Federal Register / Vol. 60, No. 11 / Wednesday, January 18, 1995 / Notices 3669

Nancy Hanks Center, 1100 Pennsylvania

Avenue, NW., Washington, DC 20506.

This meeting is for the purpose of

application evaluation, under the

National Foundation on the Arts and the

Humanities Act of 1965, as amended,

including discussion of information

given in confidence to the Agency by

grant applicants. In accordance with the

determination of the Chairman of

February 8, 1994, these sessions will be

closed to the public pursuant to

subsections (c) (4), (6) and 9(b) of

section 552b of Title 5, United States

Code.

Further information with reference to

this meeting can be obtained from Ms.

Yvonne Sabine, Advisory Committee

Management Office, National

Endowment for the Arts, Washington,

DC 20506, or call (202) 682–5439.

Dated: January 10, 1995.

Yvonne M. Sabine,

Director, Council & Panel Operations,

National Endowment for the Arts.

[FR Doc. 95–1213 Filed 1–17–95; 8:45 am]

BILLING CODE 7537–01–M

NUCLEAR REGULATORY

COMMISSION

Wisconsin Electric Power Company;

Point Beach Nuclear Power Plant, Unit

Nos. 1 and 2; Denial of Amendment to

Facility Operating License and

Opportunity for Hearing

[Docket Nos. 50–266 and 50–301]

The U.S. Nuclear Regulatory

Commission (the Commission) has

denied a request by Wisconsin Electric

Power Company, the licensee, for an

amendment to Facility Operating

Licenses DPR–24 and DPR–27 issued to

the licensee for operation of the Point

Beach Nuclear Plant, Units 1 and 2,

respectively, located in Two Creeks,

Wisconsin. Notice of Consideration of

Issuance of Amendment to Facility

Operating Licensee and Opportunity for

a Hearing was published in the Federal

Register on September 16, 1994 (59 FR

47656).

The licensee submitted the

amendment request to revise Technical

Specification (TS) Section 15.4.2, ‘‘InService Inspection of Safety Class

Components,’’ by incorporating

acceptance criteria to allow steam

generator sleeved tubes with certain

upper sleeve parent tube indications to

remain in service as described in

Westinghouse Electric Corporation’s

report WCAP–14157, ‘‘Technical

Evaluation of Hybrid Expansion Joint

(HEJ) Sleeved Tubes With Indications

Within the Upper Joint Zone.’’

The NRC staff has concluded that the

licensee’s request cannot be granted

because, based on available data,

uncertainties in: the potential locations

of cracking; crack growth rates;

allowable maximum crack size;

potential leakage rates; and the

probability of detection of cracks are too

great to demonstrate that licensing basis

criteria would be satisfied for all normal

and postulated accident conditions. The

licensee was notified of the

Commission’s denial of the proposed

change in a letter of

By February 17, 1995, the licensee

may demand a hearing with respect to

the denial described above. Any person

whose interest may be affected by this

proceeding may file a written petition

for leave to intervene.

A request for hearing or petition for

leave to intervene must be filed with the

Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555–0001, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington, DC, by

the above date.

A copy of any petition should also be

sent to the Office of the General

Counsel, U.S. Nuclear Regulatory

Commission, Washington, DC 20555,

and to Gerald Charnoff, Esq., Shaw,

Pittman, Potts, and Trowbridge, 2300 N

Street, NW., Washington, DC 20037,

attorney for the licensee.

For further details on this action, see

(1) the application for amendment of

August 26, 1994, as supplemented by

letters of September 2, 13, 22, and 29,

1994, October 5, and October 21, 1994,

and (2) the Commission’s letter to the

licensee of January 11, 1995.

These documents 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 room located at the Joseph P.

Mann Library 1516 Sixteenth Street,

Two Rivers, Wisconsin 54241.

Dated at Rockville, Maryland, this 11th day

of January 1995.

For the Nuclear Regulatory Commission.

Leif J. Norrholm,

Project Director, Project Directorate III–3,

Division of Reactor Projects—III/IV, Office of

Nuclear Reactor Regulation.

[FR Doc. 95–1173 Filed 1–17–95; 8:45 am]

BILLING CODE 7590–01–M

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 December

12, 1994, through January 5, 1995. The

last biweekly notice was published on

January 4, 1995 (60 FR 494).

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
3670 Federal Register / Vol. 60, No. 11 / Wednesday, January 18, 1995 / Notices

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 Rules Review and

Directives Branch, Division of Freedom

of Information and Publications

Services, Office of Administration, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555, 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 February 17, 1995, 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 no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington DC, by

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

Union at 1–(800) 248–5100 (in Missouri

1-(800) 342–6700). The Western Union

operator should be given Datagram

Identification Number N1023 and the

following message addressed to (Project

Director): petitioner’s name and

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555, 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
Federal Register / Vol. 60, No. 11 / Wednesday, January 18, 1995 / Notices 3671

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.

Boston Edison Company, Docket No.

50–293, Pilgrim Nuclear Power Station,

Plymouth County, Massachusetts

Date of amendment request:

November 22, 1994.

Description of amendment request:

The proposed amendment would revise

the allowable leak rate for the main

steam isolation valves (MSIVs) from the

current 11.5 standard cubic feet per

hour (scfh) for each valve, to a

maximum combined main steam line

leak rate of 46 scfh.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The operation of Pilgrim Station in

accordance with the proposed Amendment

will not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed amendment does not involve

a change to structures, components, or

systems which would affect the probability of

an accident previously evaluated in the

Pilgrim Updated Final Safety Analysis Report

(UFSAR). The proposed amendment results

in no change in radiological consequences of

the design basis LOCA [loss-of-coolant

accident] as currently analyzed for Pilgrim

Station. These analyses were calculated using

the combined total leakage factor of 46 scfh

for determining acceptance to the regulatory

limits for the offsite, control room, and

Technical Support Center (TSC) doses as

contained in 10CFR100 and 10CFR50,

Appendix A, GDC 19. The proposed change

does not compromise existing radiological

equipment qualification, since the combined

total leakage rate of 46 scfh has been factored

into our existing equipment qualification

analyses for 10 CFR 50.49.

2. The operation of Pilgrim Station in

accordance with the proposed amendment

will not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

There is no modification to the MSIVs or

other plant system or structure associated

with this amendment which could impact

their capability to perform their design

function. The total MSIV leakage rate of 46

scfh is included in the current radiological

analyses for the assessment of dose exposure

following an accident. This proposal changes

the allowable leakage rate from a per valve

to a total combined line leakage acceptance

criteria but does not change the cumulative

allowable value. Therefore, the proposed

change does not create the possibility of a

new or different kind of accident from any

accident previously analyzed.

3. The operation of Pilgrim Station in

accordance with the proposed amendment

will not involve a significant reduction in a

margin of safety.

The allowable leak rate limit specified for

the MSIVs is used to quantify the maximum

amount of bypass leakage assumed in the

LOCA radiological analysis. Results of the

analysis are evaluated against the dose

guidelines contained in GDC [General Design

Criteria] 19 and 10CFR100. The margin of

safety in this context is considered to be the

difference between the calculated dose

exposures and the guidelines provided by the

GDC 19 and 10CFR100. Therefore, since the

maximum allowable leakage for each valve

was assumed and used as the total allowable

leakage for the purpose of calculating

potential dose, the margin of safety is not

affected because the dose levels remain the

same.

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: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts

02360.

Attorney for licensee: W. S. Stowe,

Esquire, Boston Edison Company, 800

Boylston Street, 36th Floor, Boston,

Massachusetts 02199.

NRC Project Director: Walter R.

Butler.

Boston Edison Company, Docket No.

50–293, Pilgrim Nuclear Power Station,

Plymouth County, Massachusetts

Date of amendment request:

November 22, 1994.

Description of amendment request:

The proposed amendment would revise

the mode conditions under which the

Scram Discharge Instrument VolumeScram Trip Bypass in Table 3.2.C.1 is

required to be operable and changes the

associated functional test frequency

from quarterly to once per operating

cycle in Table 4.2.C.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The operation of Pilgrim Station in

accordance with the proposed amendment

will not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change to Table 3.2.C.1, and

the associated change to Table 4.2.C, removes

incorrect reactor modes listed for the Scram

Discharge Instrument Volume (SDIV)—Scram

Trip Bypass function. The Pilgrim control

rod block logic for the SDIV Bypass is not

operable nor is it required by design when in

the Run and Startup modes. The control logic

and the FSAR [Final Safety Analysis Report]

(section 7.2.3.10) specifies SDIV—Scram Trip

Bypass operability only in the Refuel and

Shutdown modes.

This change will not result in any physical

modification or operation of the control rod

block system. The change conforms the

technical specifications to the actual design

of the SDIV Scram Trip Bypass as described

in the FSAR. Changing the functional

surveillance frequency from quarterly to once

per operating cycle also conforms the

technical specifications to the applicable

mode for the function.

The change is classified as an

administrative change because it corrects an

administrative requirement that does not

reflect the logic design. It improves safety by

removing the need to install jumpers during

reactor operations to perform unnecessary

and potentially risky functional

surveillances.

Therefore, because this is an administrative

change, operation of Pilgrim will not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

2. The operation of Pilgrim Station in

accordance with the proposed amendment

will not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

This proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated because it is administrative and

requires no physical alteration of the plant

configuration, changes to setpoints, or

operating parameters.

3. The operation of Pilgrim in accordance

with the proposed amendment will not

involve a significant reduction in a margin of

safety.

The proposed change serves to enhance the

margin of safety by eliminating the potential

for error caused by installing jumpers to the

control logic during reactor operation.

Changing the functional surveillance

frequency from quarterly to once per

operating cycle also enhances the margin of

safety by allowing test performance off-line,

the mode for which the SDIV scram trip

bypass control rod blocks are designed to be

operable.

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: Plymouth Public Library, 11
3672 Federal Register / Vol. 60, No. 11 / Wednesday, January 18, 1995 / Notices

North Street, Plymouth, Massachusetts

02360.

Attorney for licensee: W. S. Stowe,

Esquire, Boston Edison Company, 800

Boylston Street, 36th Floor, Boston,

Massachusetts 02199.

