operation. The failure points of key safety

parameters are not affected.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: George L. Edgar,

Esq., Morgan, Lewis and Bockius, 1800

M Street, NW., Washington, DC 20036

NRC Section Chief: Robert A. Gramm

Wolf Creek Nuclear Operating

Corporation, Docket No. 50–482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request:

September 15, 2000 (WO 00–0036).

Description of amendment request:

The proposed amendment would revise

footnotes (b) and (c) to Table 1.1–1,

‘‘Modes,’’ of the Wolf Creek Technical

Specifications, and allow one reactor

pressure vessel (RPV) head closure bolt

to not be fully tensioned in Mode 4 (hot

shutdown) and Mode 5 (cold

shutdown). Each RPV head closure bolt

is composed of a stud, nut, and washer,

and the bolts attach the vessel head to

the vessel body. The proposed revisions

would allow the plant (1) to be in

Modes 4 and 5 with only 53 of 54 RPV

head closure bolts fully tensioned (i.e.,

to allow one head closure bolt to not be

fully tensioned), and (2) to operate with

one RPV head closure bolt less than

fully tensioned. The proposed revision

to footnote (b) requires the proposed

revision to the definition of refueling

(i.e., footnote (c) from the current ‘‘one

or more’’ RPV head closure bolts

detensioned to the proposed ‘‘two or

more’’ RPV head closure bolts

detensioned). In refueling, the RPV head

closure bolts are detensioned and

removed from the vessel body, and the

RPV head is removed from the vessel.

The licensee committed to the following

program before operating with a not

fully tensioned RPV head closure bolt:

(1) The circumstances for the closure

bolt not being fully tensioned will be

reviewed to determined that the

analysis in the application is still

applicable, (2) the RPV will not be

subject to hydrostatic test conditions

before the closure bolt is returned to

service, and (3) the plant heatup rate

will be held to 50°F per hour (half the

normal rate) until the closure bolt is

returned to service.

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.

Overall protection system performance will

remain within the bounds of the accident

analyses, since no hardware changes are

proposed. Since the stresses [in the RPV head

and body] remain within [the American

Society of Mechanical Engineers Boiler and

Pressure Vessel (ASME)] Code allowables,

the proposed change will not affect the

probability of any event initiators nor will

the proposed change affect the ability of any

safety related equipment to perform its

intended function. There will be no

degradation in the performance of nor an

increase in the number of challenges

imposed on safety related equipment

assumed to function during an accident

situation.

Therefore, the proposed change does not

involve a significant increase in the

probability or consequences of an accident

previously analyzed.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

There are no hardware changes nor are

there any changes in the method by which

any safety related plant system performs it[s]

safety function. The method of plant

operation is unaffected. Leakage would be

precluded since, as noted in the evaluation

[in the application dated September 15,

2000,] adequate compression [of the RPV

head closure bolts] remains. However, if

leakage were to result from having less than

the total number of closure studs fully

tensioned it would be detected by an increase

in the temperature on the leak-off line from

the annular space between the inner and

outer vessel head o-rings. That temperature

increase would be detected by installed

temperature indicators and alarmed in the

control room. Any leakage would be detected

as an increase in RCS [reactor coolant

system] identified LEAKAGE. Since stresses

remain within Code allowables, no new

accident scenarios, transient precursors,

failure mechanisms, or limiting single

failures are introduced as a result of this

change.

Therefore, the proposed change will not

create 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.

As indicated in Section 4.0 above [of the

application dated September 15, 2000,]

ASME Section III stress limits for effected

components are not exceeded. The

evaluations indicate that the reactor vessel

will continue to meet ASME Code allowable

stress criteria with a single untensioned

reactor vessel closure stud, or with a single

closure stud which fails in service. The

proposed change does not alter nor exceed

the acceptance criteria for any analyzed

event. There will be no effect on the manner

in which safety limits or limiting safety

system settings are determined nor will there

be any effect on those plant systems

necessary to assure the accomplishment of

protection functions.

Therefore, the proposed change to the

Technical Specifications 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. The

licensee’s reference to ‘‘closure studs’’

in the above not significant hazards

consideration is a reference to the

‘‘closure bolts’’ in the proposed

amendment.

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts and Trowbridge,

2300 N Street, NW., Washington, DC

20037.

NRC Section Chief: Stephen Dembek.

