Federal Register / Vol. 60, No. 80 / Wednesday, April 26, 1995 / Notices 20513

NUCLEAR REGULATORY

COMMISSION

[Docket No. 50–244]

Rochester Gas and Electric

Corporation; Ginna Nuclear Power

Plant; Environmental Assessment and

Finding of No Significant Impact

The U.S. Nuclear Regulatory

Commission (the Commission) is

considering issuance of an exemption

from Facility Operating License No.

DPR–18, issued to Rochester Gas and

Electric Corporation (RG&E), (the

licensee), for operation of the Ginna

Nuclear Power Plant, located in Wayne

County, New York.

Environmental Assessment

Identification of the Proposed Action

The proposed action would grant a

one-time exemption from performing

Type C tests for certain containment

isolation valves (CIVs) during the 1995

refueling outage and extend the

schedule required by 10 CFR Part 50,

Appendix J, Section III.D.3, up to 1-

month of the 2-year interval.

The proposed action is in accordance

with the licensee’s application for the

exemption dated March 15, 1995.

The Need for the Proposed Action

The proposed action is requested on

a one-time basis only to support the

current refueling outage schedule.

Requiring a plant shutdown before the

next scheduled refueling outage in April

1996, soley to perform surveillance tests

would cause an unnecessary thermal

transient on the plant and could result

in unnecessary exposure to personnel.

The performance of the CIVs and the

plant’s overall containment integrity

have been good. RG&E proposes to limit

the exemption to exclude those valves:

(1) On which maintenance has been

performed; and (2) on those valves that

have not demonstrated acceptable

leakage rate testing during the past two

leakage tests.

Environmental Impacts of the Proposed

Action

The Commission has completed its

evaluation of the proposed action and

concludes that the proposed exemption

would allow RG&E to conduct the local

leak rate tests during the next refueling

outage, an extension of up to 1 month.

There will be no changes to the facility

or the environment as a result of the

exemption.

The change will not increase the

probability or consequences of

accidents, no changes are being made in

the types of any effluents that may be

released offsite, and there is no

significant increase in the allowable

individual or cumulative occupational

radiation exposure. Accordingly, the

Commission concludes that there are no

significant radiological environmental

impacts associated with the proposed

action.

With regard to potential

nonradiological impacts, the proposed

action does involve features located

entirely within the restricted area as

defined in 10 CFR Part 20. It does not

affect nonradiological plant effluents

and has no other environmental impact.

Accordingly, the Commission concludes

that there are no significant

nonradiological environmental impacts

associated with the proposed action.

Alternatives to the Proposed Action

Since the Commission has concluded

there is no measurable environmental

impact associated with the proposed

action, any alternatives with equal or

greater environmental impact need not

be evaluated. As an alternative to the

proposed action, the NRC staff

considered denial of the proposed

action. Denial of the application would

result in no change in current

environmental impacts of the proposed

action and the alternative action are

similar.

Alternative Use of Resources

This action does not involve the use

of any resources not previously

considered in the Final Environmental

Statement for the Ginna Nuclear Power

Plant.

Agencies and Persons Consulted

In accordance with its stated policy,

on April 11, 1995, the staff consulted

with the New York State official, Donna

Ross, Acting State Liaison Officer of the

New York Energy, Research, and

Development Authority, regarding the

environmental impact of the proposed

action. The State official had no

comments.

Finding of No Significant Impact

Based upon the environmental

assessment, the Commission concludes

that the proposed action will not have

a significant effect on the quality of the

human environment. Accordingly, the

Commission has determined not to

prepare an environmental impact

statement for the proposed action.

For further details with respect to the

proposed action, see the licensee’s letter

dated March 15, 1995, 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

Rochester Public Library, 115 South

Avenue, Rochester, New York.

Dated at Rockville, Maryland, this 19th day

of April 1995.

For the Nuclear Regulatory Commission.

Ledyard B. Marsh,

Director, Project Directorate I–1, Division of

Reactor Projects—I/II, Office of Nuclear

Reactor Regulation.

[FR Doc. 95–10207 Filed 4–25–95; 8:45 am]

BILLING CODE 7590–01–M

Biweekly Notice; Applications and

Amendments to Facility Operating

Licenses Involving No Significant

Hazards Considerations

I. Background

Pursuant to Public Law 97–415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 97–415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from March 31,

1995, through April 14, 1995. The last

biweekly notice was published on

Wednesday, April 12, 1995 (60 FR

18621).

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
20514 Federal Register / Vol. 60, No. 80 / Wednesday, April 26, 1995 / Notices

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 Rules Review and

Directives Branch, Division of Freedom

of Information and Publications

Services, Office of Administration, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555, and should cite

the publication date and page number of

this Federal Register notice. Written

comments may also be delivered to

Room 6D22, Two White Flint North,

11545 Rockville Pike, Rockville,

Maryland from 7:30 a.m. to 4:15 p.m.

Federal workdays. Copies of written

comments received may be examined at

the NRC Public Document Room, the

Gelman Building, 2120 L Street, NW.,

Washington, DC. The filing of requests

for a hearing and petitions for leave to

intervene is discussed below.

By May 26, 1995, the licensee may file

a request for a hearing with respect to

issuance of the amendment to the

subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) The nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington DC, by

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

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

1-(800) 342–6700). The Western Union

operator should be given Datagram

Identification Number N1023 and the

following message addressed to (Project

Director): petitioner’s name and

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555, and to the attorney for the

licensee.
Federal Register / Vol. 60, No. 80 / Wednesday, April 26, 1995 / Notices 20515

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.

Carolina Power & Light Company, et al.,

Docket No. 50–400, Shearon Harris

Nuclear Power Plant, Unit 1, Wake and

Chatham Counties, North Carolina

Date of amendment request: March

20, 1995.

