Federal Register / Vol. 62, No. 10 / Wednesday, January 15, 1997 / Notices 2185

Commission, the presiding officer or the

presiding Atomic Safety and Licensing

Board that the petition and/or request

should be granted based upon a

balancing of the factors specified in 10

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

For further details with respect to this

action, see the application for

amendment dated November 22, 1996,

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 located at

the White Plains Public Library, 100

Martine Avenue, White Plains, New

York 10610.

Dated at Rockville, Maryland, this 9th day

of January 1997.

For the Nuclear Regulatory Commission.

George F. Wunder,

Project Manager, Project Directorate 1–1,

Division of Reactor Projects—I/II, Office of

Nuclear Reactor Regulation.

[FR Doc. 97–982 Filed 1–14–97; 8:45 am]

BILLING CODE 7590–01–P

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

20, 1996, through January 3, 1997. The

last biweekly notice was published on

January 2, 1997 (62 FR 121).

Notice Of Consideration Of Issuance Of

Amendments To Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

And Opportunity For A Hearing

The Commission has made a

proposed determination that the

following amendment requests involve

no significant hazards consideration.

Under the Commission’s regulations in

10 CFR 50.92, this means that operation

of the facility in accordance with the

proposed amendment would not (1)

involve a significant increase in the

probability or consequences of an

accident previously evaluated; or (2)

create the possibility of a new or

different kind of accident from any

accident previously evaluated; or (3)

involve a significant reduction in a

margin of safety. The basis for this

proposed determination for each

amendment request is shown below.

The Commission is seeking public

comments on this proposed

determination. Any comments received

within 30 days after the date of

publication of this notice will be

considered in making any final

determination.

Normally, the Commission will not

issue the amendment until the

expiration of the 30-day notice period.

However, should circumstances change

during the notice period such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received before

action is taken. Should the Commission

take this action, it will publish in the

Federal Register a notice of issuance

and provide for opportunity for a

hearing after issuance. The Commission

expects that the need to take this action

will occur very infrequently.

Written comments may be submitted

by mail to the Chief, Rules Review and

Directives Branch, Division of Freedom

of Information and Publications

Services, Office of Administration, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555-0001, and

should cite the publication date and

page number of this Federal Register

notice. Written comments may also be

delivered to Room 6D22, Two White

Flint North, 11545 Rockville Pike,

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

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

written comments received may be

examined at the NRC Public Document

Room, the Gelman Building, 2120 L

Street, NW., Washington, DC. The filing

of requests for a hearing and petitions

for leave to intervene is discussed

below.

By February 14, 1997, 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.
2186 Federal Register / Vol. 62, No. 10 / Wednesday, January 15, 1997 / Notices

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555-0001, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington DC, by

the above date. 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 20555-0001, and to the attorney for

the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

factors specified in 10 CFR

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

For further details with respect to this

action, see the application for

amendment which is available for

public inspection at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, and at the local public

document room for the particular

facility involved.

Commonwealth Edison Company,

Docket Nos. STN 50-454 and STN 50-

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

Ogle County, Illinois Docket Nos. STN

50-456 and STN 50-457, Braidwood

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

Illinois

Date of amendment request:

November 4, 1996, as supplemented on

December 4, 1996.

Description of amendment request:

The proposed amendment would permit

Byron, Unit 1, and Braidwood, Unit 1,

to remove sheathing filler grease in the

tendon sheathing for up to 35 tendons

in advance of the steam generator

replacement outages.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The prestressing tendons are passive

components that form part of the

containment structure. As passive

components, there are no tendon failure

modes that could act as accident initiators or

precursors.

Consequently, the proposed change to

remove a portion of the tendon sheathing

filler grease will not increase the probability

of an accident previously evaluated.

The tendons, in their passive role, function

to limit the consequences of accidents

previously evaluated, and their continued

integrity is important to the ability of the

containment to mitigate design basis

accidents. Structural degradation of the

containment is a predictable process that can

be monitored by a comprehensive

containment tendon monitoring program as

required by Technical Specification

Surveillance Requirement 4.6.1.6. The

monitoring program is based on proposed

Revision 3 of Regulatory Guide 1.35,

‘‘Inservice Surveillance of Ungrouted

Tendons in Prestressed Concrete

Containment Structures,’’ April 1979.

The tendon surveillances conducted at

both Byron and Braidwood have consistently

shown that structural integrity of the tendon

system has been maintained, including

adequate corrosion protection for the tendon

wires and end anchorage components, and

there has been no evidence of grease leakage

from the tendon sheathings. While a number

of below-grade hoop tendons have shown

signs of water intrusion, the tendons that will

have grease removed are above-grade and are

not expected to experience water intrusion.

A review of domestic nuclear facility

experience found cases where large grease

voids existed for periods longer than

requested under the proposed change

without resultant corrosion in those tendon

systems. A case where tendon wires removed

from a decommissioned plant were exposed

to an environment more severe than expected

in a sealed tendon sheath did not show signs

of corrosion. These experiences demonstrate

the effectiveness of the initial corrosion

protection systems applied to the tendons

and the effectiveness of partial grease

protection in the tendon sheathing.

Based on the above cases, it can be

concluded that the removal of the filler

grease (grease voids greater than 5 percent)

from the tendon sheathing in up to thirty-five

tendons for a limited period will not

adversely affect the integrity of the tendons

or the capability of the tendon system to

fulfill its design basis function.

The removal process will only remove the

grease not directly adhering to the tendons.

The grease remaining will be adequate to

protect the tendons during the relatively

short period of partial grease removal.

Therefore, no changes in the tendon

properties would be expected, and the

consequences of design basis accidents

previously evaluated will not be affected by

the proposed change.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.
Federal Register / Vol. 62, No. 10 / Wednesday, January 15, 1997 / Notices 2187

The proposed change only affects the

tendon sheathing filler grease void limits of

TSSR 4.6.1.6. No new equipment is being

installed and no existing equipment is being

modified. Operation with a grease void in

excess of current requirements does not alter

system configurations such that any new or

different accidents can be initiated.

Therefore, no new or different accident

initiators or precursors are being introduced,

and the proposed change will not create the

possibility of a new or different kind of

accident from any previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The margin of safety applicable to the

proposed change is defined by the difference

between the design pressure of the

containment and the point at which the

containment would actually fail. The design

pressure of the containment is 50 psi. As a

result of conservatism inherent in the design

techniques and in the material selections

made for the Byron and Braidwood

containments, a substantial margin to failure

exists in the containment. This margin is

discussed in Subsection 3.8.1.8 of the

Updated Final Safety Analysis Report. It is

noted therein that the ultimate capacity of

the concrete shell is 125 psi, corresponding

to the initiation of yield in the hoop posttensioning tendons in conjunction with

yielding of the reinforcement near the midheight of the containment wall.

It is also noted in Subsection 3.8.1.8 that

the ultimate capacity of a containment

electrical penetration is 108 psi. While this

value is substantially greater than the 50 psi

required of the design, it is lower than the

125 psi at which failure of the containment

wall section would be predicted. Therefore,

tendon strength is not the limiting factor in

the margin of safety inherent in the

containment.

As previously discussed, no degradation of

the tendons is expected to occur as a result

of the proposed TS change. Further, the

tendon strength is not the limiting factor in

the containment ultimate capacity, which is

substantially greater than the requirement

placed on the containment design by the

plant design basis. Therefore, the proposed

change will not reduce the margin of safety

designed into Byron and Braidwood.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

requested amendments involve no

significant hazards consideration.

Local Public Document Room

location: For Byron, the Byron Public

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

Box 434, Byron, Illinois 61010; for

Braidwood, the Wilmington Public

Library, 201 S. Kankakee Street,

Wilmington Illinois 60481.

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60603

NRC Project Director: Robert A. Capra

Commonwealth Edison Company,

Docket Nos. 50-237 and 50-249,

Dresden Nuclear Power Station, Units 2

and 3, Grundy County, Illinois Docket

Nos. 50-254 and 50-265, Quad Cities

Nuclear Power Station, Units 1 and 2,

Rock Island County, Illinois

Date of application for amendment

request: December 6, 1996

Description of amendment request:

The proposed amendment would allow

a single control rod to be moved when

the plant is in HOT SHUTDOWN and

COLD SHUTDOWN condition provided

the one-rod-out interlock is OPERABLE

and the reactor mode switch is in the

refuel position.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1) Involve a significant increase in the

probability or consequences of an accident

previously evaluated because of the

following:

This revision would allow a single control

rod to be withdrawn under control of the

reactor mode switch position one-rod-out

interlock in OPERATIONAL MODES 3 or 4.

