Federal Register / Vol. 65, No. 230 / Wednesday, November 29, 2000 / Notices 71131

longer wish to receive it, or would like

to be added to it, please contact the

Office of the Secretary, Attn: Operations

Branch, Washington, DC 20555 (301–

415–1661). In addition, distribution of

this meeting notice over the Internet

system is available. If you are interested

in receiving this Commission meeting

schedule electronically, please send an

electronic message to wmh@hrc.gov or

dkw@nrc.gov.

Dated: November 24, 2000.

William M. Hill, Jr.,

SECY Tracking Officer, Office of the

Secretary.

[FR Doc. 00–30465 Filed 11–27–00; 10:45

am]

BILLING CODE 7590–01–M

NUCLEAR REGULATORY

COMMISSION

Biweekly Notice; Applications and

Amendments to Facility Operating

Licenses Involving No Significant

Hazards Considerations

I. Background

Pursuant to Public Law 97–415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 97–415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from November 6,

2000, through November 16, 2000. The

last biweekly notice was published on

November 15, 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.

Written comments may be submitted

by mail to the Chief, Rules and

Directives Branch, Division of

Administrative Services, Office of

Administration, U.S. Nuclear Regulatory

Commission, Washington, DC 20555–

0001, and should cite the publication

date and page number of this Federal

Register notice. Written comments may

also be delivered to Room 6D22, Two

White Flint North, 11545 Rockville

Pike, Rockville, Maryland, from 7:30

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

Copies of written comments received

may be examined at the NRC’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 December 29, 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

71132 Federal Register / Vol. 65, No. 230 / Wednesday, November 29, 2000 / Notices

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, located at One White

Flint North, 11555 Rockville Pike (first

floor), Rockville, Maryland 20852, by

the above date. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555–0001, and to the attorney for

the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

factors specified in 10 CFR

2.714(a)(1)(i)–(v) and 2.714(d).

For further details with respect to this

action, see the application for

amendment which is available for

public inspection at the Commission’s

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 amendment request: October

6, 2000 (U–603329).

Description of amendment request:

The proposed amendment would

relocate Technical Specification Figure

3.6.4.1–1, ‘‘Secondary Containment

Drawdown Time for 1500 cfm Boundary

Leakage’’ to plant procedures.

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 any accident previously

evaluated.

The proposed Technical specification (TS)

change eliminates an inconsistency between

the Secondary Containment surveillance

requirement (SR 3.6.4.1.4) and the associated

Bases. The proposed change (1) revises the

wording of SR 3.6.4.1.4 to verify the time to

draw down the secondary containment to

≥0.25 inch water gauge for each standby gas

treatment (SGT) subsystem is within the

required time; and (2) relocates the specific

acceptance criteria (existing TS Figure

3.6.4.1–1) to plant procedures and the TS

Bases.

The scope of the proposed change is thus

limited to the affected SR. No changes to

plant equipment or the plant design are

involved. The affected SR, as are

surveillances in general, is not an initiator to

any accident previously evaluated.

Consequently, the probability of an accident

previously evaluated is not significantly

increased.

The proposed change impacts SR 3.6.4.1.4

but does not change its intent or the

associated acceptance criteria. Thus, the

components and structural integrity being

tested will still be required to be maintained

Operable and capable of performing the

accident mitigation functions assumed in the

accident analysis. As a result, the

consequences of any accident previously

evaluated are not significantly affected.

Therefore, this change does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

(2) The proposed change would not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed Technical Specification (TS)

change eliminates an inconsistency between

the Secondary Containment surveillance

requirement (SR 3.6.4.1.4) and the associated

Bases. The proposed change does not involve

a physical alteration of the plant (no new or

different type of equipment will be installed)

or a change in the methods governing normal

plant operation. No new failure modes are

thus introduced by the proposed change.

Therefore, this change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

(3) The proposed change will not involve

a significant reduction in the margin of

safety.

The proposed Technical Specification (TS)

change eliminates an inconsistency between

the Secondary Containment Integrity

surveillance requirement (SR 3.6.4.1.4) and

the associated Bases. The revised wording of

the Surveillance Requirement and the

relocation of the acceptance criteria to plant

procedures and TS Bases have been

evaluated to ensure that they are sufficient to

verify that the equipment used to meet the

LCO can perform its required functions. The

relocation of the acceptance criteria is

consistent with the Bases previously

approved in Amendment 21. Thus,

appropriate equipment continues to be tested

in a manner that gives confidence that the

equipment can perform its assumed safety

function. Therefore, this 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: Kevin P. Gallen,

Morgan, Lewis & Bockius, LLP, 1800 M

Street, NW, Washington, DC 20036–

5869

NRC Section Chief: Anthony J.

Mendiola.

AmerGen Energy Company, LLC, Docket

No. 50–461, Clinton Power Station, Unit

1, DeWitt County, Illinois

Date of amendment request: October

6, 2000 (U–603332).

Description of amendment request:

The proposed amendment would

remove from the Technical

Specification surveillance requirements

the minimum operating time specified

Federal Register / Vol. 65, No. 230 / Wednesday, November 29, 2000 / Notices 71133

for the containment/hydrogen mixing

system.

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 any accident previously

evaluated.

The proposed Technical Specification (TS)

change deletes the minimum run time

requirement (of 15 minutes) for the

Containment/Drywell Hydrogen Mixing

System as denoted in Surveillance

Requirement 3.6.3.3.1. Satisfying a TS

Surveillance Requirement ensures that the

associated system will function to mitigate

the consequences of an accident, and, as such

is not an initiator of an accident or

malfunction. Therefore, since such a test

requirement or operation is not an initiator

to any accident previously evaluated, the

probability of an accident previously

evaluated is not significantly increased.

In addition, the equipment being tested is

still required to be operable and capable of

performing its accident mitigation function

assumed in the accident analysis.

Eliminating the time requirement from the

Surveillance Requirement does not reduce or

relax the requirements to ensure that all

controls are functioning properly. Also, it

does not reduce or relax the requirements for

ensuring the degraded conditions such as

piping blockage, compressor failure or

excessive vibration can be detected for

corrective action. As a result, the

consequences of any accident previously

evaluated are not significantly affected.

Therefore, this change does not involve a

significant increase in the consequences of an

accident previously evaluated.

(2) The proposed change would not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed TS change, as denoted in

Surveillance Requirement 3.6.3.3.1, does not

involve a physical alteration of the plant (no

new or different type of equipment will be

installed) or a change in the methods

governing normal plant operation. No new

potential accident initiators are therefore

introduced by the proposed change. Thus,

this change does not create the possibility of

a new or different kind of accident from any

accident previously evaluated.

(3) The proposed change will not involve

a significant reduction in the margin of

safety.

The proposed TS change to delete the

minimum run time requirement in

Surveillance Requirement 3.6.3.3.1 does not

result in a significant reduction in the margin

of safety. As provided in the justification for

the proposed change, the 15-minute run time

acceptance criterion is not necessary to

ensure that the Containment/Drywell

Hydrogen Mixing can perform its required

function. Thus, if a margin of safety can be

ascribed to proper operation of this system

for LOCA mitigation, the system will

continue to be tested in a manner that gives

confidence that it can perform its assumed

safety function. Therefore, this 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: Kevin P. Gallen,

Morgan, Lewis & Bockius, LLP, 1800 M

Street, NW, Washington, DC 20036–

5869.

NRC Section Chief: Anthony J.

Mendiola.

AmerGen Energy Company, LLC, Inc. et

al., Docket No. 50–219, Oyster Creek

Nuclear Generating Station, Ocean

County, New Jersey

Date of amendment request: August

29, 2000.

Description of amendment request:

The proposed amendment would revise

the Technical Specifications in the

‘‘Administrative Controls’’ section to

certain position titles and the Shift

Technical Advisor (STA) staffing

requirement to allow one of the required

on-shift Senior Reactor Operator (SRO)

positions to be combined with the

required STA position so as to serve in

a dual role SRO/STA position.

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 operation of the facility in

accordance with the proposed change

involve a significant increase in the

probability or consequences of an accident

previously evaluated?

The [proposed] changes do not affect the

purpose, function, performance, operability,

inspection or testing of and does not make

any physical or procedural changes to plant

systems, structures or components. Also, all

existing technical specification limiting

conditions for operation and surveillance

requirements are retained.

TSCR [Technical Specifications Change

Request] makes administrative changes to

certain position titles without changing the

technical requirements for position

responsibilities.

TSCR also changes the Shift Technical

Advisor (STA) staffing requirement to allow

one of the required on-shift Senior Reactor

Operator (SRO) positions to be combined

with the required STA position so as to serve

in a dual-role SRO/STA position as

encouraged by the NRC in Option 1 of

Generic Letter 86–04, ‘‘Policy Statement on

Engineering Expertise On Shift’’, dated

February 13, 1986.

Implementation of the proposed dual-role

SRO/STA change will result in personnel

with enchanced operational knowledge being

assigned to perform the STA function of

providing accident assessment expertise, and

analyzing and responding to off normal

occurrences when needed. The NRC’s stated

preference in the October 28, 1985, ‘‘Policy

Statement on Engineering Expertise on

Shift’’, indicates that the NRC has concluded

that the individual filling the dual-role SRO/

STA position may perform these functions

better than a non-licensed individual filling

the STA position even when the SRO/STA is

concurrently functioning as one of the

required shift SROs.

Therefore, since no physical or procedural

changes are being made to existing plant

systems, structures or components and since

the position title changes are administrative

in nature and the function and

responsibilities of the STA will be executed

by an appropriately qualified individual

filling the dual-role SRO/STA position,

operation of the facility in accordance with

the proposed amendment would not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

2. Would operation of the facility in

accordance with the proposed change create

the possibility of a new or different kind of

accident from any accident previously

evaluated?

The [proposed] changes do not affect the

purpose, function, performance, operability,

inspection or testing of and does not make

any physical or procedural changes to plant

systems, structures or components. Also, all

existing technical specification limiting

conditions for operation and surveillance

requirements are retained.

TSCR 277 makes administrative changes to

certain position titles without changing the

technical requirements for the position

responsibilities.

TSCR 277 also changes the Shift Technical

Advisor (STA) staffing requirement to allow

one of the required on-shift Senior Reactor

Operator (SRO) positions to be combined

with the required STA position so as to serve

in a dual-role SRO/STA position as

encouraged by the NRC in Option 1 of

Generic Letter 86–04, ‘‘Policy Statement on

Engineering Expertise On Shift’’, dated

February 13, 1986.

Therefore, since no physical or procedural

changes are being made to existing plant

systems, structures or components and since

the position title changes are administrative

in nature and the function and

responsibilities of the STA will be executed

by an appropriately qualified individual

filling the dual-role SRO/STA position,

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 accident previously

evaluated.

3. Would operation of the facility in

accordance with the proposed change

involve a significant reduction in a margin of

safety?

The [proposed] changes do not affect the

purpose, function, performance, operability,

inspection or testing of and does not make

71134 Federal Register / Vol. 65, No. 230 / Wednesday, November 29, 2000 / Notices

any physical or procedural changes to plant

systems, structures or components. Also, all

existing technical specification limiting

conditions for operation and surveillance

requirements are retained.

TSCR 277 makes administrative changes to

certain position titles without changing the

technical requirements for the position

responsibilities.

TSCR 277 also changes the Shift Technical

Advisor (STA) staffing requirement to allow

one of the required on-shift Senior Reactor

Operator (SRO) positions to be combined

with the required STA position so as to serve

in a dual-role SRO/STA position as

encouraged by the NRC in Option 1 of

Generic Letter 86–04, ‘‘Policy Statement on

Engineering Expertise On Shift’’, dated

February 13, 1986.

Therefore, since no physical or procedural

changes are being made to existing plant

systems, structures or components and since

the position title changes are administrative

in nature and the function and

responsibilities of the STA will be executed

by an appropriately qualified individual

filling the dual-role SRO/STA position and

shift staffing required by TS 6.2.2.2 and 10

CFR 50.54(m)(2) will [continue] to be

maintained, operation of the facility in

accordance with the proposed amendment

will not involve a significant reduction in [a]

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Kevin P. Gallen,

Jr., Esquire, Morgan, Lewis, & Bockius

LLP, 1800 M Street, NW., Washington,

DC 20036–5869.

NRC Section Chief: M. Gamberoni.

Consolidated Edison Company of New

York, Inc., Docket No. 50–003, Indian

Point Nuclear Generating Station, Unit

1, Buchanan, New York

Date of amendment request: October

5, 2000.

Brief description of amendment: The

proposed changes would revise

Technical Specifications (TSs) Sections

3.2.1.a, 3.2.1.e, and 3.2.1.f to relocate

administrative controls to the Quality

Assurance Program Description (QAPD).

Basis for proposed no significant

hazards considerations determination:

As required by 10 CFR 50.91(a), the

licensee has provide its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) Does the proposed license amendment

involve a significant increase in the

probability or in the consequences of an

accident previously evaluated?

No. The proposed changes are

administrative in nature. The changes

involve Section 3.2.1.a, referencing the

QAPD instead of the IP#2 [Indian Point

Nuclear Generating Station, Unit 2] UFSAR

[Updated Final Safety Analysis Report],

deleting Section 3.2.1.e and having Section

3.2.1.f, refer to the QAPD for the

administrative controls. These changes do

not affect possible initiating events for

accidents previously evaluated or alter the

configuration or operation of the facility. The

Limiting Safety System Settings and Safety

Limits specified in the current Technical

Specifications remain unchanged. Therefore,

the proposed changes would not involve a

significant increase in the probability or in

the consequences of an accident previously

evaluated.

(2) Does the proposed amendment create

the possibility of a new or different kind of

accident from any accident previously

evaluated?

No. The proposed changes are

administrative in nature. The safety analysis

of the facility remains complete and accurate.

There are no physical changes to the facility

and the plant conditions for which the design

basis accidents have been evaluated are still

valid. The operating procedures and

emergency procedures are unaffected.

Consequently no new failure modes are

introduced as a result of the proposed

changes. Therefore, the proposed changes

would not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

(3) Does the proposed amendment involve

a significant reduction in a margin of safety?

No. The proposed changes are

administrative in nature. Since there are no

changes to the operation of the facility or the

physical design, the IP#1 [Indian Point

Nuclear Generating Station, Unit 1] FSAR

[Final Safety Analysis Report] or the IP#2

UFSAR design basis, accident assumptions,

or IP#1 Technical Specification Bases are not

affected. 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 considerations.

Attorney for Licensee: Brent L.

Brandenberg, Esq., Assistant General

Counsel, Consolidated Edison Company

of New York, Inc., 4 Irving Place-1830,

New York, NY 10003.

NRC Section Chief: Michael T.

Masnik.

Energy Northwest, Docket No. 50–397,

WNP–2, Benton County, Washington

Date of amendment request: October

12, 2000.

Description of amendment request:

The amendment changes the facility

name of WNP–2 to Columbia Generating

Station in all the applicable portions of

the Operating License including

Appendix A (Technical Specifications)

and Appendix B (Environmental

Protection Plan). In addition, the

proposed change makes editorial

changes to Technical Specification

Figure 4.1–1, Site Area Boundary.

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.

No. This request involves an

administrative change only. The Operating

License (OL) and Technical Specification

Figure 4.1–1, Site Area Boundary, are being

changed to reflect the new name of the

facility. In addition, editorial changes are

being made to Figure 4.1–1 for clarification.

No actual plant equipment or accident

analyses are affected by the proposed change.

Therefore, this request will have no impact

on the probability or consequence of any type

of accident previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

No. This request involves an

administrative change only. The OL and

Technical Specification Figure 4.1–1 are

being changed to reflect the new name of the

facility. In addition, editorial changes are

being made to Figure 4.1–1 for clarification.

No actual plant equipment or accident

analyses are affected by the proposed change

and no failure modes not bounded by

previously evaluated accidents will be

created. Therefore, this request will have no

impact on the possibility of any type of

accident: new, different, or previously

evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

No. Margin of safety is associated with

confidence in the ability of the fission

product barriers (i.e., fuel and fuel cladding,

Reactor Coolant System pressure boundary,

and containment structure) to limit the level

of radiation dose to the public. This request

involves an administrative change only. The

OL and Technical Specification Figure 4.1–

1 are being changed to reflect the new name

of the facility. In addition, editorial changes

are being made to Figure 4.1–1 for

clarification.

No actual plant equipment or accident

analyses are affected by the proposed change.

Additionally, the proposed change will not

relax any criteria used to establish safety

limits, will not relax any safety system

settings, or will not relax the bases for any

limiting conditions of operation. Therefore,

this proposed change will not impact 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

Federal Register / Vol. 65, No. 230 / Wednesday, November 29, 2000 / Notices 71135

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Thomas C.

Poindexter, Esq., Winston & Strawn,

1400 L Street, N.W., Washington, DC

20005–3502.

NRC Section Chief: Stephen Dembek.

Energy Northwest, Docket No. 50–397,

WNP–2, Benton County, Washington

Date of amendment request: October

30, 2000.

Description of amendment request:

Surveillance Requirement (SR) 3.6.1.3.8

currently requires verification of the

actuation capability of each excess flow

check valve (EFCV) every 24 months.

The proposed change is to relax the SR

frequency by allowing a ‘‘representative

sample’’ of reactor instrument line

EFCVs to be tested every 24 months,

such that each reactor instrument line

EFCV will be tested at least once every

10 years (nominal). The proposed

change will also result in limiting the

SR to only the reactor instrument line

EFCVs.

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

consequence of an accident previously

evaluated.

The current SR frequency requires each

reactor instrument line EFCV to be tested

every 24-months. The reactor instrument line

EFCVs at WNP–2 are designed so that they

will not close accidentally during normal

operation, but will close if a rupture of the

instrument line is indicated downstream of

the valve, and have their status indicated in

the control room. This proposed change

allows a reduced number of reactor

instrument line EFCVs to be tested every 24-

months. There are no physical plant

modifications associated with this change.

Industry operating experience demonstrates a

high reliability of these valves. Neither

reactor instrument line EFCVs nor their

failures are capable of initiating previously

evaluated accidents; therefore, there can be

no increase in the probability of occurrence

of an accident regarding this proposed

change.

Reactor instrument lines connecting to the

reactor coolant pressure boundary are

equipped with EFCVs and also have a flowrestricting orifice inside containment and

upstream of the EFCV. The consequences of

an unisolable rupture of such an instrument

line has been previously evaluated in WNP–

2 FSAR [Final Safety Analysis Report] 15.6.2.

The instrument lines that penetrate primary

containment conform to Regulatory Guide

1.11 (WNP–2 FSAR 7.1.2.4). Those

instrument lines are Seismic Category I and

terminate in instruments that are Seismic

Category I (reference WNP–2 FSAR Table

6.2–16 note 27).

The sequence of events in WNP–2 FSAR

Section 15.6.2.2 for a reactor instrument line

break assumes a continuous discharge of

reactor water through the instrument line

until the reactor vessel is cooled and

depressurized (5 hours). Although not

expected to occur as a result of this change,

the postulated failure of an EFCV to isolate

as a result of reduced testing is bounded by

this previous evaluation. Therefore, there is

no increase in the previously evaluated

consequences of the rupture of an instrument

line and there is no potential increase in the

consequences of an accident previously

evaluated as a result of this change.

The containment atmosphere and

suppression pool instrument line EFCVs are

required to remain open to sense

containment atmosphere and suppression

pool level conditions during postulated

accidents. They are not required to close

during an instrument line break assumed

during normal plant operation nor is their

design capable of closing during normal

plant conditions. These EFCVs do not meet

the criteria for inclusion in 10 CFR

50.36(c)(3) as they have no active safety

function and thus relocation of their testing

requirements to the FSAR cannot effect the

probability of an increase in the

consequences of an accident previously

evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

This proposed change allows a reduced

number of reactor instrument line EFCVs to

be tested each operating cycle and that the

testing requirements for containment

atmosphere and suppression pool instrument

line EFCVs be relocated to the FSAR. No

other changes in requirements are being

proposed. Industry operating experience

demonstrates the high reliability of these

valves. The potential failure of a reactor

instrument line EFCV to isolate by the

proposed change in testing is bounded by the

previous evaluation of an instrument line

rupture. This change will not physically alter

the plant (no new or different type of

equipment will be installed). This change

will not alter the operation of process

variables, structures, or components as

described in the safety analysis. Thus, a new

or different kind of accident will not be

created.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The consequences of an unisolable rupture

of an instrument line has been evaluated in

WNP–2 FSAR Section 15.6.2 in accordance

with the requirements of Regulatory Guide

1.11. That evaluation assumed a continuous

discharge of reactor water for the duration of

the detection and cooldown sequence (5

hours). The only margin of safety applicable

to this proposed change is considered to be

that implied by this evaluation. Since a

continuous discharge was assumed in this

evaluation, any potential failure of a reactor

instrument line EFCV to isolate as a result of

reduced testing frequency is bounded by

existing analysis and does not involve a

significant reduction in the margin of safety.

There is no accident for which the

containment atmosphere or suppression pool

instrument line EFCVs are designed to

actuate to the isolation position for

mitigation. A postulated break of a

containment atmosphere or suppression pool

instrument line under normal operating

conditions would not result in a condition

that would create the ability for these EFCVs

to operate because neither the containment

pressure nor the suppression pool level head

would be sufficient to result in their

actuation. As these EFCVs have no active

design or safety function, the relocation of

testing requirements would not involve a

significant reduction in the margin of safety.

A postulated break of any instrument line

simultaneously with a loss of coolant

accident is beyond the design basis for the

plant.

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: Thomas C.

Poindexter, Esq., Winston & Strawn,

1400 L Street, N.W., Washington, DC

20005–3502.

NRC Section Chief: Stephen Dembek.

GPU Nuclear Corporation and Saxton

Nuclear Experimental Corporation

(SNEC), Docket No. 50–146, Saxton

Nuclear Experimental Facility (SNEF),

Bedford County, Pennsylvania

Date of amendment request: February

2, 2000, as supplemented on August 11

and September 18, 2000.

Description of amendment request:

The proposed amendment would

approve the license termination plan for

the SNEF.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensees have provided their analysis of

the issue of no significant hazards

consideration, which is presented

below:

The proposed change is necessary to

achieve the decommissioning objective of

terminating the license and releasing the site

for unrestricted use. As such, the proposed

change:

1. Will not involve a significant increase in

the probability or consequences of an

accident previously evaluated since

accidents which might occur during the

active decommissioning phase of the SNEC

facility are bounded by the twelve accidents

addressed in section 3.0 of the Updated

Safety Analysis Report (USAR). The accident

analysis addressed in the USAR demonstrate

that no adverse public health and safety

impacts are expected from accidents that

71136 Federal Register / Vol. 65, No. 230 / Wednesday, November 29, 2000 / Notices

might occur during decommissioning

operations at the SNEC facility. The greater

part of radioactively contaminated materials

and components originally located in the

SNEC facility Containment Vessel are no

longer on site, having been shipped as

radioactive waste.

Implementation of the SNEC License

Termination Plan involves a continuation of

the decommissioning process including the

final status survey activity to be performed

prior to site closeout at the end of the

dismantlement phase. These activities do not

involve a significant increase in either the

probability or consequences of an accident

previously evaluated.

2. Will not create the possibility of a new

or different kind of accident from any

accident previously evaluated. Accidents

previously evaluated in the USAR access

different methods of dispersing radioactive

material to the environment, which include

a loss of support systems and external events.

Remaining dismantlement activities and final

status survey work described in the License

Termination Plan are similar to those

previously performed and will not create the

possibility of a new or different kind of

accident from any previously evaluated.

3. Will not involve a significant reduction

in a margin of safety. The Technical

Specifications currently in place at the SNEC

facility were developed to safely

decommission the SNEC facility. Issuance of

the proposed amendment would not reduce

the controls established by the technical

specifications for activities performed at the

SNEC facility. The proposed License

Amendment establishes additional controls

to ensure License Termination Plan activities

are performed effectively. Thus, this change

does not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

analysis of the licensees and, based on

this review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Attorney for the Licensee: Ernest L.

Blake, Jr., Esquire, Shaw, Pittman, Potts,

and Trowbridge, 2300 N Street, N.W.,

Washington, DC 20037.

NRC Branch Chief: Ledyard B. Marsh.

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 1,

2000.

Description of amendment request:

The licensee is proposing to relocate

Technical Specifications 3.3.3.2,

‘‘Instrumentation, Movable Incore

Detectors’’; 3.3.3.3, ‘‘Instrumentation,

Seismic Instrumentation’’; 3.3.3.4,

‘‘Instrumentation, Meteorological

Instrumentation’’; 3.3.3.8, ‘‘Loose-Part

Detection System’’; and 3.3.4, ‘‘Turbine

Overspeed Protection’’ and Index Pages

vi and vii to the Technical

Requirements Manual (TRM). The Bases

of the affected Technical Specifications

will be modified to address the

proposed changes.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed changes to relocate the

movable incore detector instrumentation,

seismic monitoring instrumentation,

meteorological monitoring instrumentation,

loose-part detection instrumentation, and

turbine overspeed protection instrumentation

from the Technical Specifications to the TRM

will have no adverse effect on plant

operation, or the availability or operation of

any accident mitigation equipment. The

plant response to the [design-basis accidents]

DBAs will not change. In addition, the

movable incore detector instrumentation,

seismic monitoring instrumentation,

meteorological monitoring instrumentation,

and loose-part detection instrumentation are

not accident initiators and cannot cause an

accident. For the turbine overspeed

protection instrumentation, the DBAs and

transients include a variety of system failures

and conditions which might result from

turbine overspeed events and potential

missiles striking various plant systems and

equipment. However, in view of the low

likelihood of the generation of turbine

missiles, the turbine overspeed protection

instrumentation does not serve a primary

protective function. Therefore, these changes

will not significantly increase the probability

or consequences of an accident previously

evaluated.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed changes to relocate the

movable incore detector instrumentation,

seismic monitoring instrumentation,

meteorological monitoring instrumentation,

loose-part detection instrumentation, and

turbine overspeed protection instrumentation

do not alter the plant configuration (no new

or different type of equipment will be

installed) or require any new or unusual

operator actions. They do not alter the way

any [structure, system, or component] SSC

functions and do not alter the manner in

which the plant is operated. The proposed

changes do not introduce any new failure

modes. Therefore, the proposed changes will

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 proposed Technical Specification and

Bases changes will relocate the requirements

for the movable incore detector

instrumentation, seismic monitoring

instrumentation, meteorological monitoring

instrumentation, loose-part detection

instrumentation, and turbine overspeed

protection instrumentation from Technical

Specifications to the TRM. Any future

changes to the relocated requirements will be

in accordance with 10 CFR 50.59 and

approved station procedures. The proposed

changes will have no adverse effect on plant

operation, or the availability or operation of

any accident mitigation equipment. The

plant response to the DBAs will not change.

In addition, the relocated requirements do

not meet any of the 10 CFR 50.36c(2)(ii)

criteria on items for which Technical

Specifications must be established.

Therefore, there will be no significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

P.O. Box 270, Hartford, Connecticut.

NRC Section Chief: James W. Clifford.

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 29,

2000 as supplemented on October 16,

2000.

Description of amendment request:

The proposed changes would modify

Technical Specification (TS) Sections

3.3.2, ‘‘Instrumentation—Engineered

Safety Features Actuation System

Instrumentation;’’ 3.7.7, ‘‘Plant

Systems—Control Room Emergency

Ventilation System;’’ 3.7.8, ‘‘Plant

Systems—Control Room Envelope

Pressurization System;’’ 3.7.9, ‘‘Plant

Systems—Auxiliary Building Filter

System;’’ 3.9.1.1, ‘‘Refueling

Operations—Boron Concentration,’’

3.9.1.2; ‘‘Refueling Operations—Boron

Concentration;’’ 3.9.2, ‘‘Refueling

Operations—Instrumentation;’’ 3.9.4,

‘‘Refueling Operations—Containment

Building Penetrations;’’ 3.9.9,

‘‘Refueling Operations—Containment

Purge and Exhaust Isolation System;’’

3.9.10, ‘‘Refueling Operations—Water

Level—Reactor Vessel;’’ and 3.9.12,

‘‘Refueling Operations—Fuel Building

Exhaust Filter System.’’ Some of these

proposed changes are associated with

the revised fuel handling accident

analysis, and integrity of the Control

Room and the Fuel Building boundaries.

Several administrative changes are also

proposed to reflect Millstone Unit 3

terminology, removal of unnecessary

information and to eliminate confusion

Federal Register / Vol. 65, No. 230 / Wednesday, November 29, 2000 / Notices 71137

by providing consistency between

limiting conditions for operation, action

requirements, and Surveillance

Requirements. The proposed Technical

Specifications changes associated with

the revised containment fuel handling

accident analysis results in an increase

in the consequences of a containment

fuel handling accident since the current

analysis of a containment fuel handling

accident does not assume the release of

any radioactive material from

containment. The revised analysis

assumes a release of radioactive material

because it assumes both personnel

access hatch doors are open and at least

one hatch door is closed within 10

minutes of a fuel handling accident

inside containment.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration. The NRC staff has

reviewed the licensee’s analysis against

the standards of 10 CFR 50.92(c). The

NRC staff’s review is presented below:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Technical Specification Changes Associated

with Analyses Changes

The proposed Technical Specification

changes associated with the revised fuel

handling accident analyses will not cause an

accident to occur and will not result in any

change in the operation of the associated

accident mitigation equipment. The design

basis accidents remain the same postulated

events described in the Millstone Unit No. 3

FSAR. Therefore, the proposed changes will

not increase the probability of an accident

previously evaluated.

The proposed Technical Specification

changes associated with the revised fuel

handling accidents analyses will increase the

associated consequences. The increased

consequences are the result of a revised plant

configuration and revised calculation

assumptions, not the result of the addition of

any new plant equipment. The current Fuel

Handling Accident Inside Containment

(FHAIC) analysis assumes the containment is

isolated, or will be isolated, prior to any

release. The revised FHAIC analysis will

allow both containment personnel access

hatch doors to remain open, under

administrative control, during core

alterations and irradiated fuel movement

inside containment. This may result in a

radioactive release if a fuel handling accident

were to occur. The revised FHAIC analysis

demonstrates that the magnitude of the

potential release is small and bounded by the

consequences of the Design Basis Loss of

Coolant Accident. The increase in the

consequences of the revised Fuel Handling

Accident Inside the Spent Fuel Pool

(FHAISFP) analysis due to the revised

calculation assumptions is small. Therefore,

the proposed changes will not result in a

significant increase in the consequences of an

accident previously evaluated.

Other Technical Specification Changes

The proposed Technical Specification

changes not associated with the revised fuel

handling accidents analyses affect the

limiting conditions for operation (LCOs),

applicability, action requirements, and

surveillance requirements of numerous

specifications associated with plant operating

restrictions, accident mitigation functions,

and accident mitigation equipment. The

affected operating restrictions, accident

mitigation functions, and accident mitigation

equipment are not accident initiators. The

proposed changes will not cause an accident

to occur and will not result in any change in

the operation of the associated accident

mitigation equipment. The design basis

accidents remain the same postulated events

described in the Millstone Unit No. 3 FSAR.

Therefore, the proposed changes will not

increase the probability of an accident

previously evaluated.

The proposed LCO and applicability

changes are consistent with the design basis

accident analyses, including the revised fuel

handling accident analyses. (The proposed

change to the LCO for containment

penetrations, which will allow both

personnel access hatch doors to remain open

during core alterations and irradiated fuel

movement inside containment will result in

an increase in the consequences of a FHAIC

as previously discussed.) This will ensure

that the accident mitigation functions and

associated equipment are available for

accident mitigation as assumed in the

associated analyses. As a result, the accident

analysis assumptions and mitigation methods

will not be adversely affected by these

changes. Therefore, the proposed changes

will not result in a significant increase in the

consequences of an accident previously

evaluated.

The additional proposed changes to the

Technical Specifications that will

standardize terminology, relocate

information to the Bases, remove extraneous

information, and make minor format changes

will not result in any technical changes to the

current requirements. Therefore, these

additional proposed changes will not result

in a significant increase in the probability or

consequences of an accident previously

evaluated.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed changes to the Technical

Specifications do not impact any system or

component that could cause an accident. The

proposed changes will not alter the plant

configuration (no new or different type of

equipment will be installed) or require any

unusual operator actions. The proposed

changes will not alter the way any structure,

system, or component functions, and will not

significantly alter the manner in which the

plant is operated. There will be no adverse

effect on plant operation or accident

mitigation equipment. The response of the

plant and the operators following an accident

will not be significantly different. In

addition, the proposed changes do not

introduce any new failure modes. Therefore,

the proposed changes will not create the

possibility of a new or different kind of

accident from any accident previously

analyzed.

3. Involve a significant reduction in a

margin of safety.

Title 10 of the Code of Federal Regulations

Part 100 establishes the accident exposure

limits (300 rem thyroid and 25 rem whole

body) for the Exclusion Area Boundary and

Low Population Zone. The radiological

consequences resulting from the Technical

Specification changes associated with the

revised fuel handling accident analyses are

well within these limits. The radiological

consequences to the Control Room Operators

resulting from the Technical Specification

changes associated with the revised fuel

handling accident analyses are also within

the GDC 19 limit. Since these limits will not

be exceeded and these limits establish the

margin of safety in the plant’s current

licensing basis, the proposed changes will

not result in a significant reduction in a

margin of safety.

The proposed Technical Specification

LCO, applicability, action requirement, and

surveillance requirement changes not

associated with the revised fuel handling

accidents analyses do not adversely affect

equipment design or operation. In addition,

the proposed allowed outage times and

shutdown times are consistent with times

already contained in the Millstone Unit No.

3 Technical Specifications. Therefore, these

changes will not result in a significant

reduction in a margin of safety.

The additional proposed changes to the

Technical Specifications that will

standardize terminology, relocate

information to the Bases, remove extraneous

information, and make minor format changes

will not result in any technical changes to the

current requirements. Therefore, these

additional changes will not result in a

significant reduction in a margin of safety.

Based on the staff’s analysis, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

P.O. Box 270, Hartford, Connecticut.

NRC Section Chief: James W. Clifford.

PSEG Nuclear LLC, Docket No. 50–354,

Hope Creek Generating Station, Salem

County, New Jersey

Date of amendment request: October

12, 2000.

Description of amendment request:

The proposed amendment would make

changes to Hope Creek Generating

Station (HCGS) Technical Specifications

(TS) and Bases associated with the

drywell vacuum breakers and the

suppression pool vacuum breakers. The

proposed changes are intended to

provide consistency between the HCGS

TS and the Standard Technical

71138 Federal Register / Vol. 65, No. 230 / Wednesday, November 29, 2000 / Notices

Specifications (STS) (NUREG–1433).

These changes include revising or

deleting specific Limiting Conditions for

Operation and Surveillance

Requirements and include relocating

information from these sections to the

Bases. In addition, a change to the

Containment Systems Surveillance

Requirements was proposed to correct

the hierarchical format of that section.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration. The NRC staff has

reviewed the licensee’s analysis against

the standards of 10 CFR 50.92(c). The

NRC staff’s analysis is presented below:

The proposed change does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The function of the suppression-chamberto-drywell vacuum breakers is to relieve

vacuum in the drywell by allowing air and

steam flow from the suppression chamber to

the drywell when the drywell is at a negative

pressure with respect to the suppression

chamber. A negative differential pressure

across the drywell wall is caused by rapid

depressurization of the drywell. Steam

condensing in the drywell as a result of a

primary system rupture results in the most

severe pressure transient.

The function of the reactor building-tosuppression chamber vacuum breakers is to

relieve vacuum when the primary

containment depressurizes below the reactor

building pressure. If the drywell

depressurizes below reactor building

pressure, the negative differential pressure is

mitigated by flow through the reactor

building-to-suppression chamber vacuum

breakers and through the suppressionchamber-to-drywell vacuum breakers. A

negative differential pressure across the

drywell wall is caused by rapid

depressurization of the drywell. The

maximum depressurization rate is a function

of the primary containment spray flow rate

and temperature and the assumed initial

conditions of the primary containment

atmosphere.

Each proposed change to the vacuum

breaker TS was categorized by the licensee as

either administrative, more restrictive, less

restrictive, or as a relocation. In addition, the

licensee made changes to the bases to capture

the information removed from the associated

Action Steps and to be consistent with the

STS. The administrative changes eliminate,

replace, or add words or phrases, to provide

clarity or to achieve consistency with the

STS. The more restrictive changes reduce the

number of vacuum breakers allowed to be

open or reduces the amount of time allowed

to close the open valves. The less restrictive

changes: (1) Eliminate the surveillance

requirements associated with the vacuum

breaker position indicators; (2) reduce the

frequency of vacuum breaker position

verification; (3) increase the time

requirement for functional testing subsequent

to steam discharged to the suppression

chamber from the safety-relief valves; (4)

increase the number of allowable inoperable

valves in one vacuum breaker assembly; (5)

eliminate repetitious visual inspections; and

(6) eliminate channel calibration as a means

to determine operability of the inboard

isolation valve auto-open control system. The

relocation changes move information from

the action steps to the bases.

The licensee stated in their October 12,

2000, application that neither the vacuum

breakers, the vacuum breaker position

indication, nor the vacuum breaker actuation

system are initiators of any analyzed event.

Therefore, the proposed change does not

involve a significant increase in the

probability of an accident previously

evaluated. In addition, the licensee’s

application states that the proposed changes

will ensure the operability of the vacuum

breakers, will provide assurance that the

containment integrity and venting capability

are maintained or restored within 1 hour, and

that sufficient vacuum breakers will remain

operable to mitigate the assumed accidents.

Therefore, there is no significant increase in

the consequences of an accident previously

evaluated. The licensee further stated that

any future changes to the licensee-controlled

documents containing relocated

requirements will be evaluated in accordance

with the PSEG Nuclear 10 CFR 50.59

program. Consequently, no significant

increase in the consequences of an accident

previously evaluated will be allowed without

prior Nuclear Regulatory Commission (NRC)

approval. Therefore, the proposed change

does not involve a significant increase in the

consequences of an accident previously

evaluated.

The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

analyzed.

The licensee stated in their application that

the proposed change does not involve a

physical alteration of the plant and does not

introduce any new modes of plant operation.

In addition, the licensee stated that any

resulting changes to the operation of the

plant will be consistent with assumptions

made in the safety analysis. Therefore the

proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

analyzed.

The proposed change does not involve a

significant reduction in a margin of safety.

The licensee stated in their application that

the proposed change does not impact any

safety analysis assumptions and will provide

assurance that the containment integrity and

venting capability are maintained or restored

within 1 hour. The licensee further stated

that Hope Creek experience has shown that

the change in surveillance frequency of

vacuum breaker position is not a significant

change in operating practice. In addition, the

operability of the vacuum breakers is not

adversely affected by steam discharged

through the safety relief valves (SRVs) and

does not pose an immediate operability

concern. Consequently, the potential impact

from the proposed increase in the amount of

time during which to perform functional

testing subsequent to an SRV lift is minimal.

Therefore, there is no significant reduction in

a margin of safety. The licensee further stated

that any future changes to the licenseecontrolled documents containing relocated

requirements will be evaluated in accordance

with the PSEG Nuclear 10 CFR 50.59

program. Consequently, no significant

reduction in a margin of safety will be

allowed without prior NRC approval.

Therefore, the proposed change does not

involve a significant reduction in a margin of

safety.

Based on the staff’s analysis, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Jeffrie J. Keenan,

Esquire, PSEG Nuclear—N21, P.O. Box

236, Hancocks Bridge, NJ 08038.

NRC Section Chief: James W. Clifford.

Tennessee Valley Authority, Docket

Nos. 50–260 and 50–296, Browns Ferry

Nuclear Plant, Units 2 and 3, Limestone

County, Alabama

Date of amendment request: October

30, 2000 (TS–407).

Description of amendment request:

The proposed amendment would revise

the Technical Specifications (TS) to

remove the term ‘‘maximum pathway’’

from the main steam isolation valve

(MSIV) leakage rate Surveillance

Requirement (SR) 3.6.1.3.10. This

proposed change would provide

consistency with 10 CFR Part 50

Appendix J leak rate testing terminology

for evaluating MSIV leakage rates.

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:

A. The proposed amendment 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, and does not

impact operator performance or procedures.

The leak rate testing of the MSIVs will

continue to be performed in accordance with

10 CFR 50 Appendix J in a manner consistent

with the guidance on leak rate testing

presented in industry guidance documents

and in the Standard TS. The change does not

impact the design basis accident analyses

presented in the Final Safety Analysis Report

(FSAR). This proposed TS change is

considered administrative in that no changes

in leak testing methods or in disposition of

leak rate results are involved. Therefore, the

proposed change does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

Federal Register / Vol. 65, No. 230 / Wednesday, November 29, 2000 / Notices 71139

B. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

No changes in accident analysis are

involved, so the consequences of accidents

will remain within the accident analysis

described in the FSAR. Therefore, the

proposed TS change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

C. The proposed amendment does not

involve a significant reduction in a margin of

safety.

The proposed change does not affect any

plant system or component, and does not

have any impact on plant operation. No

changes in accident analyses are involved,

therefore, the proposed change does not

involve a significant reduction in the margin

of safety as currently defined in the bases of

the applicable TS section or in the FSAR. For

these reasons, the proposed amendment does

not involve a significant reduction in the

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET I0H,

Knoxville, Tennessee 37902.

NRC Section Chief: Richard P.

Correia.

Tennessee Valley Authority, Docket

Nos. 50–259, 50–260 and 50–296,

Browns Ferry Nuclear Plant, Units 1, 2

and 3, Limestone County, Alabama

Date of amendment request:

November 6, 2000 (TS–411).

Description of amendment request:

The proposed amendment would revise

the technical specifications to allow two

Residual Heat Removal (RHR)

suppression pool cooling subsystems to

be inoperable for 8 hours.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

A. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The change does not result in any

hardware or operating procedure changes.

The RHR Suppression Pool Cooling

subsystems are not assumed to be initiators

of any analyzed event. This change allows an

additional 8 hours to restore required RHR

Suppression Pool Cooling subsystem(s) prior

to requiring the initiation of a unit shutdown.

The proposed 8 hour Completion Time

provides some time to restore required

subsystem(s) to Operable status, yet is short

enough that operating an additional 8 hours

is not a significant risk.

Therefore, the proposed amendment does

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

B. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The possibility of a new or different kind

of accident from any previously evaluated is

not created because the proposed change

introduces no new mode of plant operation

and it does not involve a physical

modification to the plant.

C. The proposed amendment does not

involve a significant reduction in a margin of

safety.

The increased time allowed for restoring

required inoperable RHR Suppression Pool

Cooling subsystems is acceptable based on

the small probability of an event requiring

the inoperable suppression pool cooling

subsystems to function and the desire to

restore required subsystems prior to requiring

the initiation of a plant shutdown. Delaying

a plant shutdown will minimize the potential

for a scram which then could result in a need

for a subsystem when it is inoperable. As

such, any reduction in a margin of safety will

be insignificant and offset by the benefit

gained from providing additional time to

restore required subsystem(s), thus avoiding

potential plant transients during shutdown.

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: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET I0H,

Knoxville, Tennessee 37902

NRC Section Chief: Richard P.

Correia.

Three Mile Island Nuclear Station, Unit

2, Docket No. 50–320, Middletown,

Pennsylvania

Date of amendment request: August 9,

2000.

Brief description of amendment

request: The proposed technical

specifications change request (TSCR) is

to revise Three Mile Island Nuclear

Generating Station, Unit 2 (TMI–2),

Technical Specifications Sections

6.5.3.2, 6.5.4.1, 6.5.4.2.a, 6.5.4.2.b,

6.5.4.3, 6.5.4.3.c, 6.5.4.4 and 6.5.4.6, to

eliminate the reference to Independent

Onsite Safety Review Group (IOSRG)

and to define the performance of the

IOSRG function by the nuclear quality

assurance organization. Also, two titles

that no longer exist (Manager, TMI–2

Department and division vice president)

were corrected. These administrative

changes are similar to changes that have

been already approved at other plants in

Region I.

Basis for proposed no significant

hazards consideration: As required by

10 CFR 50.91(a), the license 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 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.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

71140 Federal Register / Vol. 65, No. 230 / Wednesday, November 29, 2000 / Notices

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: Ernest L.

Blake, Jr. Esq., Shaw, Pittman, Potts &

Trowbridge 2300 N. Street, N.W.,

Washington, DC 20037.

NRC Section Chief: Mike Masnik.

Vermont Yankee Nuclear Power

Corporation, Docket No. 50–271,

Vermont Yankee Nuclear Power Station,

Vernon, Vermont

Date of amendment request: October

25, 2000.

Description of amendment request:

The proposed amendment would make

editorial and administrative changes to

the Technical Specifications (TS). These

changes correct spelling and

grammatical errors, correct references,

eliminate excessive detail related to

specifying a job title, revise position

titles, consolidate pages and generalize

statements allowing Nuclear Regulatory

Commission (NRC) approved

alternatives to specified requirements.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration which is presented below:

1. The operation of the Vermont Yankee

Nuclear Power Station in accordance with

the proposed amendment will not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes are administrative

or editorial in nature and do not involve any

physical changes to the plant. The changes

do not revise the methods of plant operation

which could increase the probability or

consequences of accidents. No new modes of

operation are introduced by the proposed

changes such that a previously evaluated

accident is more likely to occur or more

adverse consequences would result.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The operation of Vermont Yankee

Nuclear Power Station in accordance with

the proposed amendment will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

These changes are administrative or

editorial in nature and do not affect the

operation of any systems or equipment, nor

do they involve any potential initiating

events that would create any new or different

kind of accident. There are no changes to the

design assumptions, conditions,

configuration of the facility, or manner in

which the plant is operated and maintained.

The changes do not affect assumptions

contained in plant safety analyses or the

physical design and/or modes of plant

operation. Consequently, no new failure

mode is introduced.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any 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.

There are no changes being made to the TS

safety limits or safety system settings. The

operating limits and functional capabilities of

systems, structures and components are

unchanged as a result of these administrative

and editorial changes. These changes do not

affect any equipment involved in potential

initiating events or plant response to

accidents. There is no change to the basis for

any Technical Specification that is related to

the establishment or maintenance of, a

nuclear safety margin.

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.

Tennessee Valley Authority, Docket No.

50–390 Watts Bar Nuclear Plant, Unit 1,

Rhea County, Tennessee

Date of application for amendments:

October 30, 2000.

Brief description of amendments:

Modify the Technical Specifications to

allow a one-time-only increase in the

diesel generator Action Completion

Time from 72 hours to 10 days to

facilitate potential repairs to an

emergency diesel generator to improve

reliability.

Date of publication of individual

notice in the Federal Register: November

3, 2000 (65 FR 66266).

Expiration date of individual notice:

December 4, 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).

Federal Register / Vol. 65, No. 230 / Wednesday, November 29, 2000 / Notices 71141

AmerGen Energy Company, LLC, Docket

No. 50–461, Clinton Power Station, Unit

1, DeWitt County, Illinois

Date of application for amendment:

September 20, 2000.

Brief description of amendment: The

amendment allows placing a static VAR

compensator into service with just one

of the two protective subsystems

operable.

Date of issuance: November 13, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days.

Amendment No.: 136.

Facility Operating License No. NPF–

62: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: October 2, 2000 (65 FR 58829).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated November 13,

2000.

No significant hazards consideration

comments received: No.

AmerGen Energy Company, LLC, et al.,

Docket No. 50–219, Oyster Creek

Nuclear Generating Station, Ocean

County, New Jersey

Date of application for amendment:

March 21, as supplemented on June 14,

September 26, and October 16, 2000.

Brief description of amendment: The

proposed amendment revised the

Technical Specifications to delete the

reporting requirements for the core

spray sparger inspection.

Date of Issuance: November 2, 2000.

Effective date: November 2, 2000 and

shall be implemented within 30 days of

issuance.

Amendment No.: 217.

Facility Operating License No. DPR–

16: Amendment revised the Technical

Specifications and License.

Date of initial notice in Federal

Register: September 22, 2000 (65 FR

57404).

The June 14, September 26, and

October 16, 2000, letters provided

clarifying information that did not

change the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of this amendment is contained in a

Safety Evaluation dated November 2,

2000.

No significant hazards consideration

comments received: No.

AmerGen Energy Company, LLC, et al.,

Docket No. 50–219, Oyster Creek

Nuclear Generating Station, Ocean

County, New Jersey

Date of application for amendment:

June 30, as supplemented on September

26, 2000.

Brief description of amendment: The

proposed amendment revised the

Technical Specifications (TSs) to

establish that the existing Safety Limit

Minimum Critical Power Ratio

contained in TS 2.1.A is applicable for

the next operating cycle (Cycle 18).

Date of Issuance: November 3, 2000.

Effective date: November 3, 2000 and

shall be implemented within 30 days of

issuance.

Amendment No.: 218.

Facility Operating License No. DPR–

16: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: September 22, 2000 (65 FR

57406).

The September 26, 2000, letter

provided clarifying information 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 this amendment is contained in a

Safety Evaluation dated November 3,

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:

November 19, 1999, as supplemented

July 18, 2000.

Brief description of amendments: The

amendments revise Technical

Specification 5.5.11.c, ‘‘Ventilation

Filter Testing Program (VFTP),’’ to

include the requirement for laboratory

testing of Engineered Safety Feature

Ventilation System charcoal samples

per American Society for Testing and

Materials (ASTM) D3803–1989,

‘‘Standard Test Method for NuclearGrade Activated Carbon,’’ in response to

Generic Letter 99–02, ‘‘Laboratory

Testing of Nuclear-Grade Activated

Charcoal,’’ dated June 3, 1999.

Date of issuance: November 8, 2000.

Effective date: November 8, 2000, and

shall be implemented within 45 days of

the date of issuance.

Amendment Nos.: Unit 1–130, Unit

2–130, Unit 3–130.

Facility Operating License Nos. NPF–

41, NPF–51, and NPF–74: The

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 8, 2000 (65 FR 12287).

The July 18, 2000, supplement

provided clarifying information that was

within the scope of the original

application and Federal Register notice

and did not change the staff’s initial

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated November 8,

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:

February 29, 2000, as supplemented by

letters dated June 26 and August 18,

2000.

Brief description of amendments: The

amendments revised the pressuretemperature (P–T) limits for heatup,

cooldown, critical operation and

inservice leak and hydrostatic test

limitations for the reactor pressure

vessel (RPV). The amendments replaced

the current RPV P–T limit curves with

three recalculated curves that are

applicable to 32 effective full power

years. The staff has approved the

revised limits for an interim period not

to exceed December 15, 2002.

Date of issuance: November 8, 2000.

Effective date: Immediately until

December 15, 2002, to be implemented

within 30 days.

Amendment Nos.: 144 and 130.

Facility Operating License Nos. NPF–

11 and NPF–18: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

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

The June 26 and August 18, 2000,

submittals provided additional

information that did not change the

scope of the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated November 8,

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 28, 2000.

Brief description of amendments: The

amendments revise License Condition

2.C.(37) for Unit 1 and License

Condition 2.C.(21) for Unit 2, to specify

the types of fuel movements that cannot

be performed during refueling unless all

control rods are fully inserted.

Date of issuance: November 9, 2000.

Effective date: Immediately, to be

implemented within 30 days.

71142 Federal Register / Vol. 65, No. 230 / Wednesday, November 29, 2000 / Notices

Amendment Nos.: 145 and 131.

Facility Operating License Nos. NPF–

11 and NPF–18: The amendments

revised the Operating Licenses.

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 November 9,

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 3, 1999, as supplemented by

letters dated April 4, June 9, June 29,

August 2, and August 16, 2000.

Brief description of amendment: The

amendment authorized revision of the

Safety Analysis Report (SAR) to increase

the containment structural design

pressure from 54 psig to 59 psig, revised

Technical Specification (TS) Table 3.3–

3 to add a containment spray actuation

signal on high-high containment

building pressure to terminate main

feedwater and main steam flow from the

unaffected steam generator, revised TS

3.6.1.4 and Figure 3.6–1 to change the

allowable containment initial

conditions to be consistent with

analysis assumptions, and revised TS

6.15 to increase the calculated peak

accident pressure in the containment

leakage rate testing program from 54

psig to 58 psig. Related changes to the

Bases were also made.

Date of issuance: November 13, 2000.

Effective date: As of the date of

issuance to be implemented prior to the

commencement of heatup from

refueling outage 2R14.

Amendment No.: 225.

Facility Operating License No. NPF–6:

Amendment authorized revision to the

SAR and revised the TSs.

Date of initial notice in Federal

Register: February 23, 2000 (65 FR 9006).

The June 29, 2000, supplement

withdrew the proposed TS change to

clarify the allowable containment

leakage rate. The August 16, 2000,

supplement withdrew the proposed TS

change to increase the allowable

containment spray pump degradation.

The April 4, June 9, June 29, August 2,

and August 16, 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 no significant

hazards consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated November 13,

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:

June 29, 2000, as supplemented by letter

dated October 4, 2000.

Brief description of amendment: The

amendment revised the containment

cooling system Technical Specifications

to require that two independent

containment cooling groups are

operable with two operational cooling

units in each group, in Modes 1, 2, 3,

and 4.

Date of issuance: November 13, 2000.

Effective date: As of the date of

issuance to be implemented within 30

days from the date of issuance.

Amendment No.: 226.

Facility Operating License No. NPF–6:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 26, 2000 (65 FR 46008).

The October 4, 2000, supplement

provided clarifying information that was

within the scope of the original Federal

Register notice and did not change the

staff’s no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated November 13,

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:

January 19, 2000, as supplemented July

19, 2000.

Brief description of amendments: The

amendments will increase the setpoint

tolerances for the pressurizer and main

steam safety valves.

Date of Issuance: November 14, 2000.

Effective Date: November 14, 2000.

Amendment Nos.: 166 and 110.

Facility Operating License Nos. DPR–

67 and NPF–16: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: April 5, 2000 (65 FR 17915).

The July 19, 2000, submittal provided

clarifying information that did not

change the scope of the original Federal

Register Notice or change the initial no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated November 14,

2000.

No significant hazards consideration

comments received: No.

Northeast Nuclear Energy Company, et

al., Docket No. 50–423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of application for amendment:

June 26, 2000.

Brief description of amendment: The

amendment changes the Millstone

Nuclear Power Station, Unit 3,

Technical Specifications (TS) Section

1.13, Definitions, ‘‘Engineered Safety

Features Response Time’’; TS Section

1.28, ‘‘Reactor Trip System Response

Time’’; TS Section 3.3.1,

‘‘Instrumentation-Reactor Trip System

Instrumentation’’; and TS Section 3.3.2,

‘‘Instrumentation-Engineered Safety

Features Actuation System

Instrumentation’’ to provide for

verification of response time for selected

components provided that the

components and the methodology for

verification have been previously

reviewed and approved by the Nuclear

Regulatory Commission.

Date of issuance: November 3, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 60 days from the date of

issuance.

Amendment No.: 187.

Facility Operating License No. NPF–49:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 9, 2000 (65 FR 48755).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated November 3,

2000.

No significant hazards consideration

comments received: No.

PECO Energy Company, PSEG Nuclear

LLC, Delmarva Power and Light

Company, and Atlantic City Electric

Company

Docket Nos. 50–277 and 50–278, Peach

Bottom Atomic Power Station Unit Nos.

2 and 3, York County, Pennsylvania

Date of application for amendments:

November 17, 1999, as supplemented

June 15, 2000.

Brief description of amendments: The

amendments revise TS 5.5.7.c,

‘‘Ventilation Filter Testing Program

(VFTP)’’ to include the requirement for

laboratory testing of Engineered Safety

Feature Ventilation System charcoal

samples per American Society for

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.

Date of issuance: November 3, 2000.

Federal Register / Vol. 65, No. 230 / Wednesday, November 29, 2000 / Notices 71143

Effective date: As of date of issuance,

to be implemented within 30 days.

Amendments Nos.: 237 and 240.

Facility Operating License Nos. DPR–

44 and DPR–56: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: January 26, 2000 (65 FR 4288).

The June 15, 2000, letter provided

clarifying information that did not

change the initial proposed no

significant hazards consideration

determination or expand the application

beyond the scope of the original Federal

Register notice.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated November 3,

2000.

No significant hazards consideration

comments received: No.

South Carolina Electric & Gas Company,

South Carolina Public Service

Authority, Docket No. 50–395, Virgil C.

Summer Nuclear Station, Unit No. 1,

Fairfield County, South Carolina

Date of application for amendment:

June 12, 2000.

Brief description of amendment: The

proposed amendment would revise

Technical Specification 3/4.7.5,

‘‘Ultimate Heat Sink,’’ by increasing the

minimum required service water pond

level from 415 feet to 416.5 feet and

decreasing the maximum allowed

temperature at the discharge of the

service water pumps from 95 degrees

Fahrenheit to 90.5 degrees Fahrenheit.

Date of issuance: November 14, 2000.

Effective date: November 14, 2000.

Amendment No.: 149.

Facility Operating License No. NPF–12:

Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: July 26, 2000 (65 FR 46015).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated November 14,

2000.

No significant hazards consideration

comments received: No.

Notice of Issuance of Amendments to

Facility Operating Licenses and Final

Determination of No Significant

Hazards Consideration and

Opportunity for a Hearing (Exigent

Public Announcement or Emergency

Circumstances)

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 for the

amendment 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.

Because of exigent or emergency

circumstances associated with the date

the amendment was needed, there was

not time for the Commission to publish,

for public comment before issuance, its

usual 30-day Notice of Consideration of

Issuance of Amendment, Proposed No

Significant Hazards Consideration

Determination, and Opportunity for a

Hearing.

For exigent circumstances, the

Commission has either issued a Federal

Register notice providing opportunity

for public comment or has used local

media to provide notice to the public in

the area surrounding a licensee’s facility

of the licensee’s application and of the

Commission’s proposed determination

of no significant hazards consideration.

The Commission has provided a

reasonable opportunity for the public to

comment, using its best efforts to make

available to the public means of

communication for the public to

respond quickly, and in the case of

telephone comments, the comments

have been recorded or transcribed as

appropriate and the licensee has been

informed of the public comments.

In circumstances where failure to act

in a timely way would have resulted, for

example, in derating or shutdown of a

nuclear power plant or in prevention of

either resumption of operation or of

increase in power output up to the

plant’s licensed power level, the

Commission may not have had an

opportunity to provide for public

comment on its no significant hazards

consideration determination. In such

case, the license amendment has been

issued without opportunity for

comment. If there has been some time

for public comment but less than 30

days, the Commission may provide an

opportunity for public comment. If

comments have been requested, it is so

stated. In either event, the State has

been consulted by telephone whenever

possible.

Under its regulations, the Commission

may issue and make an amendment

immediately effective, notwithstanding

the pendency before it of a request for

a hearing from any person, in advance

of the holding and completion of any

required hearing, where it has

determined that no significant hazards

consideration is involved.

The Commission has applied the

standards of 10 CFR 50.92 and has made

a final determination that the

amendment involves no significant

hazards consideration. The basis for this

determination is contained in the

documents related to this action.

Accordingly, the amendments have

been issued and made effective 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 application for

amendment, (2) the amendment to

Facility Operating License, 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).

The Commission is also offering an

opportunity for a hearing with respect to

the issuance of the amendment. By

December 29, 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 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

71144 Federal Register / Vol. 65, No. 230 / Wednesday, November 29, 2000 / Notices

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. Since the Commission has

made a final determination that the

amendment involves no significant

hazards consideration, if a hearing is

requested, it will not stay the

effectiveness of the amendment. Any

hearing held would take place while the

amendment is in effect.

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:

Rulemakings and Adjudications Staff or

may be delivered to the Commission’s

Public Document Room, located at One

White Flint North, 11555 Rockville Pike

(first floor), Rockville, Maryland 20852,

by the above date. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555–0001, and to the attorney for

the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of the

factors specified in 10 CFR

2.714(a)(1)(i)-(v) and 2.714(d).

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:

November 2, 2000, as supplemented

November 3, 2000.

Description of amendment request:

The amendment revises the Technical

Specifications to allow reactor coolant

system (RCS) inservice leak and

hydrostatic testing to be performed with

the reactor in the cold shutdown mode

while the RCS temperature is greater

than 212 °F (which normally

corresponds to the hot shutdown mode).

Date of issuance: November 3, 2000.

Effective Date: As of its date of

issuance and shall be implemented

within 30 days.

Amendment No.: 267.

Facility Operating License No. DPR–59:

Amendment revises the Technical

Specifications.

Public comments requested as to

proposed no significant hazards

consideration: No.

The Commission’s related evaluation

of the amendment, finding of emergency

circumstances, and final determination

of no significant hazards consideration

are contained in a Safety Evaluation

dated November 3, 2000.

Attorney for licensee: Mr. David E.

Blabey, 1633 Broadway, New York, New

York 10019.

NRC Section Chief: Marsha

Gamberoni.

Dated at Rockville, Maryland, this 21st day

of November 2000.

For the Nuclear Regulatory Commission.

Suzanne C. Black,

Acting Director, Division of Licensing Project

Management, Office of Nuclear Reactor

Regulation.

[FR Doc. 00–30282 Filed 11–28–00; 8:45 am]

BILLING CODE 7590–01–P

SECURITIES AND EXCHANGE

COMMISSION

[Release No. 35–27280]

Filings Under the Public Utility Holding

Company Act of 1935, as amended

(‘‘Act’’)

November 21, 2000.

Notice is hereby given that the

following filing(s) has/have been made

with the Commission pursuant to

provisions of the Act and rules

promulgated under the Act. All

interested persons are referred to the

application(s) and/or declaration(s) for

complete statements of the proposed

transaction(s) summarized below. The

application(s) and/or declaration(s) and

any amendment(s) is/are available for

public inspection through the

Commission’s Branch of Public

Reference.

Interested persons wishing to

comment or request a hearing on the

application(s) and/or declaration(s)

should submit their views in writing by

December 18, 2000, to the Secretary,

Securities and Exchange Commission,

Washington, DC 20549–0609, and serve

a copy on the relevant applicant(s) and/

or declarant(s) at the address(es)

specified below. Proof of service (by

affidavit or, in the case of an attorney at

law, by certificate) should be filed with

the request. Any request for hearing

should identify specifically the issues of

facts or law that are disputed. A person

who so requests will be notified of any

hearing, if ordered, and will receive a