Description of amendment request:

The licensee proposes a revision to

Technical Specification (TS) 2.2.1,

Reactor Trip System Instrumentation

Setpoints, and to relocate cycle specific

Overpower and Overtemperature Delta

T trip setpoint parameters to the Core

Operating Limits Report (COLR).

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:

This change does not involve a significant

hazards consideration for the following

reasons:

1. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change of relocating

Overtemperature Delta T * * * and the

Overpower Delta T * * * trip setpoint

parameters to the COLR has no influence or

impact to the probability or consequences of

an accident. The revised TS will continue to

implement the Reactor Trip System

Instrumentation [Overtemperature Delta T]

and [Overpower Delta T] setpoint limits

through reference to the parameters in the

COLR. In addition, the COLR is subject to the

existing controls of TS 6.9.1.6, including the

establishment of the parameter values using

an NRC approved methodology. Given that

this change administratively relocates the

selected trip setpoint parameter values to

another TS-controlled document, there

would be no increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

No safety-related equipment, safety

function, or plant operation will be altered as

a result of this proposed change. The limits

are simply being relocated to another TScontrolled document. The TS will continue

to require operation within the required

limits as established per NRC approved

methodologies. Therefore, the proposed

changes do not create the possibility of a new

or different kind of accident from any

accident previously evaluated.

3. The proposed amendment does not

involve a significant reduction in the margin

of safety.

Relocation of the Reactor Trip System

Instrumentation [Overtemperature Delta T]

and [Overpower Delta T] setpoint limits to

the TS-controlled COLR has no effect on the

trip system setpoints currently in force in TS

2.2.1. Future revisions to the trip setpoint

parameters are governed by TS 6.9.1.6. TS

6.9.1.6 lists each TS that references values in

the COLR and the NRC approved

methodologies utilized in developing those

values. Since this change is only an

administrative relocation of the selected trip

setpoint parameter values to another TS

controlled document, the proposed changes

do not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh,

North Carolina 27605.

Attorney for licensee: R. E. Jones,

General Counsel, Carolina Power &

Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: David B.

Matthews.

Carolina Power & Light Company, et al.,

Docket No. 50–400, Shearon Harris

Nuclear Power Plant, Unit 1, Wake and

Chatham Counties, North Carolina

Date of amendment request: March

30, 1995.

Description of amendment request:

The licensee proposes to revise the

Emergency Diesel Generator (EDG)

surveillance requirements contained in

Technical Specification (TS) 4.8.1.1.2 to

be consistent with NUREG–1431,

Standard Technical Specifications for

Westinghouse Plants, and to eliminate

the need for duplicate EDG testing that

has already been implemented to satisfy

the requirements of the Station Blackout

Rule and the Maintenance Rule.

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:

This change does not involve a significant

hazards consideration for the following

reasons:

1. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

A failure of the Emergency Diesel

Generator (EDG) is not an initiator for any

previously evaluated FSAR Chapter 15

accident scenario. By committing to and

implementing an EDG reliability program

that satisfies the requirements of the Station

Blackout Rule and the Maintenance Rule, the

Shearon Harris Nuclear Power Plant (SHNPP)

will continue to ensure that target EDG

reliability and availability is being achieved

by conducting appropriate monitoring,

testing, and maintenance activities. This

program will be developed and controlled as

a Plant Operating Manual procedure and will

incorporate industry, vendor, and TDI

Owners Group recommendations. Therefore,

with commensurate levels of testing and

inspection in place to provide assurance that

the EDGs will perform their intended safety

function in the event of an accident, the

proposed changes will have no effect on the

probability or consequences of such an

accident.

2. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

A failure of the EDG is not an initiator for

any previously evaluated FSAR Chapter 15

accident scenario nor would the proposed

changes to the EDG surveillance

requirements result in the possibility of a

new or different kind of accident from any

accident previously evaluated. By

committing to and implementing an EDG

reliability program that satisfies the

requirements of the Station Blackout Rule

and the Maintenance Rule, SHNPP will

continue to ensure that target EDG reliability

and availability is being achieved by

conducting appropriate monitoring, testing,

and maintenance activities. This program

will be developed and controlled as a Plant

Operating Manual procedure and will

incorporate industry, vendor, and

TransAmerica Delaval Inc. Owners Group

recommendations. Therefore, with

commensurate levels of testing and

inspection in place to provide assurance that

the EDGs will perform their intended safety

function in the event of an accident, the

proposed changes would not increase the

possibility of a new or different kind of

accident from any previously evaluated.

3. The proposed amendment does not

involve a significant reduction in the margin

of safety.

The proposed changes will not affect any

parameters which relate to the margin of

safety as defined in the Technical

Specifications or the FSAR. Testing,

inspection and maintenance necessary to

verify the EDGs’ ability to perform their

intended safety function will continue to be
20516 Federal Register / Vol. 60, No. 80 / Wednesday, April 26, 1995 / Notices

performed. 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: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh,

North Carolina 27605.

Attorney for licensee: R. E. Jones,

General Counsel, Carolina Power &

Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: David B.

Matthews.

Commonwealth Edison Company,

Docket Nos. 50–295 and 50–304, Zion

Nuclear Power Station, Units 1 and 2,

Lake County, Illinois

Date of amendment request: March

24, 1995.

Description of amendment request:

The proposed amendments would

acknowledge the acceptability of

performing containment leakage rate

testing in accordance with 10 CFR Part

50, Appendix J, and all approved

exemptions.

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 of occurrence or consequences of

any accident previously evaluated.

The proposed changes to Technical

Specifications to add an allowance to test in

accordance with approved exemptions to 10

CFR 50 Appendix J are administrative in

nature and will not affect any accident

initiators or precursors. 10 CFR 50 Appendix

J provides the requirements to periodically

test the primary reactor containment. The

objective of these requirements is to ensure

that leakage from the primary reactor

containment structure and systems and

components that penetrate the containment

is maintained below the limits established for

containment leakage. The performance of

periodic integrated leakage rate testing (Type

A) and local penetration testing (Type B and

C) during containment life provides a current

assessment of potential leakage from

containment during accident conditions.