This interlock is explicitly assumed in the

safety analysis for control rod removal error

during refueling. A prompt reactivity

excursion could potentially result in fuel

failure. The one-rod-out interlock, together

with the requirements for adequate

SHUTDOWN MARGIN (SDM), provides

protection against prompt reactivity

excursions by preventing withdrawal of more

than one control rod and ensuring the core

remains subcritical with any one control rod

withdrawn. The addition of surveillance

requirements for the one-rod-out interlock

will assure the interlock is OPERABLE prior

to withdrawal of a control rod in

OPERATIONAL MODES 3 and 4. Although

this change will increase the frequency of

single control rod withdrawals in

OPERATIONAL MODES 3 and 4, the

probability of previously analyzed accidents,

including control rod withdrawal error, is not

affected because the same actions are

required, although they are now conducted

in different OPERATIONAL MODES.

The consequences of previously analyzed

accidents in OPERATIONAL MODES 3 and

4 are not affected by this proposed change.

The SDM requirements of TS 3.3.A assure the

reactor is maintained subcritical when all

control rods are fully inserted, without

crediting the single control rod having the

highest reactivity worth which is assumed to

be fully withdrawn. The one-rod-out

interlock of the reactor mode switch Refuel

position permits only a single control rod to

be withdrawn. The proposed change will not

affect the potential for attaining criticality in

OPERATIONAL MODES 3 and 4 or effect the

initial conditions assumed in any design

basis accident analysis.

Based on this, the probability or

consequences of any accident previously

evaluated is not increased by the proposed

changes.

2) Create the possibility of a new or

different kind of accident from any accident

previously evaluated because:

Single control rods can be withdrawn to

permit control rod recoupling in

OPERATIONAL MODES 3 and 4 under

existing TS. The proposed change will

merely expand this allowance to other

control rod maintenance and testing

activities performed in OPERATIONAL

MODES 3 and 4. The revision to

Specification 3/4.10.A provides additional

assurance that the one-rod-out interlock is

OPERABLE in OPERATIONAL MODES 3 and

4.

The additional control rod maintenance

and testing activities which could be

performed in OPERATIONAL MODES 3 and

4 are permitted by the existing TS in

OPERATIONAL MODES 1, 2 and 5.

Examples of activities which could be

performed include venting of control rods

following a reactor scram or control rod drive

system outage, normal control rod insertion/

withdrawal timing and adjustment, control

rod scram time testing and control rod

friction testing.

Based on this, the proposed changes do not

create the possibility of a new or different

kind of accident from those previously

evaluated.

Specification 3/4.10.A is revised to ensure

the one-rod-out interlock is OPERABLE,

enhancing the assurance that the plant will

prevent the withdrawal of more than one

control rod in the manner currently assumed.

Expanding the applicability of this existing

requirement to OPERATIONAL MODES 3

and 4 similarly does not create the possibility

of a new or different kind of accident from

those previously evaluated.

3) Involve a significant reduction in the

margin of safety because:

The TS currently permit single control rod

withdrawal for the purpose of control rod

recoupling when in OPERATIONAL MODES

3 or 4 if the one-rod-out interlock is

OPERABLE. This change merely allows

additional activities for which a single

control rod may be withdrawn in

OPERATIONAL MODES 3 or 4, with the

same restriction that the one-rod-out

interlock is OPERABLE.

While the TS currently allow limited

control rod withdrawal in OPERATIONAL

MODES 3 and 4 provided the one-rod-out

interlock is OPERABLE, no explicit

surveillance requirements for the one-rod-out

interlock exist while in OPERATIONAL

MODES 3 or 4. The proposed changes to the

Applicability statement in TS 3/4.10.A will

result in applicability of the Surveillance

Requirements for the one-rod-out interlock

whenever control rod withdrawal is

performed in OPERATIONAL MODES 3 and

4.

Together, the OPERABILITY requirements

for the one-rod-out interlock and the SDM

requirements of TS 3.3.A will continue to

ensure that the reactor will be maintained

subcritical during single control rod

withdrawals. Therefore, this change will not

involve a significant reduction in the margin

of safety.
2188 Federal Register / Vol. 62, No. 10 / Wednesday, January 15, 1997 / Notices

As described, the proposed amendment for

Dresden and Quad Cities Stations will not

reduce the availability of systems required to

mitigate accident conditions. Neither are new

or significantly different modes of operation

proposed. Therefore, the proposed changes

do 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: for Dresden, Morris Area

Public Library District, 604 Liberty

Street, Morris, Illinois 60450; for Quad

Cities, Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois 61021

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60603

NRC Project Director: Robert A. Capra

Entergy Operations, Inc., Docket Nos.

50-313 and 50-368, Arkansas Nuclear

One, Unit Nos. 1 and 2 (ANO-1&2),

Pope County, Arkansas

Date of amendment request: October

2, 1996

Description of amendment request:

Relocation of Radiological Effluent

Technical Specifications for Units 1 and

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:

Criterion 1 - Does not Involve a Significant

Increase in the probability or Consequences

of an Accident Previously Evaluated.

The proposed changes are considered

administrative in nature. These changes alter

only the location of programmatic controls

and procedural details relative to radioactive

effluents, radiological environmental

monitoring, solid radioactive wastes, and

associated reporting requirements.

Compliance with applicable regulatory

requirements will continue to be maintained.

In addition, the proposed changes do not

alter the conditions and assumptions in any

of the Safety Analysis Report (SAR) accident

analyses. Since the SAR accident analyses

remain bounding, the radiological

consequences previously evaluated are not

adversely affected by the proposed changes.

Therefore, this change does not involve a

significant increase in the probability or

consequences of any accident previously

evaluated.

Criterion 2 - Does Not Create the Possibility

of a New or Different Kind of Accident from

any Previously Evaluated.

The proposed changes do not involve any

changes to the configuration or method of

operation any plant equipment. The

proposed changes are considered

administrative in nature. Accordingly, no

new failure modes have been defined for any

plant system or component important to

safety nor has any new limiting single failure

have been identified as a result of the

proposed changes. Also, there will be no

change in types or increase in the amounts

of any radioactive effluents released offsite.

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 Margin of Safety.

The proposed changes do not involve nay

actual change in the methodology used in the

control of radioactive effluents, solid

radioactive wastes, or radiological

environmental monitoring. These changes are

considered administrative in nature and

provide for the relocation of procedural

details outside the Technical Specifications.

This change adds appropriate administrative

controls in the Technical Specifications to

provide continued assurance of compliance

with applicable regulatory requirements.

Therefore, this change does not involve a

significant reduction in the margin of safety.

I21Therefore, based upon the reasoning

presented above and the previous discussion

of the amendment request, Entergy

Operations has determined that the requested

change does not involve significant hazards

consideration.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

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

20005-3502

NRC Project Director: William D.

Beckner

Entergy Operations, Inc., Docket Nos.

50-313 and 50-368, Arkansas Nuclear

One, Unit Nos. 1 and 2 (ANO-1&2),

Pope County, Arkansas

Date of amendment request: October

2, 1996

Description of amendment request:

Relocation of Selected Technical

Specifications Instrumentation

Requirements Allowed by Generic

Letter 95-10

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 [Nuclear Regulatory Commission] NRC

issued Generic Letter (GL) 95-10 to allow

licensees to relocate certain instrumentation

requirements to licensee controlled

documents or programs. The staff has

concluded that the specifications listed in the

GL were not required to be included in the

technical specifications as required by 10

CFR 50.36. The staff concluded that the

instrumentation addressed in these

specifications are not related to dominant

contributors to plant risk.

The specifications included in this

amendment request are being relocated to the

Technical Requirements Manual (TRM).

Once in the TRM, future changes to these

requirements will be controlled under 10

CFR 50.59. By controlling future changes

under 10 CFR 50.59, NRC review and

approval will be requested for changes

exceeding the regulatory threshold of an

unreviewed safety question.

This amendment request does not remove

or modify any of the instrumentation

requirements for either unit. This

amendment request does not affect any of the

accident initiators, conditions or

assumptions for any of the accidents

previously evaluated. Therefore, this change

does not involve a significant increase in the

probability of any accident previously

evaluated.

This amendment request is administrative

in nature and does not affect any system or

component functional requirements. This

change does not affect the operation of the

plant or affect any component that is used to

mitigate the consequences of any accident.

Therefore, this change does not involve a

significant increase in the consequences of

any accident previously evaluated.

Therefore, this change does not involve a

significant increase in the probability or

consequences of any accident previously

evaluated.

Criterion 2 - Does Not Create the Possibility

of a New or Different Kind of Accident from

any Previously Evaluated.

The relocation of existing requirements

from the technical specifications to other

licensee controlled documents is considered

administrative in nature. This change does

not modify or remove any plant

instrumentation requirements. This proposed

change will not affect any plant system or

structure, nor will it affect any system

functional or operability requirements.

Consequently, no new failure modes are

introduced as a result of this change.

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 Margin of Safety.

The proposed amendment request

represents a relocation of a portion of the

information previously located in each unit’s

technical specification instrumentation

section to other licensee controlled

documents that ate controlled under 10 CFR

50.59. The proposed change is administrative

in nature because the instrumentation

requirements for the facility remain the same.
Federal Register / Vol. 62, No. 10 / Wednesday, January 15, 1997 / Notices 2189

The proposed change does not represent a

change in the configuration or operation of

the plant. Therefore, this change does not

involve a significant reduction in the margin

of safety.

