31354 Federal Register / Vol. 65, No. 96 / Wednesday, May 17, 2000 / Notices

Conference of Radiation Control

Program Directors (CRCPD) (public

meeting) (Contact: Paul Lohaus,

301–415–3340)

1 p.m. Meeting with Korean Peninsula

Energy Development Organization

(KEDO) and State Department

(public meeting) (Contact: Donna

Chaney, 301–415–2644)

Week of June 19—Tentative

Tuesday, June 20, 2000

9:25 a.m. Affirmation Session (public

meeting) (if needed)

9:30 a.m. Briefing on Final Rule—Part

70, Regulating Fuel Cycle Facilities

(public meeting)

1:30 p.m. Briefing on Risk-Informed

Part 50, Option 3 (public meeting)

The schedule for Commission

meetings is subject to change on short

notice. To verify the status of meetings

call (recording) (301) 415–1292.

CONTACT PERSON FOR MORE INFORMATION:

Bill Hill, (301) 415–1661.

The NRC Commission Meeting

Schedule can be found on the Internet

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

schedule.htm

This notice is distributed by mail to

several hundred subscribers; if you no

longer wish to receive it, or would like

to be added to it, please contact the

Office of the Secretary, Attn: Operations

Branch, Washington, 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@nrc.gov or

dkw@nrc.gov.

Dated: May 12, 2000.

William M. Hill, Jr.,

SECY Tracking Officer, Office of the

Secretary.

[FR Doc. 00–12505 Filed 5–15–00; 8:45 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 April 22,

2000, through May 5, 2000. The last

biweekly notice was published on May

3, 2000 (65 FR 25761).

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; (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 Review and

Directives Branch, Division of Freedom

of Information and Publications

Services, Office of Administration, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555–0001, and

should cite the publication date and

page number of this Federal Register

notice. Written comments may also be

delivered to Room 6D22, Two White

Flint North, 11545 Rockville Pike,

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

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

written comments received may be

examined at the NRC Public Document

Room, the Gelman Building, 2120 L

Street, NW., Washington, DC. The filing

of requests for a hearing and petitions

for leave to intervene is discussed

below.

By June 16, 2000, the licensee may

file a request for a hearing with respect

to issuance of the amendment to the

subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, and electronically

from the ADAMS Public Library

component on the NRC Web site, http:/

/www.nrc.gov (the Electronic Reading

Room). If a request for a hearing or

petition for leave to intervene is filed by

the above date, the Commission or an

Atomic Safety and Licensing Board,

designated by the Commission or by the

Chairman of the Atomic Safety and

Licensing Board Panel, will rule on the

request and/or petition; and the

Secretary or the designated Atomic

Safety and Licensing Board will issue a

notice of a hearing or an appropriate

order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) The nature of the

petitioner’s right under the Act to be

Federal Register / Vol. 65, No. 96 / Wednesday, May 17, 2000 / Notices 31355

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555–0001, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington DC, by

the above date. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555–0001, and to the attorney for

the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

factors specified in 10 CFR

2.714(a)(1)(i)–(v) and 2.714(d).

For further details with respect to this

action, see the application for

amendment which is available for

public inspection at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, and electronically

from the ADAMS Public Library

component on the NRC Web site,

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

Reading Room).

Arizona Public Service Company, et al.,

Docket Nos. STN 50–528, STN 50–529,

and STN 50–530, Palo Verde Nuclear

Generating Station, Units 1, 2, and 3,

Maricopa County, Arizona

Date of amendments request: May 26,

1999, as supplemented March 31, 2000.

Description of amendments request:

The licensee proposes to change the

allowable values in Technical

Specification Section 3.3.1, Table 3.3.1–

1, Item 12, ‘‘Reactor Coolant Flow,

Steam Generator No. 1-Low’’ and Item

13, ‘‘Reactor Coolant Flow, Steam

Generator No. 2-Low,’’ to reduce the

demonstrated spurious trip hazard

associated with this setpoint. This

application was originally noticed in the

Federal Register on June 30, 1999 (64

FR 35201).

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Standard 1—Does the proposed change

involve a significant increase in the

probability or consequences of an accident

previously evaluated?

No. The proposed change will change the

Reactor Protection System (RPS) reactor

coolant flow trip setpoints. The RPS

functions to mitigate the consequences of an

accident. The changes to the low reactor

coolant flow trip setpoints will reduce or

eliminate unnecessary challenges to the RPS.

Therefore, the proposed change will not

involve a significant increase in the

probability of an accident previously

evaluated.

These changes will result in an increased

time delay for the RPS low reactor coolant

flow trip. The reanalysis of the affected

Updated Final Safety Analysis Report

(UFSAR) Chapter 15 event (UFSAR 15.3.4,

Reactor Coolant Pump Shaft Break with Loss

of Offsite Power), with the increased time

delay, shows that the dose consequences for

this event remains bounded by the UFSAR

analysis. Therefore, this change does not

involve a significant increase in the

consequences of an accident previously

evaluated.

Standard 2—Does the proposed change

create the possibility of a new or different

kind of accident from any accident

previously evaluated?

No. The proposed change will change the

RPS reactor coolant flow trip setpoints. The

RPS functions to mitigate the consequences

of an accident. The changes to the low

reactor coolant flow trip setpoints will

reduce or eliminate unnecessary challenges

to the RPS. The proposed change only

changes the mitigating actions of the RPS,

without changing the required function of the

RPS. Therefore, the change to the low reactor

coolant flow trip setpoints does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

Standard 3—Does the proposed change

involve a significant reduction in a margin of

safety?

No. The proposed change will change the

RPS reactor coolant flow trip setpoints. The

reanalysis of the affected UFSAR Chapter 15

event (UFSAR 15.3.4, Reactor Coolant Pump

Shaft Break with Loss of Offsite Power), with

the revised reactor coolant flow trip

setpoints, shows that the minimum departure

from nucleate boiling ratio (DNBR) and

specified acceptable fuel design limits

(SAFDLs) for this event remains bounded by

the UFSAR analysis. 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 that

review, it appears that the three

standards of 10 CFR 50.92(c) are

31356 Federal Register / Vol. 65, No. 96 / Wednesday, May 17, 2000 / Notices

satisfied. Therefore, the NRC staff

proposes to determine that the request

for amendments involves no significant

hazards consideration.

Attorney for licensee: Nancy C. Loftin,

Esq., Corporate Secretary and Counsel,

Arizona Public Service Company, P.O.

Box 53999, Mail Station 9068, Phoenix,

Arizona 85072–3999 NRC Section Chief:

Stephen Dembek

Carolina Power & Light Company, et al.,

Docket Nos. 50–325 and 50–324,

Brunswick Steam Electric Plant (BSEP),

Units 1 and 2, Brunswick County, North

Carolina

Date of amendments request: April

26, 2000.

Description of amendments request:

The proposed amendments would

increase the maximum average ultimate

heat sink (UHS) temperature allowed by

Technical Specification (TS) 3.7.2,

‘‘Service Water System and Ultimate

Heat Sink.’’

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Operation with the maximum 24 hour

average UHS water temperature as high as

90.5°F does not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

The BSEP SW [Service Water] system is

designed to provide cooling water for the

removal of heat from equipment required for

a safe reactor shutdown following a Design

Basis Accident (DBA) or transient. This

equipment includes the Diesel Generators

(DGs), Residual Heat Removal (RHR) pump

seal coolers, room cooling units for

Emergency Core Cooling System (ECCS)

equipment, and Residual Heat Removal

Service Water (RHRSW) heat exchangers. The

SW system also provides cooling to other

components, as required, during normal

operation. The SW system is not an initiator

of any previously evaluated accident. The

safety related components associated with

SW cooling have been analyzed for a

maximum UHS temperature of 92°F. The

proposed change maintains this maximum

UHS temperature. As such, the qualification

of safety related components is not affected.

Therefore, the probability of occurrence of a

previously evaluated accident is not

increased.

The new maximum 24 hour average UHS

water temperature limit of 90.5° F has been

evaluated and it was determined that the SW

system will maintain sufficient heat removal

capability. Existing TS operability

requirements for the UHS ensure that

conservatively bounding assumptions used

in the analysis of the SW system’s heat

removal capability will be met, or the UHS

will be declared inoperable. As such, the

consequences of previously analyzed

accidents are not affected[.]

2. Operation with the maximum 24 hour

average UHS water temperature as high as

90.5°F will not create the possibility of a new

or different kind of accident from any

accident previously evaluated.

Increasing the maximum 24 hour average

UHS water temperature does not create the

possibility of an accident of a different type

than any evaluated previously in the safety

analysis report. UHS water temperature does

not represent an accident initiator. There is

no physical change to any plant structure,

system, or components. Therefore, there is no

possibility of an accident of a different type.

Increasing the maximum 24 hour average

UHS water temperature does not create the

possibility of a malfunction of a different

type than any evaluated previously. The

safety related components associated with

SW cooling have been analyzed for a

maximum UHS temperature of 92°F. This

maximum UHS temperature is maintained by

the proposed change. As such, this condition

does not introduce the possibility of a

malfunction of a different type than any

evaluated.

3. Operation with the maximum 24 hour

average UHS water temperature as high as

90.5°F does not involve a significant

reduction in a margin of safety.

UHS temperature limits are established to

ensure that the SW system is able to provide

sufficient cooling water for the removal of

heat from equipment, such as the DGs, RHR

pump seal coolers, ECCS room cooling units,

and RHRSW heat exchangers, required for a

safe reactor shutdown following a DBA or

transient. CP&L has performed an analysis

which demonstrates that this capability is not

reduced with the increased maximum 24

hour average UHS water temperature limit.

Existing TS operability requirements for the

UHS ensure that conservatively bounding

assumptions used in the analysis of the SW

system’s heat removal capability will be met,

or the UHS will be declared inoperable. As

such, the ability of the SW system to perform

its intended safety function is not affected

and the margin of safety is not reduced.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: William D.

Johnson, Vice President and Corporate

Secretary, Carolina Power & Light

Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Section Chief: Richard P.

Correia.

Duke Energy Corporation, Docket Nos.

50–269, 50–270, and 50–287, Oconee

Nuclear Station, Units 1, 2, and 3,

Oconee County, South Carolina

Date of amendment request: April 13,

2000

Description of amendment request:

The proposed amendments would

revise the Technical Specifications to

accommodate the use of Framatome

Cogema Fuels Mark-B11 fuel with M5

cladding.

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 consequence of an accident

previously evaluated?

No. The proposed change to the technical

specifications and bases incorporate the use

of Mark-B11 fuel assemblies with M5

cladding. The analyzed events are initiated

by the failure of specific plant structures,

systems, or components. The change in fuel

assembly design or cladding material does

not impact the condition or performance of

those structures, system, or components.

Therefore, the proposed changes will not

increase the probability of an accident

previously evaluated.

The accident analyses have been evaluated

to address the changes in the fuel design and

cladding material. The results of this

evaluation demonstrate that the applicable

acceptance criteria are met. Thus, the

proposed changes will not increase the

consequences of an accident previously

evaluated.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

No. The proposed changes to the technical

specifications are to support implementation

of Mark-B11 fuel assemblies with M5

cladding. The changes in fuel design and

cladding material do not alter the operating

characteristics of the plant. In addition, the

fuel handling equipment is compatible with

the Mark-B11 fuel assembly design.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Involve a significant reduction in a

margin of safety?

No. The margin of safety is established

through the design of the plant systems,

structures, components, and the parameters

within which the plant is operated. The

proposed change does not involve any

significant physical change to the plant. The

primary design changes, which enhance

nuclear, thermal-hydraulic and mechanical

performance, include the following:

1. Reduced diameter fuel rod,

2. Flow mixing vanes on five of the six

intermediate spacer grids,

3. Improved grid restraint system, and

4. M5 fuel rod cladding.

The changes in fuel design and cladding

material have been evaluated which

demonstrates that all of the applicable

acceptance criteria are met. Based on this, the

proposed changes do not involve a

significant reduction in a margin of safety.

Duke has concluded based on the above

that there are no significant hazards

considerations involved in this request.

The NRC staff has reviewed the

licensee’s analysis and, based on this

Federal Register / Vol. 65, No. 96 / Wednesday, May 17, 2000 / Notices 31357

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Anne W.

Cottington, Winston and Strawn, 1200

17th Street, NW., Washington, DC

20005.

NRC Section Chief: Richard L. Emch,

Jr.

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

219, Oyster Creek Nuclear Generating

Station, Ocean County, New Jersey

Date of amendment request:

December 1, 1999.

Description of amendment request:

The proposed amendment would revise

the standard by which GPU Nuclear

tests charcoal used in engineered

safeguards features (ESF) systems to

American Society for Testing and

Materials D3803–1989. These proposed

changes are made in accordance with

Generic Letter (GL) 99–02.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed TS change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change is in accordance with

NRC guidance in GL 99–02 which states that

new testing protocol is more accurate and

demanding than older tests. The acceptance

criteria for charcoal efficiency has been made

more stringent and there is no change to an

operating parameter of any system,

component or structure. Therefore, the

probability of occurrence of the

consequences of an accident previously

evaluated in the SAR [Safety Analysis

Report] will not increase as a result of this

change.

2. The proposed TS change does not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed change revises the testing

standard for activated charcoal efficiency to

a more conservative methodology while

increasing the acceptance criteria through the

application of a safety factor. There is no

change to an operating parameter of any

system, component, or structure. Therefore,

the proposed activity does not create the

possibility for an accident or malfunction of

a different type than any previously

identified in the SAR.

3. The proposed TS change does not

involve a significant reduction in a margin of

safety.

The proposed change does not involve a

reduction in the margin of safety. The change

is primarily administrative, adheres to NRC

guidance, and is more conservative than the

previously employed standard. The change

does not modify an operating parameter of

any system, or component structure.

Therefore, there is no 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: Ernest L. Blake,

Jr., Esquire, Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Acting Section Chief: M.

Gamberoni.

Maine Yankee Atomic Power Company,

Docket No. 50–309, Maine Yankee

Atomic Power Station, Lincoln County,

Maine

Date of amendment request: January

13, 2000.

Description of amendment request:

The amendment would add a license

condition that requires Maine Yankee

Atomic Power Company (MYAPC) to

implement and maintain in effect all

provisions of the License Termination

Plan (LTP). MYAPC submitted the LTP

in accordance with 10 CFR 50.82(a)(9)

to demonstrate that the remainder of

decommissioning activities will be

performed in accordance with Title 10

of the Code of Federal Regulations, will

not be inimical to the common defense

or security or to the health and safety of

the public, and will not have a

significant effect on the quality of the

environment.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The requested license amendment does not

authorize any plant activities beyond that

allowed by 10 CFR Chapter I or beyond that

considered in the DSAR [Defueled Safety

Analysis Report]. The bounding accident

described in the DSAR for potential airborne

activity is the postulated resin cask drop

accident in the Low Level Radioactive Waste

Storage Building. This accident is expected

to contain more potential airborne activity

than can be released from other

decommissioning events. The radionuclide

distribution assumed for the spent resin cask

has more transuranics (the major dose

contributor) than the distribution in the

components involved in other

decommissioning accidents. The accidents

considered in the DSAR include: 1)

Explosion of Liquid Petroleum Gas (LPG)

Leaked from a Front End Loader or Forklift,

2) Explosion of Oxyacetylene During

Segmenting of the Reactor Vessel Shelf, 3)

Release of Radioactivity from the RCS

Decontamination Ion Exchange Resins, 4)

Gross Leak During In-Situ Decontamination,

5) Segmentation of RCS Piping with

Unremoved Contamination, 6) Fire Involving

Contaminated Clothing or Combustible

Waste, 7) Loss of Local Airborne

Contamination Control During Blasting or

Jackhammer Operations, 8) Temporary Loss

of Services, 9) Dropping of Contaminated

Concrete Rubble, 10) Natural Phenomena and

11) Transportation Accidents. The

probabilities and consequences for these

accidents are estimated in the basis

documentation for DSAR Section 7. No

systems, structures, or components that

could initiate or be required to mitigate the

consequences of an accident are affected by

the proposed change in any way not

previously evaluated in the DSAR. Since

Maine Yankee does not exceed the salient

parameters associated with the plant

referenced in the basis documentation in any

material respects, it is concluded that these

probabilities and consequences are not

increased. Therefore, the proposed change to

the Maine Yankee License does not involve

any increase in the probability or

consequences of any accident previously

evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The requested license amendment does not

authorize any plant activities which could

precipitate or result in any accidents beyond

that considered in the DSAR. The accidents

previously evaluated in the DSAR are

described above. These accidents are

described in the basis documentation for

DSAR Section 7. The proposed change does

not affect plant systems, structures, or

components in any way not previously

evaluated in the DSAR. Since Maine Yankee

does not exceed the salient parameters

associated with the plant referenced in the

basis documentation in any material respects,

it is concluded that these accidents

appropriately bound the kinds of accidents

possible during decommissioning. Therefore,

the proposed change to the Maine Yankee

License would not create the possibility of a

new or different kind of accident from any

accident previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The margin of safety defined in Maine

Yankee’s license basis for the consequences

of decommissioning accidents has been

established as the margin between the

bounding decommissioning accident and the

dose limits associated with the need for

emergency plan offsite protection, namely

the Environmental Protection Agency

Protective Action Guidelines EPA-PAGs. As

described above, the bounding

decommissioning accident is the postulated

resin cask drop accident in the Low Level

31358 Federal Register / Vol. 65, No. 96 / Wednesday, May 17, 2000 / Notices

Radioactive Waste Storage Building. Since

the bounding decommissioning accident is

expected to contain more potential airborne

activity than can be released from other

decommissioning events and since the

radionuclide distribution assumed for the

spent resin cask has more transuranics (the

major dose contributor) than the distribution

in the components involved in other

decommissioning accidents, the margin of

safety associated with the consequences of

decommissioning accidents cannot be

reduced. The margin of safety defined in the

statements of consideration for the final rule

on the Radiological Criteria for License

Termination is described as the margin

between the 100 mrem/yr public dose limit

established in 10 CFR 20.1301 for licensed

operation and the 25 mrem/yr dose limit to

the average member of the critical group at

a site considered acceptable for unrestricted

use. This margin of safety accounts for the

potential effect of multiple sources of

radiation exposure to the critical group.

Since the license termination plan was

designed to comply with the radiological

criteria for license termination for

unrestricted use, the margin of safety cannot

be reduced. Therefore, the proposed changes

to the Maine Yankee License would not

involve a significant reduction in any margin

of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

requested amendment involves no

significant hazards consideration.

Attorney for licensee: Mary Ann

Lynch, Esquire, Maine Yankee Atomic

Power Company, 321 Old Ferry Road,

Wiscasset, Maine 04578.

NRC Section Chief: Michael T.

Masnik.

North Atlantic Energy Service

Corporation, Docket No. 50–443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request:

November 30, 1999, as supplemented on

April 28, 2000.

Description of amendment request:

The licensee proposes to change the

technical specifications (TSs) relating to

the emergency diesel generator fuel

sampling/testing surveillance

requirements (SRs). The changes would

provide a new administrative control to

establish, implement, and maintain a

diesel fuel oil testing program, relocate

fuel oil sampling/testing surveillance

requirements and fuel oil storage tank

cleaning frequency requirement to a

new diesel fuel oil testing program

which will reside in the Seabrook

Station Technical Requirements (SSTR)

Manual. The change will also add

references to the A.C. Sources—

Shutdown surveillance requirement to

perform additional activities while in

modes 5 and 6.

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. The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes do not adversely

affect accident initiators or precursors nor

alter the design assumptions, conditions, and

configuration of the facility or the manner in

which the plant is operated and maintained.

The proposed changes do not alter or prevent

the ability of structures, systems, and

components (SSCs) to perform their intended

function to mitigate the consequences of an

initiating event within the acceptance limits

assumed in the Updated Final Safety

Analysis Report (UFSAR).

The proposed changes do not affect the

source term, containment isolation or

radiological release assumptions used in

evaluating the radiological consequences of

an accident previously evaluated in the

Seabrook Station UFSAR. Further, the

proposed changes do not increase the types

and amounts of radioactive effluent that may

be released offsite, nor significantly increase

individual or cumulative occupational/

public radiation exposures. The proposed

change to SR 4.8.1.2 provides additional

requirements for operation of the facility.

These additional requirements are not

initiators of analyzed events and will not

alter assumptions relative to mitigation of

accident or transient events. The proposed

change does not adversely affect previously

evaluated accidents.

Therefore, the proposed changes do not

involve 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 do not involve

physical alteration of plant SSCs or changes

in parameters governing the manner in which

the plant is operated and maintained in a

state of readiness. The changes do not

introduce a new mode of plant operation.

As discussed in the above narrative, the

proposed change to SR 4.8.1.2 provides

additional requirements for operation of the

facility. These additional requirements are

not initiators of analyzed events and will not

alter assumptions relative to mitigation of

accident or transient events. The proposed

change does not adversely affect previously

evaluated accidents.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Involve a significant reduction in a

margin of safety.

The proposed changes do not involve a

reduction in a margin of safety because they

do not adversely affect assumptions used in

transient or safety analyses. The details

associated with the involved specifications

are not required to be in the TS to provide

adequate protection of the public health and

safety, since the TS still retains the

requirement for compliance with the

applicable standards. The level of safety of

facility operation is unaffected by the

changes since there is no change in the intent

of the TS requirements of ensuring fuel oil

is of the appropriate quality for diesel

generator use.

The proposed change to the A.C. Sources—

Shutdown SR imposes an additional level of

requirements that are more restrictive than

the current TS requirements for operation of

the facility in Modes 5 and 6. The additional

requirements being proposed enhance

assurance that the same fuel oil quality

requirements are met, and visual inspection

activities conducted, whenever a diesel

generator is required to be OPERABLE.

Therefore, the proposed changes do not

involve a significant reduction in a margin of

safety.

Based on this review, it appears that

the three standards of 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

P.O. Box 270, Hartford, CT 06141–0270.

NRC Section Chief: James W. Clifford.

North Atlantic Energy Service

Corporation, Docket No. 50–443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request: April 14,

2000.

Description of amendment request:

The licensee proposes to relocate

Technical Specification (TS) Sections

TS 3/4.9.5, ‘‘Communications’’, TS 3/

4.9.6, ‘‘Refueling Machine’’, and TS 3/

4.9.6, ‘‘Crane Travel—Spent Fuel

Storage Area’’ to the Seabrook Station

Technical Requirements Manual.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes to relocate

Technical Specifications 3/4.9.5, 3/4.9.6 and

3/4.9.7 to the Technical Requirements

Manual (TRM) are administrative in nature

and do not adversely affect accident initiators

or precursors nor alter the design

assumptions, conditions, configuration of the

Federal Register / Vol. 65, No. 96 / Wednesday, May 17, 2000 / Notices 31359

facility or the manner in which it is operated.

The proposed changes do not alter or prevent

the ability o[f] structures, systems, or

components to perform their intended

function to mitigate the consequences of an

initiating event within the acceptance limits

assumed in the Updated Final Safety

Analysis Report [UFSAR].

The subject specifications relocated to the

Technical Requirements Manual will

continue to be administratively controlled.

The TRM is a licensee-controlled document,

which contains certain technical

requirements and is the implementing

manual for the Technical Specification

Improvement Program. Changes to these

requirements are reviewed and approved in

accordance with Seabrook Station Technical

Specification, Section 6.7.1.i, and as outlined

in the TRM.

Therefore, the proposed change does not

involve 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 change does not alter the

design assumptions, conditions, or

configuration of the facility or the manner in

which the plant is operated. There are no

changes to the source term or radiological

release assumptions used in evaluating the

radiological consequences in the Seabrook

Station UFSAR. The proposed change has no

adverse impact on component or system

interactions. The proposed change will not

adversely degrade the ability of systems,

structures and components important to

safety to perform their safety function nor

change the response of any system, structure

or component important to safety as

described in the Seabrook Station Updated

Final Safety Analysis Report (UFSAR). The

proposed changes are administrative in

nature and do not change the level of

programmatic and procedural details of

assuring operation of the facility in a safe

manner. Since there are no changes to the

design assumptions, conditions,

configuration of the facility, or the manner in

which the plant is operated and surveilled,

the proposed change does not create the

possibility of a new or different kind of

accident from any previously analyzed.

3. Involve a significant reduction in a

margin of safety.

There is no adverse impact on equipment

design or operation and there are no changes

being made to the Technical Specification

required safety limits or safety system

settings that would adversely affect plant

safety. The proposed change is

administrative in nature and does not reduce

the level of programmatic or procedural

controls associated with the activities

presently performed via Technical

Specifications 3/4.9.5, 3/4.9.6 and 3/4.9.7.

Future changes to the subject technical

requirements will be reviewed and approved

in accordance with Seabrook Station

Technical Specification, Section 6.7, and as

outlined in the Technical Requirements

Manual. Specifically, all changes to the

Technical Requirements Manual require a 10

CFR 50.59 safety evaluation and will be

reviewed and approved by the Station

Operations Review Committee (SORC) prior

to implementation.

Therefore, relocation of the requirements

contained in Technical Specifications 3/

4.9.5, 3/4.9.6 and 3/4.9.7 to the Technical

Requirements Manual does not involve a

significant reduction in the margin of safety

provided in the existing specifications.

The NRC staff has reviewed the

licensee’s analysis, and based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

P.O. Box 270, Hartford, CT 06141–0270.

NRC Section Chief: James W. Clifford.

Public Service Electric & Gas Company,

Docket Nos. 50–272 and 50–311, Salem

Nuclear Generating Station, Unit Nos. 1

and 2, Salem County, New Jersey

Date of amendment request:

November 24, 1999, as supplemented by

letter dated February 10, 2000.

Description of amendment request:

The amendment will establish charcoal

filter testing requirements in the

technical specifications (TSs) for the

Auxiliary Building Ventilation (ABV)

System, the Control Room Envelope Air

Conditioning System (CREACS), and the

Fuel Handling Building Ventilation

(FHV) System that are consistent with

Generic Letter 99–02, Laboratory

Testing of Nuclear-Grade Activated

Charcoal,’’ dated June 3, 1999.

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:

The operation of the Salem units in

accordance with the proposed changes

will not involve a significant increase in

the probability or consequences of an

accident previously evaluated. The

proposed TS changes do not involve any

physical changes to plant structures,

systems, or components (SSC). The

FHV, CREACS, and ABV systems will

continue to function as designed. The

FHV, CREACS, and ABV systems are

designed to mitigate the consequences

of an accident. The proposed changes

also will not affect the sequence of any

accidents previously analyzed. The

proposed TS surveillance requirement

changes implement testing methods that

demonstrate charcoal filter capability

and establish acceptance criteria, which

ensure that Salem’s design basis

assumptions continue to be met. The

proposed surveillance requirement

acceptance criteria ensure that the FHV,

CREACS, and ABV safety functions will

be accomplished. Therefore, the

proposed TS changes would not result

in a significant increase of the

consequences of an accident previously

evaluated, nor do they involve an

increase in the probability of an

accident previously evaluated.

The operation of the Salem units in

accordance with the proposed changes

does not create the possibility of a new

or different kind of accident from any

accident previously evaluated. The

proposed TS changes do not involve any

physical changes to the design of any

plant SSC. The design and operation of

the FHV, CREACS, and ABV systems are

not changed from those currently

described in Salem’s licensing basis.

The FHV, CREACS, and ABV systems

will continue to function as designed to

mitigate the consequences of an

accident. Implementing the proposed

charcoal filter testing methods and

acceptance criteria does not change the

operation of the FHV, CREACS, and

ABV systems that would create a

different type of accident previously

evaluated. In addition, the proposed TS

changes do not alter the conclusions

described in Salem’s licensing basis

regarding the safety-related functions of

these systems. Therefore, the proposed

TS changes do not create the possibility

of a new or different kind of accident

from any previously evaluated.

The operation of the Salem units in

accordance with the proposed license

amendment will not be changed nor

result in a significant reduction to

margins of safety. The licensee is not

proposing any modifications to FHV,

CREACS, and ABV systems design or

operation, and there are no changes

being made to the TS-required safety

limits or safety system settings that

would adversely affect plant safety. The

proposed changes modify the TSs to

reference appropriate test parameters for

performing laboratory testing of nucleargrade charcoal in engineered safety

feature filtration systems in accordance

with ASTM D3803–1989. The

imposition of the more conservative

charcoal filter testing requirements

associated with ASTM D3803–1989 will

not involve a significant reduction in a

margin of safety.

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.

31360 Federal Register / Vol. 65, No. 96 / Wednesday, May 17, 2000 / Notices

Attorney for licensee: Jeffrie J. Keenan,

Esquire, Nuclear Business Unit—N21,

P.O. Box 236, Hancocks Bridge, NJ

08038.

NRC Section Chief: James W. Clifford.

Southern California Edison Company, et

al., Docket Nos. 50–361 and 50 362, San

Onofre Nuclear Generating Station,

Units 2 and 3, San Diego County,

California

Date of amendment requests: April

20, 2000 (PCN–503).

Description of amendment requests:

The amendment application proposes to

revise the San Onofre Nuclear

Generating Station, Units 2 and 3,

Technical Specification (TS) 5.5.2.5,

‘‘Reactor Coolant Pump Flywheel

Inspection Program.’’ The proposed

change would revise the required

volumetric examination frequency of

the upper flywheel on each of the

primary reactor coolant pump motors

from a 3-year to a 10-year cycle.

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?

Response: No.

Missile generation from a Reactor Coolant

Pump (RCP) flywheel could damage the

reactor coolant system, the containment, or

other equipment or systems important to

safety. The fracture mechanics analysis

performed to support the change shows that

a preexisting flaw of an initial size at the

detection threshold level will not grow to a

flaw size necessary to create flywheel

missiles within the life of the plant. The

fracture mechanics analysis conservatively

assumes minimum material toughness

properties, maximum flywheel speed,

location of flaw in the highest stress region

of the flywheel, and a number of start/stop

cycles eight times greater than the design

basis. Therefore, an existing flaw in the

flywheel will not grow to a size that exceeds

the allowable flaw size for either normal

operating or accident conditions over the

plant life. On this basis, the extension of the

3-year interval inspection to a 10-year

interval will not involve a significant

increase in the probability of an accident

previously considered. The proposed

changes do not increase the amount of

radioactive material available for release or

modify any systems used for preventing or

mitigating such releases during accident

conditions. Therefore, these changes do not

involve a significant increase in the

consequences of any accident previously

evaluated.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated?

Response: No.

The proposed changes will not change the

design configuration, or method of operation

of the plant. Therefore, the proposed changes

will not create the possibility of a new or

different kind of accident from any

previously evaluated.

3. Involve a significant reduction in a

margin of safety?

Response: No.

Significant conservatisms have been used

in the calculation of allowable flaw size

(critical flaw size) and flaw growth for each

RCP flywheel design. These include

minimum fracture toughness properties, code

reference crack growth rate curves, maximum

flywheel accident speed, postulated flaw

location at the highest stress region of the

flywheel, and a number of start/stop cycles

that is eight times the number expected in a

plant life. The final flaw size has been

determined to remain smaller than the

allowable flaw size for the flywheel under

the relevant design conditions, including

postulated accident conditions. Therefore,

the extension of the 3-year interval

inspection to a 10-year interval will not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment requests involve no

significant hazards consideration.

Attorney for licensee: Douglas K.

Porter, Esquire, Southern California

Edison Company, 2244 Walnut Grove

Avenue, Rosemead, California 91770.

NRC Section Chief: Stephen Dembek.

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of application for amendments:

February 4, 2000 (TS 99–14).

Brief description of amendments: The

proposed amendments would change

the Sequoyah Nuclear Plant (SQN)

Technical Specification Limiting

Conditions for Operations for the reactor

coolant system cold leg accumulators

(CLAs). The upper CLA water limit and

required pressure range would both be

decreased to more appropriately

account for instrument uncertainties

and instrument line tap locations.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a),

Tennessee Valley Authority (TVA), the

licensee, has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

A. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The capability of the accumulators to

perform their safety function is not affected

by this change. All components and system

functional requirements remain the same.

There are no new sequences of events which

would increase the probability of an accident

analyzed in the Final Safety Analysis Report

(FSAR). Therefore, the proposed activity does

not increase the probability of an accident

previously evaluated in the FSAR. The fuel

cladding peak temperature established by the

ECCS [Emergency Core Cooling System]

evaluation model remains below 2200

degrees Fahrenheit for a loss-of-coolant

accident (LOCA). As such, the assumptions

on fuel failure and isotope release post-LOCA

do not change from the information

presented in the FSAR.

B. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The primary function of the CLAs is in the

event of a large break LOCA to support

accident mitigation. CLAs are not a

contributor to events that could generate

accidents. The CLA system volume capability

bounds this change in operational limits and

the system is not physically changing.

Therefore, the proposed activity does not

create a possibility for an accident of a

different type than any evaluated previously.

C. The proposed amendment does not

involve a significant reduction in a margin of

safety.

The safety function provided by the CLAs

is to inject core cooling water into the reactor

coolant system when system pressure

decreases below a predetermined value

during a LOCA. The timing (function of

pressure) and amount (function of volume) of

cooling water is modeled in the ECCS

evaluation model. The proposed changes to

the accumulator operational limits have been

evaluated using the Sequoyah plant specific

ECCS model. The evaluation shows an

increase in the peak fuel cladding

temperature from 2162 degrees Fahrenheit to

2185 degrees Fahrenheit. The results confirm

that existing LOCA safety analysis

acceptance criteria (established by 10 CFR

50.46) continue to be met for the revised

accumulator limits. The safety analysis

acceptance criteria continues to be met with

the revised limits. The 23 degree increase in

the peak fuel cladding temperature

associated with accumulator operation is not

a significant reduction in the margin of

safety.

The NRC has reviewed the licensee’s

analysis and, based on this review, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 10H,

Knoxville, Tennessee 37902.

NRC Section Chief: Richard P.

Correia.

Federal Register / Vol. 65, No. 96 / Wednesday, May 17, 2000 / Notices 31361

Tennessee Valley Authority, Docket No.

50–390 Watts Bar Nuclear Plant, Unit 1,

Rhea County, Tennessee

Date of amendment request: March 6,

2000 (TS 99–09).

Description of amendment request:

The proposed amendment would revise

the Watts Bar Nuclear Plant (WBN) Unit

1 Technical Specifications (TS) and

associated TS Bases for Limiting

Condition for Operation (LCO) 3.9.4

Containment Penetrations. The revision

would permit both doors of the

containment personnel airlocks to be

open during refueling operations to

facilitate personnel and equipment

access to containment. It would also

allow containment penetration flow

paths to be open under administrative

controls to facilitate maintenance

activities during refueling operations.

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 WBN Technical

Specification LCO 3.9.4, Refueling

Operations—Containment Penetrations,

would allow both doors of the containment

personnel airlocks and certain containment

penetration flow paths to be open during core

alterations and movement of irradiated fuel

within containment under specific

administrative controls. The proposed

change is consistent with NRC approved TS

travelers TSTF–68, R2 and TSTF–312, R1,

and proposes controls similar to the

administrative controls currently allowed by

WBN TS (LCO 3.6.3) for containment

penetrations during more restrictive, higher

operational modes. The administrative

controls will ensure appropriate personnel

are aware of the open personnel airlocks and

penetration flow paths and ensure designated

individual(s) are assigned to promptly close

the airlock doors and penetration flow paths

in the event of a fuel handling accident

(FHA) inside containment. Timely closure of

penetration flow paths and closure of the

airlock doors following containment

evacuation will ensure that the unlikely

transmission of radioactive material from the

reactor building to the auxiliary building is

minimized.

In order to minimize the consequences of

any leakage of radionuclides past these open

penetrations during the period of time before

their closure, additional procedural controls

will be provided to ensure the integrity of the

WBN auxiliary building secondary

containment enclosure (ABSCE) boundary

and proper auxiliary building gas treatment

system (ABGTS) operation. These controls

will ensure that in the event of a fuel

handling accident (FHA) inside containment,

the following will be promptly

accomplished: shutdown and isolation of the

reactor building purge air ventilation system,

auxiliary building isolation, and initiation of

ABGTS. Therefore, through the use of these

controls for the proposed license

amendment, the offsite dose consequences of

a FHA inside containment with open airlock

doors and/or open penetration flow paths

remain well within the 10 CFR 100 limits

and within the limits of 10 CFR 50,

Appendix A, General Design Criteria 19 for

control room operator dose.

[The licensee’s application also states that

‘‘The results for the fuel handling analysis

inside containment with open airlock doors

and/or open penetration flow paths are

bounded by the current analysis.’’]

The containment personnel airlock doors

and containment penetration flow paths are

not initiators to any previously evaluated

accident for WBN. In addition, the position

of the airlock doors and penetration flow

paths during refueling operations has no

affect on the probability of the occurrence of

any accident previously evaluated. The

proposed revision does not alter any plant

equipment or operating practices in such a

manner that the probability of an accident is

increased. Since the probability of a accident

is not affected by the positions of the

containment personnel airlock doors, and

because the doses remain within acceptable

limits, the proposed change 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 open containment personnel airlock

doors and containment penetration flow

paths are not accident initiators and do not

represent a significant change in the

configuration of the plant. The proposed

allowance to open the containment personnel

airlock doors and penetrations during

refueling operations will not adversely affect

plant safety functions or equipment operating

practices such that a new or different

accident could be created. Therefore, since

plant safety functions are not adversely

affected and the isolation status of

containment personnel airlock doors and

penetration flow paths do not contribute to

the initiation of postulated accidents, the

proposed revision will not create 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.

WBN Technical Specification LCO 3.9.4

closure requirements for containment

penetrations ensure that the consequences of

a postulated FHA inside containment during

core alterations or fuel handling activities

remain within acceptable limits. The LCO

establishes containment closure

requirements, which limit the potential

escape paths for fission products by ensuring

that there is at least one integral barrier to the

release of radioactive material. The proposed

change to allow the containment personnel

airlock doors and containment penetration

flow paths to be open during refueling

operations under administrative controls

does not significantly affect the expected

dose consequences of a FHA because of the

absence of containment pressurization during

refueling. Without this motive force, the

potential for additional offsite dose

consequence is unlikely. The proposed

administrative controls provide assurance

that prompt closure of the airlock doors and

penetration flow paths will be accomplished

in the event of a FHA inside containment

thus minimizing the transmission of

radioactive material from the reactor building

to the auxiliary building. Under the proposed

TS change, the provisions to ensure

shutdown and isolation of the reactor

building purge air ventilation system,

auxiliary building isolation, and initiation of

ABGTS and to promptly isolate open

penetration flow paths and close the airlock

doors following containment evacuation,

provide assurance that the offsite dose

consequences of a FHA inside containment

will remain well within the 10 CFR 100

limits and within the limits of 10 CFR 50,

Appendix A, General Design Criteria 19 for

control room operator dose. Therefore, the

proposed change to the WBN Technical

Specifications does not involve a significant

reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 10H,

Knoxville, Tennessee 37902.

NRC Section Chief: Richard P.

Correia.

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.

31362 Federal Register / Vol. 65, No. 96 / Wednesday, May 17, 2000 / Notices

Florida Power Corporation, et al.,

Docket No. 50–302, Crystal River Unit

No. 3 Nuclear Generating Plant, Citrus

County, Florida

Date of application for amendment:

September 16, 1999.

Brief description of amendment: The

proposed amendment would increase

the licensed capacity for spent fuel

assembly storage in the Spent Fuel Pool

and revise the configuration for storage

of fresh fuel.

Date of publication of individual

notice in the Federal Register: December

8, 1999 (64 FR 68702).

Expiration date of individual notice:

January 7, 2000.

PECO Energy Company, Public Service

Electric and Gas Company, 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 amendment request: February

29, 2000, as supplemented on March 31,

2000.

Brief description of amendment

request: The amendments would add a

note to the completion time of

Condition A for Technical Specification

3.7.2, ‘‘Emergency Service Water (ESW)

System and Normal Heat Sink.’’ This

note would provide a one-time

extension to the completion time for one

ESW subsystem inoperable from 7 to 14

days. This note would allow the

replacement of one ESW pump

currently scheduled to occur in May

2000 and will expire on May 31, 2000.

Date of publication of individual

notice in Federal Register: March 9, 2000

(65 FR 12589).

Expiration date of individual notice:

April 10, 2000.

PP&L, Inc., Docket Nos. 50–387 and 50–

388, Susquehanna Steam Electric

Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of amendment request: March

14, 2000, as supplemented March 27,

2000. 

Brief description of amendment

request: The proposed amendment

would amend the licenses to change the

required implementation date for

previously issued license Amendment

No. 184 to Facility Operating License

NPF–14 and Amendment No. 158 to

Facility Operating License NPF–22. The

proposed amendment would not alter

any of the requirements of the SSES

Unit 1 and 2 Technical Specifications

(TSs).

Date of publication of individual

notice in Federal Register: April 27, 2000

(65 FR 24718).

Expiration date of individual notice:

May 30, 2000.

PP&L, Inc., Docket No. 50–388,

Susquehanna Steam Electric Station,

Unit 2, Luzerne County, Pennsylvania

Date of amendment request: April 10,

2000.

Brief description of amendment

request: Permits deferral of testing of

primary containment penetration flange

o-rings on spectacle flanges 2S299A and

2S299B until the Unit 2 10th refueling

outage, scheduled for spring 2001 or a

prior Unit 2 outage requiring entry into

Mode 4.

Date of publication of individual

notice in Federal Register: April 21, 2000

(65 FR 21487).

Expiration date of individual notice:

May 22, 2000.

Notice of Issuance of Amendments to

Facility Operating Licenses

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application

complies with the standards and

requirements of the Atomic Energy Act

of 1954, as amended (the Act), and the

Commission’s rules and regulations.

The Commission has made appropriate

findings as required by the Act and the

Commission’s rules and regulations in

10 CFR Chapter I, which are set forth in

the license amendment.

Notice of Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for A Hearing in

connection with these actions was

published in the Federal Register as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see: (1) The applications for

amendment, (2) the amendment, and (3)

the Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, the Gelman Building, 2120 L

Street, NW., Washington, DC, and

electronically from the ADAMS Public

Library component on the NRC Web

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

Reading Room).

AmerGen Energy Company, LLC, Docket

No. 50–461, Clinton Power Station, Unit

1, DeWitt County, Illinois

Date of application for amendment:

October 25, 1999 (U–603281).

Brief description of amendment: The

amendment revised the Technical

Specification definitions for channel

calibrations, channel functional tests,

and logic system functional tests.

Date of issuance: April 25, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days.

Amendment No.: 128.

Facility Operating License No. NPF–

62: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: January 12, 2000 (65 FR 1920).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 25, 2000.

No significant hazards consideration

comments received: No.

Carolina Power & Light Company, et al.,

Docket No. 50–400, Shearon Harris

Nuclear Power Plant, Unit 1, Wake and

Chatham Counties, North Carolina

Date of application for amendment:

November 19, 1999, as supplemented on

March 16, 2000.

Brief description of amendment: This

amendment revises the Technical

Specifications (TS) to incorporate the

American Society for Testing and

Materials (ASTM) D3803–1989,

Standard Test Method for NuclearGrade Activated Carbon,’’ in accordance

with NRC Generic Letter (GL) 99–02,

‘‘Laboratory Testing Of Nuclear-Grade

Activated Charcoal,’’ dated June 3, 1999.

Specifically, TS 4.7.6 has been revised

for the Control Room Emergency

Filtration System, TS 4.7.7 has been

revised for the Reactor Auxiliary

Building Emergency Exhaust System,

and TS 4.9.12 has been revised for the

Fuel Handling Building Emergency

Exhaust System.

Date of issuance: May 2, 2000.

Effective date: May 2, 2000.

Amendment No. 98.

Facility Operating License No. NPF–

63. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: December 15, 1999 (64 FR

70081).

The March 16, 2000, submittal

contained clarifying information only,

Federal Register / Vol. 65, No. 96 / Wednesday, May 17, 2000 / Notices 31363

and did not change the initial no

significant hazards consideration

determination. The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

May 2, 2000. No significant hazards

consideration comments received: Yes.

One comment was received, and is

addressed in the above-referenced

Safety Evaluation.

Consolidated Edison Company of New

York, Docket No. 50–247, Indian Point

Nuclear Generating Unit No. 2,

Westchester County, New York

Date of application for amendment:

March 17, 2000.

Brief description of amendment: The

amendment revises Technical

Specifications (TSs) associated with

probes used in steam generator tube

inspections, specifically TS Section

4.13.A.3.f. The proposed change would

provide more flexibility in the type of

probe used and would reflect current

technological advances in inspection

equipment, while still maintaining the

current 610-mil diameter probe

restriction. ]

Date of issuance: April 28, 2000.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 209.

Facility Operating License No. DPR–

26: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 27, 2000 (65 FR 16230).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 28, 2000.

No significant hazards consideration

comments received: No.

Entergy Operations, Inc., Docket Nos.

50–313 and 50–368, Arkansas Nuclear

One, Units 1 and 2, Pope County,

Arkansas

Date of amendment request:

November 16, 1999.

Brief description of amendments: The

proposed changes to the Arkansas

Nuclear One, Units 1 and 2, Technical

Specifications (TSs) and associated

Bases provided a 30-day allowed outage

time (AOT) for startup transformer No.

2, which is an offsite power source

shared by both units. This 30-day AOT

will be used infrequently for the

purpose of performing preventative

maintenance to increase the reliability

of the transformer. In addition, changes

have been made to the requirements

associated with demonstrating the

operability of the emergency diesel

generators (EDGs), in the event a

required power source is inoperable, to

increase the reliability of the EDGs.

Date of issuance: April 28, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment Nos.: 206 and 215.

Facility Operating License Nos. DPR–

51 and NPF–6: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: January 26, 2000 (65 FR 4271).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated April 28, 2000.

No significant hazards consideration

comments received: No.

Entergy Operations, Inc., Docket No. 50–

368, Arkansas Nuclear One, Unit No. 2,

Pope County, Arkansas

Date of application for amendment:

February 24, 2000.

Brief description of amendment: The

amendment revised Technical

Specification (TS) 4.4.11 on reactor

coolant system vent flow verification,

TS 4.6.1.1.a on containment penetration

closure verification (non-automatic),

and TS 4.6.3.1.2 on containment

isolation valve actuation verification.

The changes eliminated unnecessary

mode restrictions on these surveillance

requirements.

Date of issuance: April 26, 2000.

Effective date: As of the date of

issuance to be implemented within 30

days from the date of issuance.

Amendment No.: 214.

Facility Operating License No. NPF–6:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 22, 2000 (65 FR 15379).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 26, 2000.

No significant hazards consideration

comments received: No.

Entergy Operations, Inc., Docket No. 50–

382, Waterford Steam Electric Station,

Unit 3, St. Charles Parish, Louisiana

Date of amendment request: October

6, 1998, as supplemented by letter dated

March 3, 2000, Moderator Temperature

Coefficient test near the end of each

cycle.

Brief description of amendment: The

proposed change modifies the

requirement to perform a Moderator

Temperature Coefficient test near the

end of each cycle.

Date of issuance: April 21, 2000.

Effective date: As of the date of

issuance and shall be implemented 60

days from the date of issuance.

Amendment No.: 159.

Facility Operating License No. NPF–

38: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: August 25, 1999 (64 FR 46435).

The March 3, 2000, letter did not change

the scope of the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 21, 2000.

No significant hazards consideration

comments received: No.

Entergy Operations, Inc., Docket No. 50–

382, Waterford Steam Electric Station,

Unit 3, St. Charles Parish, Louisiana

Date of amendment request: July 15,

1999, as supplemented by letter dated

January 6, 2000.

Brief description of amendment: The

proposed change modifies plant

technical specifications to extend the

Reactor Coolant System Pressure

Temperature Curve Limit to 16 Effective

Full Power Years.

Date of issuance: April 24, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 60 days from the date of

issuance.

Amendment No.: 160.

Facility Operating License No. NPF–

38: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: January 26, 2000 (65 FR 4276).

The January 6, 2000, letter reduced

EFPY from 20 years, requested in the

July 15, 1999, letter, to 16 years. This

change is bounded by, and did not

change the scope of, the initial proposed

no significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 24, 2000.

No significant hazards consideration

comments received: No.

FirstEnergy Nuclear Operating

Company, et al., Docket Nos. 50–334

and 50–412, Beaver Valley Power

Station, Unit Nos. 1 and 2,

Shippingport, Pennsylvania

Date of application for amendments:

November 23, 1999, as supplemented

February 22, 2000.

Brief description of amendments: The

amendments make the following

changes to the Beaver Valley Power

Station, Unit Nos. 1 and 2 (BVPS–1 and

BVPS–2) Technical Specifications (TSs):

(1) For BVPS–1, surveillance

requirement (SR) 4.8.1.1.2.b.3.b is

revised to reflect a narrower required

diesel generator (DG) frequency band;

an associated footnote is deleted;

associated Bases are revised to reflect

these TS changes. (2) For BVPS–2, SR

4.8.1.1.2.f is revised to clarify that the

31364 Federal Register / Vol. 65, No. 96 / Wednesday, May 17, 2000 / Notices

DGs are only required to achieve a

minimum frequency and voltage within

the first 10 seconds of the related test,

and that the stated voltage and

frequency bands are requirements for

steady state operation of the DGs; a

footnote is also added to this SR. (3)

Page formats are revised as needed to

permit the addition or deletion of text.

Date of issuance: April 25, 2000.

Effective date: As of date of issuance

and shall be implemented within 60

days.

Amendment Nos.: 230 and 109.

Facility Operating License Nos. DPR–

66 and NPF–73: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: March 8, 2000 (65 FR 12292).

The February 22, 2000, letter provided

supplemental information and did not

change the initial proposed no

significant hazards consideration

determination or expand the

amendments beyond the scope of the

initial notice.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated April 25, 2000.

No significant hazards consideration

comments received: No.

FirstEnergy Nuclear Operating

Company, et al., Docket No. 50–412,

Beaver Valley Power Station, Unit 2,

Shippingport, Pennsylvania

Date of application for amendment:

September 22, 1999, as supplemented

April 27, 2000.

Brief description of amendment: The

amendment allowed a one-time only

extension to the surveillance interval of

the Technical Specification Surveillance

4.7.12.d for functional testing of

snubbers. The extension is limited to

the first re-entry into MODE 6 following

the defueled condition during the 8th

refueling outage or November 30, 2000,

whichever occurs sooner.

Date of issuance: May 3, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 60 days.

Amendment No.: 110.

Facility Operating License No. NPF–

73: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: November 17, 1999, (64 FR

62711).

The April 27, 2000, letter did not

change the initial proposed no

significant hazards consideration

determination or expand the

amendment beyond the scope of the

initial notice.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 3, 2000.

No significant hazards consideration

comments received: No.

FirstEnergy Nuclear Operating

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

Ottawa County, Ohio

Date of application for amendment:

November 2, 1999.

Brief description of amendment: This

amendment revised the Technical

Specifications (TSs) to modify 1) TS

Table 3.3–4, ‘‘Safety Features Actuation

System Instrumentation Trip

Setpoints,’’ to remove the ‘‘Trip

Setpoint’’ values for Instrument String

Functional Unit ‘‘f’’, Borated Water

Storage Tank (BWST) Level, 2) the

‘‘Allowable Values’’ entry for this same

Functional Unit, consistent with

updated calculations using current

setpoint methodology, 3) TS 3/4.3.2.1,

‘‘Safety Features Actuation System

Instrumentation,’’ and Bases to reflect

the removal of ‘‘Trip Setpoints’’

described above, and 4) TS 3/4.5.4,

‘‘Emergency Core Cooling Systems—

Borated Water Storage Tank,’’ and Bases

to increase the minimum volume of

water in the BWST.

Date of issuance: May 4, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 120 days.

Amendment No.: 241.

Facility Operating License No. NPF–3:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: December 15, 1999 (64 FR

70087).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 4, 2000.

No significant hazards consideration

comments received: No.

Illinois Power Company, Docket No. 50–

461, Clinton Power Station, Unit 1,

DeWitt County, Illinois

Date of application for amendment:

October 23, 1998, as supplemented

February 22 and June 24, 1999, and

March 31, 2000.

Brief description of amendment: The

amendment would allow

implementation of a feedwater leakage

control system to address leakage

through the primary containment

feedwater penetration valve.

Date of issuance: April 25, 2000.

Effective date: April 25, 2000.

Amendment No.: 127.

Facility Operating License No. NPF–

62: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: November 18, 1998 (63 FR

64118).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 25, 2000.

No significant hazards consideration

comments received: No.

Indiana Michigan Power Company,

Docket Nos. 50–315 and 50–316, Donald

C. Cook Nuclear Plant, Units 1 and 2,

Berrien County, Michigan

Date of application for amendments:

February 18, 2000, as supplemented

March 31, 2000.

Brief description of amendments: The

proposed license amendments would

approve a change to the facility

involving an unreviewed safety question

discovered by the licensee during a 10

CFR 50.59 evaluation of modifications

to the auxiliary feedwater (AFW) pump

rooms to protect the equipment in the

rooms from the environmental effects of

a postulated high-energy line break.

This will be accomplished by sealing

the AFW pump rooms to ensure that the

rooms do not communicate with the

turbine buildings or each other.

Date of issuance: April 25, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days.

Amendment Nos.: 244.

Facility Operating License Nos. DPR–

58 and DPR–74: Amendments revised

the Operating License.

Date of initial notice in Federal

Register: February 25, 2000 (65 FR 10116).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated April 25, 2000.

No significant hazards consideration

comments received: No.

North Atlantic Energy Service

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

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request:

December 13, 1999.

Description of amendment request:

The amendment changes the license to

delete expired license conditions and to

make editorial and administrative

changes to correct or clarify the license.

Date of issuance: April 27, 2000.

Effective date: As of its date of

issuance, and shall be implemented

within 60 days.

Amendment No.: 68.

Facility Operating License No. NPF–

86: Amendment revised the License.

Date of initial notice in Federal

Register: February 9, 2000 (65 FR 6408).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 27, 2000.

No significant hazards consideration

comment received: No.

Federal Register / Vol. 65, No. 96 / Wednesday, May 17, 2000 / Notices 31365

North Atlantic Energy Service

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

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request: February

18, 2000.

Description of amendment request:

The amendment revises Technical

Specifications (TSs) Surveillance

Requirements 4.0.5.a, 4.0.5.b, 4.0.5.e,

and 4.4.6.2.2.e. These changes are

required to ensure consistency between

the TSs and the second 10-year

inservice test program by approval to

use the 1995 Edition and 1996 Addenda

of the American Society of Mechanical

Engineers (ASME) Code for Operation

and Maintenance of Nuclear Power

Plants (OM Code). The revision to TSs

Surveillance Requirement 4.0.5.a also

incorporates semi-quarterly and

biennial intervals to the list of required

frequencies for performing inservice test

and inspection activities.

Date of issuance: May 8, 2000.

Effective date: As of its date of

issuance, and shall be implemented by

August 18, 2000.

Amendment No.: 69.

Facility Operating License No. NPF–

86: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: April 5, 2000 (65 FR 17917).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 8, 2000.

No significant hazards consideration

comments received: No.

Northeast Nuclear Energy Company, et

al., Docket No. 50–336, Millstone

Nuclear Power Station, Unit No. 2, New

London County, Connecticut

Date of application for amendment:

December 14, 1999, as supplemented

February 11, March 30, and April 26,

2000.

Brief description of amendment: This

amendment will revise Technical

Specifications (TSs) Sections: 3.3.2.1,

‘‘Instrumentation—Engineered Safety

Feature Actuation System

Instrumentation;’’ 3.3.3.1,

‘‘Instrumentation—Monitoring

Instrumentation—Radiation

Monitoring;’’ 3.7.6.1, ‘‘Plant Systems—

Control Room Emergency Ventilation

System;’’ 3.9.3.1, ‘‘Refueling

Operations—Decay Time;’’ 3.9.4,

‘‘Refueling Operations—Containment

Penetrations;’’ 3.9.9, ‘‘Refueling

Operations—Containment Radiation

Monitoring;’’ 3.9.10 ‘‘Refueling

Operations—Containment Purge Valve

Isolation System;’’ 3.9.13, ‘‘Refueling

Operations—Storage Pool Radiation

Monitoring;’’ 3.9.14, ‘‘Refueling

Operations—Storage Pool Area

Ventilation System—Fuel Movement;’’

3.9.15, ‘‘Refueling Operations—Storage

Pool Area Ventilation System—Fuel

Storage;’’ 3.9.16.1, ‘‘Refueling

Operations—Shielded Cask;’’ 3.9.16.2,

‘‘Refueling Operations—Shielded Cask;’’

3.9.17, ‘‘Refueling Operations—

Movement of Fuel in Spent Fuel Pool;’’

and 3.9.19.2, ‘‘Refueling Operations—

Spent Fuel Pool—Storage Pattern’’; and

add new TS 3.3.4, ‘‘Containment Purge

Valve Isolation Signal.’’ The requested

changes would make the TSs and the

Final Safety Analysis Report (FSAR)

consistent with the new analyses of the

fuel handling and cask drop accidents.

The Index Pages and the Bases for these

TSs will be modified to reflect these

proposed changes.

Date of issuance: April 28, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment No.: 245.

Facility Operating License No. DPR–

65: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 17, 2000 (65 FR 14632).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 28, 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 15, 1999, as supplemented July 20,

September 3, and November 29, 1999,

and January 18, 2000.

Brief description of amendment: The

amendment modifies the license to

change the number of owners from 14

to 13 and to remove Montaup Electric

Company as an owner as a result of the

transfer of its interest in Millstone

Nuclear Power Station, Unit No. 3 to

New England Power Company, an

existing owner.

Date of issuance: May 1, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment No.: 180.

Facility Operating License No. NPF–

49: Amendment revised the License.

Date of initial notice in Federal

Register: January 19, 2000 (65 FR 2990).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated February 24,

2000, issued with the February 24, 2000,

Order approving the transfer as noticed

in the Federal Register on March 1,

2000 (65 FR 11091).

No significant hazards consideration

comments received: No.

PECO Energy Company, Public Service

Electric and Gas Company, 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:

February 29, 2000, as supplemented on

March 31, 2000.

Brief description of amendments: The

amendments will add a note to the

completion time of Condition A for

Technical Specification 3.7.2,

‘‘Emergency Service Water (ESW)

System and Normal Heat Sink.’’ This

note will provide a one-time extension

to the completion time for one ESW

subsystem inoperable from 7 to 14 days.

This note will allow the replacement of

one ESW pump currently scheduled to

occur in May 2000 and will expire on

May 31, 2000.

Date of issuance: April 25, 2000.

Effective date: Both units, as of the

date of issuance and shall be

implemented no later than May 31,

2000.

Amendments Nos.: 231 and 236.

Facility Operating License Nos. DPR–

44 and DPR–56: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: March 9, 2000 (65 FR 12589).

The March 31, 2000, letter provided

clarifying information that did not

change the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated April 25, 2000.

No significant hazards consideration

comments received: No.

PP&L, Inc., Docket No. 50–388,

Susquehanna Steam Electric Station,

Unit 2, Luzerne County, Pennsylvania

Date of application for amendment:

April 10, 2000.

Brief description of amendment: The

amendment adds a note to Technical

Specification Surveillance Requirement

3.6.1.1.1 to defer performance of this

test on a one-time basis for spectacle

flanges 2S299A and 2S299B o-rings

until the Unit 2 10th Refueling Outage

(Spring 2001) or a prior Unit 2 outage

requiring entry into Mode 4. The change

allowed Unit 2 operation to continue

until an outage occurs where leak rate

surveillance testing on spectacle flanges

2S299A and 2S299B can be performed.

Date of issuance: May 8, 2000.

31366 Federal Register / Vol. 65, No. 96 / Wednesday, May 17, 2000 / Notices

Effective date: As of date of issuance

and shall be implemented within 30

days.

Amendment No.: 160.

Facility Operating License No. NPF–

22. This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: April 21, 2000 (65 FR 21487).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 8, 2000.

No significant hazards consideration

comments received: No.

Public Service Electric & Gas Company,

Docket No. 50–354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of application for amendment:

March 15, 2000.

Brief description of amendment: This

amendment changes Technical

Specification (TS) Definition 1.7, CORE

ALTERATION. The definition has been

revised to be similar to the definition of

CORE ALTERATION that is

documented in NUREG–1433, Revision

1, ‘‘Standard Technical Specifications,

General Electric Plants, BWR/4.’’

Date of issuance: April 25, 2000.

Effective date: As of the date of

issuance, and shall be implemented

within 3 days.

Amendment No.: 125.

Facility Operating License No. NPF–

57: This amendment revised the TSs.

Date of initial notice in Federal

Register: March 23, 2000 (65 FR 15657).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 25, 2000.

No significant hazards consideration

comments received: No.

Public Service Electric & Gas Company,

Docket No. 50–354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of application for amendment:

September 30, 1999, as supplemented

March 27, 2000.

Brief description of amendment: This

amendment revised the Technical

Specifications (TSs) associated with the

Safety Limit Minimum Critical Power

Ratios in order to support the operation

of Hope Creek Generating Station in the

upcoming Cycle 10 with a mixed core

of General Electric (GE) and Asea Brown

Bovieri/Combustion Engineering (ABB/

CE) fuel. In addition, administrative

changes have been made to the TSs to

reflect the change in fuel vendor from

GE to ABB/CE.

Date of issuance: May 1, 2000.

Effective date: As of the date of

issuance, and shall be implemented

within 60 days after completion of Cycle

9.

Amendment No.: 126.

Facility Operating License No. NPF–

57: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: November 3, 1999 (64 FR 59805).

The March 27, 2000 letter provided

clarifying information that did not

change the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 1, 2000.

No significant hazards consideration

comments received: No.

Vermont Yankee Nuclear Power

Corporation, Docket No. 50–271,

Vermont Yankee Nuclear Power Station,

Vernon, Vermont

Date of application for amendment:

July 20, 1999, as supplemented on

October 25, 1999.

Brief description of amendment: The

amendment revises Technical

Specifications to reflect the

implementation of increased core flow.

Date of Issuance: April 25, 2000.

Effective date: As of the date of

issuance, and shall be implemented

within 60 days.

Amendment No.: 187.

Facility Operating License No. DPR–

28: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 25, 1999 (64 FR 46450).

The October 25, 1999, letter provided

clarifying information that did not

change the initial proposed no

significant hazards consideration

determination or expand the scope of

the original Federal Register notice.

The Commission’s related evaluation

of this amendment is contained in a

Safety Evaluation dated April 25, 2000.

No significant hazards consideration

comments received: No.

Wisconsin Public Service Corporation,

Docket No. 50–305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of application for amendment:

March 2, 2000.

Brief description of amendment: The

amendment increases the minimum

refueling boron concentration to 2200

parts per million (ppm) from 2100 ppm

as specified in the Technical

Specification 3.8.a.5.

Date of issuance: May 1, 2000.

Effective date: Immediately upon its

date of issuance and is to be

implemented within 30 days of the date

of issuance.

Amendment No.: 147.

Facility Operating License No. DPR–

43: Amendment revised the Technical

Specifications.

Date of initial notice in Federal Register:

March 30, 2000 (65 FR 16969).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 1, 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

Federal Register / Vol. 65, No. 96 / Wednesday, May 17, 2000 / Notices 31367

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, the Gelman Building, 2120 L

Street, NW., Washington, DC, and

electronically from the ADAMS Public

Library component on the NRC Web

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

Reading Room).

The Commission is also offering an

opportunity for a hearing with respect to

the issuance of the amendment. By June

16, 2000, the licensee may file a request

for a hearing with respect to issuance of

the amendment to the subject facility

operating license and any person whose

interest may be affected by this

proceeding and who wishes to

participate as a party in the proceeding

must file a written request for a hearing

and a petition for leave to intervene.

Requests for a hearing and a petition for

leave to intervene shall be filed in

accordance with the Commission’s

‘‘Rules of Practice for Domestic

Licensing Proceedings’’ in 10 CFR Part

2. Interested persons should consult a

current copy of 10 CFR 2.714 which is

available at the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington, DC and

electronically from the ADAMS Public

Library component on the NRC Web

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

Reading Room). If a request for a hearing

or petition for leave to intervene is filed

by the above date, the Commission or an

Atomic Safety and Licensing Board,

designated by the Commission or by the

Chairman of the Atomic Safety and

Licensing Board Panel, will rule on the

request and/or petition; and the

Secretary or the designated Atomic

Safety and Licensing Board will issue a

notice of a hearing or an appropriate

order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses. 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, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, by the above date. A

copy of the petition should also be sent

to the Office of the General Counsel,

U.S. Nuclear Regulatory Commission,

Washington, DC 20555–0001, and to the

attorney for the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of the

factors specified in 10 CFR

2.714(a)(1)(i)–(v) and 2.714(d).

31368 Federal Register / Vol. 65, No. 96 / Wednesday, May 17, 2000 / Notices

Duke Energy Corporation, et al., Docket

No. 50–414, Catawba Nuclear Station,

Unit 2, York County, South Carolina

Date of application for amendments:

May 3, 2000.

Brief description of amendments: The

amendment revised the Technical

Specifications (TS) and associated Bases

Section 3.6.9 for the Hydrogen Ignition

System. Specifically, the proposed

amendment modifies Surveillance

Requirements (SRs) 3.6.9.1, 3.6.9.2, and

3.6.9.3 to exclude the two hydrogen

ignitors located beneath the reactor

vessel missile shield from the

applicability of the SRs. These two

ignitors are presently considered to be

inoperable at Unit 2 and cannot be

accessed for replacement with the unit

in its current operating mode (Mode 1).

This change is effective for Unit 2 Cycle

11 only, or until such time that the unit

enters Mode 5 (cold shutdown) such

that the inoperable ignitors can be

accessed for replacement.

Date of issuance: May 5, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment No.: 178.

Facility Operating License No. NPF–

52: Amendment revised the Technical

Specifications and associated Bases.

Public Comments requested as to

proposed no significant hazards

consideration: No.

The Commission’s related evaluation

of the amendment, finding of emergency

circumstances, consultation with the

State of South Carolina, and final no

significant hazards consideration

determination are contained in a Safety

Evaluation dated May 5, 2000.

Attorney for licensee: Ms. Lisa F.

Vaughn , Legal Department (PB05E),

Duke Energy Corporation, 422 South

Church Street, Charlotte, North Carolina

28201–1006.

NRC Section Chief: Richard L. Emch,

Jr.

Dated at Rockville, Maryland, this 10th day

of May 2000.

For the Nuclear Regulatory Commission.

John A. Zwolinski,

Director, Division of Licensing Project

Management, Office of Nuclear Reactor

Regulation.

[FR Doc. 00–12302 Filed 5–16–00; 8:45 am]

BILLING CODE 7590–01–P

POSTAL RATE COMMISSION

[Docket No. A2000–1; Order No. 1292]

Appeal of Post Office Closing

AGENCY: Postal Rate Commission.

ACTION: Notice of Docket No. A2000–1.

SUMMARY: The Commission has

established a docket to consider an

objection to the closing of the Roanoke,

WV post office. It also has issued a

procedural schedule.

DATES: See SUPPLEMENTARY INFORMATION

section for dates.

ADDRESSES: Correspondence should be

addressed to Margaret P. Crenshaw,

Secretary, Postal Rate Commission, 1333

H Street NW., Suite 300, Washington,

DC 20268–0001.

FOR MORE INFORMATION CONTACT:

Stephen L. Sharfman, General Counsel,

(202) 789–6820.

SUPPLEMENTARY INFORMATION: On May

10, 2000, the Postal Rate Commission

(PRC) issued a notice and order (No.

1292) accepting an appeal of the closing

of the Roanoke, West Virginia post

office, ZIP Code 26423. The appeal was

filed by Robert J. Conley, president of

the Lewis County Commission, on

behalf of the Commission and post

office customers, pursuant to 39 U.S.C.

404(b)(5). The appeal has been assigned

Docket No. A2000–1 and a procedural

schedule has been established.

The appeal was filed April 21, 2000.

The categories of issues apparently

raised are the effect on the community

(39 U.S.C. 404(b)(2)(A)) and effect on

postal services (39 U.S.C. 404(b)(2)(C)).

After the Postal Service files the

administrative record and the

Commission reviews it, the Commission

may find that there are more legal issues

than those set forth above. Or, the

Commission may find that the Postal

Service’s determination disposes of one

or more of those issues.

Scheduling matters. The Postal

Reorganization Act requires that the

Commission issue its decision within

120 days from the date this appeal was

filed (39 U.S.C. 404(b)(5)). The

procedural schedule has been

developed to accommodate the delay in

publication of this notice and order. In

the interest of expedition, in light of the

120-day decision schedule, the

Commission may request the Postal

Service to submit memoranda of law on

any appropriate issue. If requested, such

memoranda will be due 14 days from

the issuance of the request and the

Postal Service shall serve a copy of its

memoranda on the petitioners. The

Postal Service may incorporate by

reference in its briefs or motions any

arguments presented in memoranda it

previously filed in this docket. If

necessary, the Commission also may ask

petitioners or the Postal Service for

more information.

Ordering paragraphs. Ordering

paragraph (a) directs the Postal Service

shall file the record in this appeal by

May 25, 2000. Ordering paragraph (b)

directs the Secretary of the Postal Rate

Commission to publish this notice and

order and procedural schedule in the

Federal Register.

Procedural schedule. Key dates in this

docket include: April 21, 2000, filing of

appeal letter; May 10, 2000, issuance of

Commission notice and order on filing

of appeal; June 5, 2000, last day of filing

of petitions to intervene (see 39 CFR

3001.111(b)); June 15, 2000, deadline for

petitioner’s participant statement or

initial brief (see 39 CFR 3001.115(a) and

(b)); July 5, 2000, deadline for Postal

Service’s answering brief (see 39 CFR

3001.115(c)); July 20, 2000, deadline for

petitioner’s reply brief, should

petitioner choose to file one (see 39 CFR

3001.115(d)); July 27, 2000, deadline for

motions by any party requesting oral

argument. The Commission will

schedule oral argument only when it is

a necessary addition to the written

filings (see 39 CFR 3001.116); August

21, 2000, expiration of the

Commission’s 120-day decisional

schedule (see 39 U.S.C. 404(b)(5)).

Dated: May 11, 2000.

Margaret P. Crenshaw,

Secretary.

[FR Doc. 00–12309 Filed 5–16–00; 8:45 am]

BILLING CODE 7710–FW–P

SECURITIES AND EXCHANGE

COMMISSION

Proposed Collection; Comment

Request

Upon written request, copies available from:

Securities and Exchange Commission,

Office of Filings and Information Services,

Washington, DC 20549

Extension:

Form 8–A, OMB Control No. 3235–0056,

SEC File No. 270–54

Form 18–K, OMB Control No. 3235–0120,

SEC File No. 270–108

Notice is hereby given that pursuant

to the Paperwork Reduction Act of 1995

(44 U.S.C. 3501 et seq.) the Securities

and Exchange Commission

(‘‘Commission’’) is soliciting comments

on the collections of information

summarized below. The Commission

plans to submit these existing

collections of information to the Office

of management and Budget for

extension and approval.

Form 8–A (OMB Control No. 3235–

0056, SEC File No. 270–54) is a

registration statement for certain classes

of securities pursuant to Section 12(b)