NRC Project Director: Walter R.

Butler.

Boston Edison Company, Docket No.

50–293, Pilgrim Nuclear Power Station,

Plymouth County, Massachusetts

Date of amendment request:

November 22, 1994

Description of amendment request:

The proposed amendment would revise

the suppression chamber water level

operating range, increasing it 2 inches,

and revise the water level recorder

range.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The operation of Pilgrim Station in

accordance with the proposed amendment

will not involve a significant increase in the

probability or consequences of an accident

previously identified.

The probability of an accident is not

increased by this proposed change because

there is no relation between the Suppression

Chamber water level operating range and the

probability of an accident.

The consequences of an accident identified

are not increased. The Suppression Chamber

is an accident mitigating device. Increasing

the water level operating range has been

analyzed and does not significantly increase

the structural loads and the calculated stress

levels remain within Mark 1 Acceptance

Criteria.

We have reviewed the FSAR [Final Safety

Analysis Report] Containment Analyses and

concluded that the safety margin is not

affected. An increase in water level enhances

the Suppression Pool’s ability to mitigate an

accident by providing more water for use by

emergency cooling systems. The higher water

level increases the sink capabilities resulting

in lower torus water temperatures from steam

blowdowns. There is a minor reduction in

the free air volume of the torus which has a

negligible effect on containment post

accident pressures. Therefore, there is no

significant increase in the probability or

consequences of an accident previously

identified.

The change in water level recorder range

does not involve an increase in the

probability or consequence of an accident

because the new recording range accounts for

instrument loop uncertainties and is thus

more conservative than the previous range.

2. The operation of Pilgrim Station in

accordance with the proposed amendment

will not create the possibility of a new or

different kind of accident from any accident

previously analyzed.

An increase in the Suppression Chamber

water level operating range does not create a

new or different kind of accident from any

accident previously analyzed because the

Suppression Chamber is an accident

mitigating device. The Suppression Chamber

serves as the heat sink for any postulated

transient or accident condition when the

primary heat sink (main condenser) is

unavailable and as a source of water for the

Core Standby Cooling Systems. The

structural affects of the increase in water

volume have been analyzed and do not

significantly effect the Mark 1 containment

loads.

Revising the water level recording range is

more conservative than that previously used

and does not create the possibility of a new

or different kind of accident.

3. The operation of Pilgrim Station in

accordance with the proposed amendment

will not involve a significant reduction in the

margin of safety.

Operation with an increased Torus water

level does not affect the structure and

attached piping of the Pilgrim Suppression

Chamber and does not significantly affect the

calculated stress levels; therefore, there is no

significant reduction in the margin of safety.

The change in the water level recording

range is due to replacing the transmitter with

a smaller span. The change from 0 to 32

inches to -7 to +7 inches enhances resolution

and accuracy of the water level instrument

loop.

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: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts

02360.

Attorney for licensee: W. S. Stowe,

Esquire, Boston Edison Company, 800

Boylston Street, 36th Floor, Boston,

Massachusetts 02199.

NRC Project Director: Walter R.

Butler.

Entergy Operations, Inc., Docket Nos.

50–313, Arkansas Nuclear One, Unit

No. 1, Pope County, Arkansas

Date of amendment request: August

30, 1994.

Description of amendment request:

The proposed amendment relocates

refueling cycle specific parameters from

the technical specifications to the Core

Operating Limits Report as per

recommendations promulgated by NRC

Generic Letter 88–16. Additionally, the

amendment adds a 24 hour limit on

operations when only one reactor

coolant pump is operating in each loop.

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:

Criterion 1—Does Not Involve a Significant

Increase in the Probability or Consequences

of an Accident Previously Evaluated.

The relocation of cycle-specific variables

from the Technical Specifications to the Core

Operating Limits Report (COLR) is

considered to be administrative in nature and

has no impact on plant operation or safety.

The Technical Specifications will continue to

require operation within the core operational

limits for each cycle reload as calculated by

the NRC approved reload methodologies. The

values and setpoints placed in the COLR are

addressed in the reload report for each

particular fuel cycle. The reload report

presents the results of evaluations of

accidents addressed in the ANO–1 Safety

Analysis Report. These evaluations

demonstrate that changes in the fuel cycle

design and the corresponding COLR do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The revision of Specification 3.1.1.1.a and

addition of the footnote to Table 2.3–1 result

in additional restrictions on operation with

one reactor coolant pump in each loop with

the reactor critical. This more restrictive

specification limits operation with one

reactor coolant pump in each loop to a 24

hour period when the reactor is critical. This

change incorporates a more restrictive

control and does not affect any previously

analyzed event.

Therefore, this change does not involve a

significant increase in the probability or

consequences of any accident previously.

Criterion 2—Does Not Create the

Possibility of a New or Different Kind of

Accident from any Previously Evaluated.

This relocation of cycle-specific variables

from the Technical Specifications to the

COLR does not create the possibility of a new

or different kind of accident from any

previously analyzed. The cycle-specific

variables will continue to be calculated using

NRC approved methodologies. Technical

Specifications will continue to require

operation within the required core operating

limits and appropriate actions will be taken

if the limits are exceeded. Because plant

operation continues to be limited in

accordance with the values of cycle-specific

parameter limits that are established using

NRC approved methodologies, the

relocations included in this submittal are

considered to be administrative in nature and

have no impact on plant safety as a

consequence.

The revision of Specification 3.1.1.1.a and

addition of the footnote to Table 2.3–1 result

in additional restrictions on operation with

one reactor coolant pump in each loop with

the reactor critical. This more restrictive

specification limits operation with one

reactor coolant pump in each loop to a 24

hour period when the reactor is critical. This

proposed change introduces no new mode of

plant operation.

Therefore, this change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

Criterion 3—Does Not Involve a Significant

Reduction in the Margin of Safety.
Federal Register / Vol. 60, No. 11 / Wednesday, January 18, 1995 / Notices 3673

The proposed relocations are considered to

be administrative in nature and do not

involve a significant reduction in the margin

of safety since they only involve transferring

limits from the Technical Specifications to

the COLR. The values and setpoints placed

in the COLR are addressed in the reload

report for each particular fuel cycle. The

development of limits for future reloads will

continue to conform to methodologies

described in NRC approved documentation.

Each future reload involves a 10CFR50.59

safety review to assure that operation of the

unit within the cycle-specific limits will not

involve a significant reduction in the margin

of safety.

The revision of Specification 3.1.1.1.a and

addition of the footnote to Table 2.3–1 result

in additional restrictions on operation with

one reactor coolant pump in each loop with

the reactor critical. This more restrictive

specification limits operation with one

reactor coolant pump in each loop to a 24

hour period when the reactor is critical. This

change does not involve a significant

reduction in the margin of safety, rather, it

constitutes an additional limitation not

previously included in the Technical

Specifications.

Therefore, this change does not

involve a significant reduction in the

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas

72801.

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, N.W., Washington, D.C.

20005–3502.

NRC Project Director: William D.

Beckner.

Entergy Operations, Inc., et al., Docket

No. 50–416, Grand Gulf Nuclear

Station, Unit 1, Claiborne County,

Mississippi

Date of amendment request:

November 9, 1994.

Description of amendment request:

The proposed amendment revises those

specifications associated with various

engineered safety feature systems

following a design basis fuel handling

accident. The proposed changes affect

conditions where irradiated fuel is

handled in the primary or secondary

containment and when fuel is handled

over the reactor vessel with fuel in the

vessel. These changes are based on a

recent re-analysis of the fuel handling

accident for Grand Gulf Nuclear Station

(GGNS).

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

significantly increase the probability or

consequences of an accident previously

evaluated.

The proposed definition of RECENTLY

IRRADIATED fuel is used to establish

operational conditions where specific

activities represent situations where

significant radioactive releases can be

postulated. These operational conditions are

consistent with the design basis analysis.

Because the equipment affected by the

revised operational conditions is not

considered an initiator to any previously

analyzed accident, inoperability of the

equipment cannot increase the probability of

any previously evaluated accident. The

proposed applicability in conjunction with

existing administrative controls on light

loads, bounds the conditions of the current

design basis fuel handling accident analysis

which concludes that the radiological

consequences are within the acceptance

criteria of NUREG 0800, Section 15.7.4 and

General Design Criteria 19. Therefore, the

proposed changes do not significantly

increase the probability or consequences of

any previously evaluated accident.

Based on the above, the proposed changes

do not significantly increase the probability

or consequences of any accident previously

evaluated.

2. The proposed changes would not create

the possibility of a new or different kind of

accident from any previous analyzed.

The proposed definition is used to

establish operational conditions where

specific activities represent situations where

significant radioactive releases can be

postulated. These operational conditions are

consistent with the design basis analysis. The

proposed changes do not introduce any new

modes of plant operation and do not involve

physical modifications to the plant.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any previous analyzed.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously analyzed.

3. The proposed changes do not involve a

significant reduction in a margin of safety.

The revised definition is used to establish

operational conditions where specific

activities represent situations where

significant radioactive releases can be

postulated. These operational conditions are

consistent with the design basis analysis and

are established such that the radiological

consequences are at or below the current

GGNS licensing limit. Safety margins and

analytical conservatisms have been evaluated

and are well understood. Substantial margins

are retained to ensure that the analysis

adequately bounds all postulated event

scenarios. The proposed change only

eliminates the excess margin from the

analysis. The current margin of safety is

retained.

Specifically, the margin of safety for the

fuel handling accident is the difference

between the 10 CFR 100 limits and the

licensing limit defined by NUREG 0800,

Section 15.7.4. With respect to the control

room personnel doses, the margin of safety is

the difference between the 10 CFR 100 limits

and the licensing limit defined by 10 CFR 50,

Appendix A, Criterion 19 (GDC 19). Excess

margin is the difference between the

postulated doses and the corresponding

licensing limit.

The proposed applicability continues to

ensure that the whole-body and thyroid

doses at the exclusion area and low

population zone boundaries as well as

control room, doses are at or below the

corresponding licensing limit. The margin of

safety is unchanged; therefore, the proposed

changes do not involve a significant

reduction in a margin of safety.