Therefore, based upon the reasoning

presented above and the previous discussion

of the amendment request, Entergy

Operations has determined that the requested

change does not involve significant hazards

consideration.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

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

20005-3502

NRC Project Director: William D.

Beckner

Entergy Operations Inc., Docket No. 50-

382, Waterford Steam Electric Station,

Unit 3, St. Charles Parish, Louisiana

Date of amendment request:

December 2, 1996

Description of amendment request:

The proposed Technical Specification

(TS) Change Request will permit the use

of 10 CFR Part 50 Appendix J, Option

B, Performance-Based Containment

Leakage Testing for Type A, B and C

leak rate testing. TSs 3/4.6.1.1, 3/4.6.1.2,

3/4.6.1.3, 4.6.1.6 and 4.6.1.7 are revised

and Section 6.15 is added establishing

the Containment Leakage Rate Testing

Program. The Bases are revised to reflect

this change. Minor editorial changes are

included in this request. Waterford

Steam Electric Station is planning to

have a Containment Leakage Rate

Testing Program in place prior to the

next scheduled refueling outage. This

program will be in accordance with the

guidelines contained in Regulatory

Guide 1.163, ‘‘Performance-Based

Containment Leak-Test Program,’’ dated

September 1995.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The proposed change will not affect the

assumptions, design parameters, or results of

any accident previously evaluated. The

proposed change does not add or modify any

existing equipment. The proposed changes

will result in increased intervals between

containment leakage tests determined

through a performance based approach. The

intervals between such tests are not related

to conditions which cause accidents. The

proposed changes do not involve a change to

the plant design or operation. Therefore, this

change does not involve a significant

increase in the probability of any accident

previously evaluated.

NUREG-1493, ‘‘Performance-Based

Containment Leak-Test Program,’’

contributed to the technical bases for Option

B of 10 CFR 50 Appendix J. NUREG-1493

contains a detailed evaluation of the

expected leakage from containment and the

associated consequences. The increased risk

due to lengthening of the intervals between

containment leakage tests was also evaluated

and found acceptable. Using a statistical

approach, NUREG-1493 determined the

increase in the expected dose to the public

from extending the testing frequency is

extremely small. It also concluded that a

small increase is justifiable due to the

benefits which accrue from the interval

extension. The primary benefit is in the

reduction in occupational exposure. The

reduction in the occupational exposure is a

real reduction, while the small increase to

the public is statistically derived using

conservative assumptions. Therefore, this

change does not involve a significant

increase in the consequences of any accident

previously evaluated.

The proposed change does not involve

modifications to any existing equipment. The

proposed change will not affect the operation

of the plant or the manner in which the plant

is operated. The reduced testing frequency

will not affect the testing methodology.

Therefore, the proposed change will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed change does not change the

performance methodology of the containment

leakage rate testing program. However, the

proposed change does affect the frequency of

containment leakage rate testing. With an

increased frequency between tests, the

proposed change does increase the

probability that a increase in leakage could

go undetected for a longer period of time.

Operational experience has demonstrated the

leak tightness of the containment buildings

has been significantly below the allowable

leakage limit.

The margin of safety that has the potential

of being impacted by the proposed change

involves the offsite dose consequences of

postulated accidents which are directly

related to containment leakage rates. The

limitation on containment leakage rate is

designed to ensure the total leakage volume

will not exceed the value assumed in our

accident analysis. The margin of safety for

the offsite dose consequences of postulated

accidents directly related to containment

leakage is maintained by meeting the 1.0 La

acceptance criteria. The proposed change

maintains the 1.0 La acceptance criteria.

Therefore, the proposed change will not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, LA 70122

Attorney for licensee: N.S. Reynolds,

Esq., Winston & Strawn 1400 L Street

N.W., Washington, D.C. 20005-3502

NRC Project Director: William D.

Beckner

Florida Power and Light Company,

Docket No. 50-335, St. Lucie Plant Unit

1, St. Lucie County, Florida

Dates of amendment request:

December 9, 1996

Description of amendment request:

The licensee proposed to modify

specifications for selected cycle-specific

reactor physics parameters to refer to

the St. Lucie Unit 1 Core Operating

Limits Report (COLR) for limiting

values. Minor administrative changes

are also included. The proposed

Technical Specification (TS) changes

utilized the guidance provided in

Generic Letter 88-16 and are intended to

be consistent with the Standard

Technical Specifications for

Combustion Engineering Plants

(NUREG-1432, Revision 1).

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below.

(1) Operation of the facility in accordance

with the proposed amendment would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed amendment relocates the

calculated values of selected cycle-specific

reactor physics parameter limits from the TS

to the COLR, and includes minor editorial

changes which do not alter the intent of

stated requirements. The amendment is

administrative in nature and has no impact

on any plant configuration or system

performance relied upon to mitigate the

consequences of an accident. Parameter

limits specified in the COLR for this

amendment are not changed from the values

presently required by Technical

Specifications. Future changes to the

calculated values of such limits may only be

made using NRC approved methodologies,

must be consistent with all applicable safety

analysis limits, and are controlled by the 10

CFR 50.59 process. Assumptions used for

accident initiators and/or safety analysis

acceptance criteria are not changed by this

amendment. Therefore, operation of the

facility in accordance with the proposed

amendment will not involve a significant

increase in the probability or consequences

of an accident previously evaluated.
2190 Federal Register / Vol. 62, No. 10 / Wednesday, January 15, 1997 / Notices

(2) Operation of the facility in accordance

with the proposed

amendment would not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed amendment relocates the

calculated values of cycle specific reactor

physics limiting parameters to the COLR and

will not change the physical plant or the

modes of operation defined in the facility

license. The changes do not involve the

addition of new equipment or the

modification of existing equipment, nor do

they alter the design configuration of St.

Lucie plant systems. Therefore, operation of

the facility in accordance with the proposed

amendment would not create the possibility

of a new or different kind of accident from

any accident previously evaluated.

(3) Operation of the facility in accordance

with the proposed

amendment would not involve a

significant reduction in a margin of safety.

The cycle specific parameter limits being

relocated to the COLR by this amendment

have not been changed from the values

presently required by the TS, and a

requirement to operate the plant within the

bounds of the limits specified in the COLR

is retained in the individual specifications.

Future changes to the calculated values of

these limits by the licensee may only be

developed using NRC-approved

methodologies, must remain consistent with

all plant safety analysis limits addressed in

the Final Safety Analysis Report (FSAR), and

are further controlled by the 10 CFR 50.59

process. As discussed in Generic Letter 88-

16, the administrative controls established

for the values of cycle specific parameters

using the guidance of that letter assure

conformance with 10 CFR 50.36. Safety

analysis acceptance criteria are not being

altered by this amendment. Therefore,

operation of the facility in accordance with

the proposed amendment would not involve

a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee’s

analysis and, based on thisreview, 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: Indian River Junior College

Library, 3209 Virginia Avenue, Fort

Pierce, Florida 34954-9003

Attorney for licensee: M. S. Ross,

Attorney, Florida Power & Light, 11770

US Highway 1, North Palm Beach,

Florida 33408

NRC Project Director: Frederick J.

Hebdon

Florida Power and Light Company,

Docket No. 50-335 St. Lucie Plant Unit

1, St Lucie County, Florida

Date of amendment request:

December 20, 1996

Description of amendment request:

The licensee proposed to delete a

footnote associated with TS 2.1.1,

‘‘Reactor Core Safety Limits,’’ which

requires reactor thermal power to be

limited to 90% of 2700 Megawatts

thermal for Cycle 14 operation beyond

7000 Effective Full Power Hours

[EFPH]. The thermal power limit was

required pending completion of a Small

Break Loss of Coolant Accident

(SBLOCA) reanalysis that demonstrated

acceptable results using input

assumptions corresponding to an

increased number of steam generator

tubes being plugged. The SBLOCA

reanalysis was completed and included

with the submittal.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below.

(1) Operation of the facility in accordance

with the proposed amendment would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change will allow full Cycle

14 operation at 100% of rated power (2700

MWth), by deleting the requirement to

derate to 90% of rated power prior to

exceeding 7000 EFPH. This restriction was

imposed in the NRC transmittal letter for

License Amendment 145 for SBLOCA

considerations when considering the

increased SGTP [steam generator tube

plugging]

level of 30% plus or minus 7%. All Final

Safety Analysis Report (FSAR) events, other

than SBLOCA were evaluated at 100% of

rated thermal power and showed no

significant increases in the probability or

consequences of accidents previously

evaluated.

The SBLOCA was reanalyzed to

demonstrate continued compliance with 10

CFR 50.46 criteria. There is no impact of the

proposed change on any FSAR accident

initiator. The plant configuration and

systems remain unchanged.