10 CFR 50.12 allows the Commission to

grant specific exemptions to the requirements

of 10 CFR 50 Appendix J when those

exemptions are authorized by law, will not

present undue risk to the public, and are

consistent with the common defense and

security. In addition, special circumstances

must exist as described in Section 50.12.

Since all exemptions to 10 CFR 50 Appendix

J receive NRC review and approval prior to

being implemented, all containment leakage

rate testing will continue to be performed in

accordance with NRC approved

methodologies when relying upon the

allowance that is added to the Technical

Specifications by the proposed amendment.

The proposed changes are consistent with the

requirements provided in NUREG–1431,

‘‘Standardized Technical Specifications,

Westinghouse Plants’’ which has been

approved by the NRC.

The proposed changes will not affect any

accident initiators or precursors and will not

change or alter the design assumptions for

the systems used to mitigate the

consequences of an accident. The proposed

changes do not involve the addition of any

new or different type of equipment, nor do

they involve the operation of equipment

required for safe operation of the facility in

a manner different from those addressed in

the UFSAR. There are no changes to

parameters governing plant operation as a

result of the proposed changes. The results

and conclusions in the Zion Updated Final

Safety Analysis Report (UFSAR) are

unaffected by this proposed License

Amendment.

Based on the previous discussion, the

proposed changes do not involve a

significant increase in the probability of

occurrence or consequences of any accident

previously evaluated.

2. The proposed amendment does not

create the possibility of a new or different

kind of accident from any previously

analyzed.

The proposed changes to Technical

Specifications to add an allowance to

perform containment leakage rate testing in

accordance with approved exemptions to 10

CFR 50 Appendix J are administrative in

nature and do not involve the addition of any

new or different types of safety related

equipment, nor does it involve the operation

of equipment required for safe operation of

the facility in a manner different from those

addressed in the safety analyses. The

proposed changes may only affect the

methods used to perform containment

leakage rate testing while in a shutdown

condition. No safety related equipment or

function will be altered as a result of the

proposed changes. Also, the procedures

governing normal plant operation and

recovery from an accident are not changed by

the proposed Technical Specification

changes. Since no new failure modes or

mechanisms are added by the proposed

changes, the possibility of a new or different

kind of accident is not created.

3. The proposed changes do not involve a

significant reduction in a margin of safety.

Given the proposed changes to Technical

Specifications, testing would be allowed in

accordance with approved exemptions to

Appendix J. Exemptions are allowed by the

Commission in accordance with 10 CFR

50.12 when it is shown that the exemption

is authorized by law, will not present an

undue risk to the public health and safety,

and is consistent with the common defense

and security. In addition, special

circumstances must exist.

The proposed changes will not impact any

margin of safety and testing in accordance

with approved exemptions will not involve

a significant reduction in a level of safety

since containment leakage testing is

performed while in a shutdown condition. In

addition, it is likely that any test

methodology that significantly reduces a

margin of safety would not be approved by

the NRC.

The ability to safely shut down the

operating unit and mitigate the consequences

of all accidents previously evaluated will be

maintained. Therefore, the margin of safety is

not significantly affected.

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: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois

60085.

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60690.

NRC Project Director: Robert A. Capra.

Consumers Power Company, Docket No.

50–155, Big Rock Point Plant,

Charlevoix County, Michigan

Date of amendment request:

December 15, 1994.

Description of amendment request:

The proposed amendment would revise

Technical Specification 11.3.1.5

ACTION a. to eliminate the need to

demonstrate that the actuation circuitry

of the unaffected reactor

depressurization system channels is

operable. In addition, an editorial

change correcting a typographical error

is 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:

1. Will the proposed change involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

The proposed change will eliminate the

probability of a subsystem failure caused by

additional testing (which unnecessarily

introduces the potential for human and

equipment problems), therefore eliminating

the probability that the facility would have

to be challenged and brought to the

SHUTDOWN condition within 12 hours and

to the COLD SHUTDOWN condition within

the following 24 hours.
Federal Register / Vol. 60, No. 80 / Wednesday, April 26, 1995 / Notices 20517

2. Will the proposed change create the

possibility of a new or different kind of

accident from any accident previously

evaluated?

The proposed change does not alter the

plant configuration, systems, components, or

operation; and does not create the possibility

of a new or different kind of accident from

any accident previously evaluated. The

proposed change is expected to eliminate

unnecessary challenges to a safety system

that has already been determined to be

operable by routine surveillance testing;

therefore contributing to the overall safe

operation of the facility.

3. Will the proposed change involve a

significant reduction in the margin of safety?

The RDS [Reactor Depressurization

System] provides for both manual and

automatic depressurization of the primary

system to allow injection of the core spray

following a small-to-intermediate size break

in the primary system. This will allow core

cooling with the objective of preventing

excessive fuel clad temperatures. The design

of the system is based on the specified

initiation set points described in the

Technical Specifications. Transient analysis

demonstrated that these conditions result in

adequate safety margins for both the fuel and

the system pressure. The proposed change

does not affect these setpoints, therefore the

margin of safety is not changed.

In addition, the proposed editorial

change to correct a typographical error

is administrative in nature and,

therefore, would have no effect on the

three standards of 10 CFR 50.92

discussed above.

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: North Central Michigan

College, 1515 Howard Street, Petoskey,

Michigan 49770.

Attorney for licensee: Judd L. Bacon,

Esquire, Consumers Power Company,

212 West Michigan Avenue, Jackson,

Michigan 49201.

NRC Project Director: Cynthia A.

Carpenter, Acting.