59228 Federal Register / Vol. 65, No. 193 / Wednesday, October 4, 2000 / Notices

Previously Published Notices of

Consideration of Issuance of

Amendments to Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing

The following notices were previously

published as separate individual

notices. The notice content was the

same as above. They were published as

individual notices either because time

did not allow the Commission to wait

for this biweekly notice or because the

action involved exigent circumstances.

They are repeated here because the

biweekly notice lists all amendments

issued or proposed to be issued

involving no significant hazards

consideration.

For details, see the individual notice

in the Federal Register on the day and

page cited. This notice does not extend

the notice period of the original notice.

Duke Energy Corporation, Docket Nos.

50–269, 50–270, and 50–287, Oconee

Nuclear Station, Units 1, 2, and 3,

Oconee County, South Carolina

Date of amendment request:

September 7, 2000.

Description of amendment request:

The amendments revise Surveillance

Requirement 3.8.1.9.a by adding a note

that states the upper limits on frequency

and voltage are not required to be met

for the annual test of the Keowee Hydro

Units until the NRC issues an

amendment that removes the note in

response to an amendment request to be

submitted no later than April 5, 2001.

Date of publication of individual

notice in Federal Register: September 19,

2000 (65 FR 56600).

Expiration date of individual notice:

October 3, 2000, for comments; October

19, 2000, for hearings.

Niagara Mohawk Power Corporation,

Docket No. 50–410, Nine Mile Point

Nuclear Station Unit No. 2, Oswego

County, New York

Date of application for amendment:

July 14, 2000.

Brief description of amendment

request: The proposed amendment

would revise License Condition 2.c.(10),

‘‘Additional Condition 1,’’ which was

imposed by Amendment No. 91 dated

February 15, 2000. License Condition

2.c.(10) defines the meaning of

implementation of Improved Technical

Specifications and specifies that

implementation be completed by

August 31, 2000. The licensee has

proposed to revise the implementation

date from August 31, 2000, to December

31, 2000.

Date of publication of individual

notice in Federal Register: July 27, 2000

(65 FR 46183).

Expiration date of individual notice:

August 28, 2000.

Tennessee Valley Authority, Docket No.

50–390 Watts Bar Nuclear Plant, Unit 1,

Rhea County, Tennessee

Date of application for amendments:

June 7, as supplemented June 23, and

August 24, 2000.

Brief description of amendment:

Changes to Facility Operating License

and Technical Specifications to reflect

an increase in allowable thermal power

from 3411 to 3459 megawatts.

Date of publication of individual

notice in the Federal Register: September

7, 2000 (65 FR 54322).

Expiration date of individual notice:

October 10, 2000.

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 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 ADAMS Public Library

component on the NRC Web site, http:/

/www.nrc.gov (the Electronic Reading

Room).

Carolina Power & Light Company,

Docket No. 50–261, H. B. Robinson,

Steam Electric Plant, Unit No. 2,

Darlington County, South Carolina

Date of application for amendment:

November 30, 1999.

Brief description of amendment: This

amendment revises the testing

requirements in Technical Specification

5.5.11, ‘‘Ventilation Filter Testing

Program (VFTP),’’ in response to

Generic Letter 99–02, ‘‘Laboratory

Testing of Nuclear-Grade Activated

Charcoal.’’

Date of issuance: September 14, 2000.

Effective date: September 14, 2000.

Amendment No.: 189.

Facility Operating License No. DPR–

23. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: December 29, 1999 (64 FR

73087).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated September 14,

2000.

No significant hazards consideration

comments received: No.

Carolina Power & Light Company, et al.,

Docket No. 50–400, Shearon Harris

Nuclear Power Plant, Unit 1, Wake and

Chatham Counties, North Carolina.

Date of application for amendment:

June 7, 2000, as supplemented on

August 23, 2000.

Brief description of amendment: This

amendment revises the surveillance test

intervals and allowed outage times for

Engineered Safety Features Actuation

System (ESFAS) instrumentation in

Technical Specification (TS) 3/4.3.2. It

also revises the reactor trip system

instrumentation requirements in TS 3/

4.3.1 associated with implementing the

ESFAS relaxations.

Date of issuance: September 13, 2000.

Effective date: September 13, 2000.

Amendment No.: 101.

Facility Operating License No. NPF–

63. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: July 12, 2000 (65 FR 43044).

The supplemental submittal dated

August 23, 2000, provided clarifying

information only, and did not change

the initial no significant hazards

consideration determination.

Federal Register / Vol. 65, No. 193 / Wednesday, October 4, 2000 / Notices 59229

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated September 13,

2000.

No significant hazards consideration

comments received: No.