Therefore, operation of the facility in

accordance with the proposed amendment

would not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

(2) Operation of the facility in accordance

with the proposed amendment would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

This proposed amendment removes the

requirement in the Technical Specifications

to derate to 90% of 2700 MWth for Cycle 14

operation beyond 7000 EFPH. There will be

no change to the modes of operation of the

plant. The plant configuration and the design

functions of all the safety systems remain

unchanged.

The proposed amendment will not change

the physical plant or the modes of operation

defined in the facility license. The changes

do not involve the addition of new

equipment or the modification of existing

equipment, nor do they alter the design of St.

Lucie plant systems. Therefore, operation of

the facility in accordance with the proposed

amendment would not create the possibility

of a new or different kind of accident from

any accident previously evaluated.

(3) Operation of the facility in accordance

with the proposed amendment would not

involve a significant reduction in a margin of

safety.

The impact of the proposed change on

available margin to the acceptance criteria for

Specified Acceptable Fuel Design Limits

(SAFDL), primary and secondary overpressurization, peak containment pressure,

potential radioactive releases, 10 CFR 50.46

requirements for the large break LOCA, and

existing limiting conditions for operation has

been evaluated and addressed in the reduced

RCS [reactor coolant system] flow operating

license Amendment No. 145. A requirement

to derate to 90% of 2700 MWth was imposed

based on the SBLOCA analysis. The small

break LOCA analysis with 30% plus or

minus 7% SGTP

supported operation up to 7000 EFPH at

100% of rated thermal power. A reanalysis of

SBLOCA with the limiting end-of-cycle

conditions at 100% of rated power,

demonstrates continued compliance with 10

CFR 50.46 criteria.

Therefore, operation of the facility in

accordance with the proposed amendment

would not involve a significant reduction in

a margin of safety.The NRC staff has

reviewed the licensee’s analysis and, based

on thisreview, 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: Indian River Junior College

Library, 3209 Virginia Avenue, Fort

Pierce, Florida 34954-9003

Attorney for licensee: M. S. Ross,

Attorney, Florida Power & Light, 11770

US Highway 1, North Palm Beach,

Florida 33408

NRC Project Director: Frederick J.

Hebdon

GPU Nuclear Corporation, Docket No.

50-289, Three Mile Island, Unit 1,

Dauphine County, Pennsylvania

Date of amendment request:

December 3, 1996

Description of amendment request:

This amendment will incorporate

certain improvements from the Standard

Technical Specifications for B&W Plants

(NUREG-1430).

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:

GPU Nuclear has determined that this

Technical Specification Change Request

involves no significant hazards consideration

as defined in 10 CFR 50.92 because:

1. Operation of the facility in accordance

with the proposed amendment would not
Federal Register / Vol. 62, No. 10 / Wednesday, January 15, 1997 / Notices 2191

involve a significant increase in the

probability of occurrence or the

consequences of an accident previously

evaluated. The proposed amendment deletes

limiting condition for operation (LCOs) from

the TMI-1 Technical Specifications that are

no longer required to be addressed in

Technical Specifications per 10 CFR

50.36(c)(2)(ii). The proposed amendment also

deletes a Surveillance requirement from the

TMI-1 Technical Specifications. This

surveillance requirement has no

corresponding LCO and is formatted in the

typical LCO format. These items are

addressed in licensee controlled documents.

This proposed amendment incorporates

relaxation of selected timeclocks and

surveillances frequencies consistent with

NUREG 1430 and adds a timeclock to a

unique LCO. The proposed changes do not

modify the operation, limits or controls of

systems, structures or components relied

upon to prevent or mitigate the consequences

[of] accidents previously evaluated. Also, the

reliability of systems and components relied

upon to prevent or mitigate the consequences

of accidents previously evaluated is not

degraded by the proposed changes.

Therefore, this change does not involve a

significant increase in the probability of

occurrence or the consequences of an

accident previously evaluated.

2. Operation of the facility in accordance

with the proposed amendment would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated because no new failure

modes are created by the proposed changes.

3. Operation of the facility in accordance

with the proposed amendment would not

involve a significant reduction in a margin of

safety. The proposed amendment does not

change any operating limits for reactor

operation.

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 staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Law/Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Walnut

Street and Commonwealth Avenue, Box

1601, Harrisburg, PA 17105.

Attorney for licensee: Ernest L. Blake,

Jr., Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037

NRC Project Director: John F. Stolz

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 amendment requests: October

25, 1996

Description of amendment requests:

The proposed amendments would

incorporate the requirements of 10 CFR

Part 50, Appendix J, Option B for

containment leakage tests. In addition,

the amendments would add a new

section to Technical Specifications,

which establishes the requirements of

the containment leakage rate testing

program, consistent with the Improved

Standard Technical Specifications.

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 amendment will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed changes provide a

mechanism within the TS for implementing

a performance-based leakage rate test

program which was promulgated by the

revision to 10 CFR Part 50 to incorporate

Option B to Appendix J. The proposed

changes do not involve any physical or

operational changes to structures, systems or

components. The current safety analyses and

safety design basis for the accident mitigation

functions of the containment, the airlocks,

and the containment isolation valves are

maintained. Since the allowable containment

leakage is still maintained within the

analyzed limit assumed in the accident

analysis, there is no adverse effect on either

onsite or offsite dose consequences.

Therefore, these changes will not increase the

probability or 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 analyzed.

The proposed changes do not involve any

physical or operational changes to structures,

systems or components. No new failure

mechanisms beyond those already

considered in the current plant safety

analyses are introduced. 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 amendment will not

involve a significant reduction in the margin

of safety.

Extending containment leakage rate test

intervals from those currently provided in

the Technical Specifications to those

provided for in 10 CFR (Part) 50 Appendix

J, Option B may slightly increase the risk due

to an increased likelihood of containment

leakage corresponding to the increased

testing intervals. However, this is somewhat

compensated by the corresponding risk

reduction benefits received from the

reduction in component cycling, stress, and

wear associated with the increased intervals.

When considering the total integrated risk,

which includes all analyzed accident

sequences, the possible additional risk

associated with increasing test intervals is

negligible.

The NRC letter to NEI (Nuclear Energy

Institute) dated November 2, 1995, recognizes

that changes similar to the proposed changes

at PINGP (Prairie Island Nuclear Generating

Plant) are required to implement Option B of

10 CFR (Part) 50, Appendix J. In NUREG1493, ‘‘Performance-Based Containment

Leak-Test Program’’, dated September 1995,

which forms the basis for the Appendix J

revision, the NRC concludes that adoption of

performance-based testing will not

significantly reduce the margin of safety. The

containment leak rate data and component

performance history at PINGP are consistent

with the conclusions reached in NUREG1493 and NEI 94-01. Thus, the proposed

license amendments do not involve a

significant reduction in a margin of safety

and will continue to support the regulatory

goal of ensuring an essentially leak-tight

containment boundary.

Based on the above, it is concluded that the

proposed change does not result in a

significant reduction in margin with respect

to plant safety as defined in the USAR or the

Technical Specification Bases.

Based on the evaluation described above,

and pursuant to 10 CFR Part 50, Section

50.91, Northern States Power Company has

determined that operation of the Prairie

Island Nuclear Generating Plant in

accordance with the proposed license

amendment request does not involve any

significant hazards considerations as defined

by NRC regulations in 10 CFR Part 50,

Section 50.92.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment requests involve no

significant hazards consideration.

Local Public Document Room

location: Minneapolis Public Library,

Technology and Science Department,

300 Nicollet Mall, Minneapolis,

Minnesota 55401

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts, and Trowbridge,

2300 N Street, NW, Washington, DC

20037

NRC Project Director: John N. Hannon

Pennsylvania Power and Light

Company, Docket Nos. 50-387 and 50-

388 Susquehanna Steam Electric

Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of amendment request:

November 18, 1996

Description of amendment request:

The amendments would amend the

Technical Specifications for

Susquehanna Steam Electric Station

(SSES), Units 1 and 2 by increasing the

maximum isolation times for the reactor

core isolation cooling inboard warm-up

line isolation valves (HV129F088 and

HV249F088) from 3 seconds to 12

seconds, the high pressure core
2192 Federal Register / Vol. 62, No. 10 / Wednesday, January 15, 1997 / Notices

injection inboard warm-up line isolation

valves (HV-155F100 and HV-255F100)

from 3 seconds to 6 seconds and the

reactor recirculation process sample line

(RRPSL) isolation valves (HV143F019

and HV243F019) from 2 seconds to 9

seconds.

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.

Chapters 6,9, and 15 of the FSAR [final

safety analysis report], current operating

cycles Reload Summary Reports for Units 1

and 2, Design Basis Document DBD046

(Seismic and Hydrodynamic Loads), and

NUREG-0776 (Safety Evaluation Report for

SSES), were reviewed to determine if the

proposed action has an effect on the

spectrum of analyzed anticipated operational

transients or postulated design basis

accidents.

