21034 Federal Register / Vol. 65, No. 76 / Wednesday, April 19, 2000 / Notices

should be granted based upon a

balancing of the 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 dated March 16, 2000, as

supplemented by letter dated April 11,

2000, which are available for public

inspection at the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington, DC,

and accessible electronically through

the ADAMS Public Electronic Reading

Room link at the NRC Web site (http:/

/www.nrc.gov).

Dated at Rockville, Maryland, this 13th day

of April 2000.

For the Nuclear Regulatory Commission.

Steven D. Bloom,

Project Manager, Section #2, Project

Directorate IV and Decommissioning,

Division of Licensing Project Management,

Office of Nuclear Reactor Regulation.

[FR Doc. 00–9752 Filed 4–18–00; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Sunshine Meeting Notice

AGENCY HOLDING THE MEETING: Nuclear

Regulatory Commission

DATES: Weeks of April 17, 24, May 1, 8,

15, and 22, 2000

PLACE: Commissioners’ Conference

Room, 11555 Rockville Pike, Rockville,

Maryland

STATUS: Public and Closed

MATTERS TO BE CONSIDERED: 

Week of April 17

There are no meetings scheduled for the

Week of April 17.

Week of April 24—Tentative

There are no meetings scheduled for the

Week of April 24.

Week of May 1—Tentative

Tuesday, May 2

9:30 a.m. Briefing on Oconee License

Removal (Public Meeting) (Contact: Dave

Lange, 301–415–1730)

Wednesday, May 3

9:25 a.m. Affirmation Session (Public

Meeting) (If needed)

9:30 a.m. Briefing on Efforts Regarding

Release of Solid Material (Public

Meeting) (Contact: Frank Cardile, 301–

415–6185)

Week of May 8—Tentative

Monday, May 8

10:00 a.m. Briefing on Lessons Learned from

the Nuclear Criticality Accident at

Tokaimura and the Implications on the

NRC’s Program (Public Meeting)

(Contact: Bill Troskoski, 301–415–8076)

Tuesday, May 9

8:55 Affirmation Session (Public Meeting) (If

needed)

9:00 a.m. Meeting with Stakeholders on

Efforts Regarding Release of Solid

Material (Public Meeting) (Contact:

Frank Cardile, 301–415–6185)

Week of May 15—Tentative

Tuesday, May 16

9:25 a.m. Affirmation Session (Public

Meeting) (If needed)

Week of May 22—Tentative

Thursday, May 25

8:30 a.m. Briefing on Operating Reactors and

Fuel Facilities (Public Meeting)

10:15 a.m. Briefing on Status of Regional

Programs, Performance and Plans (Public

Meeting)

1:30 p.m. Briefing on Improvements to 2.206

Process (Public Meeting)

*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: April 16, 2000.

William M. Hill, Jr.,

SECY Tracking Officer, Office of the

Secretary.

[FR Doc. 00–9907 Filed 4–17–00; 12:48 pm]

BILLING CODE 7590–01–M

NUCLEAR REGULATORY

COMMISSION

Biweekly Notice; Applications and

Amendments to Facility Operating

Licenses Involving No Significant

Hazards Considerations

I. Background

Pursuant to Public Law 97–415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 97–415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from March 25,

2000, through April 7, 2000. The last

biweekly notice was published on April

5, 2000 (65 FR 17908).

Notice of Consideration of Issuance of

Amendments to Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a

proposed determination that the

following amendment requests involve

no significant hazards consideration.

Under the Commission’s regulations in

10 CFR 50.92, this means that operation

of the facility in accordance with the

proposed amendment would not (1)

involve a significant increase in the

probability or consequences of an

accident previously evaluated; or (2)

create the possibility of a new or

different kind of accident from any

accident previously evaluated; or (3)

involve a significant reduction in a

margin of safety. The basis for this

proposed determination for each

amendment request is shown below.

The Commission is seeking public

comments on this proposed

determination. Any comments received

within 30 days after the date of

publication of this notice will be

considered in making any final

determination.

Normally, the Commission will not

issue the amendment until the

expiration of the 30-day notice period.

However, should circumstances change

during the notice period such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received before

action is taken. Should the Commission

take this action, it will publish in the

Federal Register a notice of issuance

and provide for opportunity for a

hearing after issuance. The Commission

expects that the need to take this action

will occur very infrequently.

Federal Register / Vol. 65, No. 76 / Wednesday, April 19, 2000 / Notices 21035

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 Public Document

Room, the Gelman Building, 2120 L

Street, NW., Washington, DC. The filing

of requests for a hearing and petitions

for leave to intervene is discussed

below.

By May 19, 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 Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, and electronically

from the ADAMS Public Library

component on the NRC Web site, http:/

/www.nrc.gov (the Electronic Reading

Room). If a request for a hearing or

petition for leave to intervene is filed by

the above date, the Commission or an

Atomic Safety and Licensing Board,

designated by the Commission or by the

Chairman of the Atomic Safety and

Licensing Board Panel, will rule on the

request and/or petition; and the

Secretary or the designated Atomic

Safety and Licensing Board will issue a

notice of a hearing or an appropriate

order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) The nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555–0001, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington DC, by

the above date. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555–0001, and to the attorney for

the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

factors specified in 10 CFR

2.714(a)(1)(i)–(v) and 2.714(d).

For further details with respect to this

action, see the application for

amendment which is available for

public inspection at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, and electronically

from the ADAMS Public Library

component on the NRC Web site, http:/

/www.nrc.gov (the Electronic Reading

Room).

Commonwealth Edison Company,

Docket Nos. STN 50–454 and STN 50–

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

Ogle County, Illinois

Docket Nos. STN 50–456 and STN 50–

457, Braidwood Station, Unit Nos. 1 and

2, Will County, Illinois

Date of amendment request: January

20, 2000.

Description of amendment request:

The proposed amendment would revise

the technical specifications (TSs) to

extend the allowable completion times

for the required actions associated with

restoration of an inoperable emergency

diesel generator (EDG), and permit the

performance of the 24-hour EDG

endurance run during Modes 1 and 2

(i.e., ‘‘Power Operation’’ or ‘‘Startup’’).

A new requirement is proposed which

21036 Federal Register / Vol. 65, No. 76 / Wednesday, April 19, 2000 / Notices

will require verification of the opposite

unit’s EDGs when the affected EDG is

inoperable.

Basis for proposed no significant

hazards consideration determination.

As required by 10 CFR 50.92(c), the

staff’s analysis of the issue of no

significant hazards consideration is

presented below:

Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

The proposed changes include the

extension of the completion time for the

EDGs from 72 hours to 14 days. In

conjunction with the proposed change, a new

required action is proposed to be

incorporated into the TSs that will require

verification of the operability of the opposite

unit’s EDGs while the affected EDG is

inoperable. The proposed changes do not

significantly increase the probability of

occurrence of a previously evaluated

accident because the EDGs are not initiators

of accidents. Extending the completion times

of the EDGs would not have any impact on

the frequency of any accident previously

evaluated and, therefore, the probability of a

previously analyzed accident is unchanged.

The proposed change to the completion time

for EDGs will not result in any changes to the

plant activities associated with EDG

maintenance. The EDGs mitigate the

consequences of previously evaluated

accidents involving a loss of normal power,

the safety-related buses and as such, the

operability or availability of the EDGs could

affect accident consequences. A

configuration risk management program

(CRMP) was developed and will be used to

ensure that the risk impact of equipment out

of service is appropriately evaluated prior to

performing any maintenance activity.

Increases in risk posed by potential

combinations of equipment out of service

during the EDG extended completion time

will be managed under the CRMP. In

addition, compensatory actions have been

identified to mitigate an increase in risk.

Procedures have been developed to

implement the compensatory actions.

The proposed changes also include a

change to the TS surveillance requirement

related to the conduct of the 24-hour EDG

endurance run. Specifically, the change

would permit the endurance run to be

performed during Modes 1 and 2. The test

configuration to be used is consistent with

the configuration currently used during the

one-hour monthly EDG tests currently

conducted.

The probability of an accident is not

increased by performing the 24-hour

endurance run in Modes 1 and 2 since the

EDGs are used to support mitigation of the

consequences of an accident. The failure of

an EDG while testing is not an assumed

initiator of a previously analyzed accident.

The EDGs were designed to be tested by

running in parallel with offsite power and

design features such as protective devices

were included. The proposed change does

not affect parallel testing design features, the

consequences of postulated failures during

parallel testing, and postulated interactions

with offsite power during parallel testing. If

problems are encountered during testing, the

EDG connection to the bus will be

interrupted, allowing the offsite circuits to

continue to supply the bus. Testing of the

EDG does not affect the remainder of the

safety-related equipment analyzed to mitigate

the consequences of an accident. The control

logic prevents potential damage of the

emergency core cooling System (ECCS)

equipment powered by the EDG to ensure

that the ECCS equipment is available in the

event of an actual safety injection with or

without a Loss of Offsite Power (LOOP). Only

one EDG per unit will be tested in parallel

with the offsite sources at a time in order to

prevent any grid disturbance from potentially

affecting more than one EDG. Thus, during

the test, the remaining EDG, which is capable

of supplying power to mitigate all design

basis accidents, will be available to respond

normally to a start signal.

To fully evaluate the effect of the proposed

EDG TS changes, probabilistic risk

assessment (PRA) methods and deterministic

analyses were utilized. The results of the risk

analysis show no significant increase in Core

Damage Frequency (CDF) and Large Early

Release Frequency (LERF).

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously analyzed.

Does the change create the possibility of a

new or different kind of accident from any

accident previously evaluated?

The proposed changes do not involve a

physical change to the plant. No new

equipment is being introduced, and installed

equipment is not being operated in a new or

different manner except for the following.

The electrical lineup for performing the 24-

hour run will be the same as the lineup for

performance of the one-hour run, which is

routinely performed at least once per month

for each EDG. The difference between these

two surveillances is in their duration. There

is no change being made to the parameters

within which the plant is operated. There are

no setpoints affected by this proposed change

at which protective or mitigative actions are

initiated. This proposed changes will not

alter the manner in which equipment

operation is initiated, nor will the function

demands on credited equipment be changed.

No alteration in the procedures, which

ensure that the plant remains within

analyzed limits, is being proposed, and no

change is being made to the procedures

relied upon to respond to an off-normal

event. As such, no new failure modes are

being introduced. Other than the changes in

duration of EDG unavailability from 72 hours

to 14 days and on-line testing from 60

minutes to 24 hours, the change does not

alter assumptions made in the safety analysis

and licensing basis.

Therefore, these proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

Does the change involve a significant

reduction in a margin of safety?

The proposed changes will extend the

allowable Completion Times for the Required

Actions associated with restoration of an

inoperable EDG and allow the performance of

the 24-hour endurance run at power. In

conjunction with the proposed changes, a

new required action is proposed to be

incorporated into the TSs. The new action

will require verification of the operability of

the opposite unit’s EDGs while the affected

EDG is inoperable. These actions will be

taken to ensure the availability of the

remaining alternating current power sources

to the affected engineered safety feature bus.

The CRMP will be used to ensure that the

risk impact of equipment out of service is

appropriately evaluated prior to performing

any maintenance activity. Increase in risk

posed by potential combinations of

equipment out of service during the EDG

extended completion time will be managed

under the CRMP. In addition, compensatory

actions have been identified to mitigate

increase in risk. Procedures have been

developed to implement the compensatory

actions.

With regard to the proposed change for the

24-hour endurance run, the EDGs were

designed to be tested by running in parallel

with offsite power and, design features such

as protective devices were included. The

proposed change does not affect parallel

testing design features, the consequences of

postulated failures during parallel testing,

and postulated interactions with offsite

power during parallel testing. If problems are

encountered during testing, the EDG

connection to the bus will be interrupted

allowing the offsite circuits to continue to

supply the bus. Further, the EDG system

design includes emergency override of the

test mode for both accident conditions (safety

injection) and loss of offsite power to permit

a response to actual emergency signals and

return control of the EDG to the automatic

control system.

Therefore, implementation of the proposed

changes will not involve a significant

reduction in the margin of safety.

Based on the staff’s analysis, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Ms. Pamela B.

Stroebel, Senior Vice President and

General Counsel, Commonwealth

Edison Company, P.O. Box 767,

Chicago, Illinois 60690–0767.

NRC Section Chief: Anthony J.

Mendiola.

Florida Power Corporation, et al.,

Docket No. 50–302, Crystal River

Nuclear Generating Plant, Unit No. 3,

Citrus County, Florida

Date of amendment request: March 6,

2000.

Description of amendment request:

The proposed amendment would revise

the Improved Technical Specification

(ITS) Action Condition and Surveillance

Requirement (SR) for the safety-related

diesel-driven emergency feedwater

pump (EFP–3). The ITS required

Federal Register / Vol. 65, No. 76 / Wednesday, April 19, 2000 / Notices 21037

inventory volume for lube oil would be

revised to agree with the actual test

values and are included in the ITS

Action Condition, SR and Bases.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below.

1. Does not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

The revised lube oil requirements are being

made to ensure EFP–3 is capable of seven

days of continuous operation. The proposed

amendment provides the same functional

requirement as previously approved. The

EFW system is used for accident mitigation

and is not an initiator of design basis

accidents. Therefore, the probability of

previously analyzed events is not affected by

this change. No capability or design

functions of EFP–3 or the emergency

feedwater (EFW) system will change. The

initial conditions for accidents that require

EFW and accident mitigation capability of

the EFW system will remain unchanged.

Therefore, the proposed amendment will not

increase the consequences of evaluated

accidents.

2. Does not create the possibility of a new

or different kind of accident from any

accident previously evaluated.

The revised ITS Condition will ensure

equipment is restored to operable status in

accordance with previously approved

timeframes and functional levels. The revised

SR will assure the same functional

requirement as the previously approved SR.

Lube oil will be stored on-site and the lube

oil inventory in the sump ensures adequate

time to transfer the stored inventory into the

engine. No new plant configurations or

conditions are created by these revised ITS

Conditions or SR. Therefore, the proposed

amendment cannot create the possibility of

an accident of a different type than

previously evaluated in the Safety Analysis

Report.

3. Does not involve a significant reduction

in the margin of safety.

The proposed ITS Condition and SR

changes ensure adequate lube oil inventory is

available to operate EFP–3 for seven days.

The proposed changes replace the calculated

lube oil inventory values with a more

conservative value derived from actual test

data for EFP–3. The revised SR ensures the

same functional requirement for a seven-day

supply of lube oil for EFP–3 as was

previously approved. Similarly, the revised

ITS Condition ensures the same functional

level as currently approved by requiring that

a reduced lube oil inventory of less than

seven days but more than six days is restored

to the seven-day level within 48 hours. The

revised SR allows the lube oil inventory to

be stored off engine. The inventory in the

EFP–3 sump and auxiliaries provides

sufficient time to permit the transfer of stored

inventory into the engine. EFP–3 is designed

to allow monitoring of lube oil level and

addition of lube oil while the engine is

operating. Based on the above, the revised

ITS meets the same intent as the currently

approved specifications. Therefore, there is

no reduction in the margin of safety

associated with the proposed ITS

amendment.

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: 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 Nos. 50–275, Diablo Canyon

Nuclear Power Plant, Unit No. 1, San

Luis Obispo County, California

Date of amendment requests:

December 31, 1999, as supplemented by

letter dated January 18, 2000.

Description of amendment requests:

The amendment would revise Section

2.C.(1) of Facility Operating License No.

DPR–80 for the Diablo Canyon Power

Plant, Unit No. 1 to authorize operation

at reactor core power levels not to

exceed 3411 megawatts thermal (100

percent rated power). This amendment

would also (1) revise the definition in

Section 1.1 of the technical

specifications (TS) of rated thermal

power to reflect Unit 1 operation at the

uprated reactor core power level, (2)

change the reactor core safety limits in

TS Figure 2.1.1–1 to reflect the current

fuel type and provide additional margin

for OT∆T and OP∆T setpoint

calculations, and change the nominal

full power Tavg in the OT∆T and OP∆T

function in Notes 1 and 2 to TS Table

3.3.3–1.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

All previously evaluated accidents have

been reviewed for the proposed increase in

Unit 1 power rating, and these reviews are

summarized in WCAP–14819, ‘‘Pacific Gas

and Electric Company Diablo Canyon Power

Plant, Unit 1 3425 MWt [megawatt thermal]

Uprating Program Licensing Report.’’ The

majority of the Diablo Canyon Power Plant

(DCPP) accident analyses already bound the

higher power rating of Unit 2 combined with

the lower reactor coolant system (RCS) flow

rate of Unit 1. Hence, the uprate has no

impact on these previously evaluated

accidents. This is also true of dose

assessment, which remains based on the

original 3568 MWt core source terms and is

not impacted by the uprate.

The previously evaluated accidents that are

impacted by the uprate are large break lossof-coolant accident (LOCA), small break

LOCA, the OT∆T/ OP∆T setpoint

calculations, and accidental depressurization

of the RCS. The large break LOCA was

reanalyzed for uprated conditions using best

estimate methodology. The reanalysis

demonstrated no increase in consequence

and was approved by the NRC in License

Amendments 121 (Unit 1) and 119 (Unit 2).

The small break LOCA was also reanalyzed,

and continues to demonstrate a large margin

to peak clad temperature limits. The current

OT∆T/OP∆T setpoints are bounding for the

Unit 1 uprated power conditions based on

revising the reactor core safety limits in TS

Figure 2.1.1–1 to credit the exclusive use of

Vantage 5 fuel. The accidental RCS

depressurization reanalysis shows that the

departure from nucleate boiling ratio remains

above the applicable limit value. In

summary, no design or analysis acceptance

criteria will be exceeded, the functional

integrity of all plant systems are unaffected,

and there is no impact on the integrity of the

fission product barriers or assumed dose

source terms. Therefore, the consequences of

all previous evaluated accidents are not

substantially increased.

It was determined that there would be no

impact on any component reliabilities

assumed in the PRA model, and therefore no

impact on the resulting core damage

frequency. The PRA model envelopes both

units, based on using the originally higher

rated Unit 2 power level.

The operation impacts of the proposed

power increase were reviewed against the

unit design capability, and it was determined

that no system, structure, or component

would exceed design conditions or loads.

While the low pressure turbines see a small

(less than 1.5°F ) increase in temperature, the

effect on missile generation probability is not

significant. There is no significant increase in

the probability of component failure, offsite

power loss, or any other accident initiator.

Therefore, the probability of all previously

evaluated accidents is not substantially

increased.

Therefore, the proposed change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

Normal operation will not be substantially

impacted by increasing the Unit 1 licensed

power rating to match Unit 2. Procedures

will be essentially unchanged, or where

changes are required, they will be made to

more closely resemble those in effect at Unit

2. Training will communicate all impacts to

personnel and the plant simulator will be

updated to match the power level of both

21038 Federal Register / Vol. 65, No. 76 / Wednesday, April 19, 2000 / Notices

Units 1 and 2. There is, therefore, no

possibility of a new or different kind of

accident related to human performance.

Plant systems, structures, and components

have been evaluated for the proposed uprate.

Most have identical counterparts in operation

at Unit 2 at this higher power level. A few

are slightly different, such as the generator

cooling system, and for these the design

margins have been reviewed and found to be

acceptable. Therefore, there is no possibility

of a new or different kind of accident related

to system, structure, or component

performance.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed changes do not involve a

significant reduction in a margin of safety

because the margin of safety associated with

plant parameters as verified by the results of

the accident analyses are within acceptable

limits. As mentioned, most analyses

demonstrating adequate margins of safety

already assume the higher thermal power

rating of Unit 2 and bound Unit 1 at the

uprated thermal power conditions. The few

transients that are reanalyzed meet the

applicable acceptance criteria.

The reactor core safety limits specified in

TS Figure 2.1.1–1 envelope operation with

both 17x17 standard and 17x17 Vantage 5

fuel. The proposed change revises the reactor

core safety limits in Figure 2.1.1–1 to credit

the exclusive use of Vantage 5 fuel. These

revised safety limits will continue to satisfy

fuel design criteria. The current OT∆T and

OP∆T setpoints provide adequate margin to

the revised reactor core safety limits at the

uprated Unit 1 conditions, which include a

slightly higher nominal full power Tavg in

Notes 1 and 2 to ITS Table 3.3.3–1.

Therefore, the proposed change does not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment 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.

PECO Energy Company, Public Service

Electric and Gas Company, Delmarva

Power and Light Company, and Atlantic

City Electric Company, Dockets Nos. 50–

277 and 50–278, Peach Bottom Atomic

Power Station, Units Nos. 2 and 3, York

County, Pennsylvania

Date of application for amendments:

August 11, 1999.

Description of amendment request:

The proposed amendment clarifies the

use of containment overpressure for

ensuring adequate net positive suction

head (NPSH) for the emergency core

cooling system (ECCS) pumps.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed License Amendment

Request does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

This proposed license amendment request

does not involve any physical changes to

plant Structures, Systems, or Components

(SSC), or how the SSC are operated,

maintained, and tested. The proposed

changes involve the acceptability of taking

credit for a specific amount of containment

overpressure following the initiation of an

event. This credit involves the mitigation of

an event, and not prevention or identification

of an event. Credit for containment

overpressure is not considered a precursor to

any event.

Crediting containment overpressure does

not turn an Anticipated Operational

Occurrence (AOO) into an Abnormal

Operational Transient (AOT) or a Design

Basis Accident (DBA).

Calculations performed in support of the

license amendment request provide a

conservative estimate of the Minimum

Containment Pressure Available (MCPA)

following all design and licensing basis

events for which some amount of

containment overpressure is required. The

NPSH calculations for the Residual Heat

Removal (RHR) and Core Spray (CS) pumps

include conservative assumptions and input

values ensure that, barring beyond-designbasis loss of containment integrity, adequate

NPSH is provided to the RHR and CS pumps

for the entire duration of any of these events.

The proposed license amendment request

makes a change to the PBAPS licensing basis

to clearly define amount of containment

overpressure allowed. This value is

designated as the Containment Overpressure

License (COPL). Conservative analyses have

assured that the MCPA is always greater than

this COPL for design basis events. Therefore,

adequate NPSH is provided to the RHR and

Core Spray pumps for all design and

licensing basis events.

The evaluation for MCPA and NPSH

includes the consideration for any one single

active failure. The worst-case single active

failure is the failure of one electrical division.

There is no credible single active failure that

can compromise the containment integrity.

The evaluation for MCPA and NPSH does not

place any restrictions on system operation

following a design or licensing basis event.

The analysis concludes that adequate NPSH

will be available, even assuming the worst

single active failure.

Therefore, the proposed license

amendment request does not significantly

increase the probability or consequences of

an accident previously evaluated.

2. The proposed License Amendment

Request does not create the possibility of a

new or different kind of accident from any

accident previously evaluated.

This proposed license amendment request

does not involve any physical changes to

plant SSC, or how the SSC are operated,

maintained, and tested. This proposed

license amendment request involves the

acceptability of taking credit for some

amount of containment overpressure

following the initiation of an event. This

credit involves the mitigation of an event,

and not prevention or identification of an

event. Credit for containment overpressure is

not considered a precursor to any event.

Worst-case single active failure (i.e., loss of

one electrical division) was considered in the

assessment of MCPA and COPR [containment

overpressure required]. The supporting

calculations indicate that adequate NPSH is

provided to the RHR and CS pumps for all

design and licensing basis events, even with

the worst single active failure.

Therefore, the proposed license

amendment request does not create the

possibility of a new or different kind of

accident from any previously evaluated.

3. The proposed License Amendment

Request does not involve a significant

reduction in a margin of safety.

The MCPA and NPSH analyses supporting

this license amendment request include

conservative assumptions and use

conservative input values that are consistent

with or bound the analytical limits of the

PBAPS Technical Specifications. These

analyses indicate that adequate NPSH margin

is available for operation of the RHR and CS

systems to meet their safety functions

following any design or licensing basis event.

This includes operation of RHR in

Suppression Pool Cooling, Wetwell Spray,

Drywell Spray, and Low Pressure Coolant

Injection modes, and CS in Short Term and

Long Term Spray Cooling. Therefore, the

proposed license amendment request does

not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for Licensee: J. W. Durham,

Sr., Esquire, Sr. V.P. and General

Counsel, PECO Energy Company, 2301

Market Street, Philadelphia, PA 19101.

NRC Section Chief: James W. Clifford.

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 amendment requests: January

19, 2000 (PCN–512).

Description of amendment requests:

The amendment application proposes to

revise the San Onofre Nuclear

Federal Register / Vol. 65, No. 76 / Wednesday, April 19, 2000 / Notices 21039

Generating Station, Units 2 and 3,

technical specifications (TSs)

Surveillance Requirement (SR) 3.0.3.

SR 3.0.3 allows compliance with the

requirement to declare a limiting

condition for operation not met to be

delayed whenever it is discovered that

a surveillance was not performed within

its specified frequency (a missed

surveillance). Presently, SR 3.0.3 allows

a delay ‘‘up to 24 hours or up to the

limit of the specified Frequency,

whichever is less.’’ The licensee

proposes to revise the allowable delay

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

specified Frequency, whichever is

greater.’’

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Will operation of the facility in

accordance with this proposed change

involve a significant increase in the

probability or consequences of any accident

previously evaluated?

Response: No.

The proposed change would extend the

maximum allowable time for completing a

Surveillance not performed within its

specified Frequency (a missed Surveillance)

without declaring the affected Limiting

Condition For Operation (LCO) not met. The

presently allowed time is up to 24 hours from

the time of discovery or up to the limit of the

specified Frequency, whichever is less. The

proposed allowed time is up to 24 hours from

the time of discovery or up to the limit of the

specified Frequency, whichever is greater.

Surveillances are rarely missed. This is

demonstrated by a limited review of Licensee

Event Reports (LERs), which found very few

occurrences of missed Surveillances, given

the number of LERs submitted and the large

number of Surveillances performed.

Moreover, Surveillances, whether performed

inside or outside the required Frequency,

nearly always verify conformance with

Technical Specification requirements. This is

demonstrated by a survey of selected

licensees regarding entries into Surveillance

Requirement (SR) 3.0.3. As stated in Generic

Letter 87–09, ‘‘* * * the vast majority of

surveillances do in fact demonstrate that

systems or components are operable.’’ As

stated in the SR 3.0.3 Bases, ‘‘* * * the most

probable result of any particular Surveillance

being performed is the verification of

conformance with the requirements.’’

Therefore, it is unlikely that plant

equipment would be inoperable during the

time period of up to 24 hours or up to the

limit of the specified Frequency, whichever

is greater, that would be allowed under the

proposed change for the completion of a

missed Surveillance.

If, upon discovery of a missed

Surveillance, it is known that the

Surveillance would fail, SR 3.0.1 would

require that the affected LCO be declared not

met and the appropriate Condition(s)

entered.

Performance of some Surveillances carries

with it a slight risk, either from making some

plant equipment temporarily inoperable or

from performing plant manipulations, or

both. The increase in plant risk from

performing such Surveillances, combined

with the confidence that a Surveillance test

will be satisfactory when performed, together

provide justification for extending the

current allowable time to up to 24 hours or

up to the specified Frequency, whichever is

greater.

The foregoing discussion demonstrates that

the probability or consequences of any

accident previously evaluated will not be

significantly increased by the proposed

change.

2. Will operation of the facility in

accordance with this proposed change create

the possibility of a new or different kind of

accident from any accident previously

evaluated?

Response: No.

This amendment request is administrative

in nature and does not involve any change

to plant equipment. Therefore, it will not

create the possibility of a new or different

kind of accident from any previously

evaluated.

3. Will operation of the facility in

accordance with this proposed change

involve a significant reduction in a margin of

safety?

Response: No.

This amendment request does not change

the manner in which safety limits or limiting

safety settings are determined.

As discussed above, Surveillances are

rarely missed, and, when performed,

Surveillances nearly always verify

conformance with Technical Specification

requirements, making it unlikely that plant

equipment would be inoperable during the

time period of up to 24 hours or up to the

limit of the specified Frequency, whichever

is greater, that would be allowed under the

proposed change for the completion of a

missed Surveillance.

Therefore, the proposed change will not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment requests involve no

significant hazards consideration.

Attorney for licensee: Douglas K.

Porter, Esquire, Southern California

Edison Company, 2244 Walnut Grove

Avenue, Rosemead, California 91770.

NRC Section Chief: Stephen Dembek.

STP Nuclear Operating Company,

Docket Nos. 50–498 and 50–499, South

Texas Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request: March

17, 2000.

Description of amendment request:

Revise Technical Specification 3/4.7.4

to revise the surveillance requirements

(SRs) 4.7.4.b.1 and 4.7.4.b.2 to

incorporate the wording from the

Westinghouse Standard Improved

Technical Specifications (NUREG–1431)

and to delete SR 4.7.4.b.3. SR 4.7.4.b.3

requires verifying at least once per 18

months that each screen wash booster

pump and the traveling screen start

automatically on a safety injection test

signal. The licensee also proposed

changes to the Technical Specifications

Bases associated with the Technical

Specification changes and

administrative changes to the Bases

Index.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the proposed amendment involve

a significant increase in the probability or

consequences of an accident previously

evaluated?

NUREG 1431 related changes:

Incorporating the NUREG 1431

[Westinghouse Standard Improved Technical

Specifications] wording for SR 4.7.4.b.1 and

SR 4.7.4.b.2 does not significantly increase

the probability of an accident because the

surveillance testing of the Essential Cooling

Water system has no effect on accident

initiation probability. This change does not

significantly increase the consequences of an

accident because the surveillance

requirements still provide adequate

assurance that the Essential Cooling Water

system can provide its design function.

Screen wash system changes:

Eliminating the requirement for the

Essential Cooling Water traveling screens and

screen wash booster pumps to start on a

safety injection signal does not increase the

probability of any accident previously

evaluated. The traveling screens and the

screen wash booster pumps have no potential

for initiating an accident. Eliminating the

requirement for the traveling screens and the

screen wash booster pumps to start on a

safety injection signal does not increase the

consequences of any accident previously

evaluated. A control system is provided to

automatically start and stop the traveling

screens during normal operation. A high

differential water level sensed across any

traveling screen alarms in the control room

and automatically starts the screen wash

booster pump and, after reaching adequate

screen wash pressure, starts the traveling

screen. A safety injection signal is not needed

for this function. In addition, there are no

circumstances associated with any event

requiring a safety injection signal that would

cause a high differential water level across

the traveling screen.

The changes to the Bases Index are

administrative and have no relevance to

accident probability or consequences.

21040 Federal Register / Vol. 65, No. 76 / Wednesday, April 19, 2000 / Notices

Based on the above, STPNOC [STP Nuclear

Operating Company] concludes that the

proposed change does not increase the

probability or consequences of an accident

previously evaluated.

2. Does the proposed amendment create

the possibility of a new or different kind of

accident from any accident previously

evaluated?

NUREG 1431 related changes:

Incorporation of the NUREG 1431 wording

into the surveillance requirements does not

create the possibility of a new or different

kind of accident because the surveillance

requirements are not substantially changed

and do not involve any different operational

configurations for the station.

Screen wash system changes:

Elimination of the requirement to start the

traveling screen and screen wash booster

pump on a safety injection signal will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated. As discussed above,

the traveling screens and screen wash booster

pump have no potential to initiate an

accident. In addition, STPNOC is not

proposing any different operational

configurations for the station.

The changes to the Bases Index are

administrative and have no relevance to

accidents.

Based on the above, STPNOC concludes

that the proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. Does the proposed amendment involve

a significant reduction in a margin of safety?

NUREG 1431 related changes:

Incorporation of the NUREG 1431 wording

for SR 4.7.4.b.1 and SR 4.7.4.b.2 does not

significantly change the way the surveillance

requirements will be performed. The

Surveillance Requirements still provide

adequate assurance that the Essential Cooling

Water will perform its function. There is no

change in the operational configuration of the

plant. Consequently, the changes to these

surveillance requirements do not

significantly affect the margin of safety.

Screen wash system changes:

Elimination of the requirement for the

traveling screen and screen wash booster

pump to start on a safety injection signal will

not prevent the traveling screen and screen

wash booster pump to start when required.

The systems will start automatically without

the need for a safety injection signal. In

addition, there is no design basis or

mechanistic reason to postulate the need to

automatically start the traveling screens or

screen wash booster pump on a safety

injection signal.

The changes to the Bases Index are

administrative and have no relevance to the

safety margin.

Based on the above, STPNOC concludes

that the proposed change does not involve a

significant decrease in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the standards of

10 CFR 50.92(c) are satisfied. Therefore,

the NRC staff proposes to determine that

the request for amendments involves no

significant hazards consideration.

Attorney for licensee: Jack R.

Newman, Esq., Morgan, Lewis &

Bockius, 1800 M Street, NW.,

Washington, DC 20036–5869.

NRC Section Chief: Robert A. Gramm.

Wolf Creek Nuclear Operating

Corporation, Docket No. 50–482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request: March

31, 2000 (ET 00–0018).

Description of amendment request:

The proposed amendment would

modify the actions for Limiting

Condition for Operation (LCO) 3.7.9,

‘‘Ultimate Heat Sink (UHS),’’ of the

technical specifications (TSs). The

proposed new Action A would allow

the plant to operate with the plant inlet

water temperature of the UHS above 90

°F, if the licensee verified the required

cooling capacity within 4 hours and

once per 12 hours thereafter, but that

the plant would be shut down if the

water temperature exceeded 94 °F. This

would change the current requirement

to shut down the plant if the inlet water

temperature of the UHS exceeded the 90

°F. The time to shut down the plant is

not being changed in the 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 proposed change does not involve any

physical alteration of plant systems,

structures or components. The proposed

change provides an allowance for the plant

to continue operation with [the] plant inlet

water temperature [of the UHS] in excess of

the current Technical Specification limit of

90 °F with the verification that required

cooling capacity being maintained and [the

plant inlet water] temperature ≤ 94 °F. The

94 °F limit is less than the design limit of 95

°F for associated plant components. The

plant inlet water temperature is not assumed

to be an initiating condition of any accident

analysis evaluated in the Updated Safety

Analysis Report (USAR). Therefore, the

allowance for the [plant inlet] water

temperature to be in excess of the current

limit does not involve an increase in the

probability of an accident previously

evaluated in the USAR. The UHS supports

OPERABILITY of safety related systems used

to mitigate the consequences of an accident.

Plant operation for brief periods with [the]

plant inlet water temperature greater than 90

°F up to 94 °F will not adversely affect the

OPERABILITY of these safety related systems

and will not adversely impact the ability of

these systems to perform their safety related

functions. Therefore, the proposed change

does not involve a significant increase in the

consequences of an accident previously

evaluated in the USAR.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed change does not involve any

physical alteration of plant systems,

structures or components. The temperature of

the plant inlet water being greater than 90 °F

but less than or equal to 94 °F (with the main

cooling lake dam intact) does not introduce

new failure mechanisms for systems,

structures or components not already

considered in the USAR. The 94 °F limit is

less than the design limit of 95 °F for

associated plant components. Therefore, the

possibility of a new or different kind of

accident from any accident previously

evaluated is not created.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed change will allow an

increase in [the] plant inlet water

temperature above the current Technical

Specification limit of 90 °F for the UHS,

provided [the] UHS temperature is

maintained below 95 °F and that the required

cooling capacity is verified maintained

within 4 hours and once per 12 hours

thereafter. Additionally, the plant inlet water

temperature will be verified to be ≤94 °F once

per 12 hours. The proposed change does not

alter any safety limits, limiting safety system

settings, or limiting conditions for operation,

and the proposed changes provide continued

assurance that with a plant inlet water

temperature > 90 °F, the design temperature

of safety related equipment are maintained

within acceptable limits such that a safe

shutdown of the plant can be performed. In

addition, avoiding a plant transient during

environmental conditions that could

challenge the stability of the Electrical Power

System offsets any perceptible reduction in

the margin of safety as a result of the

proposed change. Thus, the proposed change

does not involve a significant reduction in

any margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts and Trowbridge,

2300 N Street, NW, Washington, DC

20037

NRC Section Chief: Stephen Dembek.

Federal Register / Vol. 65, No. 76 / Wednesday, April 19, 2000 / Notices 21041

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 notice was previously

published as a separate individual

notice. The notice content was the same

as above. It was published as an

individual notice either because time

did not allow the Commission to wait

for this biweekly notice or because the

action involved exigent circumstances.

It is 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.

Tennessee Valley Authority, Docket

Nos. 50–259, 50–260 and 50–296,

Browns Ferry Nuclear Plant, Units 1, 2

and 3, Limestone County, Alabama

Date of application for amendments:

March 29, 2000 (TS–402).

Brief description of amendments:

Changes Technical Specification 3/

4.6.4.1 ‘‘Secondary Containment’’ to

permit maintenance on a secondary

containment access door when one or

more units are operating and the other

door is closed.

Date of publication of individual

notice in the Federal Register: April 6,

2000 (65 FR 18141)

Expiration date of individual notice:

April 20, 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 Commission’s Public Document

Room, the Gelman Building, 2120 L

Street, NW., Washington, DC, and

electronically from the ADAMS Public

Library component on the NRC Web

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

Reading Room).

Carolina Power & Light Company, et al.,

Docket No. 50–400, Shearon Harris

Nuclear Power Plant, Unit 1, Wake and

Chatham Counties, North Carolina

Date of application for amendment:

August 26, 1999, as supplemented on

February 24, and March 14, 2000.

Brief description of amendment: This

amendment revises Technical

Specification 3/4.9.4, and its associated

bases, to allow the personnel airlock

and certain other containment

penetrations to remain open during

refueling operations provided specific

administrative controls are met. This

amendment is approved for use during

refueling outage 9 and operating cycle

10.

Date of issuance: March 27, 2000.

Effective date: March 27, 2000.

Amendment No.: 97.

Facility Operating License No. NPF–

63. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: October 6, 1999 (64 FR 54374).

The February 24, and March 14, 2000,

submittals contained clarifying

information only, and did not change

the initial no significant hazards

consideration determination. The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated March 27, 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:

May 20, 1999.

Brief description of amendments: The

amendments changed the Technical

Specification (TS) value for the

minimum suppression chamber water

level to a more conservative value.

Date of issuance: March 30, 2000.

Effective date: Immediately, to be

implemented within 60 days.

Amendment Nos.: 176 & 172.

Facility Operating License Nos. DPR–

19 and DPR–25: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: August 25, 1999 (64 FR

46426).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated March 30, 2000.

No significant hazards consideration

comments received: No.

Detroit Edison Company, Docket No.

50–341, Fermi 2, Monroe County,

Michigan

Date of application for amendment:

December 17, 1999, as supplemented

March 8, 2000.

Brief description of amendment: The

amendment revises Technical

Specification (TS) 2.1, ‘‘Safety Limits

(SLs),’’ changing the safety limit

minimum critical power ratio limits in

TS 2.1.1.2 to reflect the results of cyclespecific calculations performed for

Fermi 2 operating Cycle 8.

Date of issuance: March 30, 2000.

Effective date: As of the date of

issuance and shall be implemented

prior to the startup from the seventh

refueling outage.

Amendment No.: 138.

Facility Operating License No. NPF–

43: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: January 26, 2000 (65 FR 4269).

The March 8, 2000, letter provided

clarifying information that was within

the scope of the original Federal

Register notice and did not change the

staff’s initial proposed no significant

hazards consideration determination.

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated March 30, 2000.

No significant hazards consideration

comments received: No.

Detroit Edison Company, Docket No.

50–341, Fermi 2, Monroe County,

Michigan

Date of application for amendment:

July 30, 1999, as supplemented

December 17, 1999, and March 1, 2000.

21042 Federal Register / Vol. 65, No. 76 / Wednesday, April 19, 2000 / Notices

Brief description of amendment: The

amendment revises Technical

Specification (TS) 3.3.1.1, ‘‘Reactor

Protection System (RPS)

Instrumentation,’’ to reflect the

activation of the automatic trip

associated with the oscillation power

range monitor (OPRM). The amendment

also revises TS 3.4.1, ‘‘Recirculation

Loops Operating,’’ to remove

requirements related to the manual

detection and suppression of core

thermal-hydraulic instabilities because

these actions are no longer necessary

after the OPRM upscale function is

activated.

Date of issuance: March 31, 2000.

Effective date: As of the date of

issuance and shall be implemented

prior to the startup from the seventh

refueling outage.

Amendment No.: 139.

Facility Operating License No. NPF–

43: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: November 3, 1999 (64 FR

59800).

The December 17, 1999, and March 1,

2000, letters provided clarifying

information that was within the scope of

the original Federal Register notice and

did not change the staff’s initial

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated March 31, 2000.

No significant hazards consideration

comments received: No.

Entergy Operations, Inc., Docket No. 50–

313, Arkansas Nuclear One, Unit No. 1,

Pope County, Arkansas

Date of amendment request: August 6,

1998, as supplemented by letter dated

February 16, 2000.

Brief description of amendment: The

amendment revises the minimum and

the maximum concentration limits for

the sodium hydroxide tank. The

amendment also deletes the maximum

specified tank volume and revises the

minimum specified tank volume to refer

to the parameter used in the safety

analysis with no allowance for

instrument uncertainty.

Date of issuance: March 28, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment No.: 206.

Facility Operating License No. DPR–

51: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 10, 1999 (64 FR

6695).

The February 16, 2000, letter

provided clarifying information that did

not change the scope of the August 6,

1998, application and the initial

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated March 28, 2000.

No significant hazards consideration

comments received: No.

Entergy Operations, Inc., Docket No. 50–

368, Arkansas Nuclear One, Unit No. 2,

Pope County, Arkansas

Date of application for amendment:

January 27, 2000.

Brief description of amendment: The

amendment deleted the current

requirements of Technical Specification

(TS) 4.7.9.1.2.d, ‘‘Source installed in the

Boronometer,’’ associated with the

installed boronometer sealed source.

The source was recently removed and

stored, and the requirements of TS

4.7.9.1.2.d are no longer applicable.

Date of issuance: March 24, 2000.

Effective date: As of the date of

issuance to be implemented within 30

days from the date of issuance.

Amendment No.: 212.

Facility Operating License No. NPF–6:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 23, 2000 (65 FR

9007).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated March 24, 2000.

No significant hazards consideration

comments received: No.

Entergy Operations, Inc., Docket No. 50–

368, Arkansas Nuclear One, Unit No. 2,

Pope County, Arkansas

Date of application for amendment:

January 27, 2000.

Brief description of amendment: The

amendment relocated the schedule for

the withdrawal of reactor vessel

material surveillance specimens, from

the Technical Specifications to the

Safety Analysis Report, pursuant to the

guidance provided in Generic Letter 91–

01, ‘‘Removal of the Schedule for the

Withdrawal of Reactor Vessel Material

Specimens From Technical

Specifications.’’ Changes to the related

Bases were also made. In addition, the

proposed change to the surveillance

specimen removal schedule was

approved.

Date of issuance: April 4, 2000.

Effective date: As of the date of

issuance to be implemented within 30

days from the date of issuance.

Amendment No.: 213.

Facility Operating License No. NPF–6:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 23, 2000 (65 FR

9007).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 4, 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 26, 1999, as supplemented by

submittal dated December 7, 1999.

Brief description of amendment: This

amendment permits implementation of

10 CFR Part 50, Appendix J, Option B,

and reference Regulatory Guide 1.163,

‘‘Performance-Based Containment Leak

Test Program,’’ dated September 1995,

which specifies a method acceptable to

the NRC for complying with Option B.

These changes relate only to Type B and

C (local) leakage rate testing. The use of

Option B for Type A (integrated) leakage

rate testing was approved on February

22, 1996, by License Amendment No.

205.

Date of issuance: March 28, 2000.

Effective date: Immediately as of its

date of issuance and shall be

implemented within 120 days.

Amendment No.: 240.

Facility Operating License No. NPF–3:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 25, 1999 (64 FR

46437).

The letter of December 7, 1999,

contained clarifying information and

did not change the initial no significant

hazards consideration determination

and did not expand the scope of the

original Federal Register notice.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated March 28, 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:

September 9, 1999, as supplemented by

submittal dated February 28, 2000.

Brief description of amendment: This

amendment includes nine minor,

unrelated revisions to the technical

specifications (TSs). These revisions,

which are minor in both content and

safety significance, include

Federal Register / Vol. 65, No. 76 / Wednesday, April 19, 2000 / Notices 21043

clarifications and editorial changes to

the TSs.

Date of issuance: March 30, 2000.

Effective date: Immediately as of the

date of issuance and shall be

implemented within 90 days.

Amendment No.: 111.

Facility Operating License No. NPF–

58: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: November 3, 1999 (64 FR

59803).

The supplemental information

contained clarifying information and

did not change the initial no significant

hazards consideration determination

and did not expand the scope of the

original Federal Register notice.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated March 30, 2000.

No significant hazards consideration

comments received: No.

Indiana Michigan Power Company,

Docket Nos. 50–315 and 50–316, Donald

C. Cook Nuclear Plant, Units 1 and 2,

Berrien County, Michigan

Date of application for amendments:

December 3, 1998.

Brief description of amendments: The

amendments made administrative

changes to several Technical

Specifications to remove obsolete

information, provide consistency

between Unit 1 and Unit 2, provide

consistency with the Standard

Technical Specifications, provide

clarification, and correct typographical

errors.

Date of issuance: March 31, 2000.

Effective date: March 31, 2000, with

full implementation within 30 days.

Amendment Nos.: 243 and 224.

Facility Operating License Nos. DPR–

58 and DPR–74: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: August 31, 1999 (64 FR 47535).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated March 31, 2000.

No significant hazards consideration

comments received: No.

Nebraska Public Power District, Docket

No. 50–298, Cooper Nuclear Station,

Nemaha County, Nebraska

Date of amendment request:

December 6, 1999, as supplemented

March 17, 2000.

Brief description of amendment:

Amendment to technical specifications

changes the safety limit minimum

critical power ratio (SLMCPR) from 1.06

to 1.08 for two recirculation loop

operation and from 1.07 to 1.09 for

single recirculation loop operation.

Date of issuance: March 31, 2000.

Effective date: March 31, 2000, to be

implemented within 30 days.

Amendment No.: 182.

Facility Operating License No. DPR–

46: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: December 29, 1999 (64 FR

73093).

The March 17, 2000, letter provided

additional 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 March 31, 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:

December 6, 1999, as supplemented by

letters dated February 22 and March 14,

2000.

Brief description of amendment: The

amendment modifies the Technical

Specification (TS) surveillance

requirements associated with ensuring a

limited number of charging and high

pressure safety injection pumps are

incapable of injecting into the Reactor

Coolant System when the plant is

shutdown. In addition, the TS Bases are

modified to address these changes.

Date of issuance: March 30, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment No.: 243.

Facility Operating License No. DPR–

65: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 26, 2000 (65 FR 4285).

The February 22 and March 14, 2000,

supplemental letters provided clarifying

information that did not change the

staff’s original no significant hazards

consideration determination or expand

the scope of the application as

published.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated March 30, 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:

September 7, 1999.

Brief description of amendment: The

amendment removes the current special

exception which precludes applying the

18-month functional testing surveillance

to the Steam Generator Hydraulic

Snubbers for Technical Specification 3/

4.7.8, ‘‘Plant Systems, Snubbers.’’

Date of issuance: March 31, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 60 days from the date of

issuance.

Amendment No.: 244.

Facility Operating License No. DPR–

65: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 26, 2000 (65 FR 4283).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated March 31, 2000.

No significant hazards consideration

comments received: No.

Portland General Electric Company, et

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

Plant, Columbia County, Oregon

Date of application for amendment:

January 29, 1998.

Brief description of amendment: The

amendment deletes paragraph 2.D of

Facility Operating License No. NPF–1

and revises the Permanently Defueled

Technical Specifications (PDTS) by

deleting PDTS 5.7.1.1(b). These changes

remove the requirements for a security

program at the 10 CFR part 50 licensed

site once the spent nuclear fuel has been

relocated to the 10 CFR part 72 licensed

Independent Spent Fuel Storage

Installation.

Date of issuance: April 6, 2000.

Effective date: April 6, 2000, to be

implemented within 30 days after the

transfer of the last cask of spent nuclear

fuel from the spent fuel pool to the

independent spent fuel storage

installation is complete.

Amendment No.: 203.

Facility Operating License No. NPF–1:

The amendment changes the Operating

License and the Permanently Defueled

Technical Specifications.

Date of initial notice in Federal

Register: September 8, 1999 (64 FR

48865).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 6, 2000.

No significant hazards consideration

comments received: No.

21044 Federal Register / Vol. 65, No. 76 / Wednesday, April 19, 2000 / Notices

Public Service Electric & Gas Company,

Docket Nos. 50–272 and 50–311, Salem

Nuclear Generating Station, Unit Nos. 1

and 2, Salem County, New Jersey

Date of application for amendments:

July 23, 1999, as supplemented

September 13, 1999, and January 31,

2000.

Brief description of amendments: The

amendments revised the Technical

Specifications to remove the restriction

on performing the 24-hour endurance

run test of emergency diesel generators

(EDGs) every 18 months only during

shutdown. Additionally, for Salem Unit

1 only, a note associated with a onetime extension of a surveillance

requirement was deleted. Date of issuance: March 30, 2000.

Effective date: As of its date of

issuance and shall be implemented

within 60 days. Amendment Nos.: 229 and 210.

Facility Operating License Nos. DPR–

70 and DPR–75: The amendments

revised the Technical Specifications. Date of initial notice in Federal

Register: October 6, 1999 (64 FR 54380).

The January 31, 2000, supplement

provided clarifying information that did

not change the initial proposed no

significant hazards consideration

determination and did not expand the

scope of the original application as

published. The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated March 30, 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:

October 4, 1999.

Brief description of amendments: The

amendments revised the Technical

Specifications 5.5.6, ‘‘Prestressed

Concrete Containment Tendon

Surveillance Program,’’ to incorporate

three exceptions to Regulatory Guide

(RG) 1.35, Revision 2, 1976. The

exceptions concern the number of

tendons detensioned, inspection of

concrete adjacent to vertical tendons,

and the time during which areas

adjacent to tendons are inspected. Date of issuance: March 27, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment Nos.: Unit 1–112; Unit

2–90.

Facility Operating License Nos. NPF–

68 and NPF–81: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: February 9, 2000 (65 FR 6411). The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated March 27, 2000. No significant hazards consideration

comments received: No.

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of application for amendments:

April 29, 1999 (TS 99–04). Brief description of amendments: The

amendments delete Sequoyah Nuclear

Plant Technical Specification (TS)

monthly surveillance test on the

auxiliary feedwater suction pressure

switches.

Date of issuance: March 29, 2000.

Effective date: As of the date of

issuance, to be implemented no later

than 45 days after issuance. Amendment Nos.: 253 and 244.

Facility Operating License Nos. DPR–

77 and DPR–79: Amendments revise the

TS.

Date of initial notice in Federal

Register: May 19, 1999 (64 FR 27325). The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated March 29, 2000. No significant hazards consideration

comments received: No.

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of application for amendments:

August 30, 1999. Brief description of amendments:

Requirements related to containment

isolation valves that were located in two

different sections of the technical

specifications were consolidated into

one section. Also, conditions relating to

or usage of a check valve as an isolation

device was clarified.

Date of issuance: March 29, 2000.

Effective date: March 29, 2000.

Amendment Nos.: 254 and 245.

Facility Operating License Nos. DPR–

77 and DPR–79: Amendments revise the

technical specifications. Date of initial notice in Federal

Register: October 6, 1999 (64 FR 54382).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated March 29, 2000. No significant hazards consideration

comments received: No.

TXU Electric, Docket Nos. 50–445 and

50–446, Comanche Peak Steam Electric

Station, Unit Nos. 1 and 2, Somervell

County, Texas

Date of amendment request: January

13, 2000.

Brief description of amendments: The

amendments: (1) Revise Technical

Specification 3.8.3 (Condition B and

Surveillance Requirement (SR) 3.8.3.2)

to increase the required emergency

diesel generator (EDG) lube oil

inventory values; (2) Revise SR 3.8.3.2,

for EDG lube oil inventory, to add a note

stating that the surveillance is not

required to be performed until the diesel

has been in shutdown greater than 10

hours; and (3) Delete the footnote

associated with SR 3.8.4.7 which

provided a ‘‘one time only’’ alternative

to battery testing requirements.

Date of issuance: March 24, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment Nos.: 75 and 75.

Facility Operating License Nos. NPF–

87 and NPF–89: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: February 23, 2000 (65 FR 9012).

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated March 24, 2000.

No significant hazards consideration

comments received: No

TXU Electric, Docket Nos. 50–445 and

50–446, Comanche Peak Steam Electric

Station, Unit Nos. 1 and 2, Somervell

County, Texas

Date of amendment request: January

13, 2000.

Brief description of amendments: The

amendments add ‘‘NOTE 3’’ to

Surveillance Requirement 3.3.1.10 to

allow entry into MODES 2 or 1 without

the performance of N–16 detector

plateau verification until 72 hours after

achieving equilibrium conditions at

greater than or equal to 90 percent rated

thermal power.

Date of issuance: March 24, 2000.

Effective date:E As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment Nos.: 76 and 76.

Facility Operating License Nos. NPF–

87 and NPF–89: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: February 23, 2000 (65 FR 9013)

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated March 24, 2000.

No significant hazards consideration

comments received: No.

Federal Register / Vol. 65, No. 76 / Wednesday, April 19, 2000 / Notices 21045

TXU Electric, Docket Nos. 50–445 and

50–446, Comanche Peak Steam Electric

Station, Unit Nos. 1 and 2, Somervell

County, Texas

Date of amendment request: January

13, 2000.

Brief description of amendments: The

amendments add ‘‘NOTE 3’’ to

Surveillance Requirement 3.3.1.10 to

allow entry into MODES 2 or 1 without

the performance of N–16 detector

plateau verification until 72 hours after

achieving equilibrium conditions at

greater than or equal to 90 percent rated

thermal power.

Date of issuance: March 24, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendments Nos.: 76 and 76.

Facility Operating License Nos. NPF–

84 and NPF–89: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: February 23, 2000 (65 FR 9013).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated March 24, 2000.

No significant hazards consideration

comments received: No.

Union Electric Company, Docket No.

50–483, Callaway Plant, Unit 1,

Callaway County, Missouri.

Date of application for amendment:

December 3, 1999 (ULNRC–04158).

Brief description of amendment: The

amendment changed Section 5.6.6,

‘‘Reactor Coolant System (RCS) Pressure

and Temperature Limits Report

(PTLR),’’ of the improved Technical

Specifications (ITS) that were issued on

May 28, 1999, in Amendment No. 133.

The current Technical Specifications

(CTS) remain in effect until the ITS are

implemented on or before April 30,

2000. The changes to the ITS approve

the use of the PTLR by the licensee to

make changes to the plant pressure

temperature limits and low temperature

over pressure protection limits without

prior NRC staff approval, in accordance

with Generic Letter 96–03, ‘‘Relocation

of the Pressure Temperature Limit

Curves and Low Temperature

Overpressure Protection System

Limits,’’ dated January 31, 1996. The

changes (1) add the word criticality to

ITS subsection 5.6.6.a as one of the

reactor conditions for which RCS

pressure and temperature limits will be

determined, (2) add the phrase ‘‘and

COMS PORV,’’ where COMS PORV

stands for cold overpressure mitigation

system power operated relief valve, to

the introductory paragraph of ITS

subsection 5.6.6.b to show that the

analytical methods listed in the

subsection are also the COMS PORV,

and (3) replace the two documents

listed in ITS subsection 5.6.6.b by the

reference to the NRC letter that approves

use of the PTLR and the Westinghouse

topical report, WCAP–14040-NP-A,

Revision 2, ‘‘Methodology Used to

Develop Cold Overpressure Mitigating

System Setpoints and RCS Heatup and

Cooldown Limit Curves,’’ dated January

1996, that provides the methodology

that will be used by licensee in using

the PTLR report. The current plant

pressure temperature limits and low

temperature overpressure protection

limits are in the CTS and were approved

in Amendment No. 124, which was

issued April 2, 1998.

Date of issuance: March 24, 2000.

Effective date: March 24, 2000, to be

implemented no later than April 30,

2000.

Amendment No.: 134.

Facility Operating License No. NPF–

30: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: December 29, 1999 (64 FR

73101).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated March 24, 2000.

No significant hazards consideration

comments received: No

Union Electric Company, Docket No.

50–483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application for amendment:

January 14, 2000, as supplemented by

letter dated February 17, 2000 (ULNRC–

04172 and -04187).

Brief description of amendment: The

amendment revised several sections of

the improved Technical Specification

(ITSs) to correct 14 editorial errors made

in either (1) the application dated May

15, 1997, (and supplementary letters) for

the ITSs, or (2) the certified copy of the

ITSs that was submitted in the

licensee’s letters of May 27 and 28,

1999. The ITSs were issued as

Amendment No. 133 by the staff in its

letter of May 28, 1999, and will be

implemented by the licensee to replace

the current TSs by April 30, 2000.

Date of issuance: March 27, 2000.

Effective date: March 27, 2000, to be

implemented by April 30, 2000.

Amendment No.: 135.

Facility Operating License No. NPF–

30: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: February 23, 2000 (65 FR 9013)

and February 25, 2000 (65 FR 10118).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated March 27, 2000.

No significant hazards consideration

comments received: No.

Vermont Yankee Nuclear Power

Corporation, Docket No. 50–271,

Vermont Yankee Nuclear Power Station,

Vernon, Vermont.

Date of application for amendment:

June 15, 1999, as supplemented on

January 14, 2000.

Brief description of amendment: The

amendment revises Technical

Specifications (TSs) Sections 3.1/4.1

Reactor Protection System and 3.2/4.2

Protective Instrument Systems

instrumentation, tables, and the

associated Bases to increase the

surveillance test intervals (STIs), add

allowable out-of-service times (AOTs),

replace generic emergency core cooling

system actions for inoperable

instrument channels with functionspecific actions, and relocate selected

trip functions from the TSs to a Vermont

Yankee controlled document. In

addition, revision to TS Section 3.1/4.1

Reactor Protection System and the

associated Bases is proposed to remove

the RUN Mode APRM Downscale/IRM

High Flux/Inoperative Scram Trip

Function (APRM Downscale RUN Mode

SCRAM). The submittal also proposes to

implement editorial corrections and

administrative changes that do not alter

the meaning or intent of the

requirements.

Date of Issuance: April 3, 2000.

Effective date: As of the date of

issuance, and shall be implemented

within 90 days.

Amendment No.: 186.

Facility Operating License No. DPR–

28: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 20, 1999 (64 FR

56535).

The Commission’s related evaluation

of this amendment is contained in a

Safety Evaluation dated April 3, 2000.

No significant hazards consideration

comments received: No

Dated at Rockville, Maryland, this 12th day

of April 2000.

For the Nuclear Regulatory Commission.

John A. Zwolinski,

Director, Division of Licensing Project

Management, Office of Nuclear Reactor

Regulation.

[FR Doc. 00–9680 Filed 4–18–00; 8:45 am]

BILLING CODE 7590–01–P

21046 Federal Register / Vol. 65, No. 76 / Wednesday, April 19, 2000 / Notices

1See Securities Exchange Act Release No. 42360

(Jan. 28, 2000), 65 FR 5004 (Feb. 2, 2000)

(‘‘Decimals Order’’).

2Since the date of the Decimals Order, the

Commission approved the registration of the

International Securities Exchange (‘‘ISE’’) as a

national securities exchange. See Securities

Exchange Act Release No. 42455 (Feb. 24, 2000), 65

FR 11388 (March 2, 2000). On March 10, 2000, the

Commission included the ISE within the term

‘‘Participants’’ for purposes of the Decimals Order.

See Securities Exchange Act Release No. 42516

(March 10, 2000), 65 FR14637 (March 17, 2000)

(‘‘Extension Order’’).

3See Letters from Frank G. Zarb, Chairman and

Chief Executive Officer, NASD, to Arthur Levitt,

Chairman, Commission, dated March 6, 2000 and

March 21, 2000.

4Nasdaq has committed to stepping up its efforts

(including, at the Commission’s request, hiring an

independent consultant to advise on capacity

issues) to help ensure that it manages its growth

responsibly. The Commission expects, and has been

assured, that Nasdaq will dedicate substantial

SECURITIES AND EXCHANGE

COMMISSION

Proposed Collection; Comment

Request

Upon Written Requests, Copies

Available From: Securities and

Exchange Commission, Office of

Filings and Information Services,

Washington, DC 20549

Extension:

Rule 489 and Form F–N, SEC File No. 270–

361, OMB Control No. 3235–0411

Form 24F–2, SEC. File No. 270–399, OMB

Control No. 3235–0456

Notice is hereby given that, pursuant

to the Paperwork Reduction Act of 1995

(‘‘Act’’) [44 U.S.C. 3501 et seq.], the

Securities and Exchange Commission

(‘Commission’’) is soliciting comments

on the collections of information

summarized below. The Commission

plans to submit these existing

collections of information to the Office

of Management and Budget for

extension and approval.

Rule 489 under the Securities Act of

1993, Filing of Form by Foreign Banks

and Certain of their Holding Companies

and Finance Subsidiaries; and Form F–

N, Appointment of Agent for Service of

Process by Foreign Banks and Foreign

Insurance Companies and Certain of

Their Holding Companies and Finance

Subsidiaries Making Public Offerings of

Securities in the United States.

Rule 489 under the Securities Act of

1933 [17 CFR 230.489] requires foreign

banks and insurance companies and

holding companies and finance

subsidiaries of foreign banks and foreign

insurance companies that are excepted

from the definition of ‘‘investment

company’’ by virtue of Rules 3a–1, 3a–

5, and 3a–6 under the Investment

Company Act of 1940 to file Form F–N

to appoint an agent for service of

process in the United States when

making a public offering of securities.

Approximately seven entities are

required by Rule 489 to file Form F–N,

which is estimated to require an average

of one hour to complete. The estimated

annual burden of complying with the

rule’s filing requirement is

approximately eight hours, as one of the

entities has submitted multiple filings.

Under 17 CFR 270.24f–2, any openend management companies (‘‘mutual

funds’’), unit investment trusts (‘‘UITs’’)

or face-amount certificate companies

(collectively, ‘‘funds’’) that are deemed

to have registered an indefinite amount

of securities must, not later than 90 days

after the end of any fiscal year in which

it has publicly offered such securities,

file Form 24F–2 with the Commission.

Form 24F–2 is the annual notice of

securities sold by funds that

accompanies the payment of registration

fees with respect to the securities sold

during the fiscal year.

The Commission estimates that 8,203

funds file Form 24F–2 on the required

annual basis. The average annual

burden per respondent for Form 24F–2

is estimated to be one hour. The total

annual burden for all respondents to

Form 24F–2 is estimated to be 8,203

hours.

Compliance with the collection of

information required by Form 24F–2 is

mandatory. The Form 24F–2 filing that

must be made to the Commission is

available to the public.

The estimates of average burden hours

are made solely for the purposes of the

PRA and are not derived from a

comprehensive or even representative

survey or study of the cost of

Commission rules and forms. An agency

may not conduct or sponsor, and a

person is not required to respond to, a

collection of information unless it

displays a currently valid OMB control

number.

Written comments are invited on: (a)

Whether the proposed collection of

information is necessary for the proper

performance of the functions of the

agency, including whether the

information will have practical utility;

(b) the accuracy of the agency’s estimate

of the burden of the collection of

information; (c) ways to enhance the

quality, utility, and clarity of the

information collected; and (d) ways to

minimize the burden of the collection of

information on respondents, including

through the use of automated collection

techniques or other forms of information

technology. Consideration will be given

to comments and suggestions submitted

in writing within 60 days of this

publication.

Please direct your written comments

to Michael E. Bartell, Associate

Executive Director, Office of

Information Technology, Securities and

Exchange Commission, 450 Fifth Street,

NW., Washington, DC. 20549.

Dated: April 11, 2000.

Jonathan G. Katz,

Secretary.

[FR Doc. 00–9788 Filed 4–18–00; 8:45 am]

BILLING CODE 8010–01–M

SECURITIES AND EXCHANGE

COMMISSION

[Release No. 34–42685; File No. 4–430]

Order Staying the Deadlines for

Decimal Implementation and Notice of

Request for Comment on Revised

Decimal Implementation Schedules

April 13, 2000.

On January 28, 2000, the Securities

and Exchange Commission

(‘‘Commission’’) issued an order (the

‘‘Decimals Order’’) 1 requiring the

American Stock Exchange LLC

(‘‘AMEX’’), the Boston Stock Exchange,

Inc. (‘‘BSE’’), the Chicago Board Options

Exchange, Inc. (‘‘CBOE’’), the Chicago

Stock Exchange, Inc. (‘‘CHX’’), the

Cincinnati Stock Exchange, Inc.

(‘‘CSE’’), the National Association of

Securities Dealers, Inc. (‘‘NASD’’), the

New York Stock Exchange, Inc.

(‘‘NYSE’’), the Pacific Exchange, Inc.

(‘‘PCX’’), and the Philadelphia Stock

Exchange, Inc. (‘‘PHLX’’) (collectively

the ‘‘Participants’’) 2 to facilitate an

orderly transition to decimal pricing in

the United States securities markets.

The Decimals Order prescribed a

timetable for the Participants to begin

trading some equity securities (and

options on those equity securities) in

decimals by July 3, 2000, and all

equities and options by January 3, 2001.

On March 6, 2000, despite previous

assurances of readiness, the NASD

announced that The Nasdag Stock

Market Inc. (‘‘Nasdaq’’) would not have

sufficient capacity to meet the target

dates for implementation. 3 The NASD

also expressed concerns regarding

overall industry readiness and requested

that the Commission work with the

industry and the markets to determine

an appropriate time frame that would

not impose unnecessary risks on

investors. 4