Therefore, the proposed changes 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: Judge George W. Armstrong

Library, 220 S. Commerce Street,

Natchez, Mississippi 39120.

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, NW., 12th Floor,

Washington, DC 20005–3502.

NRC Project Director: William D.

Beckner.

Maine Yankee Atomic Power Company,

Docket No. 50–309, Maine Yankee

Atomic Power Station, Lincoln County,

Maine

Date of amendment request:

December 6, 1994.

Description of amendment request:

The proposed amendment would revise

the Technical Specifications to allow

the use of the Combustion Engineering

sleeving process for repairing steam

generator tubes. (The current

requirement specifies that degraded

steam generator tubes be repaired by

plugging.)

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration. The NRC has reviewed

the licensee’s analysis against the

standard of 10 CFR 59.92(c). The staff’s

review is presented below:

1. The proposed amendment would

not involve a significant increase in the
3674 Federal Register / Vol. 60, No. 11 / Wednesday, January 18, 1995 / Notices

probability or consequences of an

accident previously evaluated.

With the sleeve dimensions,

materials, and connecting joints

designed to the applicable American

Society of Mechanical Engineers Boiler

and Pressure Vessel Code (ASME Code),

the proposed sleeving repair becomes an

in-kind substitution for the steam

generator tube being repaired. The

design criteria for the sleeves conform to

the stress limits and safety margins of

Code Section III. Safety factors of 3

(normal operation) and 1.5 (accident

conditions) were applied to the sleeve

design. Mechanical testing using Code

stress allowables also has been

performed in support of the sleeve

design. Based on the results of vendor

test and analysis programs, the sleeves

fulfill their intended function as leak

tight structural members and meet or

exceed all design criteria.

Evaluation of the steam generator

tubes and proposed sleeves indicates no

detrimental effects on the sleeve or

sleeve-tube assembly from reactor

coolant system flow, reactor or steam

generator coolant chemistry, or thermal

or pressure conditions (including

transients) that may be experienced by

the Maine Yankee plant. Corrosion

testing of sleeve-tube assemblies

indicates no evidence of sleeve or steam

generator tube corrosion considered

detrimental under anticipated service

conditions.

Installation of the proposed sleeves

will be controlled via Combustion

Engineering’s proprietary equipment

and process. The process has been used

24 separate times since 1984 to install

approximately 4100 steam generator

sleeves in nuclear facilities worldwide.

The Maine Yankee steam generator

design has been reviewed and found

compatible with the sleeve installation

equipment and process. Installation of

the proposed sleeves will have no

significant effect on either plant

configuration or operation.

The licensee therefore concludes that

implementation of the proposed change

will not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

2. The proposed amendment would

not create the possibility of a new or

different kind of accident from any

accident previously evaluated.

As discussed above, the structural

integrity, thermal characteristics, and

material properties of the proposed

sleeves are compatible with Maine

Yankee’s steam generators. Therefore,

the functions of the steam generators

will not be significantly affected by

installation of the proposed sleeves. In

addition, the proposed sleeves do not

interact with any other plant systems.

Finally, the continued integrity of

installed sleeves is periodically verified

by the steam generator inspections

required by plant Technical

Specifications.

The licensee therefore concludes that

implementation of the proposed change

will not create a new or different kind

of accident from any accident

previously evaluated.

3. The proposed amendment would

not involve a significant reduction in a

margin of safety.

Repair of degraded steam generator

tubes via the use of the proposed sleeves

has been confirmed to restore the

structural integrity of faulted tubes

under normal operating and postulated

accident conditions. The design safety

factors used for the sleeves are

consistent with ASME Code safety

factors required in the design of Maine

Yankee’s steam generators. The repair

limit for the proposed sleeves is

consistent with that established for

Maine Yankee’s steam generators. The

design of the sleeve-to-tube joint has

been verified by testing to preclude

significant leakage during normal and

postulated accident conditions. Use of

the previously identified design safety

factors design verification testing

assures that margin to safety with

respect to installation of the proposed

sleeves is not significantly different

from the original steam generator tubes.

The licensee therefore concludes that

implementation of the proposed change

would not involve a significant

reduction in a margin of safety.

Based on this review, it appears that

the three standards of 10 CFR 50.92(c)

are satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Wiscasset Public Library, High

Street, P.O. Box 367, Wiscasset, Maine

04578.

Attorney for licensee: Mary Ann

Lynch, Esquire, Maine Yankee Atomic

Power Company, 329 Bath Road,

Brunswick, Maine 04011.

NRC Project Director: Walter R.

Butler.

Niagara Mohawk Power Corporation,

Docket No. 50–220, Nine Mile Point

Nuclear Station Unit No. 1, Oswego

County, New York

Date of amendment request:

December 23, 1994.

Description of amendment request:

The proposed amendment would revise

Technical Specifications (TSs) 2.1.2,

‘‘Fuel Cladding Integrity,’’ 3.6.2/4.6.2,

‘‘Protective Instrumentation,’’ and

associated Bases to extend the

calibration frequency of the reactor

recirculation flow transmitters from

once per quarter to once per operating

cycle and for the square rooters and

summers from once per quarter to once

per year. The proposed amendment

would revise the flow biased average

power range monitor (APRM) scram and

rod block, recirculation flow

comparator, and flow unit upscale

setpoints and the associated Bases of

TSs 2.1.2, 2.2.2, and 3.6.2/4.6.2.

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 operation of Nine Mile Point Unit 1,

in accordance with the proposed

amendment, will not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

The proposed changes extend the

calibration interval for the recirculation flow

square rooters, summers and transmitters and

revise the setpoints for the recirculation flow

upscale and comparator rod block trips. The

associated analytical limits for APRM flow

biased scram and rod block increase by 2%

and 8% respectively. Setpoints are for plant

protective functions (i.e., scram and rod

block) which respond to an accident or

transient. The scram and rod block function

responds to mitigate the consequences of an

accident or transient. Therefore, a change to

the setpoints cannot increase the probability

of these accidents or transients. Likewise,

changes to surveillance intervals for the

protective functions which respond to an

accident or transient cannot increase the

probability. In fact, the proposed increase in

the surveillance intervals reduce the

probability of an inadvertent scram by

reducing the duration that the plant is in the

one-half scram condition.

The new surveillance intervals, setpoints

and allowable setpoint deviations are

calculated using the approved GE [General

Electric Company] setpoint methodology

documented in NEDC–31336. The

methodology in NEDC–31336 provides

assurance that safety system actuation (i.e.,

reactor scram or control rod withdrawal

block) will occur prior to the associated

system parameters (neutron flux and

recirculation flow) exceeding their analytical

limits. Based upon re-evaluation of NMP1

[Nine Mile Point Nuclear Station Unit No. 1]

accidents and transients, it has been shown

that the fuel thermal limits are not

significantly impacted. Therefore, the

consequences of an accident or transient has

not significantly increased.

Thus, plant response to previously

analyzed accidents remains within

previously determined limits. Therefore, the

operation of Nine Mile Point Unit 1, in

accordance with the proposed amendment,

will not involve a significant increase in the

probability or consequences of an accident

previously evaluated.
Federal Register / Vol. 60, No. 11 / Wednesday, January 18, 1995 / Notices 3675

The operation of Nine Mile Point Unit 1,

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 to extend the

calibration frequency do not represent a

physical change to the plant as described in

the NMP1 Final Safety Analysis Report

(Updated). However, this change results in

increasing the analytical limits for the APRM

flow based scram and rod block by 2% and

8% respectively. The proposed changes do

not alter the plant configuration and the

initial conditions used for the design basis

accident analysis are still valid. Thus, no

potential initiating events are created which

would cause any new or different kinds of

accidents. As such, the plant initial

conditions utilized for the design basis

accident analysis are still valid. Therefore,

operation of Nine Mile Point Unit 1 in

accordance with the proposed change will

not create the possibility of a new or different

kind of accident from any previously

assessed.

The operation of Nine Mile Point Unit 1,

in accordance with the proposed

amendment, will not involve a significant

reduction in a margin of safety.

The analytical limits for the APRM flow

biased scram and rod block increase by 2%

and 8% respectively. The trip units in the

APRM and recirculation flow

instrumentation systems will continue to be

calibrated every three months. In addition,

the entire APRM and recirculation flow

instrumentation systems will still be subject

to Instrument Channel Tests every three

months. These tests, together with the

calibration of the flow square rooters and

summers once per year and the flow

transmitters once per operating cycle, will

assure that system reliability and availability

are maintained at their current levels.

Reanalysis of the design basis transients was

performed utilizing these new values. The

results showed that the increase had an

insignificant effect on the consequences of

these events. Therefore, the proposed

amendment will not involve a significant

reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 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: Reference and Documents

Department, Penfield Library, State

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: Michael J. Case

Niagara Mohawk Power Corporation,

Docket No. 50–410, Nine Mile Point

Nuclear Station, Unit 2, Oswego

County, New York

Date of amendment request:

December 13, 1994.

Description of amendment request:

The proposed amendment would revise

Technical Specification Table 3.6.1.2–1,

‘‘Allowable Leak Rates Through Valves

in Potential Bypass Leakage Paths,’’ to

increase the maximum allowable

leakage rate of each of the eight main

steamline isolation valves from 6.0 scfh

to 24.0 scfh.

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 operation of Nine Mile Point Unit 2,

in accordance with the proposed

amendment, will not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

The proposed changes to Technical

Specification Table 3.6.1.2–1 would allow a

maximum leakage of 24.0 scfh for each of the

eight MSIVs [main steamline isolation

valves]. The current Technical Specifications

allow a maximum leakage for an MSIV of 6.0

scfh.

Closure of one or more of the MSIVs at

rated power is a pressure transient for the

reactor coolant pressure boundary. This

pressure transient is evaluated in Section

15.2.4 of the USAR [Updated Safety Analysis

Report]. Closure of MSIV(s), as analyzed in

the USAR, could occur due to manual or

automatic actions. A change to the leakage

limit for the MSIVs does not affect either the

manual or automatic actions that would close

the MSIVs. Therefore, the proposed change to

the table cannot affect the probability of the

closure of one or more MSIVs at rated power.