The proposed modifications involve

replacing the pilot solenoid valves on the

Reactor Recirculation Loop ‘‘B’’ Process

Sample Line Isolation Valve (HV1/243F019)

and the inboard RCIC [reactor core isolation

cooling] and HPCI [high pressure core

injection] Steam Warm-Up Line Isolation

Valves (HV-1/249F088 and HV-1/255F100).

They do not alter any system operation or

control logic other than to increase the time

it takes for the associated containment

isolation valve to close. As discussed above,

the effects of the increased isolation times for

RCIC and HPCI impacted lines are bounded

by the larger parallel lines with isolation

times much greater than the new isolation

times for the smaller lines. In the case of the

Reactor Recirculation Loop ‘‘B’’ Process

Sample Line, the worst case scenario for a

line of that size is addressed in FSAR Section

15.6.2 and the results have been found

acceptable. In fact, the line breakage event

analyzed in the FSAR section postulates a

break outside containment that is not isolable

and that does not require operator action for

up to 10 minutes.

The modifications enhance isolation valve

performance by ensuring proper operation in

the event of a degraded air system.

Failures within the Process Sampling,

RCIC or HPCI systems or their components

are not postulated as causes of accident

scenarios nor is increasing the stroke time of

the subject containment isolation valves [HV1/243F019]. These systems provide safety

features utilized to mitigate the consequences

of the accidents. However, the failure mode

of the replacement solenoid valve is similar

in each case to that of the solenoid valve

being replaced in that it closes upon loss of

power or loss of air supply. The current

ability of the plant design to meet the single

failure criterion is unchanged by this

modification.

Based on the above discussion, the

proposed action does not involve a

significant increase in the probability or

consequences of an accident as previously

evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

Chapters 6, 9, and 15 of the FSAR were

reviewed to determine if the proposed action

[valve replacement with increased isolation

times for associated HPCI, RCIC, RRPSL

valves] has the potential of creating a

postulated initiating event which is different

than the analyzed anticipated operational

transients or postulated design basis accident

addressed. The review did not identify a

postulated initiating event which would

create the possibility for an accident of a

different type due to replacing the pilot

solenoid valves of the affected Reactor

Recirculation LOOP ‘‘B’’ Process Sample

Line or RCIC or HPCI Steam Warm-Up Line

isolation valves.

Also, the Reactor Recirculation Process

Sample Line, as part of the Process Sampling

System described in FSAR section 9.3.2.3,

does not perform any safety functions. It is

simply an alternate means for in line reactor

water chemistry monitoring upon the loss of

the RWCU system, and its loss does not

create any possibility for unevaluated

accidents or malfunctions.

Thus, replacing the pilot solenoid valves

on the affected Reactor Recirculation Process

Sample Line, RCIC Steam Warm-Up Line,

and HPCI Steam Warm-Up Line isolation

valves as well as relocating the Process

Sample Line solenoid valve for EQ

[equipment qualification] purposes does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed action involves replacing

existing pilot solenoid valves on containment

isolation valves for the Process Sampling,

RCIC, and HPCI Systems, as listed above,

with direct acting solenoid valves to ensure

proper valve operation in the event of a

degraded air or gas system as well as

relocating the Process Sampling pilot

solenoid valve for EQ purposes.

a. Reactor Recirculation Loop ‘‘B’’ Process

Sample Line

The limiting condition for the operation of

the Reactor Recirculation Loop ‘‘B’’ Process

Sample Line Inboard Isolation Valve (HV-1/

243F019) is governed by Technical

Specification Section 3/4.6.3 and its Bases

which presently requires this valve to close

within 2 seconds as defined in Technical

Specification Table 3.6.3-1. The proposed

modifications involve replacing the pilot

solenoid valve of the normally open isolation

valve (HV-1/243F019) with a direct acting

pilot solenoid valve as well as relocating the

pilot solenoid valve to assure an EQ life

which supports a 24 month operating cycle.

The combined effects of a lower flow

coefficient and relocating the solenoid valve

will require an increase in the Technical

Specification Table 3.6.3-1 isolation time

from 2 seconds to 9 seconds.

This increase in isolation time does not

reduce the margin of safety as defined in the

Technical Specification Section Basis,

because breakage of lines of this size is

addressed in the Susquehanna SES [steam

electric station] FSAR Section 15.6.2 and the

results found acceptable. In fact, the line

breakage event analyzed postulates a break

outside containment that is not isolable and

that does not require operator action for up

to 10 minutes. Also, it is noted that the

outboard isolation valve, HV-1/243F020, also

closes on the same containment isolation

signal, and its Technical Specification

isolation time limit remains 2 seconds.

The failure mode of the affected Reactor

Recirculation Loop ‘‘B’’ Process Sample Line

Inboard isolation valve is to close on loss of

power or air supply, therefore, the proposed

modifications do not affect the operability of

the isolation valve or reduce the margin of

safety.

b. RCIC

The limiting condition for operation of the

RCIC system is governed by Technical

Specification Section 3/4.7.3 and its Bases

which requires RCIC to be operable as the

primary non-ECCS source of emergency core

cooling. The proposed modifications involve

replacing the pilot solenoid valve of the

normally closed Steam Warm-Up Line

Isolation Valve (HV-1/249F088). This valve

can be manually opened in the absence of an

isolation signal to permit steam from the

reactor to pressurize and warm the steam

supply line downstream of the HV-1/

249F007 valve.

Installation of the direct acting solenoid

valve will require an increase in the

Technical Specification Section 3/4.6.3

isolation time for the RCIC Steam Warm-Up

Line Isolation Valve (HV-1/249F088) from 3

seconds to 12 seconds but does not reduce

the margin of safety as defined in the

Technical Specification Section Basis. The

increase in closure time for the HV-1/

249F088 isolation valve does not

compromise the overall line isolation due to

the fact that the impact of these 1’’ warm up

line valves is enveloped by the impact of the

much larger 4’’ RCIC inboard and outboard

isolation valves (HV-1/249F007 and HV-1/

249F008), which remain open an additional

8 seconds before isolating. The 4’’ valves are

the limiting components for providing

containment isolation for this line.

The failure mode of the affected RCIC

Steam Warm-Up Line Isolation Valve is to

close, if open, on loss of power or air supply,

therefore, the proposed modifications do not

affect the operability of the isolation valve or

reduce the margin of safety.

c. HPCI

The limiting condition for operation of the

HPCI system is governed by Technical

Specification Section 3/4.5.1 and its Bases

which requires HPCI to be operable for

proper Emergency Core Cooling System

operation. Operability includes the HPCI

pump and a flow path capable of taking

suction from the suppression pool and

delivering the water to the reactor vessel. The

proposed modifications involve replacing the

pilot solenoid valve of the normally closed

Steam Warm-Up Line Isolation Valve (HV-1/

255F100). This valve can be manually

opened in the absence of an isolation signal,

to permit steam from the reactor to pressurize
Federal Register / Vol. 62, No. 10 / Wednesday, January 15, 1997 / Notices 2193

and warm the steam supply line downstream

of the HV-1/255F002 valve.

Installation of the direct acting solenoid

valve will require an increase in the

Technical Specification Section 3/4.6.3

isolation time for the HPCI Steam Warm-Up

Line Isolation Valve (HV-1/255F100) from 3

seconds to 6 seconds but does not reduce the

margin of safety as defined in the Technical

Specification Section Basis. The increase in

closure time for the HV-1/255F100 isolation

valve does not compromise the overall line

isolation due to the fact that the impact of

these 1’’ warm up line valves is enveloped

by the impact of the much larger 10’’ HPCI

inboard and outboard isolation valves (HV-1/

255F002 and HV-1/255F003) which remain

open an additional 44 seconds before

isolating. The 10’’ valves are the limiting

components for providing containment

isolation for this line.

The failure mode of the affected HPCI

Steam Warm-Up Line Isolation Valve is to

close, if open, on loss of power or air supply,

therefore, the proposed modifications do not

affect the operability of the isolation valve or

reduce the margin of safety.

Thus, based on a review of the Technical

Specification, their Bases, the FSAR and

NUREG 0776 (Safety Evaluation Report for

SSES), the replacement of the pilot solenoid

valves does not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Osterhout Free Library,

Reference Department, 71 South

Franklin Street, Wilkes-Barre, PA 18701

Attorney for licensee: Jay Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street NW.,

Washington, DC 20037

NRC Project Director: John F. Stolz

Pennsylvania Power and Light

Company, Docket No. 50-388,

Susquehanna Steam Electric Station,

Unit 2, Luzerne County, Pennsylvania

Date of amendment request:

December 18, 1996

Description of amendment request:

The amendment would change the

Susquehanna Steam Electric Station

Unit 2 Technical Specifications to

reflect the use of a 24-month operating

cycle and the use of the ATRIUM-10

fuel design. The amendment includes

changes to two definitions in Section 1,

inclusion of new minimum critical

power ratio safety limits in Sections

2.1.2 and 3.4.1.1.2, changes in Section

5.3.1 to reflect the new fuel design, and

the listing of Siemens Power

Corporation topical reports in Section

6.9.3.2.

