62380 Federal Register / Vol. 65, No. 202 / Wednesday, October 18, 2000 / Notices

concurrence of the Subcommittee

Chairman; written statements will be

accepted and made available to the

Committee. Electronic recordings will

be permitted only during those portions

of the meeting that are open to the

public, and questions may be asked only

by members of the Subcommittee, its

consultants, and staff. Persons desiring

to make oral statements should notify

the cognizant ACRS staff person named

below five days prior to the meeting, if

possible, so that appropriate

arrangements can be made.

Further information regarding topics

to be discussed, the scheduling of

sessions open to the public, whether the

meeting has been canceled or

rescheduled, the Chairman’s ruling on

requests for the opportunity to present

oral statements, and the time allotted

therefor can be obtained by contacting

the cognizant ACRS staff person, Dr.

John T. Larkins (telephone: 301/415–

7360) between 7:30 a.m. and 4:15 p.m.

(EDT). Persons planning to attend this

meeting are urged to contact the above

named individual one or two working

days prior to the meeting to be advised

of any changes in schedule, etc., that

may have occurred.

Dated: October 12, 2000.

James E. Lyons,

Associate Director for Technical Support,

ACRS/ACNW.

[FR Doc. 00–26758 Filed 10–17–00; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Advisory Committee on Reactor

Safeguards Meeting of the

Subcommittee on Plant Systems;

Notice of Meeting

The ACRS Subcommittee on Plant

Systems will hold a meeting on October

31, 2000, 11545 Rockville Pike,

Rockville, Maryland.

The entire meeting will be open to

public attendance.

The agenda for the subject meeting

shall be as follows: Tuesday, October

31, 2000—8:30 a.m. until 12:00 noon.

The Subcommittee will discuss the

safety evaluation reports on the topical

reports for ABB/CE and Siemens Digital

I&C Applications. The purpose of this

meeting is to gather information,

analyze relevant issues and facts, and to

formulate proposed positions and

actions, as appropriate, for deliberation

by the full Committee.

Oral statements may be presented by

members of the public with the

concurrence of the Subcommittee

Chairman and written statements will

be accepted and made available to the

Committee. Electronic recordings will

be permitted only during those portions

of the meeting that are open to the

public, and questions may be asked only

by members of the Subcommittee, its

consultants, and staff. Persons desiring

to make oral statements should notify

the cognizant ACRS staff engineer

named below five days prior to the

meeting, if possible, so that appropriate

arrangements can be made.

During the initial portion of the

meeting, the Subcommittee, along with

any of its consultants who may be

present, may exchange preliminary

views regarding matters to be

considered during the balance of the

meeting.

The Subcommittee will then hear

presentations by and hold discussions

with representatives of the NRC staff,

and other interested persons regarding

this review.

Further information regarding topics

to be discussed, whether the meeting

has been canceled or rescheduled, and

the Chairman’s ruling on requests for

the opportunity to present oral

statements and the time allotted

therefor, can be obtained by contacting

the cognizant ACRS staff engineer, Mr.

Amarjit Singh (telephone 301/415–

6899) between 7:30 a.m. and 4:15 p.m.

(EDT). Persons planning to attend this

meeting are urged to contact the above

named individual one or two working

days prior to the meeting to be advised

of any potential changes to the agenda,

etc., that may have occurred.

Dated: October 6, 2000.

James E. Lyons,

Associate Director for Technical Support.

[FR Doc. 00–26759 Filed 10–17–00; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Biweekly Notice; Applications and

Amendments to Facility Operating

Licenses Involving No Significant

Hazards Considerations

I. Background

Pursuant to Public Law 97–415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 97–415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from September

25, 2000, through October 6, 2000. The

last biweekly notice was published on

October 4, 2000.

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. 202 / Wednesday, October 18, 2000 / Notices 62381

Written comments may be submitted

by mail to the Chief, Rules Review and

Directives Branch, Division of Freedom

of Information and Publications

Services, Office of Administration, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555–0001, and

should cite the publication date and

page number of this Federal Register

notice. Written comments may also be

delivered to Room 6D22, Two White

Flint North, 11545 Rockville Pike,

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

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

written comments received may be

examined at the NRC’s Public Document

Room, the Gelman Building, 2120 L

Street, NW, Washington, DC through

September 22, 2000. The NRC is

relocating its Public Document Room to

the NRC’s headquarters building.

Effective September 26, 2000,

documents may be examined at the

NRC’s Public Document Room, located

at One White Flint North, 11555

Rockville Pike (first floor), Rockville,

Maryland 20852. The filing of requests

for a hearing and petitions for leave to

intervene is discussed below.

By November 17, 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, located at One

White Flint North, 11555 Rockville Pike

(first Floor), Rockville, Maryland 20852.

Publicly available records will be

accessible and electronically from the

ADAMS Public Library component on

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

(the Electronic Reading Room). If a

request for a hearing or petition for

leave to intervene is filed by the above

date, the Commission or an Atomic

Safety and Licensing Board, designated

by the Commission or by the Chairman

of the Atomic Safety and Licensing

Board Panel, will rule on the request

and/or petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) The nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555–0001, Attention:

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, located at One

White Flint North, 11555 Rockville Pike

(first floor), Rockville, Maryland 20852.

Publicly available records will be

accessible and electronically from the

ADAMS Public Library component on

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

(the Electronic Reading Room).

AmerGen Energy Company, LLC, Docket

No. 50–289, Three Mile Island Nuclear

Station, Unit 1, Dauphin County,

Pennsylvania

Date of amendment request: August 9,

2000.

Description of amendment request:

The proposed amendment revises

Sections 6.5.3 and 6.5.4 of the Technical

Specifications to eliminate reference to

the Independent Onsite Safety Review

62382 Federal Register / Vol. 65, No. 202 / Wednesday, October 18, 2000 / Notices

Group (IOSRG) and to redefine the

performance of the IOSRG function by

the nuclear quality assurance

organization.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Operation of the facility in accordance

with the proposed amendment would not

involve a significant increase in the

probability of occurrence or the

consequences of an accident previously

evaluated. The proposed changes do not

affect assumptions contained in plant safety

analyses, the physical design and/or

operation of the plant, nor do they affect

Technical Specifications that preserve safety

analysis assumptions. None of the proposed

changes involve a physical modification to

the plant, a new mode of operation or a

change to the UFSAR [Updated Final Safety

Analysis Report] transient analyses. No

Technical Specification Limiting Condition

for Operation, Action Statement, or

Surveillance Requirement is affected by any

of the proposed changes. The proposed

changes do not alter the design, function, or

operation of any plant component. Therefore,

the proposed amendment does not affect the

probability of occurrence or consequences of

an accident previously evaluated.

2. Operation of the facility in accordance

with the proposed amendment would not

create the possibility of a new or different

kind of accident from any previously

evaluated. The proposed changes do not

affect assumptions contained in the plant

safety analyses, the physical design and/or

modes of plant operation defined in the plant

operating license, or Technical Specifications

that preserve safety analysis assumptions.

The proposed changes do not introduce a

new mode of plant operation or surveillance

requirement, nor involve a physical

modification to the plant. The proposed

changes do not alter the design, function, or

operation of any plant components.

Therefore, the proposed amendment does not

affect the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Operation of the facility in accordance

with the proposed amendment would not

involve a significant reduction in a margin of

safety. None of the proposed changes involve

a physical modification to the plant, a new

mode of operation or a change to the UFSAR

transient analyses. No Technical

Specification Limiting Condition for

Operation, Action Statement, or Surveillance

Requirement is affected. Therefore, the

proposed amendment does not reduce the

margin of safety.

Based upon the analysis provided herein

[the licensee’s August 9, 2000 application],

the proposed changes will not increase the

probability or consequences of an accident

previously evaluated, create the possibility of

a new or different kind of accident from any

accident previously evaluated, or involve a

reduction in a margin of safety. The

performance of safety assessment and the

IOSRG functions by a single qualified

organization will lead to efficiencies in the

performance of both functions. The training

and qualification of the personnel performing

the IOSRG functions will be unchanged from

the current requirements. Therefore, the

proposed changes meet the requirements of

10 CFR 50.92(c) and involve no significant

hazards consideration.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Edward J.

Cullen, Jr., Esq., PECO Energy Company,

2301 Market Street, S23–1,

Philadelphia, PA 19103.

NRC Section Chief: Marsha

Gamberoni.

AmerGen Energy Company, LLC, Docket

No. 50–289, Three Mile Island Nuclear

Station, Unit 1, Dauphin County,

Pennsylvania

Date of amendment request: August 9,

2000.

Description of amendment request:

The proposed amendment revises the

Three Mile Island Nuclear Station, Unit

1 (TMI–1), Updated Final Safety

Analysis Report (UFSAR), Section

14.1.2.10, ‘‘Steam Generator Tube

Failure Analysis,’’ to include the dose

resulting from the postulated postaccident steam release through the main

steam safety valves. The revised dose for

the TMI–1 steam generator tube failure

analysis would be increased above the

values previously reviewed and

approved by the NRC, but would

continue to be below the limits in Title

10 of the Code of Federal Regulations

(10 CFR) Part 100. The proposed change

to the UFSAR modifies the existing

analysis to account for release of

radioactivity to the atmosphere for the

postulated tube rupture analysis. The

existing dose calculations do not

account for this release. Editorial and

grammatical corrections are also made.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Operation of the facility in accordance

with the proposed amendment would not

involve a significant increase in the

probability of occurrence or the

consequences of an accident previously

evaluated. This change has no effect on

structures, systems or components prior to

the postulated steam generator tube failure

accident or any other accident. The proposed

change corrects the existing UFSAR Steam

Generator Tube Failure accident analysis to

account for the release to atmosphere through

the main steam safety valves (MSSVs). The

resulting revised radiological consequences

for the postulated Steam Generator Tube

Failure accident remain well below the 10

CFR 100 limits.

2. Operation of the facility in accordance

with the proposed amendment would not

create the possibility of a new or different

kind of accident from any previously

evaluated. This change has no impact on any

plant structures systems or components. The

only impact is the revised radiological

consequences of the Steam Generator Tube

Failure accident analysis to account for the

release to atmosphere through the MSSVs.

This change only corrects the existing TMI

Unit 1 UFSAR.

3. Operation of the facility in accordance

with the proposed amendment would not

involve a significant reduction in a margin of

safety. No change to any plant structure,

system or component is being made or

proposed by this change. This change does

not involve any change to safety system

setpoints for operation. The revised

radiological consequences of the Steam

Generator Tube Failure accident analysis

remain well below 10 CFR 100 limits.

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: Edward J.

Cullen, Jr., Esq., PECO Energy Company,

2301 Market Street, S23–1,

Philadelphia, PA 19103.

NRC Section Chief: Marsha

Gamberoni.

Calvert Cliffs Nuclear Power Plant, Inc.,

Docket Nos. 50–317 and 50–318, Calvert

Cliffs Nuclear Power Plant, Unit Nos. 1

and 2, Calvert County, Maryland

Date of amendments request:

September 14, 2000.

Description of amendments request:

The proposed amendment revises the

Unit 1 and Unit 2 heatup curves

(Technical Specification Figures 3.4.3–

1) and Unit 1 and Unit 2 cooldown

curves (Technical Specification Figures

3.4.3–2) to increase the allowable

heatup and cooldown rates. Use of

stress intensity factor KIC, permitted by

American Society of Mechanical

Engineers (ASME) Code Case N–640,

Federal Register / Vol. 65, No. 202 / Wednesday, October 18, 2000 / Notices 62383

made it possible to increase the heatup

and cooldown rates without changing

existing pressure-temperature (P–T)

limits. The existing P–T limits were

approved previously. Application of

Code Case N–640 to generate P–T curves

is not currently permitted by the

regulations. Therefore, pursuant to 10

CFR 50.12, a separate request for an

exemption to use Code Case N–640 was

submitted in a letter dated September

14, 2000.

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. Would not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

In accordance with 10 CFR Part 50,

Appendix G, the Calvert Cliffs pressure/

temperature (P–T) limits for material fracture

toughness requirements of the reactor coolant

pressure boundary materials were developed

using the methods of linear elastic fracture

mechanics and the guidance found in the

American Society of Mechanical Engineers

(ASME) Boiler and Pressure Vessel Code,

Section III, Appendix G. The Calvert Cliffs P–

T limits are based on fluence level. The

fluence levels are determined in the same

manner as the pressurized thermal shock

(PTS) screening criteria defined in 10 CFR

50.61 for the critical elements. Methods

described in the Nuclear Regulatory

Commission Regulatory Guide 1.99, Revision

2, are used to predict the embrittlement effect

of neutron irradiation on reactor vessel

materials. Regulatory Guide 1.99 defines

embrittlement effect in terms of adjusted

reference temperatures, which depends on

the material property of the PTS critical

elements.

The proposed higher heatup and cooldown

rates for the Technical Specification P–T

limits were made possible by the ASME Code

Case N–640 which permits use of reference

stress intensity factor KIC, in place of KIA.

Use of KIC, for the maximum stress intensity

factor that will not lead to failure, is the

correct value to use. Although conservative

in terms of developing P–T limits, use of KIA

results in a very restrictive heatup and

cooldown rate that challenges plant safety.

To bound the existing LTOP [lowtemperature overpressure protection] enable

temperatures, while increasing the heatup

and cooldown rates, the criteria described in

ASME Section XI Code Case 514 is used.

Code Case 514 is listed in Regulatory Guide

1.147 as acceptable to the Nuclear Regulatory

Commission (NRC) for this application. With

the new higher heatup and cooldown rates,

the underlying intent of the 10 CFR Part 50,

Appendix G, requirement for adequate

margin to prevent brittle failure of the reactor

coolant pressure boundary materials is

maintained. Additionally, since the

cooldown rates are not changed above 300°

F, the safety analyses and dose consequences

in the Updated Final Safety Analysis Report

are not affected.

Therefore the proposed change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Would not create the possibility of a new

or different [kind] of accident previously

evaluated.

The implementation of the proposed

revision has no significant effect on either the

configuration of the plant, or the manner in

which it is operated.

Therefore, this proposed change does not

create the possibility of a new or different

kind of accident from any previously

evaluated.

3. Would not involve a significant

reduction in a margin of safety.

As discussed above, although conservative

in terms of developing P–T limits, use of KIA

results in a very restrictive heatup and

cooldown rate that challenges plant safety.

The insignificant margin reduction in P–T

limits is more than compensated by the

safety benefits that are realized in terms of

plant component integrity as a result of the

higher heatup and cooldown rates. With the

proposed change, the underlying intent of the

10 CFR Part 50, Appendix G, requirement for

adequate margin to prevent brittle failure of

the reactor coolant pressure boundary

materials is maintained, and there is a net

gain in overall plant safety margin.

Therefore, this proposed change does not

significantly reduce the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendments request involves no

significant hazards consideration.

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Section Chief: Marsha

Gamberoni.

Calvert Cliffs Nuclear Power Plant, Inc.,

Docket Nos. 50–317 and 50–318, Calvert

Cliffs Nuclear Power Plant, Unit Nos. 1

and 2, Calvert County, Maryland

Date of amendments request:

September 14, 2000.