The radiological evaluation of the DBA–

LOCA [Design Basis Accident—Loss-ofCoolant Accident] incorporates a maximum

leakage of 24.0 scfh for each of the four main

steam lines. In addition, the revised

radiological evaluation includes the impact

of the proposed license amendment currently

under review by the Staff which would

increase the rated operation of NMP2 from

3323 to 3467 megawatts thermal (see NMPC

letter dated July 22, 1993 to the NRC). The

revised radiological evaluation also includes

the impact of License Amendment No. 56

(see NMPC letter dated July 1, 1994 to the

NRC and License Amendment No. 56, dated

August 30, 1994).

The new doses from the revised

radiological analysis for a DBA–LOCA, as

shown in Table 1 [of December 13, 1994,

amendment request], continue to remain

below 10 CFR [Part] 100 guideline values and

GDC [General Design Criterion] 19 limits.

The impact of the increased MSIV leakage on

vital area access and equipment qualification

is minimal and acceptable. Therefore,

operation with the proposed change to the

Technical Specifications will not

significantly increase the consequences of an

accident previously evaluated.

The operation of Nine Mile Point Unit 2,

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 safety function of the MSIVs is to

isolate the main steam lines in a timely

manner to preclude the uncontrolled leakage

of radioactive steam. This is accomplished by

providing the MSIVs with the capability of

rapidly closing automatically in response to

various plant conditions. The increase in the

leakage limit for the MSIVs from 6.0 scfh to

24.0 scfh will not inhibit the MSIVs’ isolation

function. Therefore, operation with the

proposed increase in the MSIV leakage will

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The operation of Nine Mile Point Unit 2,

in accordance with the proposed

amendment, will not involve a significant

reduction in a margin of safety.

The revised radiological analysis follows

the very conservative fuel failure and

instantaneous release assumptions of RG

[Regulatory Guide] 1.3, with the exception of

regulatory position C.1.f as permitted by SRP

[Standard Review Plan] Section 6.5.5,

‘‘Pressure Suppression Pool as a Fission

Product Cleanup.’’ The Staff approved the

use of SRP Section 6.5.5. as part of the

licensing basis of NMP2 in License

Amendment No. 56.

The revised radiological analysis

incorporates the maximum allowable leakage

limit of 24.0 scfh for each of the four main

steam lines. The revised radiological analysis

also includes the impacts of the proposed

power uprate of NMP2 and License

Amendment No. 56. The new doses from the

revised radiological analysis remain below

the Staff acceptance criteria of 10 CFR [Part]

100 guideline values and GDC 19 (see Table

1 [of December 13, 1994, amendment

request]). Therefore, operation with the

proposed changes to the Technical

Specifications will not significantly reduce a

margin of safety.

Accordingly, as determined by the analysis

above, this proposed amendment involves no

significant hazards consideration.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 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: Reference and Documents

Department, Penfield Library, State

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: Michael J. Case.
3676 Federal Register / Vol. 60, No. 11 / Wednesday, January 18, 1995 / Notices

Philadelphia Electric Company, Public

Service Electric and Gas Company,

Delmarva Power and Light Company,

and Atlantic City Electric Company,

Dockets Nos. 50–277 and 50–278, Peach

Bottom Atomic Power Station, Units

Nos. 2 and 3, York County,

Pennsylvania

Date of application for amendments:

August 3, 1994.

Description of amendment request:

The proposed changes would delete a

footnote in the Technical Specifications

(TS) regarding snubber functional

testing frequency and make permanent

the current one-time snubber functional

test frequency of 24 months.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated, because the probability of a

seismic or other dynamic event is

independent of the surveillance period for

snubber tests. The change does not introduce

any failure mechanisms to the previously

considered events. The consequences of an

accident previously evaluated in the SAR

[Safety Analysis Report] is not increased by

the proposed revision to [t]he snubber TS. No

physical changes are being made to the plant.

The snubbers’ role in mitigating the

consequences of an accident is to provide

restraint during seismic or other dynamic

events while permitting the slow movement

of piping and components during heatup and

cooldown. The proposed TS change will not

affect the snubbers ability to continue to

perform this role for the following reasons:

(1) Changing the inspection cycle to 24

months will not reduce the ability of the

functional testing to confirm the operability

of the snubber population. The original

interval of 18 months was selected to

accommodate the need to test snubbers that

were inaccessible during normal operation.

Since snubbers do not require preventative

maintenance during the operating cycle, the

additional time added by a 24 month

operating cycle has minimal impact, if any,

on snubber operability. (2) The requirement

to monitor service life remains part of TS.

The review of snubber service life records is

a documentation review of the snubbers

service life. If a snubber’s service life would

expire prior to the next scheduled review

then the snubber is reconditioned, replaced

or reevaluated to extend its service life. (3)

Snubber functional testing has shown no

failure mechanism which would be

aggravated by an extension of the test interval

to 24 months. A historical search of

completed snubber functional STs was

completed. The historical search indicated

that even though the snubbers did not always

meet the initial screening functional test

criteria of the ST, the piping system was

operable based on an engineering evaluation

and there was no evidence of a time

dependent failure mechanism. To ensure the

snubber remains operational during the next

operating cycle, snubbers not meeting the

screening ST acceptance criteria are either

replaced or reconditioned.

(2) The proposed change does not create

the possibility of a new or different kind of

accident from any previously evaluated

because the proposed change does not

involve operational procedure or physical

changes to the plant. Since snubbers will

continue to meet their design basis of

protecting the piping and equipment during

dynamic events, the possibility of a different

type of accident will not be created.

(3) The proposed change does not involve

a significant reduction in a margin of safety.

There may be a slight increase, if any, in the

possibility of undetected snubber failures

because of the increase in the interval of

functional testing for snubbers; however, the

historical data of previous snubber functional

surveillance testing and the supporting

engineering evaluations indicate that on

those occasions where snubbers did not meet

initial surveillance testing requirements, the

piping systems were all operable. Therefore,

the probability of occurrence of a

malfunction of equipment is minimal and

equipment important to safety (ITS) that use

snubbers will continue to meet design

requirements and the margin of safety will be

unaffected.

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: Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Education

Building, Walnut Street and

Commonwealth Avenue, Box 1601,

Harrisburg, Pennsylvania 17105.

Attorney for Licensee: J. W. Durham,

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

Counsel, Philadelphia Electric

Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101.

NRC Project Director: John F. Stolz.

Public Service Electric & Gas Company,

Docket No. 50–354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of amendment request:

September 29, 1994.

Description of amendment request:

This amendment requests revision of

Table 4.3.6–1 ‘‘Control Rod Block

Instrumentation Surveillance

Requirements.’’ The channel calibration

frequencies for the Source Range

Monitor (SRM) and the Intermediate

Range Monitor (IRM) would be changed

as follows: the up-scale and the downscale trip functions on each instrument

would be changed from Note ‘‘SA’’,

once-per-184 days to note ‘‘R’’, onceper-refuel interval.

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 not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

The proposed changes involve no

hardware changes, no changes to the

operation of any systems or components, and

no changes to existing structures. The

revision of channel calibration frequencies

for the SRM and IRM trip function portion

of the control rod block instrumentation

represent changes that do not affect plant

safety and do not alter existing accident

analyses.

2. Will not create the possibility of a new

or different kind of accident from any

previously evaluated.

The proposed changes are procedural in

nature concerning the calibration frequency

of instrumentation that have historically

shown little set point drift. The channel

calibration methodology for the SRM and

IRM control rod block trip functions remain

unchanged. The proposed changes while

slightly increasing the possibility of an

undetected instrument error will not create a

new or unevaluated accident or operating

condition.

3. Will not involve a significant reduction

in a margin of safety.

The proposed changes are in accordance

with recommendations provided by the NRC

regarding the improvement of Technical

Specifications. These changes will result in

the perpetuation of current safety margins

while reducing regulatory burden and

decreasing equipment degradation.

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, New Jersey

08070.

Attorney for licensee: M. J.

Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW.,

Washington, DC 20005–3502.

NRC Project Director: John F. Stolz.

Wisconsin Public Service Corporation,

Docket No. 50–305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of amendment request:

December 16, 1994.
Federal Register / Vol. 60, No. 11 / Wednesday, January 18, 1995 / Notices 3677

Description of amendment request:

The proposed amendment would revise

Kewaunee Nuclear Power Plant (KNPP)

Technical Specification (TS) 3.4 by

removing the Limiting Conditions for

Operation (LCOs) for the Turbine

Overspeed Protection System (TOPS).

Tables TS 4.1–1 and TS 4.1–3 would

also be revised to remove the

surveillance requirements for the TOPS

instrumentation and turbine valves. The

TOPS and related requirements would

be relocated to the Updated Safety

Analysis Report (USAR).

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:

Significant Hazards Determination for

Proposed Changes to TS 3.4.c and Table TS

4.1–1 and Associated Bases Changes

In accordance with 10 CFR Part 50, Section

50.91 and using the standards provided in

Section 50.92, the proposed change has been

reviewed to determine that no significant

hazards exist as a result of this change. The

analysis showed:

(1) The proposed amendment will not

involve a significant increase in the

probability or consequence of an accident

previously evaluated.

The purpose of the Turbine Overspeed

Protection System (TOPS) is to prevent an

overspeed event, which is a precursor to a

potential turbine-generated missile. Neither

Transient Analyses nor Design Basis

Accidents (DBAs) evaluated in the accident

analyses contained in Chapter 14 of the

Kewaunee Nuclear Power Plant (KNPP)

Updated Safety Analysis Report (USAR)

assume operation of the TOPS. The

calculations and probabilities associated with

USAR section 14.2.7, ‘‘Turbine Missile

Damage to the Spent Fuel Pool,’’ are not

affected by this amendment. This

amendment does not implement physical

changes to the plant and does not change the

KNPP’s existing requirements. As a result,

this change will not increase the probability

of a previously evaluated accident.

The purpose of the TOPS is preventative

and it serves no function to mitigate the

consequences of any accident previously

evaluated. Therefore, removing the

requirements associated with the TOPS from

the TSs will not affect the consequences of

an accident previously evaluated.