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:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The applicable sections of the FSAR [Final

Safety Analysis Report] are Chapters 5,6.3,9,

and 15 of the FSAR. Chapter 5 discusses the

results of the ASME overpressure analyses

for the reactor pressure boundary. Chapter

6.3 discusses the LOCA [loss-of-coolant

accident]. Chapter 9 discusses fuel storage

and handling. Chapter 15 describes the

transient and accident analyses, a majority of

which have been generically dispositioned to

be non-limiting. A discussion of the impact

of the Technical Specification changes is

provided below.

The change to Definitions 1.2 and 1.3

makes the definitions applicable to ATRIUM10. There are no effects on safety functions

from this change.

A cycle specific MCPR [minimum critical

power ratio] Safety Limit analysis was

performed for PP&L [Pennsylvania Power &

Light Company] by SPC [Siemens Power

Corporation]. This analysis used NRC

approved methods described in Technical

Specification Reference 13 (ANF-524(P)(A),

Revision 2 and Supplement 1 Revision 2.).

The SAFETY LIMIT MCPR calculation

statistically combines uncertainties on

feedwater flow, feedwater temperature, core

flow, core pressure, core power distribution,

and the uncertainty in the Critical Power

Correlation. The SPC analysis used cycle

specific power distributions and calculated

MCPR values such that at least 99.9% of the

fuel rods are expected to avoid boiling

transition during normal operation or

anticipated operational occurrences. The

resulting two-loop and single-loop values

(Technical Specification sections 2.1.2 and

3.4.1.1.2) are included in the proposed

change. Thus, the cladding integrity and its

ability to contain fission products is not

adversely affected.

The change to the Design Features (Section

5.3) increases the allowable enrichment.

Analyses have demonstrated that the

ATRIUM-10 fuel will remain subcritical (keffective<0.95) in both the spent fuel pool

and the new fuel vault. Thus, the change to

allowable enrichment has no impact on

safety functions. The description of a fuel

assembly (Section 5.3) is also revised to

reflect the ATRIUM-10 central water channel,

and reference to an active fuel length of 150

inches was deleted. This change reflects the

physical characteristics of the ATRIUM-10

fuel and has no impact on the probability or

consequences of an event.

Included in the revised Technical

Specifications via reference (Section 6.9.3.2)

are additional NRC approved methodology

reports. The NRC approved topical reports

contain methodology which is used to assure

safe operation of Unit 2 with ATRIUM-10

fuel. These methodologies assure that the

core meets appropriate margins of safety for

all expected plant operational conditions

ranging from refueling and cold shutdown of

the reactor through power operation. Thus,

the results obtained from the analyses will

provide assurance that the reactor will

perform its design safety function during

normal operation and design basis events.

The BASES changes for Section 2.1.1

(THERMAL POWER,Low Pressure or Low

Flow) reflect that the Safety Limit is valid for

both 9x9-2 and ATRIUM-10.

Therefore, the proposed action does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The changes to the Unit 2 Technical

Specifications (Definitions, MCPR safety

limits, Design Features, and inclusion of

methodology references) to allow use of

ATRIUM-10 fuel do not require any physical

plant modifications, physically affect any

plant components, or entail significant

changes in plant operation. Thus, the

proposed change does not create the

possibility of a previously unevaluated

operator error or a new single failure. The

referenced methodology added to Section

6.9.3.2 contains NRC approved acceptance

criteria. The consequences of transients and

accidents will remain within the criteria

approved by the NRC. Therefore, the

proposed change does not create the

possibility or a new or different kind of

accident from any accident previously

evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The applicable Technical Specification

Sections include 1.0, 2.0, 3/4.4, 5.3, and

6.9.3.2.

The changes to the Unit 2 Technical

Specifications discussed in Item 1 above

(Definitions, MCPR Safety Limits, Design

Features, and inclusion of methodology

references) to allow use of ATRIUM-10 fuel

do not require any physical plant

modifications, physically affect any plant

components, or entail significant changes in

plant operation. Therefore, the proposed

change will not jeopardize or degrade the

function or operation of any plant system or

component governed by Technical

Specifications. The NRC approved methods

detailed in the references added to Section

6.9.3.2 maintain an equivalent margin of

safety as currently defined in the bases of the

applicable Technical Specification sections.

Therefore, the proposed change does not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Osterhout Free Library,
2194 Federal Register / Vol. 62, No. 10 / Wednesday, January 15, 1997 / Notices

Reference Department, 71 South

Franklin Street, Wilkes-Barre, PA 18701

Attorney for licensee: Jay Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street NW.,

Washington, DC 20037

NRC Project Director: John F. Stolz

Tennessee Valley Authority, Docket

Nos. 50-259, 50-260 and 50-296,

BrownsFerry Nuclear Plant, Units 1, 2

and 3, Limestone County, Alabama

Date of amendment request:

December 11, 1996 (TS 386)

Description of amendment request:

The proposed amendment changes the

as-found tolerance for the main steam

system safety/relief valves (S/RV) from

plus or minus 1% to plus or minus 3%.

The licensee states that the proposed

change is consistent with methodology

submitted by the Boiling Water Reactor

Owners Group (BWROG) and approved

by the NRC.

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 amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

TVA [the Tennessee Valley Authority, the

licensee] is proposing a change to the ‘‘asfound’’ tolerances for the S/RV set points.

This proposed TS [technical specification]

amendment does not alter the frequency of

verifying the S/RV lift set points, or the

number of S/RVs required to be operable.

The amendment does not involve physical

changes or modifications to the S/RVs, or

change the operating mode or safety function

of the S/RVs. The safety lift set points will

still be required to be set within a tolerance

of plus or minus 1% following testing.

S/RV actuation is not a precursor to any

design basis accident analyzed in the BFN

[Browns Ferry Nuclear Plant] UFSAR

[Updated Final Safety Analysis Report].

Therefore, this change does not increase the

probability of any previously evaluated

accident.

Generic considerations related to the set

point tolerances were addressed in NEDC31753P [BWROG In-Service Pressure Relief

Valve Technical Specification Licensing

Topical Report] and previously reviewed by

NRC. In accordance with the NRC SER

[Safety Evaluation Report, see letter from A.

C. Thadani, NRC to C. L. Tully, BWROG,

dated March 8, 1993] on utilizing the NEDC

results, certain plant specific evaluations

were performed to support the proposed

change. Specifically, the current Unit 2

reload licensing report includes the transient

analyses for the anticipated operational

occurrences and the limiting

overpressurization transient utilizing the

plus or minus

3% S/RV set point tolerance and were

performed in accordance with NRC approved

methods. The alternate operating modes were

also included in the reload licensing report.

These analyses concluded there is adequate

margin to design core thermal limits and

pressure limits for the reactor vessel. The

corresponding Unit 3 core reload licensing

report for the next operating cycle (starts in

March 1997) is in progress and will also use

the plus or minus 3% S/RV set point

tolerance. Prior to the return of Unit 1 to

service, the same reload analysis will be

performed. Similar results to those for Unit

2 are expected.

The operation of high pressure injection

systems have been determined not to be

adversely affected by the proposed change.

LOCA [loss of coolant accident] response,

containment hydrodynamic loads, pump and

valve performance, and instrumentation

performance were likewise satisfactorily

evaluated. Therefore, this proposed change

does not significantly increase the

consequences of any previously evaluated

accident.

B. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed change does not involve a

modification to plant equipment. No new

failure modes are introduced. Plant systems

will continue to function and no new system

interactions are introduced by this proposed

change. Therefore, the proposed change does

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

C. The proposed amendment does not

involve a significant reduction in a margin of

safety.

The proposed change has been analyzed in

accordance with NRC approved methodology

and the margins of safety for the design basis

accidents and transients analyzed in Chapter

14 of the BFN UFSAR have not been

significantly reduced. Therefore, the

proposed change does not involve a

significant reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Athens Public Library, South

Street, Athens, Alabama 35611

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET llH,

Knoxville, Tennessee 37902

NRC Project Director: Frederick J.

Hebdon

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50-346, Davis-Besse Nuclear Power

Station, Unit No. 1, Ottawa County,

Ohio

Date of amendment request:

December 11, 1996

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) Section

1.0, ‘‘Definitions,’’ TS Table 1.2,

‘‘Frequency Notation,’’ TS Section 3/4.3,

‘‘Instrumentation,’’ and TS Section 3/

4.5, ‘‘Emergency Core Cooling Systems.’’

Surveillance requirements would be

modified to account for the increase in

the fuel cycle, consistent with Generic

Letter 91-04, ‘‘Changes in Technical

Specification Surveillance Intervals to

Accommodate a 24-month Fuel Cycle,’’

dated April 2, 1991. Administrative

changes consistent with the fuel cycle

change are also proposed.