Commonwealth Edison Company,

Docket Nos. 50–237 and 50–249,

Dresden Nuclear Power Station, Units 2

and 3, Grundy County, Illinois

Date of application for amendments:

February 23, 2000, as supplemented by

letters dated June 19 and July 17, 2000.

Brief description of amendments: The

amendments changed Technical

Specification (TS) 3/4.6.K to revise the

reactor pressure-temperature (P–T)

limits; changed TSs 1.0 and 3/4.12.C to

delete a special test exception that

allowed the hydrostatic test to be

performed above 212 degrees Fahrenheit

while in Mode 4; and added a condition

to the Unit 2 and 3 licenses to specify

expiration dates for the P–T limits.

Date of issuance: September 19, 2000.

Effective date: Immediately, to be

implemented within 30 days.

Amendment Nos.: 179 and 174.

Facility Operating License Nos. DPR–

19 and DPR–25: The amendments

revised the Facility Operating Licenses

and Technical Specifications.

Date of initial notice in Federal

Register: April 5, 2000 (65 FR 17911).

The June 19 and July 17, 2000, letters

are within the scope of the original

notice and did not change the original

no significant hazards consideration

determination. The Commission’s

related evaluation of the amendments is

contained in a Safety Evaluation dated

September 19, 2000.

No significant hazards consideration

comments received: No.

Commonwealth Edison Company,

Docket Nos. 50–237 and 50–249,

Dresden Nuclear Power Station, Units 2

and 3, Grundy County, Illinois

Date of application for amendments:

August 3, 1999, as supplemented by

letter dated February 25, 2000.

Brief description of amendments: The

amendments revised Technical

Specification (TS) Section 2.1.B to

reflect a change to the Minimum Critical

Power Ratio for Unit 2; added an

approved analytical method to TS

Section 6.9.A.6 for Units 2 and 3 for use

in determining core operating limits;

and added conditions to the Unit 2 and

3 licenses to limit the maximum rod

average burnup for any rod to 60 GWD/

MTU until the staff has completed an

environmental assessment supporting a

greater limit.

Date of issuance: September 21, 2000.

Effective date: Immediately, to be

implemented within 30 days.

Amendment Nos.: 180 and 175.

Facility Operating License Nos. DPR–

19 and DPR–25: The amendments

revised the Facility Operating Licenses

and Technical Specifications.

Date of initial notice in Federal

Register: September 8, 1999 (64 FR

48859). The February 25, 2000,

submittal provided additional clarifying

information that did not change the

initial proposed no significant hazards

consideration.

The Commission’s related evaluation

of the amendments is contained in an

Environmental Assessment dated

September 19, 2000, and a Safety

Evaluation dated September 21, 2000.

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:

February 29, 2000, as supplemented by

letter dated July 5, 2000.

Brief description of amendments: The

amendments revised the Technical

Specifications Table 3.3.2–1, Engineered

Safety Feature Actuation System

Instrumentation, Function 6.f, Auxiliary

Feedwater Pump Suction Pressure-Lo.

Date of issuance: September 13, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment Nos.: 193 and 174.

Facility Operating License Nos. NPF–

9 and NPF–17: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: August 23, 2000 (65 FR 51350).

The supplement dated July 5, 2000,

provided clarifying information that did

not change the scope of the February 29,

2000, application and the initial

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated September 13,

2000.

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:

January 6, 2000, as supplemented by

letter dated July 20, 2000.

Brief description of amendments: The

amendments revised the Technical

Specifications (TS) 3.3.1, ‘‘Reactor Trip

System Instrumentation;’’ TS 3.3.2,

‘‘Engineered Safety Features Actuation

System Instrumentation;’’ TS 3.3.5,

‘‘Loss of Power Diesel Generator Start

Instrumentation;’’ and TS 3.3.6,

‘‘Containment Purge and Exhaust

Isolation Instrumentation.’’

Date of issuance: September 18, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment Nos.: 194/175.

Facility Operating License Nos. NPF–

9 and NPF–17: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: March 22, 2000 (65 FR 15378).

The supplement dated July 20, 2000,

provided clarifying information that did

not change the scope of the January 6,

2000, application and the initial

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated September 18,

2000.

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 29, 2000, as supplemented by

letters dated July 27, and August 10,

2000. Other related information was

submitted by letters dated April 10,

April 17, and June 19, 2000.