(2) The proposed amendment will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

This amendment does not involve any

changes in the operational characteristics of

the surveillance tests and will impose no

new requirements. This change will simply

relocate the same testing requirements from

the KNPP Technical Specifications to the

KNPP USAR. Since this change is

administrative in nature, it will not create a

new or different kind of accident from any

accident previously evaluated.

(3) The proposed amendment will not

involve a significant reduction in the margin

of safety.

KNPP’s USAR section 14.2.7, ‘‘Turbine

Missile Damage to the Spent Fuel Pool,’’ will

not be affected by this amendment.

Relocating the TOPS and related

requirements is a change that is

administrative in nature and does not alter

the intent of any requirements. Therefore it

can be concluded that this change will not

involve a significant reduction in the margin

of safety.

Significant Hazards Determination for

Proposed Change to Table TS 4.1–3 and

associated Basis Change

In accordance with 10 CFR Part 50, Section

50.91 and using the standards provided in

Section 50.92, the proposed change has been

reviewed to determine that no significant

hazards exist as a result of this change. The

analysis showed:

(1) The proposed amendment will not

involve a significant increase in the

probability or consequence of an accident

previously evaluated.

This amendment does not involve any

changes in the operation or frequency of the

turbine valve tests. This amendment will

simply relocate the turbine valve testing

requirements from the Kewaunee Nuclear

Power Plant’s (KNPP’s) Technical

Specifications (TSs) to the Updated Safety

Analysis Report (USAR). This change is

administrative in nature and therefore will

not involve a significant increase in the

probability or consequence of an accident

previously evaluated.

(2) The proposed amendment will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

This amendment is administrative in

nature and will not change any requirements.

This change will simply relocate the

requirements from the KNPP TSs to the

USAR. The purpose of the turbine stop and

governor valves is to control steam flow to

the turbine. This amendment will not

adversely affect the steam flow control

capability of the turbine valves. Therefore,

this change will not create the possibility of

a new or different type of accident from any

accident previously evaluated.

(3) The proposed amendment will not

involve a significant reduction in the margin

of safety.

This amendment will simply relocate the

existing turbine valve testing requirements

and will not result in any changes to the

requirements. The KNPP will continue to

follow the recommendations of WCAP 11525,

‘‘Probabilistic Evaluation of Reduction in

Turbine Valve Test Frequency.’’ As a result,

KNPP will continue to maintain acceptably

low probabilities of turbine valve failure.

Since the same requirements still exist and

turbine valve testing will continue to be

consistent with the recommendations of

WCAP 11525, this amendment will not

involve a significant decrease in the margin

of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the 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

Library Learning Center, 2420 Nicolet

Drive, Green Bay, Wisconsin 54301.

Attorney for licensee: Bradley D.

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

Box 1497, Madison, Wisconsin 53701–

1497.

NRC Project Director: Leif J.

Norrholm.

Wolf Creek Nuclear Operating

Corporation, Docket No. 50–482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request:

December 12, 1994.

Description of amendment request:

This amendment request proposes

revising Technical Specifications

4.7.1.2.1.b.1 and 4.7.1.2.1.b.2 to clarify

the surveillance requirements for

verifying the correct required position

for the valves in the auxiliary feedwater

system.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes do not affect the

ability of the auxiliary feedwater system to

perform its intended safety function. The

changes are administrative in nature since

they merely clarify the demonstration of

operability required in the surveillance

requirements.

2. The proposed changes do not create the

possibility of a new or different kind of

accident from any previously evaluated.

There are no new failure modes or

mechanisms associated with the proposed

changes. The changes are administrative

changes to remove confusion when

performing surveillance requirements to

demonstrate operability.

3. The proposed changes do not involve a

significant reduction in the margin of safety.

These proposed changes do not effect [sic]

any technical specification margin of safety.

The changes only provide clarification for

performance of surveillance 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.
3678 Federal Register / Vol. 60, No. 11 / Wednesday, January 18, 1995 / Notices

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: Theodore R.

Quay

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.

GPU Nuclear Corporation, et al.,

Docket No. 50–219, Oyster Creek

Nuclear Generating Station, Ocean

County, New Jersey

Date of amendment request:

November 25, 1994.

Description of amendment request:

The proposed amendment would revise

Technical Specification 5.3.1.E to allow

2645 fuel assemblies to be stored in the

fuel pool. This is an increase of 45 fuel

assemblies from the current limit of

2600. The 45 additional storage

locations currently exist in the racks in

the fuel pool. They were included in the

re-racking project allowed by License

Amendment No. 76 but were not

incorporated in the Technical

Specifications since, at the time, it was

believed they would not be needed.

Date of publication of individual

notice in Federal Register: December 20,

1994 (59 FR 65542).

Expiration date of individual notice:

January 19, 1995.

Local Public Document Room

location: Ocean County Library,

Reference Department, 101 Washington

Street, Toms River, NJ 08753.

Nebraska Public Power District, Docket

No. 50–298, Cooper Nuclear Station,

Nemaha County, Nebraska

Date of amendment request:

December 22, 1994.

Description of amendment request:

The proposed amendment is a Line Item

Technical Specifications Improvement

and would revise the Cooper Nuclear

Station Technical Specifications,

definition 1.0.J, concerning entering an

operational condition consistent with

the wording proposed in NRC Generic

Letter 87–09, ‘‘Sections 3.0 and 4.0 of

the Standard Technical Specifications

on the Applicability of Limiting

Conditions for Operation and

Surveillance Requirements,’’ dated June

4, 1987.

Date of individual notice in the

Federal Register: January 3, 1995 (60 FR

153).

Expiration date of individual notice:

February 2, 1995.

Local Public Document Room

location: Auburn Public Library, 118

15th Street, Auburn, Nebraska 68305.

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.

Carolina Power & Light Company,

Docket No. 50–261, H. B. Robinson

Steam Electric Plant, Unit No. 2,

Darlington County, South Carolina

Date of application for amendment:

August 11, 1994.

Brief description of amendment: The

amendment deletes the requirement to

perform a 5-year interval hydrostatic

test on the auxiliary coolant system

critical headers from TS Section 4.1.3,

Table 4.1–3, Item 11.

Date of issuance: December 28, 1994.

Effective date: December 28, 1994.

Amendment No.: 155.

Facility Operating License No. DPR–

23. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: November 23, 1994 (59 FR

60379).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated December 28,

1994. No significant hazards

consideration comments received: No.

Local Public Document Room

location: Hartsville Memorial Library,

147 West College, Hartsville, South

Carolina 29550

Carolina Power & Light Company,

Docket No. 50–261, H.B. Robinson

Steam Electric Plant, Unit No. 2,

Darlington County, South Carolina

Date of application for amendment:

July 28, 1994.

Brief description of amendment: The

amendment allows an increased limit

for fuel enrichment. The changes allow

for the storage of fuel with an

enrichment not to exceed 4.95 + 0.05 w/

o U–235 in the new and spent fuel

storage racks.

Date of issuance: January 5, 1995.

Effective date: January 5, 1995.

Amendment No. 156.

Facility Operating License No. DPR–

23. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: August 31, 1994 (59 FR

45018).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated January 5, 1995.

No significant hazards consideration

comments received: No.
Federal Register / Vol. 60, No. 11 / Wednesday, January 18, 1995 / Notices 3679

Local Public Document Room

location: Hartsville Memorial Library,

147 West College, Hartsville, South

Carolina 29550.

Duke Power Company, Docket Nos. 50–

369 and 50–370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg

County, North Carolina

Date of application for amendments:

October 4, 1994.

Brief description of amendments: The

amendments revised the Technical

Specification to eliminate a compliance

conflict when swapping the Centrifugal

Changing (NV) pumps in Modes 4, 5,

and 6. In eliminating the conflict, this

amendment permits flexibility in the

operation of the NV pumps during unit

startup without a safety concern.

Date of issuance: November 17, 1994.

Effective date: To be implemented

within 30 days from the date of

issuance.

Amendment Nos.: 152 and 134.

Facility Operating License Nos. NPF–

9 and NPF–17: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: October 13, 1994 (59 FR

52003).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated November 17,

1994.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Atkins Library, University of

North Carolina, Charlotte (UNCC

Station), North Carolina 28223

Duke Power Company, Docket Nos. 50–

269, 50–270, and 50–287, Oconee

Nuclear Station, Units 1, 2, and 3,

Oconee County, South Carolina

Date of application of amendments:

November 11, 1993, as supplemented

February 23, April 12, and July 29,

1994.

Brief description of amendments: The

amendments reflect the consolidation of

the Quality Verification Department

with the Nuclear Generation

Department that realigned the Nuclear

Safety Review Board to report to the

Senior Nuclear Officer, change an

organizational unit term from ‘‘group’’

to ‘‘division,’’ modify titles of positions

designated to approve modifications,

clarify the responsibilities of the Safety

Assurance Manager, and delete the

requirement to perform an annual fire

protection audit.

Date of Issuance: January 4, 1995.

Effective date: To be implemented

within 30 days from the date of

issuance.

Amendment Nos.: 208, 208, and 205.

Facility Operating License Nos. DPR–

38, DPR–47, and DPR–55: The

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 5, 1994 (59 FR 619).

The February 23, April 12 and July 29,

1994, letters provided clarifying

information that did not change the

scope of the November 11, 1993,

application or the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated January 4, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Oconee County Library, 501

West South Broad Street, Walhalla,

South Carolina 29691.

Duquesne Light Company, et al., Docket

No. 50–412, Beaver Valley Power

Station, Unit 2, Shippingport,

Pennsylvania

Date of application for amendment:

February 16, 1994.

Brief description of amendment: This

amendment deletes the Appendix B

Section 4.2.2 requirement to perform

infrared aerial photography every other

year.

Date of issuance: January 5, 1995.

Effective date: January 5, 1995.

Amendment No: 65.

Facility Operating License No. NPF–

73. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 6, 1994 (59 FR 34663).

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated January 5, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa,

Pennsylvania 15001.

Entergy Operations, Inc., Docket No.

50–368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of application for amendment:

November 29, 1994, as supplemented by

letters dated December 20 and 21, 1994.