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:

Toledo Edison has reviewed the proposed

changes and determined that a significant

hazards consideration does not exist because

operation of the Davis-Besse Nuclear Power

Station, Unit No. 1, in accordance with these

changes would:

1a. Not involve a significant increase in the

probability of an accident previously

evaluated because no such accidents are

affected by the proposed revisions to increase

the surveillance test intervals from 18 to 24

months for the subject Technical

Specifications (TS) 3.4.3.1.1, Reactor

Protection System Instrumentation; TS 3/

4.3.2.1, Safety Features Actuation System

Instrumentation; TS 3/4.3.2.2, Steam and

Feedwater Rupture Control System

Instrumentation; TS 3/4.3.3.1, Radiation

Monitoring Instrumentation; TS 3/4.3.3.5.2,

Remote Shutdown Instrumentation; TS 3/

4.3.3.6, Post-Accident Monitoring

Instrumentation, TS 3/4.5.1, Emergency Core

Cooling Systems (ECCS), Core Flooding

Tanks; and TS 3/4.5.2, Emergency Core

Cooling Systems, ECCS Subsystems - Tavg

greater than or equal to 280°F. Initiating

conditions and assumptions remain as

previously analyzed for accidents in the

DBNPS Updated Safety Analysis Report.

These revisions do not involve any

physical changes to systems or components,

nor do they alter the typical manner in which

the systems or components are operated.

Review results of historical 18 month

surveillance data and maintenance records

support an increase in the surveillance test

intervals from 18 to 24 months (and up to 30

months on a non-routine basis) because little,

if any, potential for an increase in a failure

rate of a system or component was identified

during these reviews.

These proposed revisions are consistent

with NRC guidance on evaluating and

proposing such revisions as provided in

Generic Letter 91-04, ‘‘Changes in Technical

Specification Surveillance Intervals to

Accommodate a 24-Month Fuel Cycle,’’ dated

April 2, 1991.

The proposed revision to Technical

Specification Table 1.2, Frequency Notation,

and the related proposed revision from an
Federal Register / Vol. 62, No. 10 / Wednesday, January 15, 1997 / Notices 2195

‘‘R’’ frequency notation to an ‘‘E’’ frequency

notation for Technical Specification

Surveillance Requirements that are

remaining on an 18 month frequency, are

administrative in nature, do not change

current actual Technical Specification

requirements, and do not affect previously

evaluated accidents.

1b. Not involve a significant increase in the

consequences of an accident previously

evaluated because the source term,

containment isolation or radiological releases

are not being changed by these proposed

revisions. Existing system and component

redundancy is not being changed by these

proposed changes. Existing system and

component operation is not being changed by

these proposed changes. The assumptions

used in evaluating the radiological

consequences in the DBNPS Updated Safety

Analysis Report are not invalidated.

The proposed revision to Technical

Specification Table 1.2, Frequency Notation,

and the related proposed revision from an

‘‘R’’ frequency notation to an ‘‘E’’ frequency

notation for Technical Specification

Surveillance Requirements that are

remaining on an 18 month frequency, are

administrative in nature, do not change

current actual Technical Specification

requirements, and do not affect previously

evaluated accidents.

2. Not create the possibility of a new or

different kind of accident from any accident

previously evaluated because these revisions

do not involve any physical changes to

systems or components, nor do they alter the

typical manner in which the systems or

components are operated.

Review results of historical 18 month

surveillance data and maintenance records

support an increase in the surveillance test

intervals from 18 to 24 months (and up to 30

months on a non-routine basis) because little,

if any, potential for an increase in a failure

rate of a system or component was identified

during these reviews. No changes are being

proposed to the type of testing being

performed, only to the length of the

surveillance test interval.

The proposed revision to Technical

Specification Table 1.2, Frequency Notation,

and the related proposed revision from an

‘‘R’’ frequency notation to an ‘‘E’’ frequency

notation for Technical Specification

Surveillance Requirements that are

remaining on an 18 month frequency, are

administrative in nature, do not change

current actual Technical Specification

requirements, and do not affect the manner

in which systems and components are being

operated or tested.

3. Not involve a significant reduction in a

margin of safety because the review results of

the historical 18 month surveillance data and

maintenance records identified little, if any,

potential for an increase in a failure rate of

a system or component due to increasing the

surveillance test interval to 24 months.

Existing system and component redundancy

is not being changed by these proposed

changes.

The proposed revision to Technical

Specification Table 1.2, Frequency Notation,

and the related proposed revision from an

‘‘R’’ frequency notation to an ‘‘E’’ frequency

notation for Technical Specification

Surveillance Requirements that are

remaining on an 18 month frequency, are

administrative in nature, do not change

current actual Technical Specification

requirements, and do not reduce the margin

of safety.

There are no new or significant changes to

the initial conditions contributing to accident

severity or consequences, therefore there are

no significant reductions 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: University of Toledo, William

Carlson Library, Government

Documents Collection, 2801 West

Bancroft Avenue, Toledo, Ohio 43606

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037

NRC Project Director: Gail H. Marcus

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.

Boston Edison Company, Docket No.

50-293, Pilgrim Nuclear Power Station,

Plymouth County, Massachusetts

Date of application for amendment:

May 1, 1996, as supplementedNovember

26, 1996.

Brief description of amendment: The

proposed amendment will modify Table

3.1.1, ‘‘Reactor Protection System

(SCRAM) Instrumentation

Requirement,’’ Table 3.2.C.1,

‘‘Instrumentation That Initiates Rod

Blacks,’’ and Technical Specification 3/

4.4, ‘‘Standby Liquid Control.’’

Date of issuance: December 27, 1996

Effective date: December 27, 1996

Amendment No.: 169

Facility Operating License No. DPR35: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 6, 1995 (61 FR 28606)

The November 26, 1996, letter provided

clarifying information that did not

change the initial proposed no

significant hazards consideration.The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated December 27, 1996. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts

02360.

Commonwealth Edison Company,

Docket Nos. 50-295 and 50-304, Zion

Nuclear Power Station, Units 1 and 2,

Lake County, Illinois

Date of application for amendments:

July 26, 1996, as supplemented on

September 3, 1996, September 18, 1996,

two submittals dated October 14, 1996,

October 22, 1996, two submittals dated

November 8, 1996, and December 17,

1996.

Brief description of amendments: The

amendments allow Commonwealth

Edison Company to control the reactor

coolant system pressure and

temperature limits for heatup,

cooldown, low temperature operation

and hydrostatic testing. They also revise

the reactor vessel material surveillance

program specimen withdrawal schedule
2196 Federal Register / Vol. 62, No. 10 / Wednesday, January 15, 1997 / Notices

such that the Unit 2 removal of capsule

X is delayed until 19 Effective Full

Power Years.

Date of issuance: December 20, 1996

Effective date: Immediately, to be

implemented within 60 days.

Amendment Nos.: 177 and 164

Facility Operating License Nos. DPR39 and DPR-48: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: September 25, 1996 (61 FR

50341). The September 3, 1996,

September 18, 1996, two submittals

dated October 14, 1996, October 22,

1996, two November 8, 1996, and

December 17, 1996, submittals provided

additional clarifying information that

did not change the initial proposed no

significant hazards consideration

determination. The Commission’s

related evaluation of the amendments is

contained in a Safety Evaluation dated

December 20, 1996.No significant

hazards consideration comments

received: No

Local Public Document Room

location: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois

60085.

Detroit Edison Company, Docket No. 50-

341, Fermi-2, Monroe County, Michigan

Date of application for amendment:

March 25, 1996 (NRC-96-0003)

Brief description of amendment: The

amendment revises the testing

requirements used to determine the

operability of the charcoal in the

engineered safety feature systems.

Date of issuance: December 23, 1996

Effective date: December 23, 1996,

with full implementation within 45

days

Amendment No.: 110

Facility Operating License No. NPF43. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: July 31, 1996 (61 FR 40014)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated December 23, 1996.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Monroe County Library

System, 3700 South Custer Road,

Monroe, Michigan 48161

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 for amendments:

December 11, 1996, as supplemented

December 17, 19, and 26, 1996

Brief description of amendments: The

amendments approve changes to the

Updated Final Analysis Report

(UFSAR), and require that the changes

be submitted with the next update of the

UFSAR pursuant to 10 CFR 50.71(e).

The associated Safety Evaluation

delineates the staff’s review and

findings regarding the one-time

emergency power engineered safeguards

functional test.

Date of issuance: January 2, 1997

Effective date: January 2, 1997

Amendment Nos.: 220, 220, 217

Facility Operating License Nos. DPR38, DPR-47, and DPR-55: The

amendments revised the Updated Final

Safety Analysis Report. Public

comments requested as to proposed no

significant hazards consideration: Yes.