Brief description of amendments: The

amendments revised the Technical

Specifications (TS) to reference the

Westinghouse Best Estimate Large Break

Loss-of-Coolant Accident analysis

methodology described in WCAP–

12945–P–A, March 1998. The changes

also address corresponding TS Bases

changes.

Date of issuance: September 22, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment Nos.: 195 and 176.

Facility Operating License Nos. NPF–

9 and NPF–17: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: August 23, 2000 (65 FR 51349).

The supplements dated April 10,

April 17, June 19, July 27, and August

10, 2000, provided clarifying

information that did not change the

scope of the June 29, 2000, application

and the initial proposed no significant

hazards consideration determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated September 22,

2000.

59230 Federal Register / Vol. 65, No. 193 / Wednesday, October 4, 2000 / Notices

No significant hazards consideration

comments received: No.

Duke Energy Corporation, Docket Nos.

50–269, 50–270, and 50–287, Oconee

Nuclear Station, Units 1, 2, and 3,

Oconee County, South Carolina

Date of application of amendments:

April 26, 1999; supplemented May 15,

July 26, and August 23, 2000.

Brief description of amendments: The

amendments revised various provisions

of the Technical Specifications and

Final Safety Analysis Report related to

the steam generator tube loads following

a main steam line break and runout

protection for the turbine-driven

emergency feedwater pump.

Date of Issuance: September 18, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment Nos.: 315, 315, & 315.

Facility Operating License Nos. DPR–

38, DPR–47, and DPR–55: Amendments

revised the Technical Specifications and

Final Safety Analysis Report.

Date of initial notice in Federal

Register: May 19, 1999 (64 FR 27320).

The supplements dated May 15, July

26, and August 23, 2000, provided

clarifying information that did not

change the scope of the April 26, 2000,

application and the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated September 18,

2000.

No significant hazards consideration

comments received: No.

Energy Northwest, Docket No. 50–397,

WNP–2, Benton County, Washington

Date of application for amendment:

April 13, 2000, as supplemented by

letter dated May 15, 2000.

Brief description of amendment: The

amendment revised Surveillance

Requirements 3.3.1.1.10 for Function 8

of Table 3.3.1.1–1 and 3.3.4.1.2.a. for

reactor protection system and end of

cycle recirculation pump trip

instrumentation of the WNP–2 technical

specifications. The amendment extends

the frequency of these surveillance

requirements from 18 months to 24

months.

Date of issuance: September 15, 2000.

Effective date: September 15, 2000, to

be implemented within 30 days from

the date of issuance.

Amendment No.: 168.

Facility Operating License No. NPF–

21: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: June 14, 2000 (65 FR 37423).

The May 15, 2000, supplemental

letter provided clarifying information,

did not expand the scope of the

application as originally noticed and

did not change the staff’s original

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated September 15,

2000.

No significant hazards consideration

comments received: No.

FirstEnergy Nuclear Operating

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

Ottawa County, Ohio

Date of application for amendment:

July 28, 1999, as supplemented on June

6, 2000.

Brief description of amendment: This

amendment revised the ultimate heat

sink (UHS) average water temperature

from 85 degrees Fahrenheit (°F) to ≤90

°F and permits plant operations in

Operating Modes 1 through 4 with an

average water temperature of ≤90 °F.

Date of issuance: September 12, 2000.

Effective date: Immediately, to be

implemented within 90 days.

Amendment No.: 242.

Facility Operating License No. NPF–3:

The amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 25, 1999 (64 FR 46438).

The June 6, 2000, submittal provided

additional clarifying information that

did not change the initial proposed no

significant hazards consideration

determination. The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

September 12, 2000.

No significant hazards consideration

comments received: No.

FirstEnergy Nuclear Operating

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

Ottawa County, Ohio

Date of application for amendment:

September 7, 1999, supplemented July

14, 2000.

Brief description of amendment: This

amendment revises Technical

Specification 3/4.3.2.1, Safety Features

Actuation System Instrumentation Trip

Setpoints, to remove the ‘‘Trip

Setpoint’’ values for Instrument String

Functional Unit ‘‘b’’, Containment

Pressure-High, and Functional Unit ‘‘c’’,

Containment Pressure-High-High, and

also modifies the ‘‘Allowable Values’’

entry for these same Functional Units,

consistent with updated calculations

using current setpoint methodology.

The changes also revise Limiting

Conditions for Operation (LCO) 3.3.2.1,

and Bases 3/4.3.1 and 3/4.3.2 to reflect

the removal of the ‘‘Trip Setpoint’’

values for these Functional Units.

Date of issuance: September 14, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 90 days.

Amendment No.: 243.

Facility Operating License No. NPF–3:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: December 15, 1999 (64 FR

70086).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated September 14,

2000.

No significant hazards consideration

comments received: No.

FirstEnergy Nuclear Operating

Company, Docket No. 50–440, Perry

Nuclear Power Plant, Unit 1, Lake

County, Ohio

Date of application for amendment:

August 4, 1999, and as supplemented by

letter dated August 7, 2000.

Brief description of amendment: This

amendment revised Technical

Specification 3.9.1, ‘‘Refueling

Equipment Interlocks,’’ by introducing

an optional operator action when one or

more required refueling equipment

interlocks are inoperable. The new

operator action permits continued invessel fuel movement under specific

administrative controls.

Date of issuance: September 12, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 90 days.

Amendment No.: 116.

Facility Operating License No. NPF–

58: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: August 25, 1999 (64 FR 46439).

The August 7, 2000, supplement

contained clarifying information that

was within the scope of the original

application and Federal Register notice

and did not change the staff’s initial

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated September 12,

2000.

No significant hazards consideration

comments received: No.

Florida Power Corporation, et al.,

Docket No. 50–302, Crystal River Unit

No. 3 Nuclear Generating Plant, Citrus

County, Florida

Date of application for amendment:

September 16, 1999, as supplemented

Federal Register / Vol. 65, No. 193 / Wednesday, October 4, 2000 / Notices 59231

by letters dated May 3 and June 29,

2000.

Brief description of amendment: To

allow an increase in the spent fuel pool

(SFP) storage capacity by replacing fuel

racks in the ‘‘B’’ SFP with new highdensity fuel racks.

Date of issuance: September 13, 2000.

Effective date: September 13, 2000.

Amendment No.: 193.

Facility Operating License No. DPR–

31: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: December 8, 1999 (64 FR

68702).

The May 3 and June 29, 2000

supplements provided clarifying

information and did not affect the initial

no significant hazards determination.

The Commission’s related evaluation

of the amendment is contained in an

Environmental Assessment dated

September 5, 2000 and in a Safety

Evaluation dated September 13, 2000.

No significant hazards consideration

comments received: No.

GPU Nuclear, Inc. et al., Docket No. 50–

219, Oyster Creek Nuclear Generating

Station, Ocean County, New Jersey

Date of application for amendment:

June 18, 1999, as supplemented on June

22 and December 10, 1999, and

February 10, and May 2, 2000.

Brief description of amendment: The

amendment revised the Technical

Specifications to reflect the installation

of additional spent fuel pool storage

racks. The additional new racks will

provide 390 additional spent fuel

assembly storage locations.

Date of issuance: September 15, 2000.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 215

Facility Operating License No. DPR–

16: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 17, 1999 (64 FR 44757).

The supplemental letters dated June

22 and December 10, 1999, and

February 10 and May 2, 2000, did not

affect the proposed finding of no

significant hazards consideration, and

was within the scope of the amendment

application as noticed.

The Commission’s related evaluation

of this amendment is contained in a

Safety Evaluation dated September 15,

2000.

No significant hazards consideration

comments received: No.

North Atlantic Energy Service

Corporation, et al., Docket No. 50–443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request: February

18, 2000, as superseded by a letter dated

June 20, 2000.

Description of amendment request:

The amendment changes the Seabrook

Station Technical Specifications to

provide operational flexibility during

the shutdown modes of operation.

These enhancements include: (1) The

ability to have a standby Safety Injection

(SI) pump available during Reactor

Coolant System (RCS) reduced

inventory conditions with the RCS

pressure boundary intact; (2) realigning

a footnote to clarify the allowance of an

inoperable SI pump to be energized for

testing or filling accumulators; (3)

allowance for an additional charging

pump to be made capable of injection

during pump-swap operations; (4)

recognition that a substantial vent area

exists for cold overpressure protection

when the reactor vessel head is on and

the studs are fully detensioned; (5) limit

maneuvering the plant beyond Hot

Shutdown when one charging pump is

operable; and (6) establishes a new

value for the open permissive interlock

associated with the Residual Heat

Removal System suction isolation

valves.

Date of issuance: September 11, 2000.

Effective date: As of its date of

issuance, and shall be implemented

within 60 days.

Amendment No.: 74.

Facility Operating License No. NPF–

86: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 9, 2000 (65 FR 48752).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated September 11,

2000.

No significant hazards consideration

comments received: No.