Brief description of amendment: The

amendment deleted the requirement to

perform the full complement of steam

generator surveillances as outlined in

the technical specifications (TSs) when

the steam generators are subjected to

special inspections that are in addition

to the periodic inspections required by

the TSs. This amendment is applicable

only to the special steam generator

inspection scheduled for January 1995.

Date of issuance: January 5, 1995.

Effective date: January 5, 1995.

Amendment No.: 158.

Facility Operating License No. NPF–6.

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: December 5, 1994 (59 FR

62416). The additional information

contained in the supplemental letters

dated December 20 and 21, 1994, was

clarifying in nature and thus, within the

scope of the initial notice and did not

affect the staff’s proposed no significant

hazards consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated January 5, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas

72801.

Entergy Operations, Inc., System

Energy Resources, Inc., South

Mississippi Electric Power Association,

and Mississippi Power & Light

Company, Docket No. 50–416, Grand

Gulf Nuclear Station, Unit 1, Claiborne

County, Mississippi

Date of application for amendment:

June 17, 1994, as supplemented by letter

dated August 17, 1994.

Brief description of amendment: The

amendment removed License Condition

2.C.(25)(b) and Attachment 2 to Facility

Operating License No. NPF–29,

‘‘Transamerica Delaval Inc. (TDI) Diesel

Generator Maintenance and

Surveillance Requirements (NUREG–

1216, August 1985).’’

Date of issuance: January 4, 1995.

Effective date: January 4, 1995.

Amendment No: 114.

Facility Operating License No. NPF–

29. Amendment revises the License.

Date of initial notice in Federal

Register: September 14, 1994 (59 FR

47167).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated January 4, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Judge George W. Armstrong

Library, 220 S. Commerce Street,

Natchez, Mississippi 39120.

Florida Power and Light Company, et

al., Docket No. 50–389, St. Lucie Plant,

Unit No. 2, St. Lucie County, Florida

Date of application for amendments:

September 23, 1993 and clarified July

25, 1994. The July 25, 1994 submission

did not change the amendment

described in the initial Federal Register

notice.
3680 Federal Register / Vol. 60, No. 11 / Wednesday, January 18, 1995 / Notices

Brief description of amendments: This

amendment makes changes to Technical

Specification 6.2.3, Independent Safety

Engineering group. The change

maintains the requirement to perform

independent technical reviews while

providing increased flexibility to

accomplish this function.

Date of Issuance: December 22, 1994.

Effective Date: December 22, 1994.

Amendment Nos.: 69.

Facility Operating License Nos. DPR–

67 and NPF–16: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: October 27, 1993 (58 FR

57851). The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

December 22, 1994.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort

Pierce, Florida 34954–9003.

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:

October 20, 1994.

Brief description of amendments:

These amendments relocate the diesel

fuel oil testing program requirements to

Technical Specifications (TS) Section 6

and to the Bases section of the TS. Also

added were actions statements to

address diesel fuel oil which does not

meet the program limits.

Date of issuance: December 28, 1994.

Effective date: December 28, 1994.

Amendment Nos. 169 and 163.

Facility Operating Licenses Nos. DPR–

31 and DPR–41: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: November 9, 1994 (59 FR

55870).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated December 28,

1994.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Florida International

University, University Park, Miami,

Florida 33199.

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:

October 20, 1994.

Brief description of amendments: The

amendments remove the schedule for

withdrawal of reactor vessel material

specimens from the Technical

Specifications as discussed in Generic

Letter 91–01.

Date of issuance: December 28, 1994.

Effective date: December 28, 1994.

Amendment Nos. 170 and 164.

Facility Operating Licenses Nos. DPR–

31 and DPR–41: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: November 23, 1994 (59 FR

60381).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated December 28,

1994.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Florida International

University, University Park, Miami,

Florida 33199.

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:

September 13, 1994, as supplemented

by letter dated December 6, 1994.

Brief description of amendments: The

amendments replace Containment

Systems Technical Specification (TS)

3.6.2.2 for the Spray Additive System

with a new Emergency Core Cooling

System (ECCS) TS 3.5.5 for the ECCS

Recirculation Fluid pH Control System.

Date of issuance: January 5, 1995.

Effective date: Phase I to be

implemented following Unit 2 Cycle 4

refueling outage; Phase II to be

implemented following Unit 1 Cycle 6

refueling outage.

Amendment Nos.: 77 and 56 Phase 1;

78 and 57 Phase II.

Facility Operating License Nos. NPF–

68 and NPF–81: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: October 26, 1994 (59 FR

53840). The December 6, 1994, letter

provided clarifying information that did

not change the initial proposed no

significant hazards consideration.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated January 5, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Burke County Library, 412

Fourth Street, Waynesboro, Georgia

30830.

GPU Nuclear Corporation, et al.,

Docket No. 50–219, Oyster Creek

Nuclear Generating Station, Ocean

County, New Jersey

Date of application for amendment:

October 19, 1991, as supplemented

March 9, April 27, and December 15,

1994.

Brief description of amendment: The

amendment establishes additional

requirements for the availability of

Local Power Range Monitors (LPRMs)

associated with the Average Power

Range Monitoring (APRM) system.

These additional requirements further

restrict the allowable number of out of

service LPRM/APRM detectors in order

to ensure a sufficient response to

regional thermal hydraulic oscillations

in the reactor core to prevent violation

of the Minimum Critical Power Ratio

(MCPR) safety limit. The amendment

also identifies a lower bound MCPR

operating limit for each cycle as

identified in the Core Operating Limits

Report. This limit shall be greater than

or equal to 1.47.

Date of Issuance: December 29, 1994.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 176.

Facility Operating License No. DPR–

16. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: November 13, 1991 (56 FR

57697). The March 9, April 27, and

December 15, 1994, letters provided

clarifying information that did not

change the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of this amendment is contained in a

Safety Evaluation dated December 29,

1994.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Ocean County Library,

Reference Department, 101 Washington

Street, Toms River, NJ 08753.

Houston Lighting & Power Company,

City Public Service Board of San

Antonio, Central Power and Light

Company, City of Austin, Texas, Docket

Nos. 50–498 and 50–499, South Texas

Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request: July 18,

1994.

Brief description of amendments: The

amendments revised TS Table 4.3–1,

Reactor Trip System Instrumentation

Surveillance Requirements; TS 3.3.4,

Turbine Governor Valves; and TS
Federal Register / Vol. 60, No. 11 / Wednesday, January 18, 1995 / Notices 3681

3.7.1.2, Turbine Driven Auxiliary

Feedwater Pump, to remove one-time

amendments that are no longer

necessary. In addition, six minor

editorial changes were made.

Date of issuance: December 27, 1994.

Effective date: December 27, 1994, to

be implemented within 31 days of

issuance.

Amendment Nos.: Unit 1—

Amendment No. 67; Unit 2—

Amendment No. 56.

Facility Operating License Nos. NPF–

76 and NPF–80. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: August 31, 1994 (59 FR

45024).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated December 27,

1994.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Wharton County Junior

College, J.M. Hodges Learning Center,

911 Boling Highway, Wharton, Texas

77488.

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

Duane Arnold Energy Center, Linn

County, Iowa

Date of application for amendment:

June 30, 1994, as supplemented

November 10, 1994.

Brief description of amendment: The

proposed amendment would add

Operability Requirements, Limiting

Conditions for Operations (LCO) and

Surveillance Requirements for the

Control Building Chillers.

Date of issuance: December 29, 1994.

Effective date: Date of issuance, to be

implemented within 120 days.

Amendment No.: 205.

Facility Operating License No. DPR–

49. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 3, 1994 (59 FR 39592).

The additional information contained in

the supplemental letter dated November

10, 1994, was clarifying in nature and

did not change the NRC staff’s initial

proposed no significant hazards

consideration determination. The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated December 29, 1994.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Cedar Rapids Public Library,

500 First Street, SE., Cedar Rapids, Iowa

52401.

Indiana Michigan Power Company,

Docket Nos. 50–315 and 50–316, Donald

C. Cook Nuclear Plant, Unit Nos. 1 and

2, Berrien County, Michigan

Date of application for amendments:

November 15, 1993.

Brief description of amendments: The

amendments make various

administrative and editorial changes to

the Technical Specifications.

Date of issuance: December 30, 1994.

Effective date: December 30, 1994.

Amendment Nos.: 186 and 172.

Facility Operating License Nos. DPR–

58 and DPR–74. Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: December 22, 1993 (58 FR

67849).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated December 30,

1994.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Maud Preston Palenske

Memorial Library, 500 Market Street, St.

Joseph, Michigan 49085.

Indiana Michigan Power Company,

Docket Nos. 50–315 and 50–316, Donald

C. Cook Nuclear Plant, Unit Nos. 1 and

2, Berrien County, Michigan

Date of application for amendments:

July 19, 1994.

Brief description of amendments: The

amendments remove the specific

requirements for Types A, B, and C

containment leakage rate tests from the

Technical Specifications and replace

these requirements with a requirement

to perform Types A, B, and C testing in

accordance with Appendix J to 10 CFR

Part 50.

Date of issuance: January 5, 1995.

Effective date: January 5, 1995.

Amendment Nos.: 187/173.

Facility Operating License Nos. DPR–

58 and DPR–74. Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: September 28, 1994 (59 FR

49430).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated January 5, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Maud Preston Palenske

Memorial Library, 500 Market Street, St.

Joseph, Michigan 49085.

Indiana Michigan Power Company,

Docket Nos. 50–315 and 50–316, Donald

C. Cook Nuclear Plant, Unit Nos. 1 and

2, Berrien County, Michigan

Date of application for amendments:

November 15, 1993, and supplemented

October 7, 1994.

Brief description of amendments: The

amendments replace the current

Technical Specification testing

requirements for the Event V reactor

coolant system pressure isolation valves

with the requirements from ASME

Boiler and Pressure Vessel Code,

Section XI.

Date of issuance: January 5, 1995.

Effective date: January 5, 1995.

Amendment Nos.: 188/174.

Facility Operating License Nos. DPR–

58 and DPR–74. Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: January 5, 1994 (59 FR 623).

At the request of the NRC, the

licensee submitted the October 7, 1994,

supplement to clarify the new

requirements. This supplement did not

change the NRC’s initial proposed no

significant hazards considerations

finding; therefore, renoticing was not

warranted.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated January 5, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Maud Preston Palenske

Memorial Library, 500 Market Street, St.

Joseph, Michigan 49085.

Niagara Mohawk Power Corporation,

Docket No. 50–220, Nine Mile Point

Nuclear Station, Unit 1, Oswego

County, New York

Date of application for amendment:

August 26, 1994.

Brief description of amendment: The

amendment revises Technical

Specification 4.3.3.c.(1) to permit a onetime extension of the second 10-year

service period for the primary

containment integrated leakage rate

(Type A) test. The one-time extension

permits delaying the third Type A test

of the second 10-year service period

from the 1995 refueling outage until the

1997 refueling outage. This delay will

result in an interval of approximately 46

months between the second and third

Type A tests of the second 10-year

service period.

Date of issuance: December 29, 1994.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 151.
3682 Federal Register / Vol. 60, No. 11 / Wednesday, January 18, 1995 / Notices

Facility Operating License No. NPF–

63: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: September 28, 1994 (59 FR

49431). The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

December 29, 1994.

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:

April 22, 1994.

Brief description of amendment: The

amendment changes Table 3.3–9 of the

Technical Specifications by modifying

the indicated measurement range for the

neutron flux monitor on the remote

shutdown panel. The amendment also

includes some corrections of

typographical errors in the Technical

Specifications.

Date of issuance: December 20, 1994.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 183.

Facility Operating License No. DPR–

65. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: May 25, 1994 (59 FR 27059).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated December 20,

1994.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Learning Resources Center,

Thames Valley State Technical College,

574 New London Turnpike, Norwich,

Connecticut 06360.

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:

July 22, 1994.

Brief description of amendment: The

amendment (1) changes the title of

Figure 3.1–5 to be consistent with the

applicable Limiting Condition For

Operation (LCO), (2) relocates the

Chemical and Volume Control System

(CVCS) valve position requirements to

the Reactivity Control Systems—

Shutdown Margin specifications, and

(3) consolidates action statements to be

expressed in the LCOs rather than in

Surveillance Requirements. The

amendment also clarifies the

requirements for calculating the heat

flux hot channel factor FQ(z) when using

the base load option.

Date of issuance: December 29, 1994.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 99.

Facility Operating License No. NPF–

49. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 31, 1994 (59 FR

45029).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated December 29,

1994.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, Thames Valley Campus, 574

New London Turnpike, Norwich, CT

06360.

Northeast Nuclear Energy Company, et

al., Docket No. 50–423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of application for amendment:

June 2, 1994, as supplemented August

25, 1994.

Brief description of amendment: The

amendment revises the Technical

Specifications (TS) to remove expired

one-time extensions of surveillance,

removes an obsolete definition of

charging pump operability, and

incorporates 11 line item improvements

in accordance with the guidance

provided in Generic Letter 93–05.

Several editorial changes have been

made to renumber TS pages and delete

the blank pages from the TS.

Date of issuance: January 3, 1995.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 100.

Facility Operating License No. NPF–

49. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: September 14, 1994, (59 FR

47170).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated January 3, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, Thames Valley Campus, 574

New London Turnpike, Norwich, CT

06360.

Northern States Power Company,

Docket Nos. 50–282 and 50–306, Prairie

Island Nuclear Generating Plant, Unit

Nos. 1 and 2, Goodhue County,

Minnesota

Date of application for amendments:

October 3, 1994, as supplemented

November 30, 1994.

Brief description of amendments: The

amendments revise Prairie Island

Nuclear Generating Plant Technical

Specification 4.6, ‘‘Periodic Testing of

Emergency Power Systems.’’

Specifically, the amendments modify

the emergency diesel generator (EDG)

24-hour load test requirements to

provide an indicated load range of 103–

110 percent of the continuous rating.

These amendments also rephrase

various EDG test requirements to

provide clarity and delete the

requirements to verify that the autoconnected loads do not exceed 3000

kilowatts (Unit 2 5100 kilowatts).

Date of issuance: January 5, 1995.

Effective date: January 5, 1995, with

full implementation within 30 days.

Amendment Nos.: 113 and 106.

Facility Operating License Nos. DPR–

42 and DPR–60. Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: November 9, 1994 (59 FR

55877).

The November 30, 1994, request

provided additional clarification that

was within the scope of the initial

notice and did not affect the staff’s

proposed no significant hazards

consideration findings.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated January 5, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Minneapolis Public Library,

Technology and Science Department,

300 Nicollet Mall, Minneapolis,

Minnesota 55401.

Omaha Public Power District, Docket

No. 50–285, Fort Calhoun Station, Unit

No. 1, Washington County, Nebraska

Date of amendment request: October

7, 1994.

Brief description of amendment: The

amendment (1) deletes the surveillance

requirements contained in Technical

Specification (TS) 3.6(3)a for the raw

water backup valves to the containment

cooling coils, (2) deletes the

surveillance requirements in TS 3.2,

Table 3–5, item 6, for raw water valves,

and (3) revises the basis of TS 2.4 to

reflect these changes.
Federal Register / Vol. 60, No. 11 / Wednesday, January 18, 1995 / Notices 3683

Date of issuance: December 29, 1994.

Effective date: December 29, 1994.

Amendment No.: 166.

Facility Operating License No. DPR–

40. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: November 9, 1994 (59 FR

55879).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated December 29,

1994.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska

68102.

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 application for amendments:

April 5, 1994.

Brief description of amendments:

These amendments delete the frequency

requirements for a number of audits

listed under Technical Specification

6.5.2.8 and also remove the audit

requirements for the Emergency Plan

and the Security Plan since these

requirements have been added to the

respective plan documents. The TS

changes included in the April 5, 1994,

application were approved with the

exception of those related to the fire

protection and loss prevention

programs. These proposed changes are

still under evaluation by the staff and

will be addressed in a future safety

evaluation.

Date of issuance: December 22, 1994.

Effective date: As of the date of

issuance, to be implemented within 30

days.

Amendment Nos.: 137 and 107.

Facility Operating License Nos. NPF–

14 and NPF–22. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: May 25, 1994 (59 FR 27061).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated December 22,

1994.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Osterhout Free Library,

Reference Department, 71 South

Franklin Street, Wilkes-Barre,

Pennsylvania 18701.

Power Authority of the State of New

York, Docket No. 50–333, James A.

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of application for amendment:

September 16, 1994, as supplemented

November 29, 1994.

Brief description of amendment: The

amendment revised Technical

Specifications Section 6.0

(Administrative Controls) to reflect, in

part, licensee management changes in

the corporate organization. Specifically,

the title of Executive Vice President—

Nuclear Generation was changed to

Executive Vice President and Chief

Nuclear Officer and a new position,

Vice President Regulatory Affairs and

Special Projects, which reports to the

Executive Vice President and Chief

Nuclear Officer, was established. In

addition, the list of Safety Review

Committee (SRC) members, which was

previously by job title, was deleted and

replaced with a description of SRC

membership requirements, including

individual qualifications and the

minimum number of SRC members was

reduced from 8 to 6.

Date of issuance: December 22, 1994.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 220.

Facility Operating License No. DPR–

59: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: September 30, 1994 (59 FR

50021).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated December 22,

1994.

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.

Power Authority of the State of New

York, Docket No. 50–333, James A.

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of application for amendment:

August 4, 1994, as supplemented

November 10, 1994.

Brief description of amendment: The

amendment revises requirements in the

Technical Specifications (TSs) related to

primary containment atmosphere

monitoring and drywell to torus

differential pressure. Specifically, TS

3.7.A.6. has been revised to adopt

primary containment inerting/

deinerting requirements that are

consistent with NUREG–1433,

‘‘Standard Technical Specifications—

General Electric Plants, BWR/4.’’ TSs

4.7.A.6.a. and 4.7.A.7.a. have been

revised to provide frequencies for the

verification of primary containment

oxygen concentration and pressure

differential between the drywell and

torus. TSs 3.7.A.7.a.(1), 3.7.A.7.a.(3),

and 3.7.A.8. have been revised to

provide requirements for establishing

and maintaining differential pressure

between the drywell and torus that are

consistent with NUREG–1433. TS

3.7.A.9. has been deleted and related

requirements have been incorporated

into Notes for Table 3.2–8. Several

administrative changes to Tables 3.2–8

and 4.2–8 have also been made to

improve the overall quality of the TSs.

Date of issuance: December 28, 1994.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 221.

Facility Operating License No. DPR–

59: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 31, 1994 (59 FR

45032).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated December 28,

1994.

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.

Power Authority of The State of New

York, Docket No. 50–286, Indian Point

Nuclear Generating Unit No. 3,

Westchester County, New York

Date of application for amendment:

September 16, 1994, as supplemented

November 29, 1994.

Brief description of amendment: The

amendment revised Technical

Specifications Section 6.0

(Administrative Controls) to reflect, in

part, licensee management changes in

the corporate organization. Specifically,

the title of Executive Vice President—

Nuclear Generation was changed to

Executive Vice President and Chief

Nuclear Officer and a new position,

Vice President Regulatory Affairs and

Special Projects, which reports to the

Executive Vice President and Chief

Nuclear Officer, was established. In

addition, the list of Safety Review

Committee (SRC) members, which was

previously by job title, was deleted and

replaced with a description of SRC

membership requirements, including
3684 Federal Register / Vol. 60, No. 11 / Wednesday, January 18, 1995 / Notices

individual qualifications and the

minimum number of SRC members was

reduced from 8 to 6.

Date of issuance: December 22, 1994.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 156.

Facility Operating License No. DPR–

64: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: September 30, 1994 (54 FR

50021).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated December 22,

1994.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10610.

Public Service Electric & Gas Company,

Docket Nos. 50–272 and 50–311, Salem

Nuclear Generating Station, Unit Nos. 1

and 2, Salem County, New Jersey

Date of application for amendments:

September 29, 1994.

Brief description of amendments: The

amendments change the Technical

Specification surveillance interval for

performing an air or smoke flow test

through each containment spray header

from 5 to 10 years.

Date of issuance: December 27, 1994.

Effective date: December 27, 1994.

Amendment Nos. 163, 144.

Facility Operating License Nos. DPR–

70 and DPR–75. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: November 23, 1994 (59 FR

60385).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated December 27,

1994.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Salem Free Public Library, 112

West Broadway, Salem, New Jersey

08079.

Southern California Edison Company,

et al, Docket No. 50–206, San Onofre

Nuclear Generating Station, Unit No. 1,

San Diego County, California

Date of application for amendment:

April 18, 1994, as supplemented

October 26, 1994.

Brief description of amendment: The

amendment revises Sections 2.C and 2.D

of the San Onofre Nuclear Generating

Station, Unit 1 (SONGS 1) Operating

License. Section 2.C will be revised to

modify or delete several licensing

conditions which either no longer apply

or require revision to apply to SONGS

1 in its permanently shutdown and

defueled condition. Section 2.D will be

revised to exempt Fire Protection

reporting from the reporting

requirements of Section 2.D.

Date of issuance: December 22, 1994.

Effective date: January 21, 1995.

Amendment No.: 156.

Facility Operating License No. DPR–

13: The amendment revised the license

conditions.

Date of initial notice in Federal

Register: May 25, 1994 (59 FR 27066).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated December 22,

1994.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Main Library, University of

California, P.O. Box 19557, Irvine,

California 92713.

Southern Nuclear Operating Company,

Inc., Docket Nos. 50–348 and 50–364,

Joseph M. Farley Nuclear Plant, Units

1 and 2, Houston County, Alabama.

Date of amendments request: October

20, 1994.

Brief Description of amendments: The

amendments delete the requirements for

the control room chlorine detection

system from the TS and the associated

Bases Sections. This request is based on

the fact that all stored gaseous chlorine

has been removed from the plant site

except for containers having an

inventory of 150 pounds or less.

Date of issuance: December 28, 1994.

Effective date: December 28, 1994.

Amendment Nos.: 111 and 102.

Facility Operating License Nos. NPF–

2 and NPF–8. Amendments revise the

Technical Specifications.

Date of initial notice in Federal

Register: November 23, 1994 (59 FR

60386).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated December 28,

1994.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Houston-Love Memorial

Library, 212 W. Burdeshaw Street, Post

Office Box 1369, Dothan, Alabama

36302.

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of application for amendments:

September 9, 1994 (TS 94–04).

Brief description of amendments: The

amendments revise the technical

specifications related to the cold leg

injection accumulators.

Date of issuance: December 27, 1994.

Effective date: December 27, 1994.

Amendment Nos.: 192 and 184.

Facility Operating License Nos. DPR–

77 and DPR–79: Amendments revise the

technical specifications.

Date of initial notice in Federal

Register: October 12, 1994 (59 FR

51629).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated December 27,

1994.

No significant hazards consideration

comments received: None.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402.

Tennessee Valley Authority, Docket

Nos. 50–327, Sequoyah Nuclear Plant,

Units 1, Hamilton County, Tennessee

Date of application for amendment:

November 2, 1994 (TS 94–17).

Brief description of amendment: The

amendment adds Operating License

Condition 2.C.(25) to provide a limited

extension of the surveillance test

intervals for certain specified

instrumentation on Unit 1 to coincide

with the Cycle 7 refueling outage. The

surveillance intervals that are affected

are specified in the attached safety

evaluation and are for tests that would

be extended to October 1, 1995, and

would result in extension of the

specified 18-, 36- and 54-month

surveillances to 29.5, 48 and 71.5

months, respectively.

Date of issuance: January 3, 1995.

Effective date: January 3, 1995.

Amendment No.: 193.

Facility Operating License Nos. DPR–

77: Amendment revises the operating

license.

Date of initial notice in Federal

Register: November 23, 1994 (59 FR

60387).

The Commission’s related evaluation

of the change to the operating license is

contained in a Safety Evaluation dated

January 3, 1995.

No significant hazards consideration

comments received: None.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402

Virginia Electric and Power Company,

Docket Nos. 50–280 and 50–281, Surry

Power Station, Unit Nos. 1 and 2, Surry

County, Virginia

Date of application for amendments:

October 11, 1994.
Federal Register / Vol. 60, No. 11 / Wednesday, January 18, 1995 / Notices 3685

1 15 U.S.C. 78s(b) (1988).

2Securities Exchange Act Release No. 34679

(September 15, 1994), 59 FR 48652.

3MT provides settlement and depository services

for securities listed on Italian stock exchanges.

4Under the service agreement, ISCC upon

instructions received from MT will: (1) Accept

receives of securities by book-entry through DTC;

(2) initiate book-entry delivery of securities on

deposit at DTC; (3) initiate reclamations of

securities received at DTC; (4) make delivery of due

bill checks or payments received by ISCC with

respect to securities; (5) initiate messages to other

DTC participants through the broadcast function of

DTC; (6) process securities on deposit at DTC and

securities subject to a reorganization, takeover, or

similar action provided that requisite funds, if

applicable, have been received in advance; (7)

withdraw rights exited from DTC and deliver such

rights to the agent designated by MT along with

instructions received with respect to such rights; (8)

at the discretion of ISCC, facilitate withdrawals-bytransfer of securities on deposit at DTC; and (9)

order proxy materials for securities on deposit at

DTC and complete such proxy materials as

instructed.

5The related money settlements for the securities

movements will take place between the parties

outside of ISCC.

6As reflected on the DTC, records securities

deposits will form the basis for the bookkeeping

entries at MT on behalf of MT’s participants.

7The amount of the deposit shall be the average

of the three highest one-month fees over the prior

twelve months plus the amount ISCC is required to

deposit with DTC with respect to the sponsored

account. However, the amount of MT’s cash deposit

with ISCC cannot be less than $50,000.

Brief description of amendments: The

amendments revise the surveillance

frequencies of the hydrogen analyzer

channel functional test and channel

calibration.

Date of issuance: December 23, 1994.

Effective date: December 23, 1994.

Amendment Nos. 195 and 195.

Facility Operating License Nos. DPR–

32 and DPR–37: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: November 9, 1994 (59 FR

55893).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated December 23,

1994.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Swem Library, College of

William and Mary, Williamsburg,

Virginia 23185.

Wolf Creek Nuclear Operating

Corporation, Docket No. 50–482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request: July 22,

1994.

Brief description of amendment: This

amendment revises Section 6 of the

Technical Specifications to reflect title

changes in the Wolf Creek Nuclear

Operating Corporation organization.

Date of issuance: December 29, 1994.

Effective date: December 29, 1994.

Amendment No.: 81.

Facility Operating License No. NPF–

42. The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: October 26, 1994 (59 FR

53845).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated December 29,

1994.

No significant hazards consideration

comments received: No.

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.

Wolf Creek Nuclear Operating

Corporation, Docket No. 50–482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request: February

23, 1994.

Brief description of amendment: This

amendment revised Technical

Specifications 3.8.1.1, ‘‘AC Sources

Operating,’’ and 3.8.1.2, ‘‘AC Sources

Shutdown,’’ to increase the minimum

volume of fuel oil required for the

emergency diesel generator fuel oil day

tanks. Several other revisions are

included that make editorial corrections

and incorporate requirements that were

inadvertently omitted from previous

amendment requests that have been

approved.

Date of issuance: December 29, 1994.

Effective date: December 29, 1994.

Amendment No.: 82.

Facility Operating License No. NPF–

42. The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: April 13, 1994 (59 FR 17609).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated December 29,

1994.

No significant hazards consideration

comments received: No.

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.

Dated at Rockville, Maryland, this 10th day

of January 1995.

For the Nuclear Regulatory Commission.

Jack W. Roe,

Director, Division of Reactor Projects—III/IV,

Office of Nuclear Reactor Regulation

[FR Doc. 95–1026 Filed 1–17–95; 8:45 am]

BILLING CODE 7590–01–P

SECURITIES AND EXCHANGE

COMMISSION

[Release No. 34–35219; International Series

Release No. 770; File No. SR–ISCC–94–4]

Self-Regulatory Organizations;

International Securities Clearing

Corporation; Order Approving

Proposed Rule Change Relating to a

Data Transmission Link With Monte

Titoli, S.P.A.

January 11, 1995.

On August 9, 1994, the International

Securities Clearing Corporation

(‘‘ISCC’’) submitted a proposed rule

change (File No. SR–ISCC–94–4) to the

Securities and Exchange Commission

(‘‘Commission’’) pursuant to Section

19(b) of the Securities Exchange Act of

1934 (‘‘Act’’).1 Notice of the proposal

appeared in the Federal Register on

September 22, 1994.2 The Commission

received no comments. This order

approves the proposal.

I. Description of the Proposal

ISCC has entered into a contract to

establish a data transmission link with

Monte Titoli (‘‘MT’’).3 The link will

permit MT to hold U.S. securities listed

on Italian stock exchanges in The

Depository Trust Company (‘‘DTC’’)

through ISCC. The service agreement,

dated June 1, 1992, between ISCC and

MT provides that ISCC will sponsor an

account for MT at DTC which will

provide MT access to certain DTC

services.4

ISCC on behalf of MT will initiate

book-entry deliveries for no value and

will accept receives of securities by

book-entry for no value.5 Both the

receive and deliver functions will be

pursuant to instructions received from

MT, and such instructions will identify

the MT member for whom the receipt or

delivery is being effected.6 In special

circumstances and at ISCC’s discretion,

DTC’s withdrawal-by-transfer service

also may be utilized. In such case, the

securities will be delivered as directed

by MT. MT will deposit with ISCC

collateral to cover MT’s obligations to

ISCC.7 To the extent that any money

settlement is required, ISCC will receive

payment in the form of an official bank

check or a wire transfer through the MT

designated correspondent bank.

On each business day at about 4:00

p.m., ISCC will transmit a preliminary

settlement statement which will detail

the net amount due to ISCC from MT or

from ISCC to MT. Under the service

agreement, MT agrees to pay to ISCC all