Description of amendments request:

The proposed amendment adds two

analytical methods to the list of

approved core operating limits

analytical methods in the Technical

Specifications (TSs) for Calvert Cliffs,

Unit Nos. 1 & 2. In a letter dated March

16, 2000, from Mr. S. A. Richards, NRC

to Mr. I. C. Rickard, ABB Combustion

Engineering, the Nuclear Regulatory

Commission approved the Topical

Report CENPD–387–P, ‘‘ABB Critical

Heat Flux Correlations for [pressurizedwater reactor] PWR Fuel’’ for

referencing in licensing applications for

Asea Brown Boveri, Inc. Combustion

Engineering, Inc. (ABB–CE) plants.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Would not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

The proposed change allows the use of the

ABB–NV and ABB–TV CHF [critical heat

flux] correlations in the thermal hydraulic

analysis for Calvert Cliffs Nuclear Power

Plant. The ABB–NV is used for a non-mixing

vane fuel assembly and the ABB–TV

correlations are used for Turbo mixing vane

fuel assembly. The CHF correlations

determine the departure from nucleate

boiling ratio (DNBR). The specified

acceptable fuel design limit for DNBR will

change for ABB–NV and ABB–TV. The use

of the ABB–NV and/or ABB–TV correlations

with the appropriate DNBR limit provides

additional operating margin for those

analyses that presently use the CE–1

correlation.

The use of a different CHF correlation will

not increase the probability of an accident

because the plant systems will not be

operated outside of design limits, the plant

equipment will not be operated in a different

manner, and system interfaces will not

change.

As Turbo fuel is introduced to reactor,

transistion cores will exist in which Turbo

mixing vane grid fuel assemblies are coresidents with non-mixing vane grid fuel

assemblies. The grid hydraulic loss

coefficient in the Turbo grids is greater than

the grid hydraulic loss coefficient for the

non-mixing grids. The flow diversion that

will result does not increase the probability

of an accident previously evaluated because

assembly flow has no impact on accident

initiators, and because plant systems will not

be operated outside of design limits, plant

equipment will not be operated in a different

manner, and system interfaces will not

change.

The change in the CHF correlation was the

subject of Topical Report CENPD–387–P–A,

which was reviewed and approved by the

NRC. The use of a different CHF correlation

will not increase the consequences of an

accident because Limiting Conditions [for]

Operation (LOCs) will continue to restrict

operation to within the regions that provide

acceptable results, and Reactor Protection

System (RPS) trip setpoints will plant

transients so that the consequences of

accidents will be acceptable.

The transition cores that will exist as

Turbo fuel is introduced to the reactor will

not increase the consequences of an accident.

The TORC code accurately predicts the flow

conditions in adjacent fuel bundles that

contain grids with different designs and

coefficients. The flow diversion will be

compensated for by DNBR margin gains.

Operation within the LOCs and RPS

setpoints will continue to restrict plant

62384 Federal Register / Vol. 65, No. 202 / Wednesday, October 18, 2000 / Notices

transients so that consequences of accidents

will be acceptable.

Therefore, the proposed TS changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Would not create the possibility of a new

or different [kind] of accident from any

accident previously evaluated.

The proposed change does not add any

new equipment, modify any interfaces with

any existing equipment, alter the

equipment’s function or change the method

of operating the equipment. The proposed

change does not alter plant conditions in a

manner that could affect other plant

components. The proposed change does not

cause any existing equipment to become an

accident initiator. The Turbo grid design

does not introduce features that could initiate

an accident.

Therefore the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Would not involve a significant

reduction in the margin of safety.

Safety Limits ensure that specified

acceptable fuel design limits are not

exceeded during steady state operation,

normal operational transients, and

anticipated operational occurrences. One of

the safety limits that accomplishes this is the

DNBR limit. The CHF correlations that have

been approved for ABB–NV and ABB–TV

result in a DNBR limit that provides a 95%

probability, at a 95% confidence, that the hot

fuel rod in the core will not experience

departure from nucleate boiling. The RPS in

combination with the LCOs, will continue to

prevent any anticipated combination of

transient conditions for reactor coolant

system temperature, pressure and thermal

power level that would result in a violation

of the Safety Limits.

Therefore the margin of safety is not

significantly reduced by this proposed

change.

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

amendments request involves no

significant hazards consideration.

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Section Chief: Marsha

Gamberoni.

Calvert Cliffs Nuclear Power Plant, Inc.,

Docket Nos. 50–317 and 50–318, Calvert

Cliffs Nuclear Power Plant, Unit Nos. 1

and 2, Calvert County, Maryland

Date of amendments request:

September 14, 2000.

Description of amendments request:

Calvert Cliffs Nuclear Power Plant, Inc.

(CCNPPI) proposed an amendment to

incorporate changes described below

into the Technical Specifications (TSs)

for Calvert Cliffs Units 1 and 2.

On September 9, 1996, a final rule

amending 10 CFR 50.55a was issued

requiring owners to implement, by

September 9, 2001, the requirements of

the 1992 Edition through the 1992

Addenda of the American Society of

Mechanical Engineers Boiler and

Pressure Vessel Code (ASME Code)

Section XI, Subsections IWE and IWL,

as modified and supplemented by 10

CFR 50.55a. CCNPPI has developed a

program to effect the implementation of

Subsections IWE and IWL. This

submittal requests a license amendment

in support of the program.

The TSs change replaces the reference

to Regulatory Guide (RG) 1.35 with a

reference to Section XI of the ASME

Code, and deletes the applicability of

Surveillance Requirement 3.0.2.

Compliance with RG 1.35 is not

sufficient to comply with 10 CFR

50.55a, as amended, and inspection

frequencies will be in accordance with

Subsection IWL of Section XI; therefore,

Surveillance Requirement 3.0.2 will no

longer apply.

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. Would not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

The Containment Building is a passive

safety structure that prevents the release of

radioactive materials to the environment in

post-accident conditions. The proposed

Technical Specification change updates

requirements of the Technical Specifications

that have been made obsolete by the

improvements of the Containment [B]uilding

inspections required by the changes in the

regulations. The improved inspections

required by the American Society of

Mechanical Engineers [Boiler and Pressure

Vessel] Code serve to maintain containment

response to accident conditions, by causing

the identification and repair of defects in

Containment Buildings.

Relocating existing requirements,

eliminating requirements that duplicate

regulations, and making administrative

improvements provide Technical

Specifications that are easier to use. Because

existing requirements are controlled by

regulation, there is no reduction in

commitment and adequate control is still

maintained. Therefore, the proposed change

would not involve a significant increase in

probability or consequences of an accident

previously evaluated.

2. Would not create the possibility of a new

or different [kind] of accident from any

accident previously evaluated.

The Containment Building is a passive

safety structure designed to contain

radioactive materials released from the

reactor coolant system. The performance of

the Containment Building is not evaluated as

the causal factor in any accident at Calvert

Cliffs Nuclear Power Plant. The proposed

Technical Specification change updates

requirements of the Technical Specifications

that were made obsolete by the

improvements of the Containment [B]uilding

inspections required by the changes in the

regulations. Revising the Technical

Specifications, to comply with current

regulations and to eliminate duplication of

requirements, does not create the possibility

of a new or different [kind] of accident from

any previously evaluated.

3. Would not involve a significant

reduction in a margin of safety.

The safety function of the Containment

Building is to provide a boundary to the

release of radioactive material to the

environment during post-accident

conditions. The change to the Technical

Specifications incorporate[s] improved

inspection techniques and criteria to ensure

optimum containment integrity and,

therefore, optimum containment response in

the event of an accident resulting in a release

of radioactive material from the reactor

coolant system. Optimizing containment

integrity will result in maintaining the

margin of safety allowed by the Containment

Buildings. 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

amendments request involves no

significant hazards consideration.

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Section Chief: Marsha

Gamberoni.

Calvert Cliffs Nuclear Power Plant, Inc.,

Docket Nos. 50–317 and 50–318, Calvert

Cliffs Nuclear Power Plant, Unit Nos. 1

and 2, Calvert County, Maryland

Date of amendments request:

September 15, 2000.

Description of amendments request:

The proposed amendment revises the

Unit 1 and Unit 2 Technical

Specification Surveillance Requirement

(SR) 3.1.7.2 which verifies that each

control element assembly (CEA) not

fully inserted is capable of full insertion

when tripped from at least the 50

percent withdrawn position.

Specifically, the proposed amendment

adds a note to SR 3.1.7.2, which allows

the SR to not be performed during

initial power escalation following a

refueling outage if SR 3.1.4.6 (CEA drop

time test) has been met. In addition,

‘‘once’’ was added to the SR frequency

Federal Register / Vol. 65, No. 202 / Wednesday, October 18, 2000 / Notices 62385

as an administrative change to clarify

that the SR is only performed once and

not on a periodic basis. This proposed

license amendment is consistent with

Technical Specification Task Force

(TSTF)–134, Revision 1, which received

Nuclear Regulatory Commission (NRC)

approval on April 21, 1998.

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. Would not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

A risk assessment was performed to

support a prior license amendment request

submitted to change Surveillance

Requirement (SR) 3.1.7.2 frequency from 24

hours to 7 days. Results of a study performed

in support of the risk assessment indicated

no change in the geometry of those

components utilized in control element

assembly (CEA) insertion over the 7-day

period. The study also evaluated electronic/

electrical failures that could cause a CEA to

be stuck, concluding that the feature that

controls the movement of the CEAs is not

time-related. Since there have been no

modifications performed on the components

analyzed or changes in the manner in which

they are operated, it is reasonable to assume

that the conclusions remain valid.

The CEA drop time test SR 3.1.4.6 proves

that any work done during the refueling

outage does not prevent the rods from

tripping. Revising SR 3.1.7.2, such that it

could allow more than seven days from

successfully performing the CEA drop time

test does not change this. However, as with

any component, there will eventually be

some time-related degradation that may

impact the ability of the CEAs to drop. Thus,

when the seven days are exceeded, there is

some negligible increase in the probability

that a rod would fail to drop. This causes an

insignificant increase in core damage

frequency because it requires multiple rod

failures to cause core damage in the event of

an overcooling event (the most bounding

accident for a stuck CEA during rod worth

testing). This additional risk is believed to be

small since the degradation is the result of

core changes, which occur slowly, and not

the result of maintenance. Thus the risk

increase due to this Technical Specification

change is considered to be negligible. The

probability of an overcooling event is not

changed by the proposed change.

Therefore the proposed change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Would not create the possibility of a new

or different [kind] of accident from any

accident previously evaluated.

The proposed change to the surveillance

requirement for CEA trippability does not

result in any change to the facility or the

manner in which it is operated.

Therefore, this proposed change does not

create the possibility of a new or different

kind of accident from any previously

evaluated.

3. Would not involve a significant

reduction in a margin of safety.

Operation of the facility in accordance

with this proposed amendment does not

involve a significant reduction in a margin of

safety. Control element assembly trippability

is still demonstrated via performance of SR

3.1.4.6. The risk increase due to this change

is considered to be negligible. Thus,

appropriate equipment continues to be tested

in a manner and at a frequency necessary to

provide reasonable assurance that the

equipment can perform its assumed safety

function.

Furthermore, this change is consistent with

Technical Specification Task Force (TSTF)–

134, Revision 1, which has been approved by

the Nuclear Regulatory Commission.

Adopting testing practices consistent with

those specified in TSTF–134, Revision 1 are

acceptable based on similar design, likecomponent testing for the system application

and the availability of other Technical

Specification requirements which provide

regular checks to ensure limits are met.

Therefore, this proposed modification does

not significantly reduce the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendments request involves no

significant hazards consideration.

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW,

Washington, DC 20037.

NRC Section Chief: Marsha

Gamberoni.

Duke Energy Corporation, Docket Nos.

50–369 and 50–370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg

County, North Carolina

Date of amendment request: August 1,

2000.

Description of amendment request:

The proposed amendments would

provide revised spent fuel pool

configurations, revised spent fuel pool

storage criteria, and revised fuel

enrichment and burnup requirements

which take credit for soluble boron in

maintaining acceptable margins of

subcriticality in the spent fuel storage

pools. Also, the proposed amendments

would provide additional criteria for

ensuring acceptable levels of

subcriticality in the spent fuel storage

pools.

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 the change involve a significant

increase in the probability or consequence of

an accident previously evaluated?

No, based upon the following:

Dropped Fuel Assembly

There is no significant increase in the

probability of a fuel assembly drop accident

in the spent fuel pools when considering the

degradation of the or Boraflex panels in the

spent fuel pool racks coupled with the

presence of soluble boron in the spent fuel

pool water for criticality control. The

handling of the fuel assemblies in the spent

fuel pool has always been performed in

borated water, and the quantity of Boraflex

remaining in the racks has no affect on the

probability of such a drop accident.

The criticality analysis showed that the

consequences of a fuel assembly drop

accident in the spent fuel pools are not

affected when considering the degradation of

the Boraflex in the spent fuel pool racks and

the presence of soluble boron.

Fuel Misloading

There is no significant increase in the

probability of the accidental misloading of

spent fuel assemblies into the spent fuel pool

racks when considering the degradation of

the Boraflex in the spent fuel pool racks and

the presence of soluble boron in the pool

water for criticality control. Fuel assembly

placement and storage will continue to be

controlled pursuant to approved fuel

handling procedures to ensure compliance

with the Technical Specification

requirements. These procedures will be

revised as needed to comply with the revised

requirements which would be imposed by

the proposed Technical Specification

changes. Note that the proposed amendment

would increase the number of different

storage limits in Technical Specification

3.7.15. However, these revised storage limits

were developed with input from station

personnel. Their awareness, in conjunction

with any procedure changes as described

above, will provide additional assurance that

an accidental misloading of a spent fuel

assembly will not occur.

There is no increase in the consequences

of the accidental misloading of spent fuel

assemblies into the spent fuel pool racks

because criticality analyses demonstrate that

the pool will remain subcritical following an

accidental misloading if the pool contains an

adequate soluble boron concentration.

Current Technical Specification 3.7.14 will

ensure that an adequate spent fuel pool boron

concentration is maintained in the McGuire

spent fuel storage pools. A McGuire Station

UFSAR change will revise Chapter 16,

‘‘Selected Licensee Commitments’’, to

provide for adequate monitoring of the

remaining Boraflex in the spent fuel pool

racks. If that monitoring identifies further

reductions in the Boraflex panels which

would not support the conclusions of the

McGuire Criticality Analysis, then the

McGuire TS’s and design bases would be

revised as needed to ensure that acceptable

subcriticality are maintained in the McGuire

spent fuel storage pools.

62386 Federal Register / Vol. 65, No. 202 / Wednesday, October 18, 2000 / Notices

Significant Change in Spent Fuel Pool

Temperature

There is no significant increase in the

probability of either the loss of normal

cooling to the spent fuel pool water or a

decrease in pool water temperature from a

large emergency makeup when considering

the degradation of the Boraflex in the spent

fuel pool racks and the presence of soluble

boron in the pool water for subcriticality

control since a high concentration of soluble

boron has always been maintained in the

spent fuel pool water. Current Technical

Specification 3.7.14 will ensure that an

adequate spent fuel pool boron concentration

is maintained in the McGuire spent fuel

storage pools.

A loss of normal cooling to the spent fuel

pool water causes an increase in the

temperature of the water passing through the

stored fuel assemblies. This causes a decrease

in water density that would result in a

decrease in reactivity when Boraflex neutron

absorber panels are present in the racks.

However, since a reduction in the amount of

Boraflex present in the racks is considered,

and the spent fuel pool water has a high

concentration of boron, a density decrease

causes a positive reactivity addition.

However, the additional negative reactivity

provided by the current boron concentration

limit, above that provided by the

concentration required to maintain keff less

than or equal to 0.95 (1470 ppm), will

compensate for the increased reactivity

which could result from a loss of spent fuel

pool cooling event. Because adequate soluble

boron will be maintained in the spent fuel

pool water, the consequences of a loss of

normal cooling to the spent fuel pool will not

be increased. Current Technical Specification

3.7.14 will ensure that an adequate spent fuel

pool boron concentration is maintained in

the McGuire spent fuel storage pools.

A decrease in pool water temperature from

a large emergency makeup causes an increase

in water density that would result in an

increase in reactivity when Boraflex neutron

absorber panels are present in the racks.

However, the additional negative reactivity

provided by the current boron concentration

limit, above that provided by the

concentration required to maintain keff less

than or equal to 0.95 (1470 ppm), will

compensate for the increased reactivity

which could result from a decrease in spent

fuel pool water temperature. Because

adequate soluble boron will be maintained in

the spent fuel pool water, the consequences

of a decrease in pool water temperature will

not be increased. Current Technical

Specification 3.7.14 will ensure that an

adequate spent fuel pool boron concentration

is maintained in the McGuire spent fuel

storage pools.

2. Will the change create the possibility of

a new or different kind of accident from any

previously evaluated?

No. Criticality accidents in the spent fuel

pool are not new or different types of

accidents. They have been analyzed in

Section 9.1.2.3 of the Updated Final Safety

Analysis Report and in Criticality Analysis

reports associated with specific licensing

amendments for fuel enrichments up to 4.75

weight percent U–235. Specific accidents

considered and evaluated include fuel

assembly drop, accidental misloading of

spent fuel assemblies into the spent fuel pool

racks, and significant changes in spent fuel

pool water temperature. The accident

analysis in the Updated Final Safety Analysis

Report remains bounding.

The possibility for creating a new or

different kind of accident is not credible. The

amendment proposes to take credit for the

soluble boron in the spent fuel pool water for

reactivity control in the spent fuel pool while

maintaining the necessary margin of safety.

Because soluble boron has always been

present in the spent fuel pool, a dilution of

the spent fuel pool soluble boron has always

been a possibility, however, a criticality

accident resulting from a dilution accident

was not considered credible. For the

proposed amendment, the spent fuel pool

dilution evaluation (Attachment 7)

demonstrates that a dilution of the boron

concentration in the spent fuel pool water

which could increase the rack keff to greater

than 0.95 (constituting a reduction of the

required margin to criticality) is not a

credible event. The requirement to maintain

boron concentration in the spent fuel pool

water for reactivity control will have no

effect on normal pool operations and

maintenance. There are no changes in

equipment design or in plant configuration.

This new requirement will not result in the

installation of any new equipment or

modification of any existing equipment.

Therefore, the proposed amendment will not

result in the possibility of a new or different

kind of accident.

3. Will the change involve a significant

reduction in a margin of safety?

No. The proposed Technical Specification

changes and the resulting spent fuel storage

operating limits will provide adequate safety

margin to ensure that the stored fuel

assembly array will always remain

subcritical. Those limits are based on a plant

specific criticality analysis (Attachment 6)

based on the ‘‘Westinghouse Spent Fuel Rack

Criticality Analysis Methodology’’ described

in Reference 1. The Westinghouse

methodology for taking credit for soluble

boron in the spent fuel pool has been

reviewed and approved by the NRC

(Reference 6). This methodology takes partial

credit for soluble boron in the spent fuel pool

and requires conformance with the following

NRC Acceptance criteria for preventing

criticality outside the reactor:

(1) keff shall be less than 1.0 if fully flooded

with unborated water which includes an

allowance for uncertainties at a 95%

probability, 95% confidence (95/95) level;

and

( 2) keff shall be less than or equal to 0.95

if fully flooded with borated water, which

includes an allowance for uncertainties at a

95/95 level.

The criticality analysis utilized credit for

soluble boron to ensure keff will be less than

or equal to 0.95 under normal circumstances,

and storage configurations have been defined

using a 95/95 keff calculation to ensure that

the spent fuel rack keff will be less than 1.0

with no soluble boron. Soluble boron credit

is used to provide safety margin by

maintaining keff less than or equal to 0.95

including uncertainties, tolerances and

accident conditions in the presence of spent

fuel pool soluble boron. The loss of

substantial amounts of soluble boron from

the spent fuel pool which could lead to

exceeding a keff of 0.95 has been evaluated

(Attachment 7) and shown to be not credible.

Accordingly, the required margin to

criticality is not reduced.

The evaluations in Attachment 7, which

show that the dilution of the spent fuel pool

boron concentration from the conservative

assumed initial boron concentration (2475

ppm) to the minimum boron concentration

required to maintain keff ≤ 0.95 (730 ppm) is

not credible, combined with the 95/95

calculation which shows that the spent fuel

rack keff will remain less than 1.0 when

flooded with unborated water, provide a

level of safety comparable to the conservative

criticality analysis methodology required by

References 2, 3 and 4.

Therefore the proposed changes in this

license amendment will not result in a

significant reduction in the facility’s 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: Ms. Lisa F.

Vaughn, Duke Energy Corporation, 422

South Church Street, Charlotte, North

Carolina 28201–1006.

NRC Section Chief: Richard L. Emch,

Jr.

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 amendment requests:

September 26, 2000.

Description of amendment requests:

The proposed amendments would

revise the current licensing basis in the

Updated Final Safety Analysis Report

by requiring operator action to mitigate

the effects of a loss of seal injection

(LOSI) cooling to the reactor coolant

pumps (RCPs).

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the change involve a significant

increase in the probability of occurrence or

consequences of an accident previously

evaluated?

The proposed change to the licensing basis

recognizes that if RCP Number 1 seal leak-off

rates are low, continuous RCP operation

following a sustained LOSI may no longer be

permitted. Tripping the plant, securing the

affected RCPs, and maintaining hot standby

Federal Register / Vol. 65, No. 202 / Wednesday, October 18, 2000 / Notices 62387

conditions following a sustained LOSI will

permit adequate RCP seal cooling by readily

achievable process controls. These actions

ensure that the probability of developing

excessive seal leakage equivalent to that of a

previously evaluated loss of coolant accident

(LOCA), has not been significantly increased.

Plant and RCP tripping are anticipated

transients that do not involve plant operation

outside design limits.

The consequences of large- and small-break

(SB) LOCAs have been evaluated and it has

been shown that the radiological

consequences of these events do not result in

unacceptable exposures to members of the

public. Therefore, even if stopping of the

RCPs following a LOSI and control of process

parameters as described above does not

preclude RCP seal failures, the consequences

of such failure are bounded by the current

accident analysis.

Therefore, the probability of occurrence or

the consequences of accidents previously

evaluated are not significantly increased.

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

The leakage resulting from failed RCP seals

may be large enough to be considered a

SBLOCA and industry data on SBLOCA

initiating frequencies includes the

contribution from failed RCP seals. SBLOCAs

are a previously evaluated class of accidents.

There is no new or different kind of accident

created as a result of this change.

Therefore, the change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. Does the change involve a significant

reduction in a margin of safety?

The original design objective for the

controlled leakage seal assemblies in the

RCPs was to permit sufficient controlled

leakage following a LOSI, such that cooling

of the leakage provided by the thermal barrier

heat exchanger would be sufficient to

continue RCP operation unabated without

challenging seal integrity. This is an implied

margin of safety for seal integrity, even if not

explicitly defined in the basis for any

Technical Specification. It has been

postulated that the reduced seal leak-off will

no longer permit continuous RCP operation

following a LOSI. The proposed change to

the licensing basis recognizes this condition

and requires pump tripping if seal injection

cannot be restored prior to receiving high

temperature alarms in the leak-off return

lines. Pump tripping reduces the heat

generated in the pump and permits readily

achievable process controls to maintain

adequate seal cooling and an adequate

margin to seal failure.

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: David W.

Jenkins, Esq., 500 Circle Drive,

Buchanan, MI 49107.

NRC Section Chief: Claudia M. Craig.

Indiana Michigan Power Company,

Docket No. 50–316, Donald C. Cook

Nuclear Plant, Unit 2, Berrien County,

Michigan

Date of amendment request:

September 30, 2000.

Description of amendment request:

The proposed amendment would allow

an extension of the steam generator tube

inspection surveillance requirements of

Technical Specification (T/S)

Surveillance Requirement 4.4.5.3. The

proposed amendment would prevent a

mid-cycle shutdown to meet the

required 40-calendar month inspection

interval of SR 4.4.5.3 and would allow

the steam generator tube inspection to

be performed during the refueling

outage following the current operating

cycle.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the change involve a significant

increase in the probability of occurrence or

consequences of an accident previously

evaluated?

The accident considered applicable to the

proposed change is a steam generator tube

rupture (SGTR). The precursors/initiators of

a SGTR (degraded, defective, or leaking

tubes) are not known or expected to be

present in the CNP [Cook Nuclear Plant] Unit

2 steam generators. These steam generators

were newly installed in 1988, and include

corrosion prevention design features not

included in previous generations of steam

generators.

There are no active degradation

mechanisms present in the Unit 2 steam

generators. Any tube imperfections that may

be present or that may be initiated during the

current operating cycle are not expected to

progress to the point of tube failure before the

next refueling outage.

Considering the condition of the steam

generators and the operational time between

inspections, the proposed change will not

significantly increase the probability of

occurrence of an accident.

The proposed change will not affect the

scope, methodology, acceptance limit, or

corrective measures of the existing steam

generator examination program.

Unit 2 recently completed an extended

shutdown that effectively limited the

operational time that is the basis for the

surveillance frequency. When the reactor is

shut down and the reactor coolant system is

at a reduced temperature, the steam

generators are not subject to conditions that

lead to significant tube degradation. Based on

power operation time, the proposed

extension will not increase the operating

interval between surveillances beyond that

currently allowed by [the] T/S.

The steam generator tube inspection

interval is not used in the SGTR accident

analysis. The proposed change will,

therefore, not affect the accident analysis or

methodology.

The severity of an analyzed tube rupture

event is not related to the time interval

between inspections. The proposed change

does not affect allowable leakage rates or

source terms, and does not change the

duration of an SGTR or the response to the

event. Because the severity of an accident is

not increased by the proposed change, there

is no impact on offsite dose considerations.

Therefore, the probability of occurrence or

the consequences of accidents previously

evaluated are not significantly increased.

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

The proposed change will not result in a

change in plant configuration or operation.

Plant systems and components will not be

operated in a different manner because of

this change. The proposed change does not

affect or create new accident initiators or

precursors.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Does the change involve a significant

reduction in a margin of safety?

The T/S limit of one gallon per minute

total steam generator tube leakage ensures the

offsite dose from tube leaks is limited to a

small fraction of 10 CFR 100 limits. The T/

S leakage limit of 500 gallons per day in one

steam generator is based on ensuring tube

integrity in the event of a steam line rupture

or loss of coolant accident. Because the

offsite dose considerations from steam

generator tube failures are limited by the

primary-to-secondary leak rate program and

not the tube inspection program, the

proposed change has no impact on offsite

dose.

There are no tubes in service in any of the

Unit 2 steam generators that were found to

be degraded, and no active steam generator

tube degradation is known to be occurring.

Therefore, the available margin in tube wall

thickness is not being significantly reduced.

During the last inspection, 50% of the tubes

were inspected (more than sixteen times the

T/S requirement), and none were found to

exceed the plugging limit, providing

additional assurance that safety margins are

not being reduced. The absence of tube

degradation, along with the material and

design features and chemistry controls,

provide reasonable assurance that tube repair

limits will not be approached during the

current operating cycle.

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

62388 Federal Register / Vol. 65, No. 202 / Wednesday, October 18, 2000 / Notices

amendment requests involve no

significant hazards consideration.

Attorney for licensee: David W.

Jenkins, Esq., 500 Circle Drive,

Buchanan, MI 49107.

NRC Section Chief: Claudia M. Craig.

Northeast Nuclear Energy Company, et

al., Docket No. 50–423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of amendment request: June 30,

2000, as supplemented September 22,

2000.

Description of amendment request:

The proposed changes would modify

Sections 2.4.13.5, ‘‘Design Bases for

Subsurface Hydrostatic Loading’’

2.5.4.6.1, ‘‘Design Basis for

Groundwater’’ 3.4.1.2, ‘‘Permanent

Dewatering System’’ 3.8.1.6.4,

‘‘Waterproofing Membrane’’ 3.8.1.6.5,

‘‘Steel Liner and Penetrations’’ 9.3.3.1,

‘‘Reactor Plant Vent and Drain Systems,

Design Bases’’ 9.3.3.2.4, ‘‘Reactor Plant

Aerated Drains System’’ 9.3.3.2.4.1,

‘‘Safety-Related Containment

Recirculation Cubicle Sump’’ 9.3.3.3,

‘‘Safety Evaluation’’ 9.3.3.4, ‘‘Tests and

Inspections’’ and 12.3.1.3.2, ‘‘PostAccident Access to Vital Areas’’ Tables

1.8–1, 3.2–1, 8.3–3, 12.3–3, and 12.3–4;

and Figures 3.8–67 and 9.3–6 of the

Final Safety Analysis Report (FSAR) to

reflect the addition of the new

subsystem and its impact on other

safety-related systems. The new sump

pump system creates the possibility of

a malfunction of a different type than

previously evaluated in the FSAR

because of the system’s dependence on

electrical power; only one nonenvironmentally qualified, non-safetyrelated pump is provided; and portions

of the Engineered Safety Feature

Building structure are now credited

with preventing Recirculation Spray

System (RSS) cubicle flooding.

Additionally, the proposed changes

involve deviations from safety

classification and ‘‘code & standards,’’

Standard Review Plan 3.4.1 and

Regulatory Guide (RG) 1.26.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

This license amendment request deals with

changes in Millstone Unit No. 3 Final Safety

Analysis Report (FSAR) due to the

installation of a new sump pump system in

the Engineered Safety Features Building

(ESFB). The sump pump system which

prevents inleakage through the containment

basemat is not connected to and is fully

independent of the reactor coolant system.

Therefore, the proposed changes to this

system will not increase the probability of

occurrence of a Loss of Coolant Accident

(LOCA). The new system is a support system

for the Recirculation Spray System (RSS) and

containment protective boundary which are

mitigation design features. Therefore, the

new system does not increase the probability

of occurrence of accidents previously

evaluated.

The proposed changes to the groundwater

sump system separate the sump from the RSS

pump cubicle. As such, the proposed

changes would preclude flooding of the RSS

cubicles and a potential malfunction of the

RSS pumps. The RSS pumps function to

provide containment and core cooling, as

early as 11 minutes and 30 minutes,

respectively, post LOCA. Operability of the

RSS pumps is required long term. Since the

changes do not affect the operation of the

RSS pumps, they will not increase the

consequences of a LOCA.

The new collection tank 3SRW–TK1 will

be installed in the location of the existing

abandoned in place Chemical Addition Tank

(CAT) 3QSS*TK2, by the Refueling Water

Storage Tank (RWST). The tank will be

seismically supported utilizing similar struts

and attachments to the RWST as the removed

CAT. A calculation has confirmed that there

is no impact on the seismic qualification of

the RWST as a result of the new tank. The

RWST provides water to the Emergency Core

Cooling System (ECCS) and Containment

Quench Spray (QSS) which are credited to

mitigate the consequences of a LOCA.

Therefore, the proposed changes do not

increase the consequences of a LOCA.

In the proposed design, the installation of

the new safety related collection sump and

casing pipe will result in a change in the

Supplemental Leak Collection and Release

System (SLCRS) boundary within the ESFB.

This modification will be performed to meet

the SLCRS design requirements. Testing will

be performed post modification and routinely

to satisfy SLCRS Technical Specification 3/

4.6.6 requirements. Per Technical

Specification 3/4.6.6 basis, the SLCRS is

credited post LOCA to limit the release of

fission products from the containment. Since

the proposed changes do not affect

operability of SLCRS, it does not increase the

consequences of a LOCA.

In the proposed changes, sumps

3DAS*SUMP7A/B inflow pathways will be

restored such that it may become potentially

contaminated. Emergency operating

procedures shall contain operator actions to

ensure that power to 3DAS*SUMP7A/B

sump pumps 3DAS–P8A/B is isolated post

LOCA. As such, the proposed changes will

continue to ensure that potentially

contaminated water is not discharged from

3DAS*SUMP7A/B. Therefore, the changes

will not increase the consequences of a

LOCA.

The design change, per NUREG–0737

Section II.B.2 requirement, has been

evaluated by a calculation to ensure that the

required operator actions post LOCA can be

performed within a 5 rem whole body dose

requirement, and has been found to be

acceptable.

Therefore, these changes will not

significantly increase the consequences of an

accident previously evaluated.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

This license amendment request is

associated with the installation of a new

sump pump system in the ESFB. The current

and new sump pump systems are not

accident initiators since neither system is

connected to, and both are fully independent

of any system that could cause an accident

to occur. The new system, which collects

groundwater from beneath the Containment

Structure and ESFB, is a support system for

the RSS and the containment protective

boundary, which are design basis accident

mitigation design features. Therefore, the

proposed changes will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Involve a significant reduction in a

margin of safety.

The Millstone Unit No. 3 FSAR changes

reflect the installation of a new sump pump

system in the ESFB. The proposed changes

do not affect operation of the RWST, ECCS,

QSS, RSS, SLCRS, Containment, EDG or any

Class 1E component required for safety. The

additional load on the Train A EDG and fuel

oil consumption are within the calculated

allowance. Therefore, these changes do not

significantly reduce the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

P.O. Box 270, Hartford, Connecticut.

NRC Section Chief: James W. Clifford.

Omaha Public Power District, Docket

No. 50–285, Fort Calhoun Station, Unit

No. 1, Washington County, Nebraska

Date of amendment request: July 28,

2000.

Description of amendment request:

The proposed amendment would revise

the Fort Calhoun Station Unit 1

Technical Specifications to allow

installation of tube sleeves as an

alternative to plugging to repair

defective steam generator tubes.

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

Federal Register / Vol. 65, No. 202 / Wednesday, October 18, 2000 / Notices 62389

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 CE Leak Tight Sleeves are designed

using the applicable American Society of

Mechanical Engineers (ASME) Boiler and

Pressure Vessel Code and, therefore, meet the

design objectives of the original steam

generator tubing. The applicable design

criteria for the sleeves conform to the stress

limits and margins of safety of Section III of

the ASME code. Mechanical testing has

shown that the structural strength of repair

sleeves under normal, upset, and faulted

conditions provides margin to the acceptance

limits.

These acceptance limits bound the most

limiting (three times normal operating

pressure differential) burst margin

recommended by Regulatory Guide 1.121.

Burst testing of sleeved tubes has

demonstrated that no unacceptable levels of

primary-to-secondary leakage are expected

during any plant condition.

Evaluation of the repaired steam generator

tubes indicates no detrimental effects on the

sleeve or sleeve-tube assembly from reactor

coolant system flow, primary or secondary

coolant chemistries, thermal conditions or

transients, or pressure conditions as may be

experienced at Fort Calhoun Station.

Corrosion testing of sleeve-tube assemblies

indicates no evidence of sleeve or tube

corrosion considered detrimental under

anticipated service conditions.

The installation of the proposed sleeves is

controlled via the sleeving vendor’s

proprietary processes and equipment. The CE

process has been in use since 1984 and has

been implemented more than 24 times for the

installation of over 4,200 sleeves. The FCS

steam generator design was reviewed and

found to be compatible with the installation

processes and equipment.

The implementation of the proposed

amendment has no significant effect on either

the configuration of the plant or the manner

in which it is operated. The consequences of

a hypothetical failure of the sleeved tube is

bounded by the current steam generator tube

rupture analysis described in Fort Calhoun

Station’s USAR, Section 14.14. Due to the

slight reduction in diameter caused by the

sleeve wall thickness, primary coolant

release rates would be slightly less than

assumed for the steam generator tube rupture

analysis, depending on the break location,

and therefore, would result in lower total

primary fluid mass release to the secondary

system. A main steam line break or feed line

break will not cause a SGTR since the sleeves

are analyzed for a maximum accident

differential pressure greater than that

predicted in the Fort Calhoun Station safety

analysis.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

As discussed above, the CE Leak Tight

Sleeves are designed using the applicable

ASME Code as guidance; therefore, they meet

the objectives of the original steam generator

tubing. As a result, the functions of the steam

generators will not be significantly affected

by the installation of the proposed sleeves.

The proposed repair sleeves do not interact

with any other plant systems. Any accident

as a result of potential tube or sleeve

degradation in the repaired portion of the

tube is bounded by the existing tube rupture

accident analysis. The continued integrity of

the installed sleeve is periodically verified by

the Technical Specification requirements.

The implementation of the proposed

amendment has no significant effect on either

the configuration of the plant or the manner

in which it is operated. Therefore, Omaha

Public Power District concludes that this

proposed change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The repair of degraded steam generator

tubes with CE Leak Tight Sleeves restores the

structural integrity of the degraded tube

under normal operating and postulated

accident conditions. The design safety factors

utilized for the repair sleeves are consistent

with the safety factors in the ASME Code

used in the original steam generator design.

The portions of the installed sleeve assembly

that represents the reactor coolant pressure

boundary can be monitored for the initiation

and progression of sleeve/tube wall

degradation. Use of the previously identified

design criteria and design verification testing

assures that the margin of safety is not

significantly different from the original steam

generator tubes. Therefore, OPPD concludes

that the proposed change does not involve a

significant reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Perry D.

Robinson, Winston & Strawn, 1400 L

Street, N.W., Washington, DC 20005–

3502.

NRC Section Chief: Stephen Dembek.

PECO Energy Company, Docket Nos.

50–352 and 50–353, Limerick

Generating Station (LGS), Units 1 and 2,

Montgomery County, Pennsylvania

Date of amendment request: July 31,

2000.

Description of amendment request:

The proposed changes will revise LGS

Technical Specifications (TSs) to

replace the existing Automatic

Depressurization System (ADS) TS

Surveillance Requirement (SR) 4.5.1.d.1,

a 31-day channel functional test of the

accumulator backup compressed gas

system low pressure alarm system, with

a 31-day verification of the ADS

accumulator gas supply header

pressure. The existing TS SR 4.5.1.d.1

and SR 4.5.1.d.2.c, a 24-month channel

calibration of the accumulator backup

compressed gas system low pressure

alarm system, will be relocated to the

Technical Requirements Manual (TRM).

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed TS changes do not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed TS changes have no physical

impact on plant equipment or the normal

operation of plant systems. The ADS and the

ADS accumulator backup compressed gas

system affected by the proposed testing

changes are normally in a standby mode and

there are no existing credible system failures

that are accident initiators. The ability of the

ADS to depressurize the vessel following a

small break Loss of Coolant Accident (LOCA)

so that flow from low pressure Emergency

Core Cooling Systems (ECCS) can enter the

core in time to limit fuel cladding

temperatures is maintained by the operability

of the ADS accumulators and their inlet

check valves. The ADS accumulator backup

compressed gas low pressure alarm system

has no impact on the ability of the ADS

accumulators and associated check valves to

maintain an adequate gas supply required to

mitigate an accident. Therefore, the removal

of the alarm system testing from the TS has

no impact on the ability of the ADS to cope

with the small break LOCA as previously

evaluated. The replacement of the monthly

alarm channel functional test with the

monthly verification of the ADS accumulator

gas supply header pressure will assure that

the ADS accumulators are pressurized as

required to support ADS operability and the

ability of ADS to mitigate the accident as

previously analyzed is maintained.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed TS changes do not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes have no physical

impact on plant equipment or the normal

operation of plant systems. The changes are

limited to changes in administrative testing

requirements for the existing ADS and ADS

accumulator backup compressed gas low

pressure alarm systems, and the long term

gas supply to the ADS valves. The changes

do not impact the methods of operation or

manipulation of these systems or

components. The impact of these changes has

been evaluated to assure that the changes are

in conformance with the required design and

licensing basis, and that system performance

is not degraded. The changes do not affect

the operation of the ADS or the ADS

accumulator backup gas system and do not

create any new system failure modes or

62390 Federal Register / Vol. 65, No. 202 / Wednesday, October 18, 2000 / Notices

accident initiators not previously considered.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed TS changes do not involve

a significant reduction in a margin of safety.

The proposed changes maintain the safety

design basis of the ADS and the ADS

accumulator backup gas systems. The ADS

accumulator backup compressed gas low

pressure alarm system does not support the

operability of the ADS accumulators which

are required to maintain an adequate gas

supply for ADS vessel depressurization.

Therefore, the Channel Functional Test and

Channel Calibration of backup gas system

alarms can be removed from the TS and have

no impact on the ability of the ADS to

depressurize the reactor and maintain current

safety margins defined in the design basis for

this TS. The availability of the ADS

accumulator backup gas system to perform its

long term cooling function after an accident

or other event is not addressed in any TS or

Bases. The proposed changes in testing also

do not impact any of the Inservice

Inspections or Tests currently performed on

the ADS or ADS accumulator backup gas

system components. Therefore, the proposed

changes do not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: 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.

PPL Susquehanna, LLC, Docket Nos. 50–

387 and 50–388, Susquehanna Steam

Electric Station, Units 1 and 2, Luzerne

County, Pennsylvania

Date of amendment request: February

29, 2000.

Description of amendment request:

The amendment would incorporate

Supplement 3 to PL–NF–90–001,

‘‘Application of Reactor Analysis

Methods for BWR Design and Analysis:

Application Enhancements,’’ into

Technical Specification Section 5.6.5,

Core Operating Limits Report. The

supplement describes alternative

methods for the analysis of the rotated

bundle event, the control rod

withdrawal error event, and the

recirculation flow controller event.

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 alternative analysis

methods do not involve an increase in the

probability or consequences of an accident

previously evaluated. The alternative

analysis methods affect the analysis methods

used to perform the Rotated Bundle Analysis,

the Rod Withdrawal Error Analysis, and the

Recirculation Flow Controller Failure

Analysis. These events are analyzed on a

cycle specific basis to ensure that the

operating limits contained in the COLR [Core

Operating Limits Report] will provide

acceptable consequences to the health and

safety of the public consistent with NRC

guidelines. No physical changes are being

made to plant systems, structures or

components. The alternative analysis

methods ensure that the [offsite] dose

consequences of the postulated events

remain within the NRC guidelines.

Based on the above, it is concluded that the

alternative analysis methods do not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The alternative analysis methods do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated. The proposed

alternative analysis methods affect the

analysis methods for the Rotated Bundle, Rod

Withdrawal Error and Recirculation Flow

Controller Failure Events. Since these

alternative analysis methods affect analytical

methods and do not affect any plant systems,

structures, or components, it is concluded

that the proposed alternative analysis

methods do not create the possibility for any

new or different kind of accident from any

accident previously evaluated.

3. The proposed change does not involve

a significant reduction in the margin of

safety.

The alternative analysis methods do not

involve a significant reduction in the margin

of safety.

The Rotated Bundle Methodology is

currently analyzed as a moderate frequency

event. The alternative methods will instead

analyze the Rotated Bundle Event as an

Infrequent Event. Analysis of this event as an

infrequent event is consistent with NRC

guidance (provided in the Standard Review

Plan) and the frequency classification of the

event as described in the SSES [Susquehanna

Steam Electric Station] FSAR [Final Safety

Analysis Report]. The proposed analysis

methodology limits the analytical [offsite]

dose to a small fraction of 10 CFR 100

guidelines consistent with the NRC

guidelines. Therefore, the proposed

alternative analysis methods do not represent

a significant reduction in the margin of

safety.

The Rod Withdrawal Error Analysis

currently does not credit the Rod Block

Monitor System to limit the extent of the

inadvertent rod withdrawal. The alternative

proposed methods will allow credit in the

analysis for the Rod Block Monitor to limit

the extent of the inadvertent control rod

withdrawal. Several plant and procedural

improvements have been implemented that

have improved the reliability of the Rod

Block Monitor System. The analytical

acceptance criteria for the event is not

affected. Therefore, the proposed alternative

analysis methods do not affect the margin of

safety.

The Recirculation Flow Controller Failure

analysis is currently analyzed using the

RETRAN code. The proposed alternative

analysis methods use [PPL Susquehanna,

LLC’s] approved steady state nodal

simulation methodology instead of the

RETRAN code. The [PPL Susquehanna, LLC,]

steady state nodal simulation methodology

produces final operating limits that are

consistent with the RETRAN methodology.

The analytical acceptance criteria is not

affected by the alternative analysis

methodology. Use of [PPL Susquehanna,

LLC’s] methodology does not affect the

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Bryan A. Snapp,

Esquire, Assoc. General Counsel, PPL

Services Corporation, 2 North Ninth St.,

GENTW3, Allentown, PA 18101–1179.

NRC Section Chief: Marsha

Gamberoni.

PPL Susquehanna, LLC, Docket Nos. 50–

387 and 50–388, Susquehanna Steam

Electric Station, Units 1 and 2, Luzerne

County, Pennsylvania

Date of amendment request: July 31,

2000.

Description of amendment request:

The amendment would remove the

phrase ‘‘maximum pathway’’ from

Surveillance Requirement 3.6.1.3.12 in

Technical Specification Section 3.6.1.3,

‘‘Primary Containment Isolation

Valves.’’

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 to eliminate the

words ‘‘maximum pathway’’ does not affect

any plant system or component. The change

does not impact operator performance or

procedures. The leak rate testing of the

MSIVs [main steam isolation valves] will

continue to be performed in accordance with

Federal Register / Vol. 65, No. 202 / Wednesday, October 18, 2000 / Notices 62391

10 CFR 50 Appendix J. The change does not

impact the design basis accident analyses

presented in the FSAR [Final Safety Analysis

Report]. The change only affects how the asfound leakage is used to evaluate operability

and reportability. This change is consistent

with the guidance on leak rate testing

presented in NEI 94–01 [Nuclear Energy

Institute Guideline for Implementing

Performance-Based Option of 10 CFR Part 50,

Appendix J] and the Standard Technical

Specifications. 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.

As discussed above, the proposed change

to the Technical Specifications does not

affect any plant system or component and

does not affect plant operation. The

consequences of accidents will remain

within the accident analysis described in the

FSAR. 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 the margin of

safety.

The proposed change does not affect any

plant system or component, and does not

have any impact on plant operation. The

proposed change does not involve a

significant reduction in the margin of safety

as currently defined in the bases of the

applicable Technical Specification section.

Therefore, the proposed change does not

involve a significant reduction in the margin

of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Bryan A. Snapp,

Esquire, Assoc. General Counsel, PPL

Services Corporation, 2 North Ninth St.,

GENTW3, Allentown, PA 18101–1179.

NRC Section Chief: Marsha

Gamberoni.

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:

September 6, 2000 (PCN–274,

Supplement 1). This application

supersedes the licensee’s application of

November 24, 1999.

Description of amendment requests:

The U.S. Nuclear Regulatory

Commission (the Commission) has

granted the request of Southern

California Edison Company to withdraw

its November 24, 1999, application for

proposed amendments. The

Commission had previously issued a

Notice of Consideration of Issuance of

Amendments published in the Federal

Register on December 29, 1999 (64 FR

73098). However, by letter dated

September 6, 2000, the licensee

withdrew the proposed change. TAC

Nos. MA7289 and MA7290 used for the

review of the November 24, 1999,

application have been closed.

As submitted by the licensee on

September 6, 2000, the proposed

amendments would modify the

Technical Specifications (TSs) for the

San Onofre Nuclear Generating Station,

Units 2 and 3, to revise TS 3.3.11, ‘‘Post

Accident Monitoring Instrumentation

(PAMI).’’ Specifically, the proposed

change would extend the PAMI channel

calibration surveillance frequency from

18 months to 24 months to

accommodate a 24-month fuel cycle for

all PAMI instruments with the

exception of the reactor coolant system

(RCS) temperature instrumentation.

Surveillance Requirement (SR) 3.3.11.4

relating to RCS temperature

instrumentation channel calibration

every 18 months will remain in place.

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:

Do the proposed amendments:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated?

Response: No

The proposed license amendment to

extend the calibration surveillance frequency

of Post Accident Monitoring Instrumentation

(PAMI) (excluding RCS temperature

instrumentation) is being made to support

plant operation with a 24-month fuel cycle.

Increasing the calibration intervals for PAMI

instrumentation to 30 months [24 months

plus the 25 percent surveillance interval

extension allowed by SR 3.0.2] (excluding

RCS temperature instrumentation) does not

affect the initiation or probability of any

previously analyzed accident. Increasing the

calibration interval will not affect the

integrity of any of the principal barriers

against radiation release (fuel cladding,

reactor vessel, and containment building).

The ability of the plant to mitigate the

consequences of any previously analyzed

accidents is not adversely affected.

PAMI instrumentation provides to the

operators both qualitative and quantitative

information used in accident mitigation and

for the safe shutdown of the plant.

Instrumentation which provides qualitative

information is unaffected by a change in

instrument accuracy induced by drift due to

the increased surveillance interval because

no explicit value is required by the

Emergency Operating Instructions (EOIs).

Instrumentation that provides quantitative

information (i.e., decision points) in the EOIs

have been evaluated. This evaluation

resulted in no changes to any operating

instructions. This evaluation of the proposed

change to the surveillance interval

demonstrates that licensing basis safety

analyses acceptance criteria and San Onofre

Nuclear Generating Station (SONGS) Units 2

and 3 EOI criteria will continue to be met.

The proposed new surveillance frequency

for these instrument channels was evaluated

using the guidance of Generic Letter 91–04

[‘‘Changes in Technical Specification

Surveillance Intervals To Accommodate a 24-

Month Fuel Cycle’’]. The basis for the change

includes a quantitative evaluation of

instrument drift for PAMI instrumentation

(excluding RCS temperature instrumentation)

providing quantitative information to the

EOIs. Also, loop accuracy/setpoint

calculations for these instruments were

updated to accommodate the extended

surveillance period. Analyses and

evaluations completed to assess the proposed

increase in the surveillance interval

demonstrate that the effectiveness of these

instruments in fulfilling their respective

functions is maintained. Technical

Specifications Channel Checks and Channel

Functional Checks for the subject channels,

will continue to be performed to provide

assurance of instrument channel

OPERABILITY.

Therefore, the proposed amendment does

not involve a significant increase in the

probability or consequences of any

previously analyzed accident.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated?

Response: No

The increased calibration surveillance

interval for PAMI instrumentation (excluding

RCS temperature instrumentation) is justified

based on evaluation of past equipment

performance and does not require any plant

hardware changes or changes in normal

system operation. Changing the calibration

interval for this instrumentation has no

means of creating the possibility of a new or

different kind of accident. There are no new

decision points or operator responses

required to support existing accident

mitigation strategies.

Therefore, there are no new failure modes

introduced as a result of extending these

surveillance intervals, and the proposed

amendment does not create the possibility of

a new or different kind of accident from any

accident previously evaluated.

3. Involve a significant reduction in a

margin of safety?

Response: No

The proposed change to the calibration

surveillance interval (excluding RCS

temperature instrumentation) was evaluated

using the criteria of 95% probability/95%

confidence level for process sensor drift.

PAMI instrumentation are used to provide

indication following certain hypothetical

accident conditions and are used in EOIs for

trending and to initiate operator action at

certain decision points. Instrument

uncertainty calculations have been updated

for PAMI instrumentation used for EOI

62392 Federal Register / Vol. 65, No. 202 / Wednesday, October 18, 2000 / Notices

decision points as appropriate. Updated

calculations show that the total loop

uncertainty for PAMI evaluated either

decreased or remained the same. These

updated calculations demonstrate that

applicable accuracy requirements for SONGS

2 and 3 are satisfied with the proposed new

surveillance intervals.

Changing the calibration interval for these

channels does not affect the margin of safety

for previously analyzed accidents. Therefore,

the proposed amendment does not involve a

significant reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment 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.

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of application for amendments:

August 31, 2000 (TS 99–17).

Brief description of amendments: The

proposed amendment would revise the

Sequoyah Nuclear Plant (SQN)

Technical Specifications (TSs). The

revision would revise TS Section 5.6,

‘‘Fuel Storage,’’ to allow credit for

soluble boron in the fuel storage pools.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a),

Tennessee Valley Authority (TVA), the

licensee, has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

A. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The presence of soluble boron in the spent

fuel pool (SFP) water for criticality control

does not increase the probability of a fuel

assembly misplacement accident in the SFP.

The handling of the fuel assemblies in the

SFP has always been performed in borated

water. The proposed change does allow

greater flexibility for fuel storage

configurations in the SFP. The increased

flexibility does not introduce any greater

complexity than the 3-zone configuration

now in use. Fuel assembly placement will

continue to be controlled pursuant to

approved fuel handling procedures and will

be in accordance with the TS limitations.

There is no increase in the probability of a

fuel placement accident.

The criticality analysis shows the

consequences of the most serious fuel

assembly misplacement accident in the SFP

are not affected when considering the

presence of soluble boron. Under normal

conditions, the rack keff [k effective] remains

subcritical as required by 10 CFR 50.68

[Section 50.68 of Title 10 of the Code of

Federal Regulations], and is less than 0.95

with only 300 ppm [parts per million]

soluble boron concentration. In the event of

a postulated fuel assembly misplacement, the

presence of sufficient soluble boron in the

SFP precludes criticality as a result of the

misplacement. The criticality analysis

demonstrates that the pool keff will remain

less than 0.95 following an accidental

misplacement due to 2000 parts per million

(ppm) boron concentration of the pool. In

fact, concentration of only 700 ppm soluble

boron is sufficient to maintain keff less than

0.95 with 95% probability at 95% confidence

level for the most serious fuel assembly

misplacement. The proposed TS will ensure

that an adequate SFP boron concentration is

maintained. There is no significant increase

in the consequences of the accidental

misplacement of spent fuel assemblies in the

SFP.

There is no increase in the probability of

the loss of normal cooling to the SFP water

when considering the presence of soluble

boron in the pool water for subcriticality

control since a high concentration of soluble

boron has always been maintained in the SFP

water.

Reactivity changes due to SFP temperature

changes have been evaluated. The base case

criticality analysis used a SFP temperature of

20°C. The SFP reactivity uncertainty due to

temperature changes was considered for SFP

temperatures ranging from 4°C to 120°C. The

reactivity increment between 4°C and 20°C is

taken into account as additional uncertainty

in the analysis. In all spent fuel temperature

cases, the temperature (and void) coefficients

of reactivity are negative. Therefore there is

no requirement for additional soluble boron

above the base case level. Because the

coefficients of reactivity are negative, the

consequences of the loss of normal cooling to

the SFP will not be increased.

Therefore, based on the conclusions of the

above analysis, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

B. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

Spent fuel handling accidents are not new

or different types of accidents and have been

evaluated in the criticality analysis,

Reference 1.

The boron concentration in the SFP water

is maintained at a minimum of 2000 ppm.

The proposed changes to the TS do not

change boron concentration requirements for

the SFP water. A dilution of the SFP soluble

boron has always been a possibility;

however, it was shown in the SFP dilution

evaluation (Reference 2) that there are no

credible dilution events for which the SFP

keff could reach criticality. Therefore, the

implementation of proposed changes to the

TS will not result in the of a new kind of

accident.

The proposed changes for re-rack storage

management continue to specify

requirements for the spent fuel rack

configurations. Since the proposed SFP

storage configuration limitations are

comparable to those used in the past, the new

limitations will not have any significant

effect on normal SFP operations and

maintenance and will not create any

possibility of a new or different kind of

accident. Verifications will continue to be

performed to ensure that the SFP loading

configuration meets specified requirements.

The misplacement of a fuel assembly in the

revised storage configurations has been

evaluated. In all cases, the rack keff remains

subcritical and less than 0.95 with 700 ppm

boron in the water.

As discussed above, the proposed changes

will not create the possibility of a new or

different kind of accident. There is no

significant change in plant configuration,

equipment design, or equipment.

Under the proposed amendment, no

changes are being made to the racks

themselves, any other systems, or to the

physical fuel handling structures in the

Auxiliary Building itself. Therefore, the

proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

C. The proposed change does not involve

a significant reduction in a margin of safety.

The TS changes proposed by this License

Amendment Request and the resulting spent

fuel storage configuration limitations will

provide adequate safety margin to ensure that

the storage fuel assembly array will always

remain subcritical. Those limits are based on

a plant specific criticality analysis (Reference

1) performed in accordance with accepted

spent fuel rack criticality analysis

methodology.

While the criticality analysis utilized

partial credit for soluble boron, storage

configurations have been defined to ensure

that the spent fuel rack keff will be less than

1.0 with no soluble boron. Soluble boron

credit is used to provide subcritical margin

such that the SFP keff is maintained less than

0.95 under all credible conditions.

The loss of substantial amounts of soluble

boron from the SFP, which could lead to keff

exceeding 0.95, has been evaluated

(Reference 2) and shown to be not credible.

This evaluation also shows that dilution of

the SFP boron concentration from 2000 ppm

to 800 ppm is not credible. Also, the spent

fuel storage pool keff remains less than 1.0 at

a 95/95 probability/confidence level with the

pool filled with unborated water. Therefore,

the proposed change does not involve a

significant reduction in the margin of safety.

Based on the above evaluation, TVA

concludes that the proposed changes to the

TSs does [sic] not result in a significant

reduction in a margin of safety.

The NRC has reviewed the licensee’s

analysis and, based on this review, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

Federal Register / Vol. 65, No. 202 / Wednesday, October 18, 2000 / Notices 62393

400 West Summit Hill Drive, ET 10H,

Knoxville, Tennessee 37902.

NRC Section Chief: Richard P.

Correia.

Vermont Yankee Nuclear Power

Corporation, Docket No. 50–271,

Vermont Yankee Nuclear Power Station,

Vernon, Vermont

Date of amendment request:

September 14, 2000, as supplemented

on September 22, 2000.

Description of amendment request:

This proposed change revises the

Technical Specification to clarify the

valve isolation signal information in

Table 4.7.2 and makes an administrative

change to the table main steam isolation

valves component identification to

include all the valves.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration which is presented below:

1. The operation of the Vermont Yankee

Nuclear Power Station in accordance with

the proposed amendment will not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

No changes are being made to plant design,

method of operation or method of testing.

This change will not alter the basic operation

of process variables, systems, or components

as described in the safety analysis. No new

equipment is introduced.

The proposed change does not affect the

ability of the primary containment isolation

system or ECCS [emergency core cooling

system] systems to perform their required

safety functions. The essential safety

functions of providing primary containment

integrity and providing water to cool the core

in the event of an accident are maintained.

There is no physical or operational change

being made which would alter the sequence

of events, plant response, or conclusions of

existing safety analyses. This proposed

change results in no impact on analyzed

accident event precursors or effects.

Therefore, the proposed change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The operation of Vermont Yankee

Nuclear Power Station in accordance with

the proposed amendment will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed change does not involve any

physical alteration of plant equipment and

does not change the method by which any

safety-related system performs its function.

As such, no new or different types of

equipment will be installed, and the basic

operation of installed equipment is

unchanged. There is no change in plant

operation that involves failure modes other

than those previously evaluated. The

methods governing plant operation and

testing remain consistent with current safety

analysis assumptions.

Therefore, the proposed change will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The operation of Vermont Yankee

Nuclear Power Station in accordance with

the proposed amendment will not involve a

significant reduction in a margin of safety.

No changes are being made to plant design,

method of operation or method of testing.

This change will not alter the basic operation

of process variables, systems, or components

as described in the safety analysis. No new

equipment is introduced.

The proposed change does not affect the

ability of the primary containment isolation

system or ECCS systems to perform their

required safety functions. The essential safety

functions of providing primary containment

integrity and providing water to cool the core

in the event of an accident are maintained.

There is no physical or operational change

being made which would alter the sequence

of events, plant response, or conclusions of

existing safety analyses. This proposed

change results in no impact on analyzed

accident event precursors or effects.

This proposed change does not alter the

physical design of the plant, methods or

modes of operation, testing or analyses,

thereby resulting in no impact on safety

functions. Since the proposed change does

not alter the means by which primary

containment isolation is maintained and

containment cooling valves are isolated in

support of RHR [residual heat removal] LPCI

[low pressure coolant injection] actuation,

there is no significant reduction in the

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Mr. David R.

Lewis, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037–1128.

NRC Section Chief: James W. Clifford.

Vermont Yankee Nuclear Power

Corporation, Docket No. 50–271,

Vermont Yankee Nuclear Power Station,

Vernon, Vermont

Date of amendment request:

September 19, 2000.

Description of amendment request:

This proposed change revises Technical

Specification (TSs) 3.5.H.3 and 3.5.H.4

related to low pressure Emergency Core

Cooling System (ECCS) injection/spray

subsystem operability during cold

shutdown and refueling conditions.

Two circumstances are considered: (1)

when no operations with the potential

for draining the reactor vessel (OPDRV)

are in progress (addressed in TS

3.5.H.3), and (2) when OPDRVs are in

progress (addressed in TS 3.5.H.4). The

proposed change provides completeness

in the TS for the defined conditions and

also provides for the operation of an

alternative combination of low pressure

ECCS injection/spray subsystems to

ensure adequate coolant inventory and

sufficient heat removal capability for the

irradiated fuel during cold shutdown

and refueling conditions when OPDRVs

are in progress.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration which is presented below:

1. The operation of Vermont Yankee

Nuclear Power Station in accordance with

the proposed amendment will not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

No changes are being made to plant design

or method of operation. This change only

affects the plant in a cold shutdown or

refueling condition and will not alter the

basic operation of process variables,

structures, systems, or components as

described in the safety analyses. No new

equipment is introduced.

The proposed change does not affect the

ability of low pressure ECCS injection/spray

systems to perform their required safety

functions. The essential safety function of

providing water to reflood the reactor vessel

following an inadvertent vessel draindown is

maintained. There is no physical or

operational change being made which would

alter the sequence of events, plant response,

or conclusions of existing safety analyses.

Therefore, the proposed change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The operation of Vermont Yankee

Nuclear Power Station in accordance with

the proposed amendment will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed change does not involve any

physical alteration of plant equipment and

does not change the method by which any

safety-related system performs its intended

safety function. As such, no new or different

types of equipment will be installed, and the

basic operation of installed equipment is

unchanged. There is no change in plant

operation that involves failure modes other

than those previously evaluated. The

methods governing plant operation and

testing remain consistent with current safety

analysis assumptions. Therefore, the

proposed change will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The operation of Vermont Yankee

Nuclear Power Station in accordance with

the proposed amendment will not involve a

significant reduction in a margin of safety.

62394 Federal Register / Vol. 65, No. 202 / Wednesday, October 18, 2000 / Notices

During refueling and cold shutdown

conditions with operations having the

potential for draining the reactor vessel

(OPDRV) in progress, any one ECCS

injection/spray subsystems is adequate to

reflood the reactor vessel in the event of an

inadvertent draindown. Since the proposed

change provides an equivalent means for

achieving this safety function, there is no

reduction in reflood capability. The

additional flexibility, to maintain a

combination of one core spray subsystem and

one LPCI [low pressure coolant injection]

subsystem (provided by this change), is

equivalent to the safety margin provided by

the existing TS since a single active failure

affecting one subsystem results in the same

remaining capability of one ECCS subsystem.

Since the changed TS provides equivalent

low pressure ECCS injection/spray capability

and protection from loss of coolant

inventory, the risk of an inadvertent

draindown event is unchanged, thus

preserving previously existing margins of

safety.

For circumstances involving no OPDRVs

during refueling and cold shutdown

conditions, no ECCS or containment cooling

equipment is required to meet safety

objectives. Thus, the margins of safety for

such situations are maintained.

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 request involves no

significant hazards consideration.

Attorney for licensee: Mr. David R.

Lewis, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037–1128.

NRC Section Chief: James W. Clifford.

Vermont Yankee Nuclear Power

Corporation, Docket No. 50–271,

Vermont Yankee Nuclear Power Station,

Vernon, Vermont

Date of amendment request:

September 26, 2000.

Description of amendment request:

This proposed change revises Technical

Specification (TS) requirements

regarding secondary containment

systems, including the Standby Gas

Treatment System (SBGTS). The

affected TS sections are 1.0, Definitions;

3/4.7.B, Standby Gas Treatment System;

and 3/4.7.C, Secondary Containment

System. In addition, a new TS section,

3/4.7.E, Reactor Building Automatic

Ventilation System Isolation Valves

(RBAVSIVs), is proposed. Some of the

proposed changes are administrative in

nature and do not affect the technical

aspects of the requirements. Associated

changes to the TS Bases are also being

made to conform to the changed TS. The

proposed changes provide certain

additional flexibility in operations when

equipment is made or found to be

inoperable, while also ensuring

appropriate actions are taken to place

the plant in a safe condition under such

conditions.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration which is presented below:

1. The operation of Vermont Yankee

Nuclear Power Station in accordance with

the proposed amendment will not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

No changes are being made to the plant

design, physical system configuration, or

basic method of operation as a result of the

proposed amendment. The Standby Gas

Treatment System (SBGTS) and secondary

containment are not assumed to be initiators

of any analyzed event. The circumstances for

which operability of SBGTS and secondary

containment are required are unchanged, and

would not occur at any greater frequency as

a result of this change. Therefore, the

probability of a design basis loss-of-coolant

accident or fuel handling accident (the

applicable accidents) previously evaluated is

not increased.

The proposed change does not increase the

consequences of an accident because system

operability requirements are being

maintained. In lieu of suspending refueling

activities when one train of SBGTS is

inoperable beyond seven days, placing the

operable train of SBGTS in operation ensures

that no failures that could prevent automatic

actuation have occurred and that any other

failure would be readily detected. Operation

of one train of the SBGTS is sufficient to

mitigate the consequences of any analyzed

event. The secondary containment systems

assumed to operate following a design basis

accident continue to function as assumed in

accident analyses to mitigate the

consequences of postulated accidents.

Therefore, the proposed change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The operation of Vermont Yankee

Nuclear Power Station in accordance with

the proposed amendment will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed change does not involve any

physical alteration to the plant structures,

systems, or components (SSCs), or the basic

manner in which these SSCs are operated or

maintained. The methods by which these

systems perform their safety function are

unchanged and remain consistent with

current safety analysis assumptions. There is

no change in plant operation that involves

failure modes other than those previously

evaluated. Therefore, the proposed change

will not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

3. The operation of Vermont Yankee

Nuclear Power Station in accordance with

the proposed amendment will not involve a

significant reduction in a margin of safety.

The proposed change does not result in a

significant reduction in a margin of safety

because restrictions placed on operations

which have the potential for releasing

radioactive material to the secondary

containment continue to be in accordance

with the assumptions and conditions of

existing safety analyses. Operations with

inoperable equipment have the proper

restrictions to maintain existing margins or to

place the plant in a safe condition such that

inoperable equipment is not required to meet

safety analysis assumptions. Ensuring

operability of one train of SBGTS together

with required secondary containment

integrity is sufficient to mitigate the

consequences of any analyzed event. Since

current analyses are unaffected in this regard,

margins of safety are maintained.

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 request involves no

significant hazards consideration.

Attorney for licensee: Mr. David R.

Lewis, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037–1128.

NRC Section Chief: James W. Clifford.

Previously Published Notices of

Consideration of Issuance of

Amendments to Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing

The following notices were previously

published as separate individual

notices. The notice content was the

same as above. They were published as

individual notices either because time

did not allow the Commission to wait

for this biweekly notice or because the

action involved exigent circumstances.

They are repeated here because the

biweekly notice lists all amendments

issued or proposed to be issued

involving no significant hazards

consideration.

For details, see the individual notice

in the Federal Register on the day and

page cited. This notice does not extend

the notice period of the original notice.

FirstEnergy Nuclear Operating

Company, et al., Docket No. 50–412,

Beaver Valley Power Station, Unit 2,

Shippingport, Pennsylvania

Date of amendment request:

September 1, 2000.

Federal Register / Vol. 65, No. 202 / Wednesday, October 18, 2000 / Notices 62395

Description of amendment request:

The proposed amendment would revise

certain 18-month surveillance

requirements in the technical

specifications by eliminating the

condition that testing be conducted

during shutdown, or during cold

shutdown or refueling mode. The

systems that would be affected are the

emergency core cooling system,

containment depressurization and

cooling system, chemical addition

system, and containment isolation valve

system.

Date of publication of individual

notice in Federal Register: September

12, 2000 (65 FR 55056).

Expiration date of individual notice:

October 12, 2000.

Power Authority of The State of New

York, Docket No. 50–286, Indian Point

Nuclear Generating Unit No. 3,

Westchester County, New York

Date of application for amendment:

April 27, 2000.

Brief description of amendment

request: The amendment seeks to extend

the applicability of the current pressuretemperature and overpressure

protection system limit curves from 13.3

effective full-power years (EFPY) to 16.2

EFPYS.

Date of publication of individual

notice in Federal Register: August 29,

2000 (65 FR 52451).

Expiration date of individual notice:

September 28, 2000.

Power Authority of the State of New

York, Docket No. 50–333, James A.

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of amendment request: August

29, 2000, as supplemented by letter

dated September 8, 2000.

Description of amendment request:

The amendment proposes to change

Technical Specifications 3.0.D and 4.0.D

to be equivalent to the Boiling-Water

Reactor NUREG–1433 guidance for the

Improved Technical Specifications

limiting condition for operation 3.0.4,

which is currently under review.

Date of publication of individual

notice in Federal Register: September

14, 2000 (65 FR 55650).

Expiration date of individual notice:

October 16, 2000.

Vermont Yankee Nuclear Power

Corporation, Docket No. 50–271,

Vermont Yankee Nuclear Power Station,

Vernon, Vermont

Date of amendment request:

September 14, 2000.

Brief description of amendment

request: The amendment would clarify

the valve isolation signal information in

the Technical Specification Table 4.7.2

and make an administrative change to

the table main steam isolation valves

component identification.

Date of publication of individual

notice in Federal Register: September 27,

2000 (65 FR 68111).

Expiration date of individual notice:

October 27, 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, located at One White Flint North,

11555 Rockville Pike (first floor),

Rockville, Maryland 20852. Publicly

available records will be accessible

electronically from the ADAMS Public

Library component on the NRC Web

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

Reading Room).

AmerGen Energy Company, LLC, Docket

No. 50–461, Clinton Power Station, Unit

1, DeWitt County, Illinois

Date of application for amendment:

June 19, 2000, as supplemented August

8, 2000.

Brief description of amendment: The

amendment allows some emergency

diesel generator Technical Specification

surveillance requirements to be

performed during plant operation

instead of during plant shutdown.

Date of issuance: October 2, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days.

Amendment No.: 132.

Facility Operating License No. NPF–

62: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: July 26, 2000 (65 FR 46006).

The supplemental information did not

change the application or affect the

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated October 2, 2000.

No significant hazards consideration

comments received: No.

AmerGen Energy Company, LLC, Docket

No. 50–289, Three Mile Island Nuclear

Station, Unit 1, Dauphin County,

Pennsylvania

Date of application for amendment:

April 1, 1999, as supplemented June 14,

and July 27, 2000.

Brief description of amendment: The

amendment revised the TMI–1

Technical Specifications (TSs) 1.4.2,

1.4.3, 1.4.4, 3.3.1.2.b, 3.3.1.3.b, and c,

3.3.2.1, Table 4.1–1 (Items 14, 25, 31,

and 32), Table 4.1–3 (Items 4 and 6),

Table 4.1–5, and TSs 4.1.5, 4.5.2.1.a and

b, 4.5.2.3.a, and 4.5.3.1.b.1 and 2, to:

add limiting condition for operation

(LCO) action statements and make LCOs

and surveillance requirements more

consistent with the revised ‘‘Standard

Technical Specifications for Babcock &

Wilcox Plants,’’ (NUREG–1430,

Revision 1); correct conflicts or

inconsistencies; and revise spent fuel

pool sampling frequency from monthly

and after adding chemicals, to weekly.

TS 3.3.1.2.d is deleted as a result of the

LCO additions described above. Also, a

Bases statement for surveillance testing

was added to Section 4.1 of the TSs and

a revised Bases to Section 4.4.4 is

included as well.

Date of issuance: September 25, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 60 days.

Amendment No.: 225.

62396 Federal Register / Vol. 65, No. 202 / Wednesday, October 18, 2000 / Notices

Facility Operating License No. DPR–50.

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 28, 1999 (64 FR 40906) and

August 23, 2000 (65 FR 51349).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated September 25,

2000.

No significant hazards consideration

comments received: No.

Arizona Public Service Company, et al.,

Docket Nos. STN 50–528, STN 50–529,

and STN 50–530, Palo Verde Nuclear

Generating Station, Units Nos. 1, 2, and

3, Maricopa County, Arizona

Date of application for amendments:

December 1, 1999 (102–04378).

Brief description of amendments: The

amendments to the operating licenses

delete or update outdated

administrative information and delete

license conditions that are no longer

applicable or have been completed.

Date of issuance: September 29, 2000.

Effective date: September 29, 2000, to

be implemented within 30 days of the

date of issuance.

Amendment Nos.: Unit 1—128, Unit

2—128, Unit 3—128.

Facility Operating License Nos. NPF–

41, NPF–51, and NPF–74: The

amendments revised the Operating

Licenses.

Date of initial notice in Federal

Register: March 8, 2000 (65 FR 12288).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated September 29,

2000.

No significant hazards consideration

comments received: No.

Arizona Public Service Company, et al.,

Docket Nos. STN 50–528, STN 50–529,

and STN 50–530, Palo Verde Nuclear

Generating Station, Units Nos. 1, 2, and

3, Maricopa County, Arizona

Date of application for amendments:

June 6, 2000, as supplemented June 29

and July 3, 2000.

Brief description of amendments: The

amendments restrict the emergency

diesel generator (DG) acceptance criteria

for steady-state voltage and frequency in

several surveillance requirements (SRs)

involving DG starts in Technical

Specification (TS) 3.8.1, ‘‘AC Sources—

Operating,’’ of the TSs for the three

units. The amendments also add a note

to each SR that states ‘‘The steady state

voltage and frequency limits are

analyzed values and have not been

adjusted for instrument error.’’ The

restricted acceptance criterion is to

ensure proper DG operation.

Date of issuance: October 4, 2000.

Effective date: October 4, 2000, to be

implemented within 45 days of the date

of issuance. For surveillance

requirements associated with the

revised steady-state voltage and

frequency limits in Technical

Specifications 3.8.1 and 3.8.2, the first

performance is due at the end of the first

surveillance interval that began on the

date the surveillance was last performed

prior to the date of implementation of

the amendments.

Amendment Nos.: Unit 1–129, Unit

2–129, Unit 3—129.

Facility Operating License Nos. NPF–

41, NPF–51, and NPF–74: The

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 12, 2000 (65 FR 43043).

The June 29 and July 3, 2000,

supplements provided clarifying

information that was within the scope of

the application and the Federal Register

notice, and did not change the staff’s

initial proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated October 4, 2000.

No significant hazards consideration

comments received: No.

Commonwealth Edison Company,

Docket Nos. 50–237 and 50–249,

Dresden Nuclear Power Station, Units 2

and 3, Grundy County, Illinois

Date of application for amendments:

February 18, 2000.

Brief description of amendments: The

amendments remove the anticipatory

reactor scram signal for turbine electrohydraulic control (EHC) low oil pressure

trip from the reactor protection system

(RPS) trip function.

Date of issuance: September 27, 2000.

Effective date: Immediately, to be

implemented within 90 days.

Amendment Nos.: 181 and 176.

Facility Operating License Nos. DPR–

19 and DPR–25: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: April 5, 2000 (65 FR 17910).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated September 27,

2000.

No significant hazards consideration

comments received: No.

Commonwealth Edison Company,

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

County Station, Units 1 and 2, LaSalle

County, Illinois

Date of application for amendments:

April 25, 2000.

Brief description of amendments: The

amendments revised Technical

Specification 3/4.9.5,

‘‘Communications’’ to allow the

movement of a control rod in a fueled

core cell in Operational Condition 5 to

be exempt from the requirement that

direct communication be maintained

between the control room and the

refueling platform personnel when the

rod is moved with its normal drive

system.

Date of issuance: October 5, 2000.

Effective date: Immediately, to be

implemented within 30 days.

Amendment Nos.: 141 and 127.

Facility Operating License Nos. NPF–

11 and NPF–18: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: June 14, 2000 (65 FR 37422).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated October 5, 2000.

No significant hazards consideration

comments received: No.

Consumers Energy Company, Docket

No. 50–255, Palisades Plant, Van Buren

County, Michigan

Date of application for amendment:

June 27, 2000, as supplemented August

18 and 30, 2000.

Brief description of amendment: The

amendment changes Improved

Technical Specification Sections 3.5.1,

‘‘Safety Injection Tanks (SITs),’’ and

3.5.2, ‘‘ECCS [Emergency Core Cooling

System]—Operating,’’ regarding

completion times for restoring an

inoperable SIT and for restoring a lowpressure safety injection train.

Date of issuance: October 2, 2000.

Effective date: As of the date of

issuance and shall be implemented on

or before December 31, 2000.

Amendment No.: 191.

Facility Operating License No. DPR–

20: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 26, 2000 (65 FR 46007)

(two notices).

The August 18 and 30, 2000,

supplemental letters provided clarifying

information that was within the scope of

the original application 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 October 2, 2000.

No significant hazards consideration

comments received: No.

Duke Energy Corporation, et al., Docket

Nos. 50–413 and 50–414, Catawba

Nuclear Station, Units 1 and 2, York

County, South Carolina

Date of application for amendments:

June 29, 2000, as supplemented by

Federal Register / Vol. 65, No. 202 / Wednesday, October 18, 2000 / Notices 62397

letters dated July 27, and August 10,

2000.

Brief description of amendments: The

amendments revised the Technical

Specifications (TS) to reference the

Westinghouse Best Estimate Large Break

Loss-of-Coolant Accident analysis

methodology described in WCAP–

12945–P–A, March 1998. These

amendments also address corresponding

TS Bases changes.

Date of issuance: October 2, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment Nos.: 188 and 181.

Facility Operating License Nos. NPF–

35 and NPF–52: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

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

The letter dated August 10, 2000,

provided additional information that

did not change the scope of the

application and the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated October 2, 2000.

No significant hazards consideration

comments received: No.

Duke Energy Corporation, et al., Docket

Nos. 50–413 and 50–414, Catawba

Nuclear Station, Units 1 and 2, York

County, South Carolina

Date of application for amendments:

May 25, 2000, as supplemented by

letters dated July 31, August 8, and

August 17, 2000.

Brief description of amendments: The

amendments temporarily revise TS

3.5.2, ‘‘Emergency Core Cooling

System;’’ TS 3.6.6, ‘‘Containment Spray

System;’’ TS 3.6.17, ‘‘Containment

Valve Injection Water System;’’ TS

3.7.5, ‘‘Auxiliary Feedwater System;’’

TS 3.7.7, ‘‘Component Cooling Water

System;’’ TS 3.7.8, ‘‘Nuclear Service

Water System;’’ TS 3.7.10, ‘‘Control

Room Area Ventilation System;’’ TS

3.7.12, ‘‘Auxiliary Building Filtered

Ventilation Exhaust System;’’ and TS

3.8.1, ‘‘AC Sources’’.

Date of issuance: October 4, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment Nos.: 189 and 182.

Facility Operating License Nos. NPF–

35 and NPF–52: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: August 25, 2000 (65 FR 51860).

The supplements dated July 31,

August 8, and August 17, 2000,

provided clarifying information that did

not change the scope of the May 25,

2000, application and the initial

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated October 4, 2000.

No significant hazards consideration

comments received: No.

Duke Energy Corporation, Docket Nos.

50–269, 50–270, and 50–287, Oconee

Nuclear Station, Units 1, 2, and 3,

Oconee County, South Carolina

Date of application of amendments:

September 7, 2000.

Brief description of amendments: The

amendments revise Surveillance

Requirement 3.8.1.9.a by adding a note

stating that the upper limits on

frequency and voltage are not required

to be met for the annual test of the

Keowee Hydro Units until the NRC

issues an amendment that removes the

note in response to an amendment

request to be submitted no later than

April 5, 2001.

Date of Issuance: October 4, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment Nos.: 316, 316, & 316.

Facility Operating License Nos. DPR–

38, DPR–47, and DPR–55: Amendments

revised the Technical Specifications.

Public comments requested as to

proposed no significant hazards

consideration: Yes (65 FR 56600 dated

September 19, 2000). That notice

provided an opportunity to submit

comments on the Commission’s

proposed no significant hazards

consideration determination. No

comments have been received. The

notice also provided for an opportunity

to request a hearing by October 19,

2000, but indicated that if the

Commission makes a final no significant

hazards consideration.

The Commission’s related evaluation

of the amendment, finding of exigent

circumstances, and a final no significant

hazards consideration determination are

contained in a Safety Evaluation dated

October 4, 2000.

Entergy Gulf States, Inc., and Entergy

Operations, Inc., Docket No. 50–458,

River Bend Station, Unit 1, West

Feliciana Parish, Louisiana

Date of amendment request: May 8,

2000, as supplemented by letter dated

August 30, 2000.

Brief description of amendment: The

amendment revises Technical

Specifications to remove the fuel

building (FB) and the FB ventilation

system from the requirements associated

with secondary containment during

power operation (except during

movement of recently irradiated fuel

assemblies in the FB).

Date of issuance: September 22, 2000.

Effective date: As of the date of

issuance and shall be implemented 30

days from the date of issuance.

Amendment No.: 113.

Facility Operating License No. NPF–

47: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: June 14, 2000 (65 FR 37424).

The August 30, 2000, supplemental

letter provided additional information

to support staff review of the original

application, and did not affect the initial

finding of no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated September 22,

2000.

No significant hazards consideration

comments received: No.

Entergy Operations, Inc., Docket No. 50–

368, Arkansas Nuclear One, Unit Nos. 1

and 2 (ANO–1 and ANO–2), Pope

County, Arkansas

Date of application for amendments:

September 17, 1999, as supplemented

by letters dated June 29, August 3, and

September 15, 2000.

Brief description of amendment: The

amendments change heavy load

handling requirements and

transportation provisions that would

permit the movement of the original and

replacement steam generators (SGs)

through the ANO–2 containment

construction opening during the SG

replacement outage.

Date of issuance: September 25, 2000.

Effective date: As of the date of

issuance to be implemented within 30

days from the date of issuance.

Amendment Nos.: 209 & 221.

Facility Operating License Nos. DRP–

51 and NPF–6: The amendments revise

the licenses.

Date of initial notice in Federal

Register: February 23, 2000 (65 FR 9004).

The additional information provided

in the June 29 and August 3, 2000,

supplemental letters was noticed in the

Federal Register on August 23, 2000 (65

FR 51352). The September 15, 2000,

supplement provided clarifying

information that was within the scope of

the Federal Register notice published

August 23, 2000, and did not change the

staff’s initial no significant hazards

consideration determination.

62398 Federal Register / Vol. 65, No. 202 / Wednesday, October 18, 2000 / Notices

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated September 25,

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:

November 29, 1999, as supplemented by

letters dated January 26, May 17 (2

letters), May 31, and August 4, 2000.

Brief description of amendment: The

amendment revised the License and

Technical Specifications (TSs), and

corresponding Bases have been changed

to maintain consistency with the

transient and accident analyses which

evaluated the impact of the replacement

steam generators (SGs) that are being

used for Cycle 15 operation. The

License was revised to incorporate a

new methodology employed in

calculating radiological doses for some

non-loss-of-coolant accident events. TS

changes were made to the Reactor

Protection System (RPS) and Engineered

Safety Features Actuation System

(ESFAS) low pressurizer pressure

setpoints, the RPS and ESFAS low SG

pressure setpoints, the RPS and ESFAS

low SG level setpoints, the reactor

coolant flow rate limit, and the high

linear power trip setpoints with

inoperable main steam safety valves

(MSSVs). The amendment also made

changes to the TSs and corresponding

Bases have been changed that are not

directly related to the replacement SGs.

These changes revised the allowed

outage time of the MSSVs in Modes 1

and 2 to allow up to 12 hours to reduce

the high linear power level-high trip

setpoint when one or more MSSVs are

inoperable, and revised the action

statement in Mode 3 to maintain at least

two MSSVs operable on each SG.

Date of issuance: September 29, 2000.

Effective date: As of the date of

issuance to be implemented prior to

startup from the 2R14 refueling outage.

Amendment No.: 222.

Facility Operating License No. NPF–6:

Amendment revised the License and

TSs.

Date of initial notice in Federal

Register: February 9, 2000 (65 FR 6405).

The January 26, May 17 (2 letters),

May 31, and August 4, 2000,

supplemental 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 September 29,

2000.

No significant hazards consideration

comments received: No.

Entergy Operations, Inc., Docket No. 50–

382, Waterford Steam Electric Station,

Unit 3, St. Charles Parish, Louisiana

Date of amendment request: January

12, 2000, as supplemented by letters

dated June 15, 2000, and September 7,

2000.

Brief description of amendment: The

proposed changes modify Technical

Specification (TS) 3.9.4, ‘‘Containment

Building Penetrations,’’ to allow the

containment equipment door, airlocks,

and other penetrations to remain open,

but capable of being closed, during core

alterations or movement of irradiated

fuel in containment. Additionally, a

note, Bases changes, and Surveillance

Requirements changes provide further

enhancements to clarify equipment

door, airlock, and penetration closure

capability.

Date of issuance: October 2, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 60 days from the date of

issuance.

Amendment No.: 169.

Facility Operating License No. NPF–

38: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: February 23, 2000 (65 FR 9008).

The June 15, 2000, and September 7,

2000, supplemental letters provided

clarifying information that did not

expand the scope of the original Federal

Register notice, or change the scope of

the initial proposed no significant

hazards consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated October 2, 2000.

No significant hazards consideration

comments received: No.

FirstEnergy Nuclear Operating

Company, et al., Docket Nos. 50–334

and 50–412, Beaver Valley Power

Station, Unit Nos. 1 and 2,

Shippingport, Pennsylvania

Date of application for amendments:

September 20, 1999, as supplemented

May 12, 2000.

Brief description of amendments: The

amendments revised the standard to

which the control room ventilation

charcoal and Supplementary Leak

Collection and Release System (SLCRS)

charcoal must be laboratory tested as

specified in: BVPS–1 Technical

Specification (TS) 4.7.7.1.1.c.2 for the

Control Room Emergency Habitability

Systems; BVPS–1 TS 4.7.8.1.b.3 for the

SLCRS; BVPS–2 TS 4.7.7.1.d for the

Control Room Emergency Air Cleanup

and Pressurization System; and BVPS–

2 TS 4.7.8.1.b.3 for the SLCRS. Nuclear

Regulatory Commission Generic Letter

99–02, ‘‘Laboratory Testing of NuclearGrade Activated Charcoal,’’ dated June

3, 1999, requested licensees to revise

their TS criteria associated with

laboratory testing of ventilation charcoal

to a valid test protocol, which included

American Society for Testing and

Materials (ASTM) D3803–1989. These

license amendments revised the

charcoal laboratory standard to follow

ASTM D3803–1989 for each BVPS Unit.

These license amendments also: (1)

Revised the minimum amount of output

in kilowatts needed for the control room

emergency ventilation system heaters at

each BVPS unit; (2) revised BVPS–1

SLCRS surveillance testing criteria to be

consistent with American Nuclear

Standards Institute/American Society of

Mechanical Engineers N510–1980, the

BVPS–1 control room ventilation

testing, and BVPS–2 SLCRS/control

room ventilation testing; and (3) made

minor typographical corrections.

Date of issuance: September 29, 2000.

Effective date: As of date of issuance

and shall be implemented within 60

days.

Amendment Nos.: 234 and 117.

Facility Operating License Nos. DPR–

66 and NPF–73: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: August 29, 2000 (65 FR 52449).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated September 29,

2000.

No significant hazards consideration

comments received: No.

FirstEnergy Nuclear Operating

Company, et al., Docket No. 50–412,

Beaver Valley Power Station (BVPS–2),

Unit 2, Shippingport, Pennsylvania

Date of application for amendment:

May 1, 2000, as supplemented July 21,

2000.

Brief description of amendment: The

amendment: (1) Revised Technical

Specification (TS) requirements

regarding the minimum number of

radiation monitoring instrumentation

channels required to be operable during

movement of fuel within the

containment; (2) revised the Modes in

which the surveillance specified by

Table 4.3–3, ‘‘Radiation Monitoring

Instrumentation Surveillance

Requirements,’’ Item 2.c.ii is required;

(3) revised TS 3.9.4, ‘‘Containment

Building Penetrations,’’ to allow both

personnel air lock (PAL) doors and

Federal Register / Vol. 65, No. 202 / Wednesday, October 18, 2000 / Notices 62399

other containment penetrations to be

open during movement of fuel

assemblies within containment,

provided certain conditions are met; (4)

revised applicability and action

statement requirements of TS 3.9.4. to

be for only during movement of fuel

assemblies within containment; (5)

revised periodicity and applicability of

Surveillance Requirement (SR) 4.9.4.1;

(6) revised SR 4.9.4.2 to verify flow rate

of air to the supplemental leak

collection and release system (SLCRS)

rather than verifying the flow rate

through the system; (7) added two new

SRs, 4.9.4.3 and 4.9.4.4, for verification

and demonstration of SLCRS

operability; (8) modified TS 3/4.9.9 for

the containment purge exhaust and

isolation system to be applicable only

during movement of fuel assemblies

within containment; (9) revised

associated TS Bases and made editorial

and format changes; and, (10) revised

the BVPS–2 Updated Final Safety

Analysis Report (UFSAR) description of

a fuel-handling accident (FHA) and its

radiological consequences. The changes

to the BVPS–2 UFSAR reflect a revised

FHA analysis that the licensee

performed to evaluate the potential

consequences of having containment

penetrations and/or the PAL open

during movement of fuel assemblies

within containment. These UFSAR

revisions include potential exclusion

area boundary, low population zone,

and control room operator doses as a

result of an FHA.

Date of issuance: September 28, 2000.

Effective date: As of date of issuance.

Technical Specification changes shall be

implemented within 60 days.

Amendment No: 116.

≤Facility Operating License No. NPF–

73. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 23, 2000 (65 FR 51342).

The July 21, 2000, letter provided

clarifying information that did not

expand the scope of the amendment and

did not change the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated September 28,

2000.

No significant hazards consideration

comments received: No

Florida Power and Light Company, et

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

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

County, Florida

Date of application for amendments:

February 16, 2000.

Brief description of amendments: (1)

Accident monitoring instrumentation

for both St. Lucie Units 1 and 2, (2)

motor operated valve thermal overload

protection bypass device TS for Unit 2,

and (3) an administrative change to the

Unit 2 Technical Specification (TS)

Index.

Date of Issuance: October 4, 2000.

Effective Date: October 4, 2000.

Amendment Nos.: 165 and 109.

Facility Operating License Nos. DPR–

67 and NPF–16: Amendments revised

the TS.

Date of initial notice in Federal

Register: April 5, 2000 (65 FR 17916).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated October 4, 2000.

No significant hazards consideration

comments received: No.

IES Utilities Inc., Docket No. 50–331,

Duane Arnold Energy Center, Linn

County, Iowa

Date of application for amendment:

November 22, 1999, as supplemented

August 14, 2000.

Brief description of amendment: The

amendment would adopt selected NRC

approved generic changes to the

Improved Technical Specifications (ITS)

NUREGs. The 16 changes come from the

Technical Specification Task Force

(TSTF) process developed by the

Industry and the NRC. Three of these

changes are Bases-only changes but are

included for completeness relative to

the TSTF process.

Date of issuance: October 3, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days.

Amendment No.: 234.

Facility Operating License No. DPR–

49: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: January 12, 2000 (65 FR 1924).

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 October 3, 2000.

No significant hazards consideration

comments received: No.

Northeast Nuclear Energy Company, et

al., Docket No. 50–423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of application for amendment:

November 29, 1999, as supplemented by

letter dated May 2, 2000.

Brief description of amendment: The

amendment changes Technical

Specification (TS) 3/4.6.6,

‘‘Supplementary Leak Collection and

Release System’’; TS 3/4.7.7, ‘‘Control

Room Emergency Ventilation System’’;

TS 3/4.7.9, ‘‘Auxiliary Building Filter

System’’; and TS 3/4.9.12, ‘‘Fuel

Building Exhaust System’’; in response

to Generic Letter 99–02, ‘‘Laboratory

Testing of Nuclear-Grade Activated

Charcoal.’’

Date of issuance: October 4, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 90 days from the date of

issuance.

Amendment No.: 184.

Facility Operating License No. NPF–

49: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 26, 2000 (65 FR 4287).

The letter dated May 2, 2000,

provided clarifying information and did

not change the staff’s initial proposed

no significant hazards consideration

determination or expand the scope of

the application as published in the

Federal Register.

The Commission’s related evaluation

of this amendment is contained in a

Safety Evaluation dated October 4, 2000.

No significant hazards consideration

comments received: No.

Nuclear Management Company, Docket

No. 50–263, Monticello Nuclear

Generating Plant, Wright County,

Minnesota

Date of application for amendment:

July 18, 2000.

Brief description of amendment: The

amendment changes the Technical

Specifications to add operability

requirements for the No. 12 residual

heat removal service water pump.

Date of issuance: October 2, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 45 days.

Amendment No.: 113.

Facility Operating License No. DPR–

22: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 23, 2000 (65 FR 51361).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated October 2, 2000.

No significant hazards consideration

comments received: No.

62400 Federal Register / Vol. 65, No. 202 / Wednesday, October 18, 2000 / Notices

PECO Energy Company, PSEG Nuclear

LLC, Delmarva Power and Light

Company, and Atlantic City Electric

Company, Docket No. 50–277, Peach

Bottom Atomic Power Station, Unit No.

2, York County, Pennsylvania

Date of application for amendment:

June 14, 2000, as supplemented August

9, 2000.

Brief description of amendment: This

amendment revised the TSs for safety

limit Minimum Critical Power Ratio

from its current value of 1.10 to 1.09 for

two recirculation-loop operation, and

from 1.12 to 1.10 for single

recirculation-loop operation.

Date of issuance: September 22, 2000.

Effective date: As of date of issuance,

and shall be implemented prior to

startup for Cycle 14 operations,

scheduled for October 2000.

Amendment No.: 236.

Facility Operating License No. DPR–

44: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: July 26, 2000 (65 FR 46012).

The August 9, 2000, letter provided

clarifying information that did not

change the initial proposed no

significant hazards consideration

determination or expand the scope of

the original Federal Register notice.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated September 22,

2000.

No significant hazards consideration

comments received: No.

Power Authority of The State of New

York, Docket No. 50–286, Indian Point

Nuclear Generating Unit No. 3,

Westchester County, New York

Date of application for amendment:

April 27, 2000.

Brief description of amendment: The

amendment would extend the

applicability of the current pressuretemperature limit curves and

overpressure protective system setpoints

from 13.3 to 16.2 effective full-power

years.

Date of issuance: October 5, 2000.

Effective date: October 5, 2000.

Amendment No.: 202.

Facility Operating License No. DPR–

64: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 29, 2000 (65 FR 52431).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated October 5, 2000.

No significant hazards consideration

comments received: No.

Power Authority of the State of New

York, Docket No. 50–333, James A.

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of application for amendment:

August 29, 2000, as supplemented

September 8, 2000.

Brief description of amendment: The

amendment adapts the provisions of the

Boiling Water Reactor Standard

Technical Specifications (STS)

regarding applicability of Technical

Specifications 3.0.D and 4.0.D in the

event of plant shutdown.

Date of issuance: September 29, 2000.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 262.

Facility Operating License No. DPR–

59: Amendment revised the Technical

Specifications.

Public comments requested as to

proposed no significant hazards

consideration: Yes September 14, 2000

(65 FR 55650). That notice provided an

opportunity to submit comments on the

Commission’s proposed no significant

hazards consideration determination.

No comments have been received. The

notice also provided for an opportunity

to request a hearing by October 16,

2000, but indicated that if the

Commission makes a final no significant

hazards consideration determination

any such hearing would take place after

issuance of the amendment.

The Commission’s related evaluation

of the amendment finding of exigent

circumstances, state consultation, and

final determination of no significant

hazards consideration determination are

considered in a Safety Evaluation dated

September 29, 2000.

Power Authority of the State of New

York, Docket No. 50–333, James A.

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of application for amendment:

April 27, 2000, as supplemented

September 5, 2000.

Brief description of amendment: The

amendment changes the Trip Level

Settings for the Residual Heat Removal

and Core Spray Start Timers as well as

the Automatic Depressurization System

Auto-Blowdown Timer.

Date of issuance: October 4, 2000.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 263.

Facility Operating License No. DPR–

59: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 14, 2000 (65 FR 37428).

The September 5, 2000, supplement

did not change the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated October 4, 2000.

No significant hazards consideration

comments received: No.

PSEG Nuclear LLC, Docket Nos. 50–272

and 50–311, Salem Nuclear Generating

Station, Unit Nos. 1 and 2, Salem

County, New Jersey

Date of application for amendments:

March 2, 2000.

Brief description of amendments: The

amendments modify the requirements

in Technical Specifications Section 3/

4.6.3, ‘‘Containment Isolation Valves,’’

by changing limiting conditions for

operation (LCO) 3.6.3.1 and 3.6.3 for

Unit Nos. 1 and 2, respectively. The

changes delete the asterisk (*)

modifying the word OPERABLE in LCO

3.6.3.1 (Unit 1) and LCO 3.6.3 (Unit 2),

and relocate its associated footnote to

the Action portion of the LCO.

Date of issuance: October 2, 2000.

Effective date: As of the date of

issuance, and shall be implemented

within 60 days of issuance.

Amendment Nos.: 235 and 216.

Facility Operating License Nos. DPR–

70 and DPR–75: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: June 28, 2000 (65 FR 39959).

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated October 2, 2000.

No significant hazards consideration

comments received: No.

PSEG Nuclear LLC, Docket Nos. 50–272

and 50–311, Salem Nuclear Generating

Station, Unit Nos. 1 and 2, Salem

County, New Jersey

Date of application for amendments:

March 13, 2000

Brief description of amendments: The

amendments revise TS Table 3.3–6,

‘‘Radiation Monitoring

Instrumentation,’’ by changing the

Containment Gaseous Activity Monitor

(R12A) alarm and trip setpoint for the

containment purge and pressure relief

system isolation for Mode 6 (Refueling)

operations.

Date of issuance: October 2, 2000.

Effective date: As of the date of

issuance, and shall be implemented

within 60 days of issuance.

Amendment Nos.: 236 and 217.

Facility Operating License Nos. DPR–

70 and DPR–75: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: July 26, 2000 (65 FR 46013).

Federal Register / Vol. 65, No. 202 / Wednesday, October 18, 2000 / Notices 62401

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated October 2, 2000.

No significant hazards consideration

comments received: No.

Southern California Edison Company et

al., Docket Nos. 50–361 and 50–362,

San Onofre Nuclear Generating Station,

Units 2 and 3, San Diego County,

California

Date of application for amendments:

May 3, 2000 (PCN–516), as

supplemented August 25, 2000.

Brief description of amendments: The

amendments consist of changes to the

Technical Specifications that revise the

pressure temperature (P–T) limits for 20

effective full power years and reduce

the minimum boltup temperature from

86 °F to 65 °F. The P–T limits

calculations are based on the 1989

American Society of Mechanical

Engineers Appendix G methodology.

Date of issuance: September 28, 2000.

Effective date: September 28, 2000, to

be implemented within 30 days of

issuance.

Amendment Nos.: Unit 2—172; Unit

3—163.

Facility Operating License Nos. NPF–

10 and NPF–15: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: May 31, 2000 (65 FR 34749).

The supplemental letter dated August

25, 2000, provided clarifying

information that was within the scope of

the May 3, 2000, application and 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 amendments is

contained in a Safety Evaluation dated

September 28, 2000.

No significant hazards consideration

comments received: No.

Southern Nuclear Operating Company,

Inc., Georgia Power Company,

Oglethorpe Power Corporation,

Municipal Electric Authority of Georgia,

City of Dalton, Georgia, Docket Nos. 50–

321 and 50–366, Edwin I. Hatch Nuclear

Plant, Units 1 and 2, Appling County,

Georgia

Date of application for amendments:

November 17, 1999, as supplemented by

letter dated August 21, 2000.

Brief description of amendments: The

amendments revise TS 5.5.7,

‘‘Ventilation Filter Testing Program’’ to

include the requirements for laboratory

testing of Engineered Safety Feature

Ventilation System charcoal samples in

accordance with American Society

Testing and Materials D3803–1989 and

the application of a safety factor of 2.0

to the charcoal filter efficiency assumed

in the plant design-basis dose analyses.

In addition, editorial revisions are being

made to some portions of TS Section 5.0

to reference the correct sections of

Regulatory Guide 1.52.

Date of issuance: October 3, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment Nos.: 223 and 164.

Facility Operating License Nos. DPR–

57 and NPF–5: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: December 15, 1999 (64 FR

70091). The supplemental letter dated

August 21, 2000, provided clarifying

information that did not change the

scope of the November 17, 1999,

application and the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated October 3, 2000.

No significant hazards consideration

comments received: Yes. One comment

was received, and is addressed in the

above-referenced Safety Evaluation.

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:

September 28, 1998, as supplemented

on April 22, 1999, April 27, 2000, and

August 15, 2000.

Brief description of amendments: The

amendments revise the technical

specifications (TSs) to eliminate the

need to enter TS 3.0.3 when multiple

trains of either the control room makeup

and cleanup filtration system or the fuel

handling building exhaust air system

are inoperable by providing an allowed

outage time of up to 12 hours to restore

at least one train to an operable status.

Date of issuance: September 26, 2000.

Effective date: September 26, 2000, to

be implemented within 60 days.

Amendment Nos.: Unit 1—125; Unit

2—113.

Facility Operating License Nos. NPF–

76 and NPF–80: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: July 26, 2000 (65 FR 46016).

The August 15, 2000, submittal

provided clarifying information that was

within the scope of the revised

application and Federal Register notice

and did not change the staff’s revised

proposed no significant hazards

considerations determination issued on

July 26, 2000. The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

September 26, 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:

May 16, 2000.

Brief description of amendments:

These amendments change the

Technical Specifications (TSs) by

replacing Surveillance Requirement

(SR) 4.8.1.1.2.c, for evaluating fuel oil

for the emergency diesel generators,

with a Diesel Fuel Oil Program in

Section 6. The revision also deletes the

portion of the SRs that specifies the use

of sodium hypochlorite solution in

cleaning of the fuel oil storage tanks,

deletes the SR to perform a pressure test

on the diesel generator fuel oil system

designed to American Society of

Mechanical Engineers Section III

requirements, and corrects various

typographical errors in the TS and

Bases. Two Bases pages are also added

to each units TS. The applicable TS

Bases are also revised.

Date of issuance: October 2, 2000.

Effective date: October 2, 2000.

Amendment Nos.: 261 and 252.

Facility Operating License Nos. DPR–

77 and DPR–79: Amendments revise the

TS.

Date of initial notice in Federal

Register: August 9, 2000 (65 FR 48758).

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated October 2, 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:

February 4, 2000.

Brief description of amendments:

These amendments change the

Technical Specifications (TS) to revise

the cold leg accumulator volume and

pressure limits based on

instrumentation changes, instrument

inaccuracies, and instrumentation tap

locations. The applicable TS bases are

also revised.

Date of issuance: October 6, 2000.

Effective date: October 6, 2000.

Amendment Nos.: 262 and 253.

Facility Operating License Nos. DPR–

77 and DPR–79: Amendments revise the

TS.

Date of initial notice in Federal

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

The Commission’s related evaluation of

62402 Federal Register / Vol. 65, No. 202 / Wednesday, October 18, 2000 / Notices

the amendment is contained in a Safety

Evaluation dated October 6, 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:

May 25, 2000 (ULNRC–04258).

Brief description of amendment: The

amendment expands (1) The range of

acceptable lift settings for the

pressurizer safety valves (PSVs), and (2)

the tolerance (from ±1% to ±2%) of the

as-found, measured lift settings of tested

PSVs, to be operable. Following testing,

however, the lift settings of the PSVs

would remain nor more than the current

±1%. The amendment revises Technical

Specifications (TS) 3.3.2, ‘‘Engineered

Safety Features Actuation System

(ESFAS) Instrumentation,’’ 3.4.10,

‘‘Pressurizer Safety Valves,’’ and 3.4.11,

‘‘Pressurizer Power Operated Relief

Valves (PORVs),’’ of the Callaway TS.

For TS 3.3.2, a new Action H for one or

more trains inoperable is added, the

note for surveillance requirement (SR)

3.3.2.14 is revised to identify another

slave relay that the SR would be

applicable to, and the automatic PORV

actuation is added to Table 3.3.2–1,

‘‘Engineered Safety Features Actuation

System Instrumentation.’’ For TS 3.4.10,

the range of allowable PSV lift settings

in the limiting condition for operation

(LCO) is expanded from ≥2460 and

≤2510 to ≥2411 and ≤2509, and SR

3.4.10.1 is revised to state that,

following testing, the lift settings shall

be ‘‘within 1% of 2460 psig’’ instead of

simply ‘‘within 1%.’’ The nominal PSV

lift setting would be changed from 2485

psig to 2460 psig. For TS 3.4.11, Actions

A and B is revised to be actions for

inoperable PORVs either solely due to

excessive PORV seat leakage (Action A)

or for reasons other than excessive seat

leakage (Action B), and Action E would

remain an action for two inoperable

PORVs, but would be only for reasons

other than excessive seat leakage.

Date of issuance: September 25, 2000.

Effective date: September 25, 2000, to

be implemented (including issuing the

revised EOP E–O and training all the

control room operator crews on the

revised procedure) before the restart

from refueling outage 11, the next

refueling outage for Callaway Plant,

Unit 1, scheduled to begin in Spring

2001.

Amendment No.: 137.

Facility Operating License No. NPF–

30: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: June 28, 2000 (65 FR 29964).

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 25, 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:

July 21, 2000 (ULNRC–04285), as

supplemented August 16, 2000.

Brief description of amendment: The

amendment revises Limiting Condition

for Operation (LCO) 3.9.4,

‘‘Containment Penetrations,’’ of the

Callaway Technical Specifications (TS)

to allow containment penetrations with

direct access to the outside atmosphere

to be open under administrative

controls during refueling operations, by

adding a note to the LCO that states

‘‘containment penetration flow path(s)

providing direct access from the

containment atmosphere to the outside

atmosphere may be unisolated under

administrative controls.’’ In addition,

there is a format and editorial correction

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

and Start Air,’’ to correct an error in the

conversion to the improved TS issued

May 28, 1999, in Amendment No. 133.

Date of issuance: September 26, 2000.

Effective date: September 26, 2000, to

be implemented (including the

completion of the administrative

procedures that ensure that open

containment penetrations, with direct

acess to the outside atmosphere during

refueling operations with core

alterations and irradiated fuel

movement inside containment, will be

promptly closed in the event of a fuel

handling accident inside containment)

before refueling operations during

refueling outage 11.

Amendment No.: 138.

Facility Operating License No. NPF–30:

The amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 23, 2000 (65 FR 51364).

The August 16, 2000, supplement

provided additional clarifying

information, did not expand the scope

of the application as originally noticed,

and did not change the staff’s original

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated September 26,

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:

September 8, 1999.

Brief description of amendment: The

amendment authorizes revisions to the

descriptions of the steam generator tube

rupture and main steam line break

accidents in the Callaway Plant, Unit 1

Final Safety Analysis Report (FSAR) to

reflect increases in the radiological dose

consequences calculated by the licensee

for these accidents.

Date of issuance: September 27, 2000.

Effective date: September 27, 2000, to

be implemented in the next periodic

update to the FSAR in accordance with

10 CFR 50.71(e). Implementation of the

amendment is the incorporation into the

FSAR the changes to the description of

the facility as described in the licensee’s

application dated September 8, 1999,

and evaluated in the staff’s Safety

Evaluation attached to the amendment.

Amendment No.: 139.

Facility Operating License No. NPF–30:

The amendment revised the Final Safety

Analysis Report.

Date of initial notice in Federal

Register: October 6, 1999 (64 FR 54383).

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 27, 2000.

No significant hazards consideration

comments received: No.

Dated at Rockville, Maryland, this 11th day

of October 2000.

For the Nuclear Regulatory Commission.

John A. Zwolinski,

Director, Division of Licensing Project

Management. Office of Nuclear Reactor

Regulation.

[FR Doc. 00–26645 Filed 10–17–00; 8:45 am]

BILLING CODE 7590–01–P

PRESIDIO TRUST

Presidio Theatre (Building 99), The

Presidio of San Francisco, California,

Notice of Termination of

Environmental Impact Statement

Process

AGENCY: The Presidio Trust.

ACTION: Notice of termination of

Environmental Impact Statement (EIS)

process for the rehabilitation and

expansion of the Presidio Theatre

(Building 99) within The Presidio of San

Francisco, San Francisco, California

(Presidio).

RATIONALE: The Presidio Trust (Trust) is

terminating the EIS process for the

Presidio Theatre in order to complete an

update of the 1994 General Management