North Atlantic Energy Service

Corporation, et al., Docket No. 50–443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request:

December 3, 1999.

Description of amendment request:

The amendment changes the Technical

Specifications by incorporating

reference to the American Society for

Testing and Materials (ASTM) Standard

D3803–1989, ‘‘Standard Test Method for

Nuclear-Grade Activated Charcoal,’’ as

the test protocol for charcoal filter

laboratory testing. In addition, there is

a change to Surveillance Requirement

4.7.6.1d.5) and 4.9.12d.4) specifying a

minimum required heater output based

on design rated voltage.

Date of issuance: September 19, 2000.

Effective date: As of its date of

issuance, and shall be implemented

within 60 days.

Amendment No.: 75

Facility Operating License No. NPF–

86: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 26, 2000 (65 FR 4282).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated September 19,

2000.

No significant hazards consideration

comments received: No.

Northeast Nuclear Energy Company, et

al., Docket No. 50–336, Millstone

Nuclear Power Station, Unit No. 2, New

London County, Connecticut

Date of application for amendment:

February 1, 2000, as supplemented on

June 1 and July 13, 2000.

Brief description of amendment: The

amendment modifies Technical

Specification (TS) 3.0.3 to state that this

specification is not applicable in

MODES 5 or 6. The amendment also

makes various changes to TSs 3/4.1

‘‘Reactor Coolant System—Coolant

Loops and Coolant Recirculation’’ and

3/4.9.8, ‘‘Refueling Operations—

Shutdown Cooling and Coolant

Circulation.’’ In addition, various

corrections and formats are revised to

achieve consistency of the structure and

wording of the TSs. The Bases for the

affected TSs have also been revised

accordingly.

Date of issuance: September 14, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 60 days from the date of

issuance.

Amendment No.: 249.

Facility Operating License No. DPR–

65: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 31, 2000 (65 FR 46748).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated September 14,

2000.

No significant hazards consideration

comments received: No.

Northeast Nuclear Energy Company, et

al., Docket No. 50–423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of application for amendment:

January 18, 1999, and supplemented by

letters dated April 5 and December 21,

1999; and May 2 and August 10, 2000.

59232 Federal Register / Vol. 65, No. 193 / Wednesday, October 4, 2000 / Notices

Brief description of amendment: The

amendment revises the Technical

Specifications (TSs) and the Final Safety

Analysis Report for Millstone 3 to allow

an entire reactor core to be offloaded to

the spent fuel pool (SFP) and an

increase in the maximum design basis

normal SFP water temperature limit

from 140 °F to 150 °F during planned

refueling outages. The increase in

maximum design basis normal SFP

water temperature up to 150 °F affects

certain Fuel Building area TS

temperature limits that require a

revision to the TSs.

Date of issuance: September 12, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 60 days from the date of

issuance.

Amendment No.: 182.

Facility Operating License No. NPF–

49: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 10, 1999 (64 FR 11962).

The letters dated April 5 and

December 21, 1999, and May 2, and

August 10, 2000, provided clarifying

information and did not change the

staff’s initial proposed no significant

hazards consideration determination or

expand the scope of the application as

published in the Federal Register.

The Commission’s related evaluation

of this amendment is contained in a

Safety Evaluation dated September 12,

2000.

No significant hazards consideration

comments received: No.

Northeast Nuclear Energy Company, et

al., Docket No. 50–423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of application for amendment:

February 3, 2000

Brief description of amendment: The

amendment changes the Millstone Unit

No. 3 Final Safety Analysis Report to

show that the configuration of valves

3CHS*V61 and 3CHS*V62 takes

exception to the American Society of

Mechanical Engineers Section III code

requirements for class 2 components.

Date of issuance: September 15, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 60 days from the date of

issuance.

Amendment No.: 183.

Facility Operating License No. NPF–

49: Amendment revised the Final Safety

Analysis Report.

Date of initial notice in Federal

Register: June 28, 2000 (65 FR 39958).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated September 15,

2000.

No significant hazards consideration

comments received: No.

PECO Energy Company, Docket Nos.

50–352 and 50–353, Limerick

Generating Station, (LGS) Units 1 and 2,

Montgomery County, Pennsylvania

Date of amendment request:

November 5, 1999, as supplemented

July 17, 2000.

Description of amendment request:

The amendments make changes to

Technical Specifications Sections

4.6.5.3.b.2 and 4.6.5.3.c, ‘‘Standby Gas

Treatment System,’’ 4.6.5.4.b.2 and

4.6.5.4.c, ‘‘Reactor Enclosure

Recirculation System,’’ and 4.7.2.c.2

and 4.7.2.d, ‘‘Control Room Emergency

Air System.’’

Date of Issuance: September 8, 2000.

Effective Date: September 8, 2000.

Amendment Nos.: 144 & 106.

Facility Operating License Nos. NPF–

39 and NPF–85. The amendments

revised the Technical Specifications.

Date of Initial notice in Federal

Register: January 26, 2000 (65 FR 4287)

The July 17, 2000, letter provided

clarifying information that did not

change the initial no significant hazards

consideration determination or expand

the scope of the original Federal

Register Notice.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated September 8,

2000.

No significant hazards consideration

comments received: No.

PECO Energy Company, Docket No. 50–

352, Limerick Generating Station, Unit

1, Montgomery County, Pennsylvania

Date of application for amendment:

May 15, 2000, as supplemented August

10, 2000.

Brief description of amendment: The

amendment revises the pressuretemperature limit curves.

Date of issuance: September 15, 2000.

Effective date: As of date of issuance,

to be implemented within 30 days.

Amendment No.: 145.

Facility Operating License No. NPF–

39. This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: July 12, 2000 (65 FR 43051).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated September 15,

2000. The August 10, 2000, letter

provided clarifying information that did

not change the initial no significant

hazard consideration determination or

expand the scope of the original Federal

Register notice.

No significant hazards consideration

comments received: No.

Southern California Edison Company, et

al., Docket Nos. 50–361 and 50–362,

San Onofre Nuclear Generating Station,

Units 2 and 3, San Diego County,

California

Date of application for amendments:

April 20, 2000 (PCN–503), and

supplemented by letter dated June 6,

2000.

Brief description of amendments: The

amendments revise TS 5.5.2.5, ‘‘Reactor

Coolant Pump Flywheel Inspection

Program’’ by changing the volumetric

examination frequency of the upper

flywheel on each of the primary reactor

coolant pump motors from a 3-year to a

10-year cycle.

Date of issuance: September 8, 2000.

Effective date: September 8, 2000, to

be implemented within 30 days of

issuance.

Amendment Nos.: Unit 2—170; Unit

3—161.

Facility Operating License Nos. NPF–

10 and NPF–15: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: May 17, 2000 (65 FR 31360).

The supplemental letter dated June 6,

2000, provided clarifying information

that was within the scope of the April

20, 2000, application and the Federal

Register notice and did not change the

staff’s initial proposed no significant

hazards consideration determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated September 8,

2000.

No significant hazards consideration

comments received: No.

Southern California Edison Company, et

al., Docket Nos. 50–361 and 50–362,

San Onofre Nuclear Generating Station,

Units 2 and 3, San Diego County,

California

Date of application for amendments:

September 12, 2000.

Brief description of amendments: The

amendments revise TS 3.6.6.1,

‘‘Containment Spray and Cooling

Systems,’’ to change the allowed outage

time (AOT) for a single inoperable train

of the containment spray system from

72 hours to 7 days. Also, the combined

AOT that appears in both Conditions A

and C of TS 3.6.6.1 is revised from 10

days to 14 days.

Date of issuance: September 12, 2000.

Effective date: September 12, 2000, to

be implemented within 30 days of

issuance.

Amendment Nos.: Unit 2—171; Unit

3—162.

Facility Operating License Nos. NPF–

10 and NPF–15: The amendments

revised the Technical Specifications.

Federal Register / Vol. 65, No. 193 / Wednesday, October 4, 2000 / Notices 59233

Date of initial notice in Federal

Register: May 3, 2000 (65 FR 25769).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated September 12,

2000.

No significant hazards consideration

comments received: No.

Southern Nuclear Operating Company,

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

425, Vogtle Electric Generating Plant,

Units 1 and 2, Burke County, Georgia

Date of application for amendments:

March 6, 2000, as supplemented by

letter dated July 7, 2000.

Brief description of amendments: The

amendments revised Technical

Specification (TS) 3.9.4, ‘‘Containment

Penetration,’’ allowing the equipment

hatch to be open during core alteration

and/or during movement of irradiated

fuel within the containment, provided

the capability for closure is maintained.

Date of issuance: September 11, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment Nos.: 115 and 93.

Facility Operating License Nos. NPF–

68 and NPF–81: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: June 28, 2000 (65 FR 39961),

July 20, 2000 (65 FR 45115). The

supplemental letter dated July 7, 2000,

provided clarifying information that did

not change the scope of the March 6,

2000, application and the initial

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated September 11,

2000.

No significant hazards consideration

comments received: No.

Tennessee Valley Authority, Docket No.

50–3a0, Watts Bar Nuclear Plant, Unit 1,

Rhea County, Tennessee

Date of application for amendment:

July 10, 2000 (TS 00–08).

Brief description of amendment:

Regarding the need to conduct channel

operational tests within 12 hours prior

to physics tests and the placing of a

reactor trip instrumentation channel

used in physics tests in a bypassed

condition instead of a tripped condition.

Date of issuance: September 13, 2000.

Effective date: September 13, 2000.

Amendment No.: 28.

Facility Operating License No. NPF–

90: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: August 9, 2000 (65 FR 48759).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated September 13,

2000.

No significant hazards consideration

comments received: No.

Wolf Creek Nuclear Operating

Corporation, Docket No. 50–482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request: June 23,

2000, and supplements dated July 21

and 26, 2000.

Brief description of amendment: The

amendment revises TS 3.9.4,

‘‘Containment Penetrations,’’ to allow

containment penetrations (with direct

access to the outside atmosphere) to be

unisolated under administrative

controls during refueling operations

with core alterations or irradiated fuel

movement inside containment. The

amendment (1) revises the note in the

Limiting Condition for Operation 3.4.9

for containment penetrations that may

be unisolated under administrative

controls, deleting the reference to

penetrations P–63 and P–98, and (2)

deletes the exception for penetrations

P–63 and P–98 in Surveillance

Requirement 3.9.4.1. In addition, there

are format and editorial corrections to

TS 3.8.3, ‘‘Diesel Fuel Oil, Lube Oil, and

Start Air,’’ and TS 5.2.2.b,

‘‘Administrative Controls,’’ to correct

errors issued in Amendment No. 123,

issued March 31, 1999.

Date of issuance: September 12, 2000.

Effective date: September 12, 2000, to

be implemented within 30 days of the

date of issuance, including the

completion of the administrative

procedures that ensure that open

containment penetrations, with direct

access to the outside atmosphere during

refueling operations with core

alterations and irradiated fuel

movement inside containment, will be

promptly closed in the event of a fuel

handling accident inside containment.

Amendment No.: 135.

Facility Operating License No. NPF–

42. The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: July 12, 2000 (65 FR 43053).

The July 21 and 26, 2000, supplements

provided additional clarifying

information, did not expand the scope

of the application as originally noticed,

and did not change the staff’s original

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated September 12,

2000.

No significant hazards consideration

comments received: No.

Dated at Rockville, Maryland, this 28th day

of September 2000.

For the Nuclear Regulatory Commission.

Elinor G. Adensam,

Acting Director, Division of Licensing Project

Management, Office of Nuclear Reactor

Regulation.

[FR Doc. 00–25377 Filed 10–3–00; 8:45 am]

BILLING CODE 7590–01–P

OFFICE OF PERSONNEL

MANAGEMENT

Excepted Service

AGENCY: Office of Personnel

Management.

ACTION: Notice.

SUMMARY: This gives notice of positions

placed or revoked under Schedules A

and B in the excepted service, as

required by Civil Service Rule VI,

Exceptions from the Competitive

Service.

FOR FURTHER INFORMATION CONTACT:

Janice Reid, Staffing Policy Division,

Employment Service (202) 606–0830.

SUPPLEMENTARY INFORMATION: The Office

of Personnel Management publishes this

monthly notice to update appointing

authorities established or revoked under

the Excepted Service provisions of 5

CFR 213. Individual authorities

established or revoked under Schedules

A and B between August 1, 2000, and

August 31, 2000, appear in the

following listing. A consolidated listing

of all authorities as of June 30 is

published annually.

Schedule A

No Schedule A authorities were

established or revoked during August

2000.

Schedule B

The following Schedule B authority

was amended effective August 17, 2000.

Schedule B 213.3209

‘‘(a) Not to exceed six

interdisciplinary positions for the

Airpower Research Institute at the Air

University, Maxwell Air Force Base,

Alabama, for employment to complete

studies proposed by candidates and

acceptable to the Air Force. Initial

appointments are made not to exceed 3

years, with an option to renew or extend

the appointments in increments of 1, 2,

or 3 years indefinitely thereafter.’’

Authority: 5 U.S.C. 3301 and 3302; E.O.

10577, 3 CFR 1954–1958 Comp., P.218.