(61 FR 66699 December 18, 1996) The

notice provided an opportunity to

submit comments on the Commission’s

proposed no significant hazards

consideration determination. No

comments have been received. The

notice also provided for an opportunity

to request a hearing by January 2, 1997,

as corrected to read January 17, 1997,

but indicated that if the Commission

makes a final no significant hazards

consideration determination, any such

hearing would take place after issuance

of the amendments.

The December 17, 19, and 26, 1996,

letters provided additional information

that did not change the scope of the

December 11, 1996, application and

initial proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendments, finding of exigent

circumstances, and a final no significant

hazards consideration determination are

contained in a Safety Evaluation dated

January 2, 1997.

Local Public Document Room

location: Oconee County Library, 501

West South Broad Street, Walhalla,

South Carolina 29691

Omaha Public Power District, Docket

No. 50-285, Fort Calhoun Station, Unit

No. 1, Washington County, Nebraska

Date of amendment request: May 31,

1996

Brief description of amendment: The

amendment revises the technical

specifications to increase the amount of

trisodium phosphate (TSP)

dodecahydrate located in the

containment sump storage baskets.

Date of issuance: December 30, 1996

Effective date: December 30, 1996

Amendment No.: 179

Facility Operating License No. DPR40. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 31, 1996 (61 FR 40025)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated December 30, 1996.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska

68102

Philadelphia Electric Company, Docket

No. 50-353, Limerick Generating

Station, Unit 2, Montgomery County,

Pennsylvania

Date of application for amendment:

August 1, 1996

Brief description of amendment: This

amendment revised the Technical

Specifications Section 3/4.4.6 (i.e.,

Figure 3.4.6.1-1) to reflect the addition

of two hydrotest curves, effective for 6.5

and 8.5 Effective Full Power Years

(EFPY), to the existing PressureTemperature Operating Limit (PTOL)

curves for LGS Unit 2.

Date of issuance: December 30, 1996

Effective date: As of date of issuance,

to be implemented within 30 days.

Amendment No.: 80

Facility Operating License No. NPF85. This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: November 6, 1996 (61 FR

57490) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

December 30, 1996.No significant

hazards consideration comments

received: No

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464

Notice Of Issuance Of Amendment To

Facility Operating License And FinalNo

Significant Hazards Consideration

Determination

During the period since publication of

the last biweekly notice, individual

notices of issuance of amendments have

been issued for the facilities as listed

below. These notices were previously

published as separate individual

notices. They are repeated here because

this biweekly notice lists all

amendments that have been issued for

which the Commission has made a final

determination that an amendment

involves no significant hazards

consideration.

In this case, a prior Notice of

Consideration of Issuance of

Amendment, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing was

issued, a hearing was requested, and the

amendment was issued before any

hearing because the Commission made

a final determination that the
Federal Register / Vol. 62, No. 10 / Wednesday, January 15, 1997 / Notices 2197

amendment involves no significant

hazards consideration.

Details are contained in the

individual notice as cited.

Philadelphia Electric Company, Docket

Nos. 50-352 and 50-353, Limerick

Generating Station, Unit 2, Montgomery

County, Pennsylvania

Date of amendment request:

December 6, 1996

Brief description of amendment

request: The amendment would revise

Technical Specification (TS) Section 2.1

and its associated TS Basis to reflect the

change in the Minimum Critical Power

Ratio Safety Limit due to the use of

GE13 fuel product line and the cyclespecific analysis performed by General

Electric Company (GE), for Limerick

Generating Station, Unit 2, Cycle 5.

Date of publication of individual

notice in Federal Register: December 23,

1996 (61 FR 67582)

Expiration date of individual notice:

January 22, 1997

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464

Dated at Rockville, Maryland, this 8th day

of January 1997.

For the Nuclear Regulatory Commission

Jack W. Roe,

Director, Division of Reactor Projects - III/

IV, Office of Nuclear Reactor Regulation

[Doc. 97–848 Filed 1–14–97; 8:45 am]

BILLING CODE 7590–01–F

OFFICE OF PERSONNEL

MANAGEMENT

Proposed Collection; Comment

Request for Reclearance of

Information Collection, OPM Form 805

Series

AGENCY: Office of Personnel

Management.

ACTION: Notice.

SUMMARY: In accordance with the

Paperwork Reduction Act of 1995 (Pub.

L. 104–13, May 22, 1995), this notice

announces that OPM will submit a

request to the Office of Management and

Budget for reclearance of the OPM Form

805 Series that collects information

from the public. OPM Form 805,

Application to be Listed Under the

Voting Rights Act of 1965, is used to

elicit information from persons applying

for voter registration under the authority

of the Voting Rights Act of 1965. The

requirements for voter eligibility vary

from State to State; therefore, OPM

Form 805 is a blanket number covering

a number of forms which conform to the

individual State’s requirements. For a

number of years, there have been forms

for 10 States: Alabama, Arizona,

Georgia, Louisiana, Mississippi, New

Mexico, North Carolina, South Carolina,

Texas (English and Spanish language

versions), and Utah. Because OPM has

never been asked to list voters in

Arizona, New Mexico, North Carolina,

and Utah, the approval of these four

forms is being permitted to lapse at the

request of the Voting Rights Section in

the Civil Rights Division of the

Department of Justice. The form requires

20 minutes to complete. Approximately

10 individuals complete the form

annually for a total public burden of 4

hours. For copies of this proposal call

James M. Farron on (202) 418–3208 or

e-mail to jmfarron@opm.gov.

DATES: Comments on this proposal

should be received on or before March

17, 1997.

ADDRESSES: Send or deliver comments

to—Steven R. Cohen, Assistant Director

for Merit Systems Oversight, Office of

Personnel Management, 1900 E Street,

NW., Room 7677, Washington, DC

20415–0001.

FOR FURTHER INFORMATION CONTACT: P.

Kaziah Clayton on (202) 606–2531 or email to pkclayto@opm.gov.

U.S. Office of Personnel Management.

Lorraine A. Green,

Deputy Director.

[FR Doc. 97–993 Filed 1–14–97; 8:45 am]

BILLING CODE 6325–01–M

PHYSICIAN PAYMENT REVIEW

COMMISSION

Sunshine Act Meeting

AGENCY: Physician Payment Review

Commission.

ACTION: Notice of meeting.

SUMMARY: The Commission will hold its

next public meeting on Thursday,

January 23, 1997, and Friday, January

24, 1997, at the Washington Marriott,

1221 22nd Street NW., Washington, DC,

in the DuPont Salon. The meetings are

tentatively scheduled to begin at 9:00

a.m. each day. In preparation for its

March 31 report, the Commission

expects to discuss such issues as

vulnerable populations, academic

health centers, quality of care, and

federal premium contributions. It will

also review draft chapters on PSOs,

access in Medicare managed care,

Medicare PPOs, risk adjustment,

secondary insurance for Medicare

beneficiaries, consumer protections in

managed care, and Medicare Fee

Schedule issues. Final agendas will be

mailed on January 17, 1997 and will be

available on the Commission’s web site

(WWW.PPRC.GOV) at that time.

ADDRESS: 2120 L Street NW., Suite 200,

Washington, DC 20037. The telephone

number is 202–653–7220.

FOR FURTHER INFORMATION CONTACT:

Annette Hennessey, Executive

Assistant, at 202–653–7220.

SUPPLEMENTARY INFORMATION: If you are

not on the Commission mailing list and

wish to receive an agenda, please call

202–653–7220 after January 16, 1997.

Lauren LeRoy, Ph.D.,

Executive Director.

[FR Doc. 97–1129 Filed 1–13–97; 3:08 pm]

BILLING CODE 6820–SE–M

UNITED STATES POSTAL SERVICE

BOARD OF GOVERNORS

Sunshine Act Meeting; Notification of

Items Added to Meeting Agenda

DATE OF MEETING: January 6, 1997.

STATUS: Closed.

PREVIOUS ANNOUNCEMENTS: 61 FR 65092,

December 10, 1996; and 61 FR 68081,

December 26, 1996.

CHANGE: At its meeting on January 6,

1997, the Board of Governors the United

States Postal Service voted unanimously

to add two items to the agenda of its

closed meeting held on that date:

4. Consideration of Personnel and

Compensation Issues.

5. Changes to the FY 1997 Advertising

Budget.

CONTACT PERSON FOR MORE INFORMATION:

Thomas J. Koerber, Secretary of the

Board, U.S. Postal Service, 475 L’Enfant

Plaza, S.W., Washington, D.C. 20260–

1000. Telephone (202) 268–4800.

Thomas J. Koerber,

Secretary.

Certified to be a true copy of the original

document.

Neva R. Watson,

Alternate Certifying Officer.

[FR Doc. 97–1057 Filed 1–10–97; 4:41 pm]

BILLING CODE 7710–12–M

SECURITIES AND EXCHANGE

COMMISSION

Submission for OMB Review;

Comment Request

Upon Written Request, Copies Available

From: Securities and Exchange

Commission, Office of Filings and

Information Services, Washington, DC

20549

Existing collection in use without an

OMB Number: