Federal Register / Vol. 63, No. 116 /Wednesday, June 17, 1998 / Notices 33103

organizational and personnel matters

that relate solely to internal personnel

rules and practices of ACRS, and

information the release of which would

constitute a clearly unwarranted

invasion of personal privacy.

The agenda for the subject meeting

shall be as follows: Tuesday, July 7,

1998—12:15 p.m.—1:30 p.m.

The Subcommittee will discuss

proposed ACRS activities and related

matters. It may also discuss the

qualifications of candidates for

appointment to the ACRS. The purpose

of this meeting is to gather information,

analyze relevant issues and facts, and to

formulate proposed positions and

actions, as appropriate, for deliberation

by the full Committee.

Oral statements may be presented by

members of the public with the

concurrence of the Subcommittee

Chairman; written statements will be

accepted and made available to the

Committee. Electronic recordings will

be permitted only during those portions

of the meeting that are open to the

public, and questions may be asked only

by members of the Subcommittee, its

consultants, and staff. Persons desiring

to make oral statements should notify

the cognizant ACRS staff person named

below five days prior to the meeting, if

possible, so that appropriate

arrangements can be made.

Further information regarding topics

to be discussed, the scheduling of

sessions open to the public, whether the

meeting has been canceled or

rescheduled, the Chairman’s ruling on

requests for the opportunity to present

oral statements, and the time allotted

therefor can be obtained by contacting

the cognizant ACRS staff person, Dr.

John T. Larkins (telephone: 301/415–

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

(EDT). Persons planning to attend this

meeting are urged to contact the above

named individual one or two working

days prior to the meeting to be advised

of any changes in schedule, etc., that

may have occurred.

Dated: June 11, 1998.

Sam Duraiswamy,

Chief, Nuclear Reactors Branch.

[FR Doc. 98–16095 Filed 6–16–98; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Sunshine Act Meeting

DATE: Weeks of June 15, 22, 29, and July

6, 1998.

PLACE: Commissioners’ Conference

Room, 11555 Rockville Pike, Rockville,

Maryland.

STATUS: Public and Closed.

MATTERS TO BE CONSIDERED:

Week of June 15

Wednesday, June 17

10:00 a.m.—Briefing by National

Mining Association on Regulation

of the Uranium Recovery Industry

(PUBLIC MEETING)

11:30 A.M.—Affirmation Session

(PUBLIC MEETING) (If needed)

2:00 p.m.—Meeting with Advisory

Committee on Medical Uses of

Isotopes (ACMUI) and Briefing on

Part 35 QM Rule (PUBLIC

MEETING) (Contact: Larry Camper,

301–415–7231)

Week of June 22—Tentative

Thursday, June 25

9:30 a.m.—Briefing by IG on Results

of NRC Organization Safety Culture

and Climate Survey (PUBLIC

MEETING)

11:30 a.m.—Affirmation Session

(PUBLIC MEETING) (If needed)

2:00 p.m.—Briefing on EEO Program

(PUBLIC MEETING)

Week of June 29—Tentative

Tuesday, June 30

10:00 a.m.—Meeting with

Commonwealth Edison (PUBLIC

MEETING) (Contact: Bob Capra,

301–415–1430)

11:30 a.m.—Affirmation Session

(PUBLIC MEETING) (if needed)

2:00 p.m.—Briefing on Performance

Assessment Progress in HLW, LLW,

and SDMP (PUBLIC MEETING)

Week of July 6—Tentative

Thursday, July 9

11:30 a.m.—Affirmation Session

(PUBLIC MEETING) (if needed)

*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.

* * * * *

ADDITIONAL INFORMATION: By a vote of 4–

0 on June 5, the Commission

determined pursuant to U.S.C. 552b(e)

and 10 CFR Sec. 9.107(a) of the

Commission’s rules that ‘‘Affirmation of

(a) Hydro Resources, Inc. Docket No.

40–8968–ML, Memorandum and Order

(Denying Motion for Stay and Request

for Prior Hearing, Lifting Temporary

Stay Denying Motions to Strike and for

Leave for Reply), LBP–98–5, (b)

Proposed Licenses to Export High

Enriched Uranium (HEU) for Production

of Medical Isotopes at the Canadian

NRU (XSNM3012) and Maple Reactors

(XSNM3013), and (c) Hydro Resources,

Inc. Docket No. 40–8968–ML,

Memorandum and Order (Denial Of

Motion to Disqualify Presiding Officer),

LBP–98–11’’ be held on June 5, and on

less than one week’s notice to the

public.

* * * * *

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.

* * * * *

William M. Hill, Jr.,

SECY Tracking Officer, Office of the

Secretary.

[FR Doc. 98–16175 Filed 6–12–98; 4:44 pm]

BILLING CODE 7590–01–M

NUCLEAR REGULATORY

COMMISSION

Biweekly Notice; Applications and

Amendments to Facility Operating

Licenses Involving no Significant

Hazards Considerations

I. Background

Pursuant to Public Law 97–415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 97–415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from May 22,
33104 Federal Register / Vol. 63, No. 116 /Wednesday, June 17, 1998 / Notices

1998, through June 5, 1998. The last

biweekly notice was published on June

3, 1998 (63 FR 30261).

Notice of Consideration of Issuance of

Amendments to Facility Operating

Licenses, Proposed no Significant

Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a

proposed determination that the

following amendment requests involve

no significant hazards consideration.

Under the Commission’s regulations in

10 CFR 50.92, this means that operation

of the facility in accordance with the

proposed amendment would not (1)

involve a significant increase in the

probability or consequences of an

accident previously evaluated; or (2)

create the possibility of a new or

different kind of accident from any

accident previously evaluated; or (3)

involve a significant reduction in a

margin of safety. The basis for this

proposed determination for each

amendment request is shown below.

The Commission is seeking public

comments on this proposed

determination. Any comments received

within 30 days after the date of

publication of this notice will be

considered in making any final

determination.

Normally, the Commission will not

issue the amendment until the

expiration of the 30-day notice period.

However, should circumstances change

during the notice period such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received before

action is taken. Should the Commission

take this action, it will publish in the

Federal Register a notice of issuance

and provide for opportunity for a

hearing after issuance. The Commission

expects that the need to take this action

will occur very infrequently.

Written comments may be submitted

by mail to the Chief, Rules and

Directives Branch, Division of

Administration 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 July 17, 1998, 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 at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact.

Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,
Federal Register / Vol. 63, No. 116 /Wednesday, June 17, 1998 / Notices 33105

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

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 at the local public

document room for the particular

facility involved.

Commonwealth Edison Company,

Docket Nos. 50–237 and 50–249,

Dresden Nuclear Power Station, Units

2 and 3, Grundy County, Illinois

Docket Nos. 50–254 and 50–265, Quad

Cities Nuclear Power Station, Units 1

and 2, Rock Island County, Illinois

Date of application for amendment

request: May 18, 1998.

Description of amendment request:

Change various technical specification

(TS) values to conservatively reflect

design values. These TS values affect:

(1) 125/250 volts direct current (Vdc)

electrolyte temperature; (2) control rod

drive accumulator pressure; (3) standby

liquid control solution temperature; (4)

ultimate heat sink minimum water

level; (5) shutdown suppression

chamber level (Quad Cities only); and

(6) degraded voltage setpoint (Quad

Cities only).

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:

Does the change involve a significant

increase in the probability or

consequences of an accident previously

evaluated?

The change does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated. The proposed changes to

certain Technical Specification

acceptance values are conservative and

serve to ensure operability of equipment

important to safety. By ensuring

equipment availability, the probability

or consequences of an accident

previously evaluated are not increased.

In addition, the proposed changes have

no impact on any initial condition

assumptions for accident scenarios.

Onsite or offsite dose consequences

resulting from an event previously

evaluated are not affected by this

proposed amendment request.

Accordingly, there is no significant

change in the probability or

consequences of an accident previously

evaluated.

Does the change create the possibility

of a new or different kind of accident

from any accident previously evaluated?

The proposed amendment does not

create the possibility of a new or

different kind of accident from any

accident previously evaluated. The

proposed license amendment provides

changes in certain Technical

Specification values to restore margin

and ensure equipment operability. Each

proposed change is conservative with

respect to current requirements. The

proposed amendment does not involve

any plant physical changes that would

create the possibility of a new or

different kind of accident from any

accident previously evaluated.

Therefore, the proposed amendment

does not create the possibility of a new

or different kind of accident from any

accident previously evaluated.

Does the change involve a significant

reduction in a margin of safety?

The proposed change does not

involve a significant reduction in a

margin of safety. In fact, the proposed

changes restore margin and ensure

equipment operability. Since the

changes maintain the necessary level of

system reliability, they do not involve a

significant reduction in the margin of

safety.

Therefore, the change does not

involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: for Dresden, Morris Area

Public Library District, 604 Liberty

Street, Morris, Illinois 60450; for Quad

Cities, Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois

61021.

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60603.

NRC Project Director: Stuart A.

Richards.

Niagara Mohawk Power Corporation,

Docket No. 50–220, Nine Mile Point

Nuclear Station Unit No. 1, Oswego

County, New York

Date of application for amendment:

May 2, 1998, as supplemented May 21,

and 23 (three letters), 1998.

Brief description of amendment: This

amendment changed Technical

Specification (TS) 3/4.6.2, ‘‘Protective

Instrumentation,’’ and its associated

Bases to reflect modifications to the

initiation instrumentation for the

Control Room Air Treatment System. It

also changed TS 3.2.4a, ‘‘Reactor

Coolant Activity,’’ and added an

additional condition to the operating

license.

Date of issuance: May 23, 1998.

Effective date: As of the date of

issuance to be implemented prior to

resumption of power operation.

Amendment No.: 161.

Facility Operating License No. DPR–

63: Amendment revises the Technical

Specifications. Public comments

requested as to proposed no significant

hazards consideration: Yes (63 FR 27601

dated May 19, 1998. The notice

recognized the existence of exigent

circumstances pursuant to 10 CFR

50.91(a)(6) and provided an opportunity

to submit comments on the

Commission’s proposed no significant

hazards consideration determination.

The notice published May 19, 1998, also

provided for an opportunity to request

a hearing by June 1, 1998 (this will be

corrected to June 18, 1998, by a notice

to be published in the near future), but

indicated that if the Commission makes

a final no significant hazards

consideration determination, any such

hearing would take place after issuance

of the amendment. Subsequent to

publishing the notice, and due to

schedule improvements which have

occurred at the plant, the Commission

has determined that the amendment

should be issued on an emergency basis

pursuant to 10 CFR 50.91(a)(5). The

Commission’s related evaluation of the

amendment, finding of emergency

circumstances, consultation with the

State of New York, and final no

significant hazards consideration

determination are contained in a Safety

Evaluation date May 23, 1998.

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State
33106 Federal Register / Vol. 63, No. 116 /Wednesday, June 17, 1998 / Notices

University of New York, Oswego, New

York 13126.

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston & Strawn,

1400 L Street, NW, Washington, DC

20005–3502.

NRC Project Director: S. Singh Bajwa.

Northeast Nuclear Energy Company, et

al., Docket No. 50–336, Millstone

Nuclear Power Station, Unit No. 2, New

London County, Connecticut

Date of amendment request: May 14,

1998.

Description of amendment request:

The proposed amendment would

change the Technical Specifications

(TSs) for the Reactor Protection System

(RPS) and the Engineered Safety

Features Actuation System (ESFAS)

instrumentation by restricting the time

most RPS and ESFAS actuation

channels can be in the bypass position

to 48 hours. The current TSs have no

time limit. The proposed amendment

would also modify the TS action

requirements and the channel

calibration requirements for the loss of

turbine load reactor trip function, and

the channel calibration requirements for

the wide range logarithmic neutron flux

monitors; add a note to exclude the

neutron detectors from the channel

calibration requirements; correct a

reference to a TS surveillance

requirement; and correct errors that

have been identified.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Involve a significant increase in the

probability or consequences of an

accident previously evaluated.

The proposed change to restrict the

time most of the reactor protection or

engineered safety feature actuation

channels can be in the bypass position

to 48 hours, from an indefinite period of

time, has no effect on the design of the

Reactor Protection System (RPS) or the

Engineered Safety Feature Actuation

System (ESFAS), and does not affect

how these systems operate. In addition,

this will minimize the susceptibility of

these systems to the remote possibility

of fault propagation between channels.

The pressurizer high pressure reactor

protection channels will not be required

to be placed in the tripped condition

after 48 hours. A failed pressurizer high

pressure channel will be allowed to

remain in the bypassed condition for up

to 30 days. If the failed pressurizer high

pressure channel was placed in the

tripped condition, and then a high

failure of another pressurizer high

pressure channel occurred, the reactor

would trip and both pressurizer power

operated relief valves (PORVs) would

open, resulting in an undesired loss of

primary coolant. Limiting the time that

a failed pressurizer high pressure

reactor protection channel can be in

bypass to 30 days will minimize the risk

of the inadvertent opening of both

PORVs, as well as the risk associated

with fault propagation between

channels. These systems will still

function as designed to mitigate design

basis accidents. Therefore, this change

does not significantly increase the

probability or consequences of an

accident previously evaluated.

The proposed change to increase the

time a second RPS or ESFAS channel

can be removed from service (from 2

hours to 48 hours), provided one of the

inoperable channels is placed in the

tripped condition, has no effect on the

design of the RPS or ESFAS and does

not affect how these systems operate.

These systems will still function as

designed to mitigate design basis

accidents.

However, one of the proposed

changes will allow two pressurizer

pressure reactor protection channels to

be removed from service (one channel

in the tripped condition and one

channel in the bypassed condition) for

48 hours instead of the current 2 hour

time limit. With a pressurizer pressure

channel in the tripped condition, the

high failure of a second pressurizer

pressure channel would initiate a

reactor trip, open both pressurizer

PORVs, and cause an undesired loss of

primary coolant. Thus, this change will

increase the probability of occurrence of

a previously evaluated accident (FSAR

[Final Safety Analysis Report] Section

14.6.1—Inadvertent Opening of a

Pressurized Water Reactor Pressurizer

Pressure Relief Valve). However, since

this configuration will only be allowed

for an additional 46 hours, the increase

in the probability of occurrence of a

previously evaluated accident will be

limited to an acceptable value.

Therefore, this change does not

significantly increase the probability or

consequences of an accident previously

evaluated.

The proposed change to apply a more

restrictive action statement to the loss of

turbine load reactor trip function has no

effect on the design of this trip function

and does not affect how this trip

function operates. Also, this trip

function is not assumed to operate to

mitigate any design basis accident.

Therefore, this change does not

significantly increase the probability or

consequences of accident previously

evaluated.

The proposed change to require a

channel calibration every 18 months for

the loss of turbine load reactor trip

function and for the wide range

logarithmic neutron flux monitors has

no effect on the design of either the loss

of turbine load reactor trip function or

the wide range logarithmic neutron flux

monitors. Also, neither of these are

assumed to operate to mitigate any

design basis accident. Therefore, this

change does not significantly increase

the probability or consequences of an

accident previously evaluated.

The proposed change to exclude the

neutron detectors from the channel

calibration requirement has no effect on

the design of the neutron detectors and

has no significant effect on how these

detectors operate. The detectors are

passive devices with minimal drift. In

addition, slow changes in the sensitivity

of the linear power range flux detectors

is compensated for by performing the

daily calorimetric calibration and the

monthly calibration using the incore

detectors. These detectors will still

function as designed to mitigate design

basis accidents. Therefore, this change

does not significantly increase the

probability or consequences of an

accident previously evaluated.

The proposed change to correct the

surveillance requirement referenced in

an action statement has no effect on the

design of the ESFAS and does not affect

how this system operates. The ESFAS

will still function as designed to

mitigate design basis accidents.

Therefore, this change does not

significantly increase the probability or

consequences of an accident previously

evaluated.

The proposed change to add a

reference to the reactor coolant pump

low speed reactor trip function to a note

that states this trip may be bypassed

when [less than] 5 [percent] power, and

that the bypass must be automatically

removed when [greater than or equal to]

5 [percent] power will not effect this

reactor trip function. This bypass

capability currently exists in the design

of the Millstone Unit No. 2 RPS, and is

the same bypass feature referenced for

the reactor coolant flow low reactor trip

function. Both of these reactor trip

functions provide protection for a

reduction in RCS [Reactor Coolant

System] flow. The addition of this note

will not result in any technical change

to the Millstone Unit No. 2 RPS. The

RPS will continue to function as before.

Therefore, this change does not

significantly increase the probability or

consequences of an accident previously

evaluated.
Federal Register / Vol. 63, No. 116 /Wednesday, June 17, 1998 / Notices 33107

The proposed change to correct the

power level high trip setpoint on

Technical Specification Page 2–4 will

not result in any change to the actual

plant setpoint for this RPS trip function.

As a result of this proposed change, the

setpoint listed on Page 2–4 will agree

with the setpoint previously approved

by the NRC, and currently used by the

RPS. The change has no effect on the

design of the RPS and does not affect

how this system operates. Therefore,

this change does not significantly

increase the probability or consequences

of an accident previously evaluated.

The information added to the Bases of

the Technical Specifications to provide

a discussion of how the RPS and ESFAS

are affected by the proposed changes,

the effect the action statements have on

the operation of the RPS and ESFAS,

and to discuss the impact of

surveillance testing on RPS operability

will have no effect on equipment

operation. The RPS and ESFAS will

continue to function as designed to

mitigate design basis accidents.

Therefore, this change does not

significantly increase the probability or

consequences of an accident previously

evaluated.

Thus, this License Amendment

Request does not impact the probability

of an accident previously evaluated nor

does it involve a significant increase in

the 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 alter the

plant configuration (no new or different

type of equipment will be installed) or

require any new or unusual operator

actions. They do not alter the way any

structure, system, or component

functions and do not alter the manner

in which the plant is operated. The

proposed changes do not introduce any

new failure modes. They will not alter

assumptions made in the safety analysis

and licensing basis. The RPS and the

ESFAS will still function as designed to

mitigate design basis accidents.

Therefore, these 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 will not reduce

the margin of safety since they have no

impact on any safety analysis

assumption. The proposed changes do

not decrease the scope of equipment

currently required to be operable or

subject to surveillance testing, nor do

the proposed changes affect any

instrument setpoints or equipment

safety functions.

The effectiveness of Technical

Specifications will be maintained since

the changes will not alter the operation

of any RPS or ESFAS function. In

addition, most of the changes are

consistent with the Calvert Cliffs RPS

and ESFAS Technical Specifications

mode provided in Enclosure 3 of the

NRC correspondence dated April 16,

1981 (R. A. Clark letter to W. G. Counsil,

Evaluation of the Reactor Protection

System Inoperable Channel Condition at

Millstone Nuclear Power Station, Unit

No. 2, dated April 16, 1981) and the

new, improved Standard Technical

Specifications (STS) for Combustion

Engineering plants (NUREG–1432).

Therefore, there is no significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, Connecticut, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

P.O. Box 270, Hartford, Connecticut.

NRC Deputy Director: Phillip F.

McKee.

Power Authority of The State of New

York, Docket No. 50–286, Indian Point

Nuclear Generating Unit No. 3,

Westchester County, New York

Date of amendment request: June 25,

1997.

Description of amendment request:

The proposed amendment would

change the Indian Point 3 Technical

Specifications to allow the use of

zirconium alloy or stainless steel filler

rods in fuel assemblies to replace failed

or damaged fuel rods.

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:

Consistent with the criteria of 10 CFR

50.92, the enclosed application is

judged to involve no significant hazards

based on the following information:

(1) Does the proposed license

amendment involve a significant

increase in the probability or

consequences of an accident previously

analyzed?

Response: The proposed changes

modify the technical specification only

to the extent that the reconstitution is

recognized as acceptable under limited

circumstances. Reconstitution is limited

to substitution of zirconium alloy or

stainless steel filler rods, and must be in

accordance with approved applications

of fuel rod configurations. Although

these changes permit reconstitution to

occur without the need for a specific

technical specification change, use of an

approved methodology is required prior

to its application. Since the changes will

allow substitution of filler rods for

leaking, potentially leaking rods or

damaged rods, the changes may actually

reduce the radiological consequences of

an accident. It is noted that the specific

changes requested in this letter have

previously been found acceptable by the

NRC in GL [Generic Letter] 90–02,

Supplement 1. For these reasons, we

conclude that the changes will not

involve a significant increase in the

probability or consequences of an

accident previously evaluated.

(2) Does the proposed license

amendment create the possibility of a

new or different kind of accident from

any previously evaluated?

Response: The proposed changes will

not create the possibility of a new or

different kind of accident from any

accident previously evaluated because

they will only affect the assembly

configuration and can only be

implemented if demonstrated to meet

current plant requirements in

accordance with an NRC-approved

methodology. The other aspects of plant

design, operation limitations, and

responses to events will remain

unchanged. It is noted that the changes

have previously been determined

acceptable by the NRC in GL 90–02,

Supplement 1.

(3) Does the proposed amendment

involve a significant reduction in a

margin of safety?

Response: The proposed change will

not involve a reduction in a margin of

safety because the changes can only be

implemented if demonstrated to meet

current plant requirements in

accordance with an NRC-approved

methodology. It is noted that the

changes have previously been

determined acceptable by the NRC in

GL 90–02, Supplement 1.

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.
33108 Federal Register / Vol. 63, No. 116 /Wednesday, June 17, 1998 / Notices

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10601.

Attorney for licensee: Mr. David

Blabey, 10 Columbus Circle, New York,

New York 10019.

NRC Project Director: S. Singh Bajwa.

Public Service Electric & Gas Company,

Docket No. 50–354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of amendment request: April 28,

1998.

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) 3.4.2.1 to

replace the plus or minus 1 percent

setpoint tolerance limit for safety/relief

valves (SRVs) with a plus or minus 3

percent setpoint tolerance limit. In

addition, the proposed amendment

would revise TS 4.4.2.2 to state that all

SRVs must be certified to be within plus

or minus 1 percent of the TS setpoint

prior to returning the valves to service.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) The proposed changes do not

involve a significant increase in the

probability or consequences of an

accident previously evaluated.

The proposed TS revisions involve:

(1) no significant hardware changes; (2)

no significant changes to the operation

of any systems or components in normal

or accident operating conditions; and (3)

no changes to existing structures,

systems, or components. Therefore these

changes will not increase the probability

of an accident previously evaluated.

These proposed changes were

developed in accordance with the

provisions contained in an NRC Safety

Evaluation Report, dated 3/8/93, for the

‘‘BWR Owners Group Inservice Pressure

Relief Technical Specification

[Revision] Licensing Topical Report’’,

NEDC–31753P as described in General

Electric report NEDC–32511P, ‘‘Safety

Review for Hope Creek [Generating

Station] Safety/Relief Valve Tolerance

Analyses’’. Since the plant systems

associated with these proposed changes

will still be capable of: (1) meeting all

applicable design basis requirements;

and (2) retain the capability to mitigate

the consequences of accidents described

in the HC [Hope Creek] UFSAR

[Updated Final Safety Analysis Report],

the proposed changes were determined

to be justified. Therefore, these changes

will not involve a significant increase in

the consequences of an accident

previously evaluated.

(2) The proposed change does not

create the possibility of a new or

different kind of accident from any

accident previously evaluated.

Establishment of the [plus or minus]

3 [percent] SRV setpoint tolerance limit

will not adversely impact the operation

of any safety related component or

equipment. Since the proposed changes

involve: (1) no significant hardware

changes; (2) no significant changes to

the operation of any systems or

components; and (3) no changes to

existing structures, systems, or

components, there can be no impact on

the occurrence of any accident. These

proposed changes were developed in

accordance with the provisions

contained in an NRC Safety Evaluation

Report, dated 3/8/93, for the ‘‘BWR

Owners Group Inservice Pressure Relief

Technical Specification [Revision]

Licensing Topical Report’’, NEDC–

31753P as described in General Electric

report NEDC–32511P, ‘‘[Safety Review

for Hope Creek Generating Station]

Safety/Relief Valve Tolerance

Analyses’’. Furthermore, there is no

change in plant testing proposed in this

change request which could initiate an

event. Therefore, these changes will 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.

Establishment of the [plus or minus]

3 [percent] SRV setpoint tolerance limit

will not adversely impact the operation

of any safety related component or

equipment. General Electric analyses

performed for Hope Creek and

contained in General Electric report

NEDC–32511P, ‘‘[Safety Review for

Hope Creek Generating Station] Safety/

Relief Valve Tolerance Analyses,’’

concluded that there is no significant

impact on fuel thermal limits, no

significant impact on safety related

systems, structures or components, and

no significant impact on the accident

analyses associated with the proposed

changes. Therefore, the changes

contained in this request do not result

in a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070.

Attorney for licensee: Jeffrie J. Keenan,

Esquire, Nuclear Business Unit—N21,

P.O. Box 236, Hancocks Bridge, NJ

08038.

NRC Project Director: Robert A. Capra.

Southern Nuclear Operating Company,

Inc., Georgia Power Company,

Oglethorpe Power Corporation,

Municipal Electric Authority of Georgia,

City of Dalton, Georgia, Docket Nos. 50–

424 and 50–425, Vogtle Electric

Generating Plant, Units 1 and 2, Burke

County, Georgia

Date of amendment request: May 8,

1998.

Description of amendment request:

The proposed amendments would

change the Vogtle Electric Generating

Plant (VEGP) Technical Specification

(TS) 5.5.7, ‘‘Reactor Coolant Pump

Flywheel Inspection Program,’’ to

provide an exception to the examination

requirements of Regulatory Position

C.4.b of Regulatory Guide (RG) 1.14,

Revision 1, August 1975.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) Operation of the facility in

accordance with the proposed

amendment would not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The safety function of the RCP

[reactor coolant pump] flywheel is to

provide sufficient rotational inertia to

ensure reactor coolant flow through the

core during coastdown following a loss

of offsite power and subsequent reactor

trip. FSAR [Final Safety Analysis

Report] Chapter 15 analysis for a

complete loss of forced reactor coolant

flow demonstrates that the reactor trip

together with the flow sustained by the

inertia of the RCP impeller will be

sufficient to prevent the most limiting

fuel assembly from exceeding the DNBR

[departure from nucleate boiling ratio]

limits.

The maximum mechanical loading on

the RCP motor flywheel results from

overspeed following a LOCA [loss-ofcoolant accident]. The analysis

presented in WCAP–14535A

demonstrates that the revised inspection

program proposed by this license

amendment will ensure the integrity of

the RCP flywheels will be maintained.
Federal Register / Vol. 63, No. 116 /Wednesday, June 17, 1998 / Notices 33109

Based upon the findings of WCAP–

14535A, the ability of the RCP flywheel

to perform its intended safety function

will be unaffected by the license

amendment and the FSAR Chapter 15

analysis will remain valid. Therefore,

these proposed changes do not involve

a significant increase in the probability

or consequences of an accident

previously evaluated.

(2) Operation of the facility in

accordance with the proposed

amendment would not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed license amendment will

not change the physical plant

configuration nor the modes of

operation of any plant equipment. Based

upon the results of WCAP–14535A, no

new failure mechanism will be

introduced by the revised RCP flywheel

inspection program. Therefore, the

proposed amendment will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

(3) Operation of the facility in

accordance with the proposed

amendment would not involve a

significant reduction in a margin of

safety.

The operating limits and functional

capabilities of the affected systems,

structures, and components will be

unchanged by the proposed

amendment. The results of the RCP

flywheel inspections performed

throughout the industry and at VEGP

have identified no indications which

would affect its integrity. As presented

in WCAP–14535A, detailed stress

analysis and risk assessments have been

completed with the results indicating

that there would be no change in the

probability of failure for RCP flywheels

if all inspections were eliminated.

Therefore, these changes do not involve

a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Burke County Public Library,

412 Fourth Street, Waynesboro, Georgia.

Attorney for licensee: Mr. Arthur H.

Domby, Troutman Sanders,

NationsBank Plaza, Suite 5200, 600

Peachtree Street, NE., Atlanta, Georgia.

NRC Project Director: Herbert N.

Berkow.

Tennessee Valley Authority, Docket No.

50–390 Watts Bar Nuclear Plant, Unit 1,

Rhea County, Tennessee

Date of amendment request: May 6,

1998.

Description of amendment request:

The proposed amendment would

replace the two percent penalty

addressed in surveillance requirement

(SR) 3.2.1.2(a) with a burnup-dependent

factor to be specified in the Watts Bar

Core Operating Limits Report (COLR).

Specifically, the following changes are

being proposed:

1. SR 3.2.1.2(a) and its associated

BASES will have the phrase ‘‘by a factor

of 1.02’’ deleted and replaced with the

phrase ‘‘by the appropriate factor

specified in the COLR.’’

2. Technical Specification (TS)

Section 5.9.5(b)(3) would be updated to

reference the revised WCAP (10216–P–

A, Revision 1A, 1994) that details the

analytical methods utilized for the new

penalty factor.

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 involves only

the manner in which the penalty factors

for FQ(Z) would be specified (i.e.,

burnup-dependent factor specified in

the Core Operating Limits Report

[COLR] versus a constant factor

specified in the TS). This is simply used

to account for the fact that FQ C(Z) may

increase between surveillance intervals.

These penalty factors are not assumed

in any of the initiating events for the

accident analyses. Therefore the

proposed change will have no effect on

the probability of any accidents

previously evaluated. The penalty

factors specified in the COLR will be

calculated using NRC-approved

methodology and will continue to

provide an equivalent level of

protection as the existing TS

requirement. Therefore, the proposed

change will not affect the consequences

of any accident previously evaluated.

B. The proposed amendment does not

create the possibility of a new or

different kind of accident from any

accident previously evaluated.

The proposed change does not

involve a physical alteration to the plant

(no new or different kind of equipment

will be installed) or alter the manner in

which the plant would be operated.

Thus, this change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

C. The proposed amendment does not

involve a significant reduction in a

margin of safety.

The proposed change will continue to

ensure that potential increases in FQ C(Z)

over a surveillance interval will be

properly accounted for. The penalty

factors will be calculated using an NRCapproved methodology. Therefore, the

proposed change will not involve a

reduction in 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.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga,

TN 37402.

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET l0H,

Knoxville, Tennessee 37902.

NRC Project Director: Frederick J.

Hebdon.

Vermont Yankee Nuclear Power

Corporation, Docket No. 50–271,

Vermont Yankee Nuclear Power Station,

Vernon, Vermont

Date of amendment request: May 1,

1998.

Description of amendment request:

The proposed amendment would make

several editorial changes to the

Administrative Controls section of the

Technical Specifications. The changes

include revisions due to organizational

changes, quality assurance changes,

editorial changes, and typographical

corrections.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration which is presented below:

1. Will the proposed changes involve

a significant increase in the probability

or consequences of an accident

previously evaluated?

The administrative change proposed

herein will have no effect on plant

hardware, plant design, safety limit

setting or plant system operation and

therefore do[es] not modify or add any

initiating parameters that would

significantly increase the probability or

consequences of any previously

analyzed accident. The proposed

amendment changes the reference to the

VYNPS QA program and makes other
33110 Federal Register / Vol. 63, No. 116 /Wednesday, June 17, 1998 / Notices

administrative changes, such as title

changes and correction/clarification of

errors. Therefore, there is no increase in

the probability or consequence of an

accident previously evaluated.

2. Will the proposed changes create

the possibility of a new or different kind

of accident from any accident

previously evaluated?

This change does not affect any

equipment nor does it involve any

potential initiating events that would

create any new or different kind of

accident. The proposed change involves

[ ] wording changes in the Technical

Specifications identifying the name of

the QA program and makes other

administrative changes, such as title

changes and corrective/clarification of

errors. Therefore no new or different

kind of accident has been introduced.

3. Will the proposed changes involve

a significant reduction in a margin of

safety?

This change does not affect any

equipment involved in potential

initiating events or safety limits. The

proposed change has no significant

impact on margin of safety, as it is

comprised of only administrative

changes.

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.

Local Public Document Room

location: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

Attorney for licensee: Mr. David R.

Lewis, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037–1128.

NRC Project Director: Cecil

O.Thomas.

Virginia Electric and Power Company,

Docket Nos. 50–338 and 50–339, North

Anna Power Station, Units No. 1 and

No. 2, Louisa County, Virginia

Date of amendment request:

September 1, 1995, as supplemented

April 8, 1996, April 22, 1996, April 23,

1996, November 18, 1997, February 9,

1998, March 25, 1998 and May 5, 1998.

This notice supersedes the Federal

Register notice of September 27, 1995

(60 FR 49949)

Description of amendment request:

The originally (September 1, 1995)

proposed changes to the Technical

Specifications (TS) would permit a

single outage of up to 14 days for each

emergency diesel generator (EDG) once

every 18 months in order to perform

preventive maintenance. The amended

request will permit a single outage of up

to 14 days for each EDG for any reason;

TS change to incorporate a

Configuration Risk Management

Program (CRMP) in the Administrative

Section in the TS, in support of the

previous submittal for the 14-day

Allowed Outage Time (AOT) for the

EDGs and would permit an increase in

the TS maintenance interval of the EDG

from 18 to 24 months, based on the

recommendation from the EDG owners

group (Fairbanks Morse Owners Group).

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. Specifically, operation of

North Anna Power Station in

accordance with the proposed Technical

Specification changes will not:

a. Involve a significant increase in the

probability or consequences of an

accident previously evaluated.

A probabilistic safety analysis (PSA)

has been performed which demonstrates

that a 14-day AOT for each EDG, results

in a small change in core damage

frequency assuming adequate

compensatory measures are in place.

The compensatory measures include

requirements that the other EDGs, offsite power supply, and the alternate

A.C. diesel (AAC DG) be operable

whenever the action statement is

entered.

The effect of the proposed change has

been calculated to be an increase in core

damage frequency of approximately 1

E–6 per year from the baseline core

damage frequency of 4.1 E–5.

Considering that credit was not taken

for the AAC DG previously in the IPE

nor was the AAC DG specified in

Technical Specifications, the proposed

changes remain bounded by the core

damage frequency identified in the

Individual Plant Examination.

Credit for the AAC DG was previously

not taken nor was the AAC DG

previously included in the Technical

Specifications. Furthermore, the

probabilistic safety analysis (PSA)

demonstrates that the increase in core

damage frequency due to extending the

EDG AOT of a 14-day period is not

significant as long as the AAC DG is

operable to act as a source of emergency

power to replace the EDG. The period of

time during which the EDG is

unavailable is short enough to limit the

impact of using the manually operated

AAC DG as a replacement for the

automatically operated EDG.

The plant design and operation are

not changed by the incorporation of a

CRMP into the Administrative Section

of Technical Specifications. Further,

with the proposed change to the

preventive maintenance interval, the

EDG reliability remains adequate to

perform its function of supporting

accident mitigation equipment with

emergency electrical power.

Therefore, neither the probability of

occurrence nor the consequences of an

accident or malfunction of equipment

important to safety previously evaluated

in the safety analysis report are

increased due [to] the proposed changes

to permit a 14-day allowed outage time

and a 24 month preventive maintenance

interval for the EDGs.

b. Create the possibility of a new or

different kind of accident from any

accident previously evaluated.

No new initiators are defined as a

result of a review of the PSA model. The

proposed Technical Specifications

changes only modify the AOT of an

EDG. The UFSAR [Updated Final Safety

Analysis Report] accidents are analyzed

assuming that the EDG is the worst

single failure. This assumption is more

severe than the proposed Technical

Specifications changes, which [replace]

the EDG with the AAC DG. Similarly,

the PSA performed to evaluate the

proposed Technical Specifications

changes considered all of the initiating

events defined for the PSA performed

for the Individual Plant Examination.

No new initiators were defined as a

result of a review of the PSA model.

Adding the CRMP and changing the

EDG preventive maintenance interval in

the Technical Specifications does not

change any method of operation or

create any new modes of operation or

accident precursors.

Therefore, it is concluded that no new

or different kind of accident or

malfunction from any previously

evaluated has been or will be created by

the proposed changes to permit a 14-day

allowed outage time and a 24 month

preventive maintenance interval for the

EDGs.

c. The proposed Technical

Specifications changes do not result in

a reduction in margin of safety as

defined in the basis for any Technical

Specifications.

The PSA was performed to evaluate

the concept of a one-time outage. The

results of the analyses show a small

change in the core damage frequency.

As described above the proposed

Technical Specifications changes only

modify the AOT of an EDG. Thus,

operation with slightly increased EDG

unavailability due to maintenance is

acceptable given the operability of the

AAC DG and the other EDG.

Incorporating the CRMP and changing

the EDG preventive maintenance

interval in the Technical Specifications
Federal Register / Vol. 63, No. 116 /Wednesday, June 17, 1998 / Notices 33111

does not affect any accident analysis

assumptions or change any Technical

Specifications criteria.

Therefore, the margin of safety is not

changed.

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.

Local Public Document Room

location: The Alderman Library, Special

Collections Department, University of

Virginia, Charlottesville, Virginia

22903–2498.

Attorney for licensee: Michael W.

Maupin, Esq., Hunton and Williams,

Riverfront Plaza, East Tower, 951 E.

Byrd Street, Richmond, Virginia 23219.

NRC Project Director: Gordon E.

Edison, Acting.

Wisconsin Public Service Corporation,

Docket No. 50–305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of amendment request: May 7,

1998.

Description of amendment request:

Technical Specification 5.4, ‘‘Fuel

Storage,’’ would be changed to increase

the allowable mass of uranium-235, per

axial centimeter, for fuel storage in new

fuel and spent fuel storage racks. This

change will allow use of new Siemens

heavy fuel assemblies.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The proposed change was reviewed in

accordance with the provisions of 10

CFR 50.92 to show no significant

hazards exist. The proposed change will

not:

(1) Involve a significant increase in

the probability or consequence of an

accident previously evaluated.

The mass of the fuel assembly is

increased by a small amount (30

pounds, or 2.4%), from that of the fuel

assemblies now in the core. Even with

this increase, the load on the fuel

handling equipment is still well within

design limits. Therefore, the

probabilities of a fuel handling accident

inside containment (FHAIC) and the

fuel handling accident outside

containment (FHAOC) are not changed.

The total core mass, with Siemens

heavy fuel, is less than that assumed in

the original plant safety analysis. The

proposed change does not alter the plant

configuration, operating set points, or

overall plant performance. The

probability of other accidents is

therefore not changed.

Attachment 4 (of the application)

shows that the consequences of a fuel

handling accident or a large break loss

of coolant accident are not significantly

affected.

Any changes in the nuclear properties

of the reactor core that may result from

a higher mass of fuel U235 per axial

centimeter will be analyzed and shown

to meet acceptance criteria in the

appropriate reload analysis, which

would be completed prior to use.

(2) Create the possibility of a new or

different kind of accident from any

previously evaluated.

As discussed above, the only safety

issue significantly affected by the

proposed change is the criticality

analysis of the spent fuel storage racks

and new fuel storage racks. Since it has

been demonstrated that keff remains

below the keff acceptance criteria, no

new or different accident would be

created through the use of fuel with up

to 56.067 grams of U235 per axial

centimeter at the Kewaunee Nuclear

Power Plant.

The proposed change does not alter

the plant configuration, operating set

points, or overall plant performance and

therefore does not create a new or

different kind of accident from any

accident previously evaluated.

(3) Involve a significant reduction in

the margin of safety.

The criticality analysis in Reference 3

(of the application) demonstrates that

adequate margins to criticality can be

maintained with up to 56.067 grams of

U235 per axial centimeter stored in either

the new fuel storage racks or the spent

fuel storage racks.

The bounding cases of the analysis

demonstrate that keff remains less than

0.95 in the spent fuel storage racks and

the new fuel storage racks if flooded

with unborated water. The bounding

cases of the analysis also demonstrate

that keff remains less than 0.98 in the

new fuel storage racks if moderated by

optimally misted moderator. Therefore,

the 56.067 grams of U235 per axial

centimeter limit is acceptable for storage

in both the new fuel storage racks and

the spent fuel storage racks.

Any changes in the nuclear properties

of the reactor core that may result from

a higher mass of fuel U235 per axial

centimeter will be analyzed in the

appropriate reload analysis to ensure

compliance with applicable reload

considerations and requirements.

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.

Local Public Document Room

location: University of Wisconsin,

Cofrin Library, 2420 Nicolet Drive,

Green Bay, WI 54311–7001.

Attorney for licensee: Bradley D.

Jackson, Esq., Foley and Lardner, P.O.

Box 1497, Madison, WI 53701–1497.

NRC Acting Project Director: Ronald

R. Bellamy.

Wisconsin Electric Power Company,

Docket No. 50–301, Point Beach Nuclear

Plant, Unit 2, Town of Two Creeks,

Manitowoc County, Wisconsin

Date of amendment request: May 15,

1998 (NPL–98–0303).

Description of amendment request:

The proposed amendment revises the

schedule for implementing the boron

concentration changes related to the

planned conversion of Unit 2 to 18-

month fuel cycles.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration which is presented below:

1. Operation of the Point Beach

Nuclear Plant in accordance with the

proposed amendment will not result in

a significant increase in the probability

or consequences of an accident

previously evaluated.

The proposed changes are

administrative only. There are no

physical changes to the facility or its

operation. All Limiting Conditions of

Operation, Limiting Safety System

Settings, and Safety Limits specified in

the Technical Specification remain

unchanged. Additionally, there are no

changes in the Quality Assurance

Program, Emergency Plan, Security

Plan, and Operator Training and

Requalification Program. Therefore, an

increase in the probability or

consequences of an accident previously

evaluated cannot occur.

2. Operation of the Point Beach

Nuclear Plant in accordance with the

proposed amendment will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes are

administrative only. No changes to the

facility structures, systems and

components or their operation will

result. The design and design basis of

the facility remain unchanged. The

plant safety analyses remain current and

accurate. No new or different failure

mechanisms are introduced. Therefore,
33112 Federal Register / Vol. 63, No. 116 /Wednesday, June 17, 1998 / Notices

the possibility of a new or different kind

of accident from any accident

previously evaluated is not introduced.

3. Operation of the Point Beach

Nuclear Plant in accordance with the

proposed amendment does not involve

a significant reduction in a margin of

safety.

The proposed [amendment is]

administrative only. All safety margins

established through the design and

facility license including the Technical

Specifications remain unchanged.

Therefore, all margins of safety are

maintained.

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.

Local Public Document Room

location: The Lester Public Library,

1001 Adams Street, Two Rivers,

Wisconsin 54241.

Attorney for licensee: John H. O’Neill,

Jr., Shaw, Pittman, Potts, and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Cynthia A.

Carpenter.

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.

Detroit Edison Company, Docket No.

50–341, Fermi 2, Monroe County,

Michigan

Date of amendment request: May 20,

1998 (NRC–98–0099).

Description of amendment request:

The proposed amendment would

modify the scram discharge volume

(SDV) vent and drain valve action

requirements to be consistent with those

contained in NUREG–1433, Revision 1,

‘‘Standard Technical Specifications

General Electric Plants, BWR/4.’’

Detroit Edison is requesting that this

license amendment request be

processed in an exigent manner in

accordance with 10 CFR 50.91(a)(6)

because delay in granting this

amendment could lead to a plant

shutdown.

Date of publication of individual

notice in Federal Register: May 28, 1998

(63 FR 29254).

Expiration date of individual notice:

Comments: June 11, 1998; hearing: June

29, 1998.

Local Public Document Room

location: Monroe County Library

System, Ellis Reference and Information

Center, 3700 South Custer Road,

Monroe, Michigan 48161.

Duke Energy Corporation, Docket Nos.

50–413 and 50–414, Catawba Nuclear

Station, Units 1 and 2, York County,

South Carolina

Date of amendment request: May 22,

1998.

Description of amendment request:

The proposed amendments would

revise Surveillance Requirement Section

4.4.3.3 of the Technical Specifications.

Section 4.4.3.3 currently requires that

the emergency power supply for the

pressurizer heaters be demonstrated

OPERABLE at least once per 18 months

by manually transferring power from the

normal to the emergency power supply.

The licensee proposed to delete the

‘‘manual’’ requirement because the

power supply transfer at the unit was

designed to be automatic. The proposed

requirement is to verify that required

pressurizer heaters are capable of being

powered from an emergency power

supply once per 18 months.

Date of publication of individual

notice in Federal Register: June 1, 1998

(63 FR 29759).

Expiration date of individual notice:

July 1, 1998.

Local Public Document Room

location: York County Library, 138 East

Black Street, Rock Hill, South Carolina.

Niagara Mohawk Power Corporation,

Docket No. 50–220, Nine Mile Point

Nuclear Station Unit No. 1, Oswego

County, New York

Date of application for amendment:

May 2, 1998.

Brief description of amendment: The

amendment changes the Technical

Specifications 3/4.6.2, ‘‘Protective

Instrumentation,’’ to reflect

modifications to the initiation

instrumentation for the Control Room

Air Treatment system.

Date of publication of individual

notice in Federal Register: May 19, 1998

(63 FR 27601).

Expiration date of individual notice:

June 18, 1998.

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Niagara Mohawk Power Corporation,

Docket No. 50–220, Nine Mile Point

Nuclear Station Unit No. 1, Oswego

County, New York

Date of application for amendment:

May 15, 1998 (two letters).

Brief description of amendment: The

amendment changes administrative

sections of the Technical Specifications

to reflect a restructuring of upper

management organization.

Date of publication of individual

notice in Federal Register: June 2, 1998

(63 FR 30026).

Expiration date of individual notice:

July 2, 1998.

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Pennsylvania Power and Light

Company, Docket Nos. 50–387 and 50–

388 Susquehanna Steam Electric

Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of amendment request: May 12,

1998.

Brief description of amendment

request: These amendments relocate

certain requirements related to fire

protection from the TSs to the Updated

Final Safety Analysis Report. The TS

sections to be relocated are: 3/4.3.7.9,

Fire Detection Instrumentation; 3/4.7.6,

Fire Suppression Systems; 3/4.7.7, Fire

Rated Assemblies; and 6.2.2e, Fire

Brigade Staffing. The amendments also

replace License Condition 2.C.(6) for

Unit 1 and License Condition 2.C.(3) for

Unit 2. These amendments are

consistent with the guidance of NRC

Generic Letter (GL) 86–10,

‘‘Implementation of Fire Protection

Requirements,’’ and GL 88–12,

‘‘Removal of Fire Protection

Requirements from Technical

Specifications.’’

Date of publication of individual

notice in Federal Register: May 21, 1998

(63 FR 28010).

Expiration date of individual notice:

June 22, 1998.

Local Public Document Room

location: Osterhout Free Library,

Reference Department, 71 South

Franklin Street, Wilkes-Barre, PA 18701.
Federal Register / Vol. 63, No. 116 /Wednesday, June 17, 1998 / Notices 33113

Rochester Gas and Electric Corporation,

Docket No. 50–244, R. E. Ginna Nuclear

Power Plant, Wayne County, New York

Date of amendment request: March

31, 1997, as supplemented June 18,

1997, October 10, 1997, October 20,

1997, November 11, 1997, December 22,

1997, January 15, 1998, January 27,

1998, March 30, 1998, April 23, 1998,

and April 27, 1998.

Brief description of amendment

request: The proposed amendment

would revise the Ginna Station

Improved Technical Specifications to

reflect a planned modification to the

spent fuel pool storage racks.

Date of publication of individual

notice in Federal Register: May 12, 1998

(63 FR 26213). This notice supersedes

the March 31, 1997, application

published on April 30, 1997 (62 FR

23502).

Expiration date of individual notice:

June 11, 1998.

Local Public Document Room

Location: Rochester Public Library, 115

South Avenue, Rochester, New York

14610.

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 at the

local public document rooms for the

particular facilities involved.

Baltimore Gas and Electric Company,

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

Cliffs Nuclear Power Plant, Unit Nos. 1

and 2, Calvert County, Maryland

Date of application for amendments:

January 31, 1997, as supplemented

February 13, February 28, March 25,

April 16, August 19, and September 29,

1997, January 22, March 17, April 8,

April 21, 1998, and May 22, 1998.

Brief description of amendments: The

amendments revise the TS for a

reduction of the total reactor coolant

system flow limit from 370,000 gallons

per minute (gpm) to 340,000 gpm in

support of increased steam generator

tube plugging.

Date of issuance: May 23, 1998.

Effective date: As of the date of

issuance Unit 1 to be implemented

within 60 days and Unit 2 prior to

startup from the spring 1999 refueling

outage.

Amendment Nos.: 228 and 202.

Facility Operating License Nos. DPR–

53 and DPR–69: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: February 26, 1997 (62 FR

8780).

The February 13, February 28, March

25, April 16, August 16, and September

29, 1997, January 22, March 17, April 8,

and April 21, 1998, and May 22, 1998,

letters provided clarifying information

that did not change the initial proposed

no significant hazards consideration.

The Commission’s related evaluation

of these amendments is contained in a

Safety Evaluation dated May 23, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Calvert County Library, Prince

Frederick, Maryland 20678.

Carolina Power & Light Company, et al.,

Docket Nos. 50–325 and 50–324,

Brunswick Steam Electric Plant, Units 1

and 2, Brunswick County, North

Carolina

Date of application for amendments:

July 18, 1997.

Brief description of amendments: The

amendments revise the listed design

suppression chamber temperature of

200°F to 220°F and the listed total water

and steam volume of the reactor coolant

system from 18,670 cubic feet to 18,320

cubic feet, respectively.

Date of issuance: May 27, 1998.

Effective date: May 27, 1998.

Amendment Nos.: 195 and 225.

Facility Operating License Nos. DPR–

71 and DPR–62: Amendments revise the

facility’s Technical Specifications.

Date of initial notice in Federal

Register: August 27, 1997 (62 FR

45454).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated May 27, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of North Carolina at

Wilmington, William Madison Randall

Library, 601 S. College Road,

Wilmington, North Carolina 28403–

3297.

Carolina Power & Light Company, et al.,

Docket Nos. 50–325 and 50–324,

Brunswick Steam Electric Plant, Units 1

and 2, Brunswick County, North

Carolina

Date of application for amendments:

June 12, 1997, as supplemented

February 2, 1998. The February 2, 1998,

submittal contained clarifying

information only and did not change the

initial proposed no significant hazards

consideration or expand the scope of the

original Federal Register Notice.

Brief Description of amendments: The

amendments consist of changes to the

Technical Specifications (TS) to revise

the Limiting Condition for Operation of

the TS to limit the drywell average air

temperature rather than primary

containment air temperature.

Additionally, the amendments require

that the drywell average air temperature

be maintained less than or equal to 150

°F during plant operation. The current

primary containment average

temperature limit is 135 °F.

Date of issuance: May 28, 1998.

Effective date: May 28, 1998.

Amendment Nos.: 196 and 226.

Facility Operating License Nos. DPR–

71 and DPR–62: Amendments change

the Technical Specifications.

Date of initial notice in Federal

Register: August 27, 1997 (62 FR 45454)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated May 28, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of North Carolina at

Wilmington, William Madison Randall

Library, 601 S. College Road,

Wilmington, North Carolina 28403–

3297.
33114 Federal Register / Vol. 63, No. 116 /Wednesday, June 17, 1998 / Notices

Carolina Power & Light Company, et al.,

Docket Nos. 50–325 and 50–324,

Brunswick Steam Electric Plant, Units 1

and 2, Brunswick County, North

Carolina

Date of application for amendments:

October 28, 1997

Brief Description of amendments: The

amendments revise certain

instrumentation allowable values in the

current technical specifications to the

Improved Technical Specifications

format.

Date of issuance: May 28, 1998.

Effective date: May 28, 1998.

Amendment Nos.: 197 and 227.

Facility Operating License Nos. DPR–71

and DPR–62: Amendments change the

Technical Specifications.

Date of initial notice in Federal

Register: December 31, 1997 (62 FR

68304)

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated May 28, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of North Carolina at

Wilmington, William Madison Randall

Library, 601 S. College Road,

Wilmington, North Carolina 28403–

3297.

Carolina Power & Light Company, et al.,

Docket Nos. 50–325 & 50–324,

Brunswick Steam Electric Plant, Units 1

& 2, Brunswick County, North Carolina

Date of amendment request:

November 15, 1995.

Brief description of amendment: The

amendments modify the channel

functional test interval in the Technical

Specifications Surveillance

Requirements for the Electrical

Protective Assemblies in the Reactor

Protection System.

Date of issuance: May 29, 1998.

Effective date: May 29, 1998.

Amendment No.: 198 and 228.

Facility Operating License Nos. DPR–

71 and DPR–62: Amendments revise the

Technical Specifications.

Date of initial notice in Federal

Register: July 3, 1996 (61 FR 34887).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 29, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of North Carolina at

Wilmington, William Madison Randall

Library, 601 S. College Road,

Wilmington, North Carolina 28403–

3297.

Carolina Power & Light Company, et al.,

Docket Nos. 50–325 & 50–324,

Brunswick Steam Electric Plant, Units 1

& 2, Brunswick County, North Carolina

Date of amendment request:

November 16, 1994, as supplemented by

letters dated February 14, 1995, and

April 9, 1998.

Brief description of amendment: The

amendments change the Technical

Specifications (TS) for Units 1 and 2 to

revise the basis for removing the

suppression chamber water temperature

monitoring instrumentation

requirements from the TS. This change

is being processed in parallel with the

Improved Technical Specification

conversion.

Date of issuance: May 29, 1998.

Effective date: May 29, 1998.

Amendment Nos.: 199 and 229.

Facility Operating License Nos. DPR–

71 and DPR–62: Amendments revise the

Technical Specifications.

Date of initial notice in Federal

Register: January 4, 1995 (60 FR 497)

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 29, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of North Carolina at

Wilmington, William Madison Randall

Library, 601 S. College Road,

Wilmington, North Carolina 28403–

3297.

Carolina Power & Light Company, et al.,

Docket Nos. 50–325 and 50–324,

Brunswick Steam Electric Plant, Units 1

and 2, Brunswick County, North

Carolina

Date of application for amendments:

April 4, 1996, as supplemented January

24, 1997, March 31, 1997, April 2, 1997,

April 14, 1997, March 24, 1998, and

May 20, 1998.

Brief Description of amendments: The

amendments modify Technical

Specifications (TS) 3.0.4, 4.0.3, and

4.0.4, and their associated Bases in

accordance with the guidance provided

in Generic Letter 87–09, ‘‘Sections 3.0

and 4.0 of the Standard Technical

Specifications (STS) on the

Applicability of Limiting Conditions for

Operation and Surveillance

Requirements.’’

Date of issuance: June 2, 1998.

Effective date: June 2, 1998.

Amendment Nos.: 200 and 230.

Facility Operating License Nos. DPR–

71 and DPR–62: Amendments change

the Technical Specifications.

Date of initial notice in Federal

Register: July 17, 1996 (61 FR 37297).

The supplemental submittals

contained clarifying information only,

and did not change the initial no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated June 2, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of North Carolina at

Wilmington, William Madison Randall

Library, 601 S. College Road,

Wilmington, North Carolina 28403–

3297

Carolina Power & Light Company, et al.,

Docket Nos. 50–325 & 50–324,

Brunswick Steam Electric Plant, Units 1

& 2, Brunswick County, North Carolina

Date of amendment request: April 30,

1997, as supplemented October 28,

1997, and May 15, 1998.

Brief description of amendment: The

amendments revise surveillance

requirements 4.7.2.b.2 and 4.7.2.c to

require testing of the control room

emergency ventiliation system charcoal

adsorber in accordance with the

American Society for Testing and

Material D3803–1989, ‘‘Standard Test

Method for Nuclear-Grade Activated

Carbon.’’

Date of issuance: June 2, 1998.

Effective date: June 2, 1998.

Amendment Nos.: 201 and 231.

Facility Operating License Nos. DPR–

71 and DPR–62: Amendments revise the

Technical Specifications.

Date of initial notice in Federal

Register: July 30, 1997 (62 FR 40846).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated June 2, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of North Carolina at

Wilmington, William Madison Randall

Library, 601 S. College Road,

Wilmington, North Carolina 28403–

3297

Carolina Power & Light Company, et al.,

Docket Nos. 50–325 and 50–324,

Brunswick Steam Electric Plant, Units 1

and 2, Brunswick County, North

Carolina

Date of application for amendments:

April 3, 1998

Brief description of amendments: The

amendments revise the specified total

volume of the condensate storage tank

capacity requirements from 150,000

gallons to 228,200 gallons to ensure the

Core Spray System requirement of

50,000 gallons.

Date of issuance: June 5, 1998.

Effective date: June 5, 1998.

Amendment Nos.: 202 and 232.
Federal Register / Vol. 63, No. 116 /Wednesday, June 17, 1998 / Notices 33115

Facility Operating License Nos. DPR–

71 and DPR–62: Amendments revise the

facility’s Technical Specifications.

Date of initial notice in Federal

Register: May 6, 1998 (63 FR 25103).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated June 5, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of North Carolina at

Wilmington, William Madison Randall

Library, 601 S. College Road,

Wilmington, North Carolina 28403–

3297.

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:

October 29, 1997.

Brief description of amendment: This

amendment changes Technical

Specifications (TS) 3.8.1.1.a.3,

3.8.1.1.b.4, and 3.8.1.1.d.2 by

eliminating the plant shutdown

requirements in these TS, and allowing

the applicable redundant feature TS to

direct the plant shutdown when

required.

Date of issuance: May 22, 1998.

Effective date: May 22, 1998.

Amendment No.: 78.

Facility Operating License No. NPF–

63: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: December 31, 1997 (62 FR

68305).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 22, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh,

North Carolina 27605

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:

April 24, 1998, as supplemented by

letter dated May 15, 1998.

Brief description of amendment: This

amendment revises TS 3.3.2,

‘‘Engineered Safety Features Actuation

System Instrumentation,’’ such that

surveillance of the undervoltage relays

may be performed without entry into TS

3.0.3. Specifically, the change modifies

Table 3.3–3 to allow operation with

more than one channel of the emergency

bus undervoltage relays inoperable.

Date of issuance: June 3, 1998.

Effective date: June 3, 1998.

Amendment No.: 79.

Facility Operating License No. NPF–

63: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: May 4, 1998 (63 FR 24574).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated June 3, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh,

North Carolina 27605.

Commonwealth Edison Company,

Docket Nos. STN 50–454 and STN 50–

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

Ogle County, Illinois Docket Nos. STN

50–456 and STN 50–457, Braidwood

Station, Unit Nos. 1 and 2, Will County,

Illinois

Date of application for amendments:

September 24, 1997.

Brief description of amendments: The

amendments revise the surveillance

frequency for the turbine throttle valves

and the turbine governor valves from

monthly to quarterly.

Date of issuance: May 26, 1998.

Effective date: Immediately, to be

implemented within 30 days.

Amendment Nos.: 103 and 93.

Facility Operating License Nos. NPF–

37, NPF–66, NPF–72 and NPF–77: The

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 11, 1998 (63 FR 11917).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated May 26, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: For Byron, the Byron Public

Library District, 109 N. Franklin, P.O.

Box 434, Byron, Illinois 61010; for

Braidwood, the Wilmington Public

Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481.

Detroit Edison Company, Docket No.

50–341, Fermi 2, Monroe County,

Michigan

Date of application for amendment:

January 28, 1998 (NRC–98–0006), as

supplemented on March 10, 1998 (NRC–

98–0036).

Brief description of amendment: The

amendment revises technical

specification surveillance requirement

4.4.3.2.2.a for the leak rate test of the

pressure isolation valves, extending it

from the current 18-month interval to a

24-month interval.

Date of issuance: May 28, 1998.

Effective date: May 28, 1998, with full

implementation within 90 days.

Amendment No.: 118.

Facility Operating License No. NPF–

43: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: February 25, 1998 (63 FR

9598).

The March 10, 1998, supplement

requested a change in the

implementation period. This

information was within the scope of the

original Federal Register notice and did

not change the staff’s initial proposed

no significant hazards considerations

determination. The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

May 28, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Monroe County Library

System, 3700 South Custer Road,

Monroe, Michigan 48161.

Detroit Edison Company, Docket No.

50–341, Fermi 2, Monroe County,

Michigan

Date of application for amendment:

November 22, 1995 (NRC–95–0124), as

supplemented February 19, April 19,

May 3, June 12, and December 4, 1996,

January 30 and August 7, 1997, and

April 27 and May 22, 1998.

Brief description of amendment: The

amendment revises technical

specification (TS) 3.8.1.1 to change the

emergency diesel generator (EDG)

allowed outage time from 3 to 7 days

and add a requirement to verify that

combustion turbine-generator 11–1 is

available prior to removing an EDG from

service. In addition, in accordance with

draft staff guidance for risk-informed

amendments, a section is added to the

Administrative Controls Section of the

TS describing the licensee’s

configuration risk management program.

The associated Bases are also revised.

The November 22, 1995, submittal also

requested changes to the testing and

reporting requirements for the EDGs.

These aspects were addressed in

Amendment No. 107 to the TS issued on

June 20, 1996. The staff’s action on the

licensee’s request is now complete.

Date of issuance: June 2, 1998.

Effective date: June 2, 1998, with full

implementation within 60 days.

Amendment No.: 119.

Facility Operating License No. NPF–

43. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: February 28, 1996 (61 FR

7550) with a supplemental notice on

May 1, 1998 (63 FR 24195).
33116 Federal Register / Vol. 63, No. 116 /Wednesday, June 17, 1998 / Notices

The February 19, April 19, May 3,

June 12, and December 4, 1996, August

7, 1997, and May 22, 1998, submittals

provided clarifying information within

the scope of the Federal Register notices

and did not change the staff’s initial

proposed no significant hazards

considerations determinations.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated June 2, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Monroe County Library

System, Ellis Reference and Information

Center, 3700 South Custer Road,

Monroe, Michigan 48161.

Duquesne Light 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:

March 17, 1998, as supplemented May

14, 1998.

Brief description of amendments:

These amendments revise Action 34 of

technical specification (TS) Table 3.3–3,

‘‘Engineered Safety Feature Actuation

System Instrumentation.’’ Action 34 is

applicable to Functional Units 6.b.,

‘‘Grid Degraded Voltage (4.16 kV Bus),’’

and 6.c., ‘‘Grid Degraded Voltage (480 v

Bus).’’ Revised Action 34 requires that

with one degraded grid voltage

monitoring channel inoperable, the

inoperable channel be placed in the

tripped condition within one hour;

otherwise, immediately enter the

applicable action statement(s) for the

associated emergency diesel generator

made inoperable by the degraded

voltage start instrumentation. The

revision to Action 34 also requires that

with two degraded grid voltage

monitoring channels inoperable, within

one hour restore at least one of the

channels to operable status and place

the other channel in the tripped

condition; otherwise, the associated

emergency diesel generator would be

declared inoperable and its applicable

action statement(s) entered.

Corresponding changes have also been

made in the bases for TS 3/4.3.2 and the

BVPS–2 TS Index pages.

Date of issuance: May 27, 1998.

Effective date: Effective immediately,

to be implemented within 60 days (both

units).

Amendment Nos.: 214 and 91.

Facility Operating License Nos. DPR–

66 and NPF–73: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: April 22, 1998 (63 FR 19969).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated May 27, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, PA

15001.

Duquesne Light Company, et al., Docket

Nos. 50–334 and 50–412, Beaver Valley

Power Station, Unit Nos. 1 and 2,

(BVPS–1 and BVPS–2) Shippingport,

Pennsylvania

Date of application for amendments:

March 16, 1998, as supplemented May

14, 1998.

Brief description of amendments:

These amendments revise technical

specification (TS) Table 4.3–1 to add

footnote 6 to the channel calibration

requirement for all instrument channels

that are provided with an input from

neutron flux detectors. Footnote 6

provides that neutron detectors may be

excluded from channel calibrations. In

addition, BVPS–1 TS Table 4.3–1 is

being revised to add channel calibration

requirements to items 2.b. (Power

Range, Neutron Flux, Low Setpoint), 5.

(Intermediate Range, Neutron Flux), 6.

(Source Range, Neutron Flux (Below P–

10)), and 23. (Reactor Trip System

Interlocks P–6, P–8, P–9, and P–10).

Furthermore, changes are being made to

correct page numbers in the BVPS–2 TS

Index and to add corresponding changes

to the TS Bases for both units.

Date of issuance: May 28, 1998.

Effective date: Both units, effective

immediately, to be implemented within

60 days.

Amendment Nos.: 215 and 92.

Facility Operating License Nos. DPR–

66 and NPF–73: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: April 22, 1998 (63 FR 19969).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated May 28, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, PA

15001.

Florida Power and Light Company,

Docket Nos. 50–250 and 50–251, Turkey

Point Plant Units 3 and 4, Dade County,

Florida

Date of application for amendments:

January 9, 1998, as supplemented by

letter dated April 20, 1998.

Brief description of amendments: The

amendments permit the use of fuel with

ZIRLO cladding.

Date of issuance: May 12, 1998.

Effective date: May 12, 1998.

Amendment Nos. 196 and 190.

Facility Operating Licenses Nos. DPR–

31 and DPR–41: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: February 25, 1998 (63 FR

9605).

The April 20, 1998 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 May 12, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Florida International

University, University Park, Miami,

Florida 33199.

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

219, Oyster Creek Nuclear Generating

Station, Ocean County, New Jersey

Date of application for amendment:

December 10, 1997.

Brief description of amendment: The

amendment clarifies sections of the

Technical Specifications that have been

demonstrated to be unclear or

conflicting.

Date of Issuance: June 4, 1998.

Effective date: June 4, 1998, to be

implemented within 30 days.

Amendment No.: 195.

Facility Operating License No. DPR–

16: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 28, 1998 (63 FR 4313).

The Commission’s related evaluation

of this amendment is contained in a

Safety Evaluation dated June 4, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Ocean County Library,

Reference Department, 101 Washington

Street, Toms River, NJ 08753.

Niagara Mohawk Power Corporation,

Docket No. 50–410, Nine Mile Point

Nuclear Station Unit No. 2, Oswego

County, New York

Date of application for amendment:

December 15, 1997, as supplemented by

letter dated April 24, 1998.

Brief description of amendment: This

amendment changes Technical

Specifications 2.1.2 and 3.4.1.1 to revise

the minimum critical power ratio safety

limits for fuel operating cycle 7 for twoloop and single-loop recirculation

operation.

Date of issuance: June 4, 1998.

Effective date: As of the date of

issuance to be implemented before
Federal Register / Vol. 63, No. 116 /Wednesday, June 17, 1998 / Notices 33117

startup of the Unit 2 reactor to begin

fuel operating cycle 7.

Amendment No.: 82.

Facility Operating License No. DPR–

63: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: January 28, 1998 (63 FR 4314).

The April 24, 1998, submittal

provided clarifying information that did

not alter the initial no significant

hazards consideration determination.

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated June 4, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126

Northeast Nuclear Energy Company, et

al., Docket No. 50–336, Millstone

Nuclear Power Station, Unit No. 2, New

London County, Connecticut

Date of application for amendment:

September 2, 1997.

Brief description of amendment: The

amendment corrects several compliance

issues as identified in Licensee Event

Report 97–022–00 ‘‘Technical

Specification Violations’’ dated July 9,

1997, by rewording the text; changing

terminology and numbering; combining

two Technical Specifications (TSs) into

one; changing the allowed outage times;

specifying guidance for entering into TS

3.0.3; changing a definition; changing

surveillance requirments, and updating

the TS Bases section to reflect changes.

Date of issuance: May 26, 1998.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 215.

Facility Operating License No. DPR–

65: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: September 24, 1997 (62 FR

50008).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 26, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, Connecticut, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut

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:

October 15, 1997, as supplemented

January 23 and April 8, 1998.

Brief description of amendment: The

amendment revises the action

statements and the instrumentation trip

setpoint tables in the Technical

Specifications for the reactor trip system

and engineered safety feature actuation

system instrumentation. In addition, the

amendment (1) decreases the reactor

trip setpoint for the reactor coolant

pump low shaft speed (underspeed trip

setpoint) from 95.8 percent to 92.4

percent of rated speed, (2) makes

editorial changes, and (3) changes the

Bases to reflect the new methodology.

Date of issuance: May 26, 1998.

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment No.: 159.

Facility Operating License No. NPF–

49: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: November 19, 1997 (62 FR

61842).

The January 23 and April 8, 1998,

submittals provided clarifying and

additional information that did not

change the scope of the October 15,

1997, application and the initial

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 26, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, Connecticut, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut.

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:

April 7, 1998.

Brief description of amendment: The

amendment replaces the pressurizer

maximum water inventory requirement

with a pressurizer maximum indicated

level requirement. The amendment also

makes editorial changes and modifies

the associated Bases section.

Date of issuance: May 27, 1998.

Effective date: As of the date of

issuance, to be implemented within 30

days.

Amendment No.: 160.

Facility Operating License No. NPF–

49: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: April 23, 1998 (63 FR 20219).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 27, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, Connecticut, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut.

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:

April 14, 1998, as supplemented May 7,

1998, and two letters dated June 4, 1998.

Brief description of amendment: The

amendment changes Technical

Specification 3/4.4.4, Relief Valves, to

ensure that the automatic capability of

the power-operated relief valves

(PORVs) to relieve pressure is

maintained when these valves are

isolated by closure of the block valves.

The amendment also makes editorial

changes, adds PORV surveillance

requirements, and modifies the

associated Bases section.

Date of issuance: June 5, 1998.

Effective date: As of the date of

issuance, to be implemented within 30

days.

Amendment No.: 161.

Facility Operating License No. NPF–

49: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: April 20, 1998 (63 FR 19532).

The May 7, 1998, letter and the two

letters dated June 4, 1998, provide

clarifying information that did not

change the scope of the April 14, 1998,

application and the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated June 5, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, Connecticut, and the
33118 Federal Register / Vol. 63, No. 116 /Wednesday, June 17, 1998 / Notices

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut.

Pacific Gas and Electric Company,

Docket Nos. 50–275 and 50–323, Diablo

Canyon Nuclear Power Plant, Unit Nos.

1 and 2, San Luis Obispo County,

California

Date of application for amendments:

December 23, 1997.

Brief description of amendments: The

amendments changed the combined

Technical Specifications (TS) for the

Diablo Canyon Power Plant (DCPP) Unit

Nos. 1 and 2 to revise TS 3/4.7.1.1,

Table 3.7–1, ‘‘Maximum Allowable

Power Range Neutron Flux High

Setpoint With Inoperable Steam Line

Safety Valves.’’ The power range (PR)

neutron flux high setpoints were

changed based on revised calculational

methodologies for 1, 2, or 3 inoperable

MSSVs per steam generator (SG). The

proposed TS change lowered the PR

neutron flux high setpoints when 2 or

3 MSSV are inoperable per loop such

that the maximum power level allowed

would be within the heat removing

capability of the remaining operable

MSSVs. Although the method for

calculating the maximum power level

allowed when one MSSV per loop is

inoperable was revised, the results were

not and the limit remained the same.

The associated Bases were also revised.

Date of issuance: May 28, 1998.

Effective date: May 28, 1998, to be

implemented within 30 days of

issuance.

Amendment Nos.: Unit 1–125; Unit

2–123.

Facility Operating License Nos. DPR–

80 and DPR–82: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: April 22, 1998 (63 FR 19975).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated May 28, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: California Polytechnic State

University, Robert E. Kennedy Library,

Government Documents and Maps

Department, San Luis Obispo, California

93407.

Public Service Electric & Gas Company,

Docket No. 50–272, Salem Nuclear

Generating Station, Unit No. 1, Salem

County, New Jersey

Date of application for amendment:

March 26, 1998.

Brief description of amendment: The

amendment revises Technical

Specification 3.1.3.3, ‘‘Rod Drop Time,’’

to change the applicability from Mode 3

(hot shutdown) to Modes 1 and 2

(startup and power operation).

Date of issuance: June 4, 1998.

Effective date: As of date of issuance

to be implemented within 60 days.

Amendment No.: 211.

Facility Operating License No. DPR–

70: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: April 22, 1998 (63 FR 19978).

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated June 4, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Salem Free Public Library, 112

West Broadway, Salem, NJ 08079.

Southern Nuclear Operating Company,

Inc., Georgia Power Company,

Oglethorpe Power Corporation,

Municipal Electric Authority of Georgia,

City of Dalton, Georgia, Docket Nos. 50–

321 and 50–366, Edwin I. Hatch Nuclear

Plant, Units 1 and 2, Appling County,

Georgia

Date of application for amendments:

May 30, 1997, as supplemented April 1,

1998.

Brief description of amendments: The

amendments revise the Technical

Specification requirements to reflect a

design modification that changes the

power sources to valves associated with

the low pressure coolant injection mode

of the residual heat removal system.

Date of issuance: June 2, 1998.

Effective date: As of the date of

issuance to be implemented prior to

startup from the next refueling outage

for both units.

Amendment Nos.: Unit 1–211; Unit

2–152.

Facility Operating License Nos. DPR–

57 and NPF–5: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: July 16, 1997 (62 FR 38139).

The April 1, 1998, submittal provided

clarifying information that did not

change the scope of the May 30, 1997,

application and the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated June 2, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Appling County Public

Library, 301 City Hall Drive, Baxley,

Georgia.

Southern Nuclear Power Company, Inc.,

Georgia Power Company, Oglethorpe

Power Corporation, Municipal Electric

Authority of Georgia, City of Dalton,

Georgia, Docket Nos. 50–424 and 50–

425, Vogtle Electric Generating Plant,

Units 1 and 2, Burke County, Georgia

Date of application for amendments:

November 20, 1997, as supplemented by

letter dated April 16, 1998.

Brief description of amendments: The

proposed changes to the Technical

Specifications (TS): (1) Remove the

inequalities applied to the ‘‘Trip

Setpoint’’ column of TS Table 3.3.1–1,

‘‘Reactor Trip System Instrumentation’’

and TS Table 3.3.2–1, ‘‘Engineered

Safety Feature Actuation System

Instrumentation’’ and revise the ‘‘Trip

Setpoint’’ column to read ‘‘Nominal

Trip Setpoint;’’ (2) Add footnotes (n)

and (i) to TS Tables 3.3.1–1 and 3.3.2–

1, respectively, to include criteria for

channel operability, reset, and

calibration tolerance about the trip

setpoint. These footnotes also allow for

the trip setpoint to be set more

conservatively than the Nominal Trip

Setpoint value as necessary in response

to plant conditions; (3) The Allowable

Value for TS Table 3.3.1–1, Function

14.b, Turbine Trip—Turbine Stop Valve

Closure, would be revised from ‘‘[greater

than or equal to] 96.7% open’’ to

‘‘[greater than or equal to] 90% open;’’

(4) Revise footnotes (l) and (m) of TS

Table 3.3.1–1 to refer to Nominal Trip

Setpoint and delete the inequalities

applied to the trip setpoints; (5) Delete

the superscript ‘‘(a)’’ from the ‘‘Trip

Setpoint’’ column on page 6 of 8 of

Table 3.3.1–1; (6) Revise the inequality

for the Engineered Safety Feature

Actuation System Allowable Value for

Steam Line Pressure—Low (Table 3.3.2–

1, Function 1.e) from ‘‘[less than or

equal to]’’ to ‘‘[greater than or equal to];’’

and (7) Revise associated TS Bases to

reflect the TS revisions.

Date of issuance: June 1, 1998.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment Nos.: Unit 1–101; Unit

2–79.

Facility Operating License Nos. NPF–

68 and NPF–81: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: December 31, 1997 (62 FR

68318).

The supplement dated April 16, 1998,

provided clarifying information that did

not change the scope of the November

20, 1997, application and the initial

proposed no significant hazards

determination.
Federal Register / Vol. 63, No. 116 /Wednesday, June 17, 1998 / Notices 33119

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated June 1, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Burke County Library, 412

Fourth Street, Waynesboro, Georgia.

TU Electric Company, Docket Nos. 50–

445 and 50–446, Comanche Peak Steam

Electric Station, Unit Nos. 1 and 2,

Somervell County, Texas

Date of amendment request: May 1,

1995 (TXX–95090).

Brief description of amendments:

These amendments revise section 3/

4.8.1 of the Technical Specifications

(TSs) to reduce the minimum fuel oil

volume requirement during MODES 5

and 6 for an operable emergency diesel

generator (EDG) and allow continued

OPERABLE status of diesel generators

during all MODES for 48 hours with

greater than a 6 day supply of diesel fuel

for a given EDG.

Date of issuance: May 22, 1998.

Effective date: May 22, 1998, to be

implemented within 30 days.

Amendment Nos.: Unit 1—

Amendment No. 60; Unit 2—

Amendment No. 46.

Facility Operating License Nos. NPF–

87 and NPF–89: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: June 21, 1995 (60 FR 32373).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated May 22, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of Texas at

Arlington Library, Government

Publications/Maps, 702 College, P.O.

Box 19497, Arlington, TX 76019.

Washington Public Power Supply

System, Docket No. 50–397, Nuclear

Project No. 2, Benton County,

Washington

Date of application for amendment:

December 4, 1997, as supplemented by

letters dated January 28, 1998, March 3,

1998, March 9, 1998, and April 24,

1998.

Brief description of amendment: The

amendment permits the continued used

of the existing Siemens Power

Corporation minimum critical power

ratio (MCPR) safety limits for WNP–2

Fuel Cycle 14 and changes the ASEA

Brown Boveri (ABB) MCPR safety limit

for single loop operation from 1.08 for

Cycle 13 to 1.09 for Cycle 14.

Date of issuance: May 29, 1998.

Effective date: May 29, 1998, to be

implemented within 30 days from the

date of issuance.

Amendment No.: 154.

Facility Operating License No. NPF–

21: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: January 14, 1998 (63 FR 2284).

The January 28, 1998, March 3, 1998,

March 9, 1998, and April 24, 1998,

supplemental letters provided

additional clarifying information and

did not change the original no

significant hazards consideration. The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated May 29, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Richland Public Library, 955

Northgate Street, Richland, Washington

99352.

Wisconsin Public Service Corporation,

Docket No. 50–305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of application for amendment:

February 25, 1998.

Brief description of amendment: The

amendment revises the Technical

Specifications to implement

performance-based containment leakage

testing under Option B of 10 CFR 50,

Appendix J.

Date of issuance: May 28, 1998.

Effective date: May 28, 1998.

Amendment No.: 136.

Facility Operating License No. DPR–

43: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: April 8, 1998 (63 FR 17237).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 28, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of Wisconsin,

Cofrin Library, 2420 Nicolet Drive,

Green Bay, WI 54311–7001.

Wolf Creek Nuclear Operating

Corporation, Docket No. 50–482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request: May 8,

1998, as supplemented by letter dated

May 11, 1998.

Brief description of amendment: The

amendment adds a new Action

Statement to Technical Specification 3/

4.3.2, Table 3.3–3, Functional Unit 7.b.,

Refueling Water Storage Tank Level—

Low-Low Coincident With Safety

Injection.

Date of issuance: May 28, 1998.

Effective date: May 28, 1998.

Amendment No.: 117.

Facility Operating License No. NPF–

42: The amendment revised the

Technical Specifications.

Public comments requested as to

proposed no significant hazards

consideration: Yes (63 FR 26829 dated

May 14, 1998). The notice provided an

opportunity to submit comments on the

Commission’s proposed no significant

hazards consideration determination.

No comments have been received. The

notice also provided for an opportunity

to request a hearing by June 15, 1998,

but indicated that if the Commission

makes a final no significant hazards

consideration determination any such

hearing would take place after issuance

of the amendment. The Commission’s

related evaluation of the amendment,

finding of exigent circumstances,

consultation with the State of Kansas

and final determination of no significant

hazards consideration are contained in

a Safety Evaluation dated May 28, 1998.

Local Public Document Room

locations: Emporia State University,

William Allen White Library, 1200

Commercial Street, Emporia, Kansas

66801 and Washburn University School

of Law Library, Topeka, Kansas 66621.

Attorney for Licensee: Jay Silberg,

Esq., Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, N.W.,

Washington, D.C. 20037.

NRC Project Director: William H.

Bateman.

Dated at Rockville, Maryland, this 10th day

of June 1998.

For the Nuclear Regulatory Commission.

Elinor G. Adensam,

Acting Director, Division of Reactor Projects—

III/IV, Office of Nuclear Reactor Regulation.

[FR Doc. 98–16012 Filed 6–16–98; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Meeting Between the American Society

for Quality and NRC to Discuss Quality

Assurance Principles

AGENCY: Nuclear Regulatory

Commission.

ACTION: Notice of a meeting between the

American Society for Quality, Energy

and Environmental Division, Power

Production Committee (ASQ EED) and

the Nuclear Regulatory Commission

(NRC) on quality assurance principles of

mutual interest.

SUMMARY: The ASQ EED and the NRC

have met periodically to discuss

technical matters of mutual interest.

Topics at this meeting will cover, codes

and standards, graded QA, and more