Florida Power Corporation, et al.,

Docket No. 50–302, Crystal River

Nuclear Generating Plant, Unit No. 3,

Citrus County, Florida

Date of amendment request: January

26, 1995, as supplemented March 9,

1995.

Description of amendment request:

The proposed amendment would revise

the technical specifications (TS) to

increase the allowable nominal fuel

enrichment from 4.2 to 5.0 weight

percent for reload fuel assemblies. TS

impose a limit on fuel enrichment of

stored fuel assemblies to prevent

inadvertent criticality. Presently, the

Crystal River Unit 3 (CR3) TS specify a

maximum enrichment of 4.5 weight

percent for storage pool A and dry fuel

(new fuel) storage racks, and 4.2 weight

percent for fuel pool B. The licensee

proposed to revise TS 3.7.15, 4.2, and

4.3, and associated TS bases to allow

increasing the enrichment limits from

4.2 to 5.0 weight percent for the dry fuel

storage racks and for A and B fuel pools.

Additionally, a typographical error in

TS 4.3.1.2.b will also be corrected.

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. This amendment will not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

An increase in fuel enrichment will not by

itself affect the mixture of fission product

nuclides. A change in fuel cycle design

which makes use of an increased enrichment

may result in fuel burnup consisting of a

somewhat different mixture of nuclides. The

effect in this instance is insignificant

because:

a. The isotopic mixture of the irradiated

assembly is relatively insensitive to the

assembly’s initial enrichment.

b. Most accident doses are such a small

fraction of 10 CFR 100 limits, a large margin

exists before any change becomes significant.

c. The change in Pu content which would

result from an increase in burnup would

produce more of some fission product

nuclides and less of other nuclides. Small

increases in some doses are offset by

reductions in other doses. The radiological

consequences of accidents are not

significantly changed.

2. This amendment will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

An unplanned criticality event will not

occur as Keff [effective neutron multiplication

factor] will not exceed 0.95 with the

maximum allowable enriched fuel in Pool A

and Pool B, when flooded with unborated

water, and Keff will not exceed 0.98 in the

new fuel storage racks assuming conditions

of optimum hypothetical low density

moderation. The new fuel storage racks have

two rows of storage cells physically blocked

to ensure reactivity limits are not exceeded.

Administrative controls assure fuel is stored

in configurations which meet the

requirements of the safety analysis.

3. This amendment will not involve a

significant reduction in a margin of safety.

While the increased enrichment in Pool A,

Pool B, and the dry storage racks may lessen

the margin to criticality, this reduction is not

significant because the overall safety margin

is within NRC criteria of Keff [less than or

equal to] 0.95 (NRC Standard Review Plan,

Section 9.1.2.)

Therefore, this amendment request satisfies

the criteria specified in 10 CFR 50.92 for

amendments which do not involve a

significant hazards consideration.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida

32629.

Attorney for licensee: A. H. Stephens,

General Counsel, Florida Power

Corporation, MAC–A5D, P. O. Box

14042, St. Petersburg, Florida 33733.

NRC Project Director: David B.

Matthews.

Houston Lighting & Power Company,

City Public Service Board of San

Antonio, Central Power and Light

Company, City of Austin, Texas, Docket

Nos. 50–498 and 50–499, South Texas

Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request: March

16, 1995.

Description of amendment request:

The proposed amendment would revise

Technical Specification 4.6.1.2,

regarding the overall integrated

containment leakage rate tests, so that it

would reference 10 CFR Part 50,

Appendix J directly, rather than

paraphrase the regulation, and allow

approved exemptions to the test

frequency requirements. In addition,

there is an associated proposed

exemption, from the requirements of 10

CFR Part 50, Appendix J, to provide a

one-time interval extension for the Unit

2 Type A test (containment integrated

leak rate test) from the current

scheduled 48 months to approximately

66 months.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Criterion 1—Does Not Involve a Significant

Increase in the Probability or Consequences

of an Accident Previously Evaluated

The proposed change specific to Unit 2

will provide a onetime exemption from the

10 CFR 50, Appendix J Section III.D.I.(a) leak

rate test schedule requirement. This change

will allow for a one-time test interval for

Type A Integrated Leak Rate Tests of

approximately 66 months.

Leak rate testing is not an initiating event

in any accident; therefore, this proposed
20518 Federal Register / Vol. 60, No. 80 / Wednesday, April 26, 1995 / Notices

change does not involve a significant

increase in the probability of a previously

evaluated accident.

Type A tests are capable of detecting both

local leak paths and gross containment

failure paths. Experience at South Texas

Project Unit 2 demonstrates that excessive

containment leakage paths are local leakage

detected by Type B and C Local Leak Rate

Tests.

Administrative controls govern

maintenance and testing of containment

penetrations such that the probability of

excessive penetration leakage due to

improper maintenance or valve misalignment

is very low. Following maintenance on any

containment penetration, a Local Leak Rate

Test is performed to ensure acceptable

leakage levels. Following a Local Leak Rate

Test on a containment isolation valve, an

independent valve alignment check is

performed. Therefore, Type A testing is not

necessary to ensure acceptable leakage rates

through containment penetrations.

While Type A testing is not necessary to

ensure acceptable leakage rates through

containment penetrations, Type A testing is

necessary to demonstrate that there are no

gross containment failures. Structural failure

of the containment is considered to be a very

unlikely event, and in fact, since South Texas

Project Unit 2 has been in operation, it has

successfully passed each Type A Integrated

Leak Rate Test. Therefore, a one-time

exemption increasing the interval for

performing an Integrated Leak Rate Test

results [sic] in a significant decrease in the

confidence in the leak tightness of the

containment structure. Therefore, this change

does not involve a significant increase in the

consequences of an accident previously

evaluated.

The proposed amendment revised

Technical Specification 4.6.1.2 to reference

the testing frequency requirements of 10 CFR

50, Appendix J, and to state that Nuclear

Regulatory Commission approved

exemptions to the applicable regulatory

requirements are permitted. This portion of

the proposed change is applicable to Unit 1

and Unit 2. The current language of

Technical Specification 4.6.1.2 paraphrases

the requirements of Section III,D.I.(a) [sic] of

Appendix J. The proposed administrative

revision simply deletes the paraphrased

language and directly references Appendix J.

No new requirements are added, nor are any

existing requirements deleted. Any specific

changes to the requirements of Section

III.D.I.(a) will require a submittal from

Houston Lighting & Power under 10 CFR

50.12 and subsequent review and approval

by the Nuclear Regulatory Commission prior

to implementation.

The proposed amendment, in itself, does

not affect reactor operations or accident

analysis and has no radiological

consequences. The change provides

clarification so that future Technical

Specification changes will not be necessary

to correspond to applicable Nuclear

Regulatory Commission-approved

exemptions from the requirements of

Appendix J.

Therefore, this proposed amendment 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 Unit 2 exemption request

does not affect normal plant operations or

configuration, nor does it affect leak rate test

methods. The proposed change allows a onetime test interval of approximately 66 months

for the Integrated Leak Rate Test. Because the

test history of South Texas Project Unit 2

demonstrates no Type A test failures during

plant lifetime, the relaxation in schedule

should not significantly decrease the

confidence in the leak tightness of the

containment.

The proposed Technical Specification

amendment for Units 1 and 2 provides

clarification to a specification that

paraphrases a codified requirement.

Since the proposed change and

amendment would not change the design,

configuration or method of operation of the

plant, they would not create the possibility

of a new or different kind of accident from

any previously evaluated.

Criterion 3—Does Not Involve a Significant

Reduction in the Margin of Safety

The purpose of the existing schedule for

Integrated Leak Rate Tests is to ensure that

release of radioactive materials will be

restricted to those leak paths and leak rates

assumed in accident analyses. The relaxed

schedule for Integrated Leak Rate Tests does

not allow for relaxation of Type B and C

Local Leak Rate Tests. Therefore, methods for

detecting local containment leak paths and

leak rates are unaffected by this proposed

change. A one-time increase of the South

Texas Project Unit 2 test interval does not

leak to a significant probability of creating a

new leakage path or increased leakage rates

because the test history for Integrated Leak

Rate Tests shows no failure during plant life.

The margin of safety inherent in existing

accident analyses is maintained.

The proposed Technical Specification

amendment for Units 1 and 2 is

administrative and clarifies the relationship

between the requirements of Technical

Specification 4.6.1.2, Appendix J, and any

approved exemptions to Appendix J. It does

not, in itself, change a safety limit, a Limiting

Condition of Operation, or a surveillance

requirement on equipment required to

operate the plant. Nuclear Regulatory

Commission approval of any proposed

change or exemption to III.D.1.(a) of

Appendix J will be required prior to

implementation.

Therefore, this change and amendment 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 standards of

10 CFR 50.92(c) are satisfied. Therefore,

the NRC staff proposes to determine that

the request for amendments involves no

significant hazards consideration.

Local Public Document Room

location: Wharton County Junior

College, J. M. Hodges, Learning Center,

911 Boling Highway, Wharton, TX

77488.

Attorney for licensee: Jack R.

Newman, Esq., Newman & Holtzinger,

P.C., 1615 L Street, NW., Washington,

DC 20036.

NRC Project Director: William D.

Beckner.

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

Duane Arnold Energy Center, Linn

County, Iowa

Date of amendment request: March

10, 1995.

Description of amendment request:

The proposed amendment would

remove redundant Limiting Conditions

of Operation and Surveillance

Requirements for the containment

hydrogen and oxygen monitors in the

Technical Specifications (TS).

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. No physical changes

will result from this amendment. This change

deletes requirements that are redundant and

unduly restrictive. The annual surveillance

deleted by this amendment is redundant to

the semi-annual surveillance required in

Table 4.2–H. The Limiting Conditions for

Operation are not changed by the proposed

amendment.

2. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated. No physical changes

will result from this amendment. Functional

tests are performed on the hydrogen and

oxygen analyzers semiannually as required in

TS Table 4.2–H. Deleting the annual

requirement for a functional test of the same

equipment will not reduce the amount of

testing performed or increase the possibility

of degraded equipment being undetected.

3. The proposed amendment does not

involve a significant reduction in a margin of

safety. No physical changes will result from

this amendment. The existing requirement

for a semi-annual test of the hydrogen and

oxygen analyzer in Table 4.2–H exceeds the

requirements to be deleted in Section 3.7/

4.7–H. The frequency of testing of the

hydrogen and oxygen analyzers will not be

reduced as a result of this amendment.

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.
Federal Register / Vol. 60, No. 80 / Wednesday, April 26, 1995 / Notices 20519

Local Public Document Room

location: Cedar Rapids Public Library,

500 First Street, S.E., Cedar Rapids,

Iowa 52401.

Attorney for licensee: Jack Newman,

Kathleen H. Shea, Morgan, Lewis &

Bockius, 1800 M Street, NW.,

Washington, DC 20036–5869.

NRC Project Director: Gail H. Marcus.

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

Duane Arnold Energy Center, Linn

County, Iowa

Date of amendment request: March

28, 1995.

Description of amendment request:

The proposed amendment would revise

and clarify Technical Specification

Table 3.2–A that lists allowable out-ofservice times and surveillance test

intervals for instrumentation.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed changes to TS Table 3.2–

A will not significantly increase the

probability or consequences of an accident

previously evaluated. The changes do not

alter the physical design or operation of the

plant and serve to describe more accurately

and clearly the actual logic configurations.

The existing logic designs are in conformance

with the Architect/Engineer’s design

documentation since plant startup. These

changes will assure that the information in

the tables is clearer and more consistent with

the column headings of the table. The

proposed changes do not affect assumptions

contained in the plant safety analysis.

The Bases changes provide additional

information about the logic arrangements as

appropriate to identify unique or different

logic configurations. Changes to the Allowed

Outage Time (AOT) descriptions for the MSL

Flow—High and MSL Tunnel Temperature—

High provide clarification regarding

application of the AOT to these logic

arrangements, since multiple instrument

channels provide input into multiple logic

channels. This application conforms to the

single failure criterion of the design basis

(NEDO–10139, Compliance of Protection

Systems to Industry Criteria: General Electric

BWR Nuclear Steam Supply System, dated

June 1970) and to the analytical basis for the

TS (NEDC–31677P–A, Technical

Specification Improvement Analysis for BWR

Isolation Actuation Instrumentation, dated

July 1990).

2. The proposed changes to Table 3.2–A

will not introduce a new or different kind of

accident from any accident previously

evaluated. The changes do not alter the

physical design of the plant or affect any

modes of operation. The proposed changes

serve to clarify the existing information to

better assure that the trip instrumentation

will be maintained as assumed in the

accident analyses contained in the Updated

Final Safety Analysis Report.

3. The proposed changes do not involve a

significant reduction in a margin of safety.

Clarification of the logic arrangements in

both Table 3.2–A and the TS Bases and how

the AOT is applied does not affect the ability

of the isolation logic to perform its intended

function. No physical changes to the plant

are being made as part of this amendment.

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: Cedar Rapids Public Library,

500 First Street, S.E., Cedar Rapids,

Iowa 52401.

Attorney for licensee: Jack Newman,

Kathleen H. Shea, Morgan, Lewis &

Bockius, 1800 M Street, NW.,

Washington, DC 20036–5869.

NRC Project Director: Gail H. Marcus.

Indiana Michigan Power Company,

Docket No. 50–315, Donald C. Cook

Nuclear Plant, Unit No. 1, Berrien

County, Michigan

Date of amendment request: March

17, 1995.

Description of amendment request:

The proposed amendment would defer

performance of the Type A containment

integrated leakage rate test until the next

refueling outage.

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:

Per 10 CFR 50.92, a proposed change does

not involve a significant hazards

consideration if the change does not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated,

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated, or

3. Involve a significant reduction in a

margin of safety.

Criterion 1

The Cook Nuclear Plant Type A test history

provides substantial justification for the

proposed test schedule. Three Type A tests

were performed over a seven year period

with successful results. The tests indicate

that the Cook Nuclear Plant has a low leakage

containment. In addition, there are no

adverse trends in the results from the

previous Types A, B, and C tests or visual

inspections that indicate a gradual

degradation of the containment boundary.

Further, there are no structural modifications

planned which would adversely affect the

structural capability of the containment and

that would be a factor in deferring the Type

A test one refueling outage. Containment leak

rate testing is not an initiator of any accident.

The proposed interval extension does not

affect reactor operations or the accident

analysis and has no radiological

consequences, except for the dose savings

associated with not performing the test.

There will be no changes to 10 CFR 100 dose

limits or the control room dose limits.

Extending the test interval will not increase

the probability of a malfunction of equipment

important to safety. Based on these

considerations, it is concluded that the

change does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

Criterion 2

The proposed change does not involve

physical changes to the plant or changes in

plant operating configuration. The proposed

change only relaxes the schedular

requirements for conducting one Type A test

from the T/Ss and defers performance of the

test one cycle. The purpose of the test is to

provide periodic verification of the leak-tight

integrity of the primary reactor containment,

and systems and components which

penetrate containment. The tests assure that

leakage through containment and systems

and components penetrating containment

will not exceed the allowable leak rate values

established in 10 CFR 50, Appendix J. Thus,

it is concluded that the proposed change

does not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

Criterion 3

The proposed change to the schedule for

performing the Type A test does not reduce

the margin of safety assumed in the accident

analysis for any release of radioactive

materials or reduce any margin of safety

preserved by the technical specifications.

The methodology, acceptance criteria, and

the technical specification leak rate limits for

the performance of the Type A test will not

change. Type A tests will continue to be

performed in accordance with 10 CFR 50,

Appendix J and the applicable Cook Nuclear

Plant Technical Specifications beginning in

1997. In addition, there are no adverse trends

in the results from the previous Type A, B,

and C tests or visual inspections that indicate

a gradual degradation of the containment

boundary. Based on these considerations, it

is concluded that the change does not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Maud Preston Palenske

Memorial Library, 500 Market Street, St.

Joseph, Michigan 49085.

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts and
20520 Federal Register / Vol. 60, No. 80 / Wednesday, April 26, 1995 / Notices

Trowbridge, 2300 N Street, NW,

Washington, DC 20037.

NRC Project Director: Cynthia A.

Carpenter, Acting.

Indiana Michigan Power Company,

Docket Nos. 50–315 and 50–316, Donald

C. Cook Nuclear Plant, Unit Nos. 1 and

2, Berrien County, Michigan

Date of amendment requests: March

31, 1995.

Description of amendment requests:

The proposed amendments would

modify the Containment Ventilation

System Technical Specifications (and

associated Bases) to allow limited

containment purge operation in Modes

1, 2, 3, and 4 for pressure control,

ALARA [as low as is reasonably

achievable], and respirable air quality

considerations.

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

The purpose of this amendment is to allow

flexibility in the use of the containment

purge system during MODES 1, 2, 3, and 4.

The use of this system during these modes

of operation has previously been approved

(Amendment No. 66). Therefore, this

amendment request does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated because the proposed change to the

T/Ss does not affect the assumptions,

parameters, or results of any UFSAR

[Updated Final Safety Analysis Report]

accident analysis. Based on the existing

system design and demonstrated closure

capability it is concluded that the proposed

changes do not modify the response of the

containment during a design basis accident.

The proposed amendment does not add or

modify any existing equipment. Based on

these considerations, it is concluded that the

changes do not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

Criterion 2

The proposed change does not involve

physical changes to the plant or changes in

the plant operating configuration. Thus, it is

concluded that the proposed change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

Criterion 3

The margin for safety presently provided is

not reduced by the proposed change. As

discussed previously, the containment purge

valves have been designed and demonstrated

capable of closure against the dynamic forces

resulting from a loss of coolant accident. The

proposed amendment does not impact the

ability of the purge valves to perform their

intended function (i.e. achieve closure) in the

event of an accident. Based on these

considerations, it is concluded that the

changes do not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment requests involve no

significant hazards consideration.

Local Public Document Room

location: Maud Preston Palenske

Memorial Library, 500 Market Street, St.

Joseph, Michigan 49085.

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW,

Washington, DC 20037.

NRC Project Director: Cynthia A.

Carpenter, Acting.

Northeast Nuclear Energy Company

(NNECO), Docket No. 50–245, Millstone

Nuclear Power Station, Unit 1, New

London County, Connecticut

Date of amendment request: March

31, 1995.

Description of amendment request:

The proposed amendment would revise

the Technical Specifications (TS) to

increase the as-found setpoint tolerance

of the safety/relief valves (SRVs) from

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

In addition, the proposed amendment

(1) would allow the as-found condition

of one SRV to be inoperable, (2) clarifies

the 1325 psig safety limit wording, (3)

increases the number of SRVs to be

tested during each refueling outage, (4)

makes editorial changes to reflect the TS

changes, and (5) revises the bases for the

applicable sections.

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:

NNECO has reviewed the proposed

changes in accordance with 10 CFR 50.92

and concluded that the changes do not

involve a significant hazards consideration

(SHC). The basis for this conclusion is that

the three criteria of 10 CFR 50.92(c) are not

compromised. The proposed changes do not

involve an SHC because the changes would

not:

1. Involve a significant increase in the

probability or consequences of an accident

previously analyzed.

The safety function of the SRVs is to

mitigate the effects of a RPV [reactor pressure

vessel] overpressurization, therefore a failure

to open until the upper setpoint limit (+3%)

is reached cannot affect the probability of an

accident. The lowest allowable limit (¥3%)

is still above normal operating pressure and

therefore does not significantly increase the

probability of an inadvertent opening.

Should the SRVs open in response to an

RCS [reactor coolant system] overpressure

event, opening of the SRVs below the

nominal setpoints does not adversely affect

the consequences of an accident. The fuel

reload analysis demonstrates that actuation

of five valves at or below 103% of nominal

provides sufficient pressure reduction to

maintain peak RCS pressure below the safety

limit of 1375 psig and to maintain vessel

steam space pressure below 1325 psig. The

hydrodynamic loads on the SRV discharge

pipe (i.e., tail pipe) and the torus remain

within the design limits.

The performance of the high pressure

systems; FWCI [feedwater coolant injection],

SLC [standby liquid control] and IC [isolation

condenser] remain acceptable. There is also

no adverse impact on the operability of the

APR [automatic pressure relief] system.

The SRV setpoints will continue to be

required to be within [plus or minus] 1%

prior to plant startup from a refueling outage.

This ensures that the SRVs are restored to the

optimal conditions at the start of each fuel

cycle.

Therefore, increasing the ‘‘as-found’’

tolerance from [plus or minus] 1% to [plus

or minus] 3% does not result in a significant

increase in the probability or consequences

of a previously analyzed accident.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

Revising the acceptable as-found setpoint

tolerance from [plus or minus] 1% to [plus

or minus] 3% does not change the type of

action that these valves are expected to

perform, nor does it change the initial ‘‘asleft’’ requirements for the valves. Plant

operating parameters have not changed.

Therefore, this change cannot create the

possibility of a new or different kind of

accident.

3. Involve a significant reduction in the

margin of safety.

The margin of safety established and stated

in the Millstone Unit No. 1 Technical

Specifications, is a peak RCS pressure of

1375 psig and a peak vessel steam space

pressure of 1325 psig. While allowing the

SRV setpoint tolerance to increase to [plus or

minus] 3% would allow peak pressures from

an MSIV [main steam isolation valve] closure

event to approach that safety limit, the safety

limit will not be exceeded. Therefore, this

change does not involve a significant

reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Learning Resource Center,

Three Rivers Community-Technical

College, Thames Valley Campus, 574

New London Turnpike, Norwich, CT

06360.
Federal Register / Vol. 60, No. 80 / Wednesday, April 26, 1995 / Notices 20521

Attorney for licensee: Ms. L. M.

Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company,

Post Office Box 270, Hartford, CT

06141–0270.

NRC Project Director: Phillip F.

McKee.

PECO Energy Company, Public Service

Electric and Gas Company, Delmarva

Power and Light Company, and Atlantic

City Electric Company, Dockets Nos. 50–

277 and 50–278, Peach Bottom Atomic

Power Station, Units Nos. 2 and 3, York

County, Pennsylvania

Date of application for amendments:

February 10, 1995.

Description of amendment request:

The proposed changes provide for the

correction of administrative errors made

in the past during the processing of

technical specification changes related

to control room ventilation filter

surveillance testing.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated because the changes are purely

administrative and do not involve any

physical changes to plant SSC [systems,

structures, or components]. Therefore, these

changes will not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

2. The proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated because the changes will not alter

the plant or the manner in which the plant

is operated. The changes do not allow plant

operation in any mode that is not already

evaluated in the safety analysis. The changes

will not alter assumptions made in the safety

analysis and licensing bases. Therefore, these

changes will not create the possibility of a

new or different kind of accident from any

accident previously evaluated.

3. The proposed changes do not involve a

significant reduction in a margin of safety

because they are purely administrative and

have no impact on any safety analysis

assumptions.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Education

Building, Walnut Street and

Commonwealth Avenue, Box 1601,

Harrisburg, Pennsylvania 17105.

Attorney for licensee: J. W. Durham,

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

Counsel, PECO Energy Company, 2301

Market Street, Philadelphia,

Pennsylvania 19101.

NRC Project Director: John F. Stolz.

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: March

15, 1994.

Description of amendment request:

This amendment would reflect an

exemption from 10 CFR Part 50,

Appendix J, Section II.H.4, concerning

the scope of Type ‘C’ testing on

specified emergency core cooling

system and reactor core isolation

cooling containment isolation valves by

revising Technical Specification Table

3.6.3–1, Primary Containment Isolation

Valves. The subject valves on systems

which terminate below the minimum

water level of the suppression pool and

are associated with closed systems

would be tested using requirements of

the American Society of Mechanical

Engineers’ Section XI Code.

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:

I. This proposal does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed change to the scope of Type

‘C’ testing for the subject valves does not

affect the probability of the design basis

accidents. The valves will continue to be

maintained in an operable state, and in their

current design configuration. There is no

correlation between the scope of the Type ‘C’

testing and accident probability.

PP&L reviewed the postulated

consequences of design basis events on

primary containment isolation under the

proposed change. GDC 50 design

conformance states that the primary

containment structure, including access

openings, penetrations and the containment

heat removal system, is designed so that the

containment structure and its internal

compartments can withstand, without

exceeding the design leakage rate (1.0% per

day), the peak accident pressure and

temperature that could occur during any

postulated LOCA.

For the purposes of considering the

consequences of LOCAs under the proposed

change, a single active failure of a CIV or a

passive failure of the closed system were

reviewed, within the limits of the existing

licensing basis. Under the existing licensing

basis, a pipe rupture of seismically qualified

ECCS piping does not have to be assumed

concurrent with the LOCA, except if it is a

consequence of the LOCA. Consequential

failures can be eliminated, since a LOCA

inside containment is separated from the

ECCS piping by the containment structure.

Consequential failures of the ECCS piping

from LOCA’s outside containment are

outside the Appendix J design

considerations, although they are adequately

addressed through the redundancy and

separation of the ECCS design. A single

active failure of the CIV, under the LOCA

condition, can be accommodated since the

closed and filled system piping remains as

the leakage barrier. The ECCS passive failure

criterion does require consideration of

system leaks, but not pipe breaks, beyond the

initiating LOCA. Pipe leakage, equivalent to

the leakage from a valve or pump seal failure,

should be considered at 24 hours or greater

post-LOCA. The capability to make-up

inventory to the suppression pool is adequate

to ensure that postulated seat leakage and

pipe leakage does not result in a condition

that jeopardizes pool level. Make-up

capability exists to the suppression pool via

the Condensate Storage Tank and Spray

Pond. Actions to make-up to the suppression

pool are delineated in Emergency Operating

Procedures.

Therefore, the proposal to eliminate the

subject Type ‘C’ tests does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

II. This proposal does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The acceptability of the proposed change

to the scope of Type ‘C’ testing for the subject

valves is based on maintaining the existing

barriers to primary containment leakage, and

ensuring that the suppression pool level is

assured for 30 days during all design basis,

post-accident modes of operation. By meeting

these dual objectives, the plant response to

the design basis events will be unchanged,

and no new accident scenarios will be

encountered. These two objectives are

related, in that, the suppression pool

inventory creates a passive barrier to primary

containment atmospheric leakage for

penetrations which are located below the

minimum water level of the pool. The subject

valve lines terminate below the minimum

suppression pool water level.

The subject valves are all single isolation

valves associated with lines that penetrate

the primary containment, but are not

connected directly to the primary

containment atmosphere or the reactor

coolant pressure boundary. The redundant

isolation boundary for each of the affected

valves is the closed system associated with

the valve. This configuration is described in

General Design Criteria (GDC) 57. The

proposed exemption, and Technical

Specification change, does not alter the

configuration of these systems. The valves

will continue to be tested and maintained to

ensure their operability. The closed system
20522 Federal Register / Vol. 60, No. 80 / Wednesday, April 26, 1995 / Notices

piping meets PP&L’s design conformance to

GDC 56 and is verified via a 10CFR50

Appendix J Type ‘A’ test. The integrity of the

closed systems is also monitored and

controlled via Technical Specification

6.8.4.a.

The subject valves may be open, or change

state, postaccident to support the design

function of their associated ECCS systems

(HPCI, Core Spray, RHR) or RCIC. The subject

valves function as system valves during the

periods when they are open or in an

intermediate state, not as containment

isolation valves. Reliance is placed on the

suppression pool seal and the closed system

piping to maintain the barrier between

primary and secondary containment

atmospheres.

Therefore, with the valve and closed

system configuration unaffected by the

proposed change, the existing barriers to

primary containment atmospheric leakage are

maintained, so long as the suppression pool

level is ensured.

The suppression pool is designed and

operated so that it is filled with water in

accordance with Technical Specifications 3/

4.5.3, ‘‘Suppression Chamber,’’ 3/4.6.2,

‘‘Depressurization Systems—Suppression

Chamber,’’ and the associated Bases. The

supply of water in the suppression pool is

assured for 30 days during all design basis,

post-accident modes of operation. Type ‘C’

leak rate testing has historically been

performed on valves associated with lines

that connect to the suppression pool. The

acceptance criteria for combined leakage

from these penetrations is 3.3 gpm. This

leakage rate is at a level which ensures the

30 day post-accident suppression pool level.

However, for the valves discussed in this

change, seat leakage past the CIV is into a

closed and filled system. Thus ‘‘leakage’’

from the suppression pool, past the CIV, is

a function of closed system leakage.

As mentioned above, the integrity of the

closed system piping is verified via a

10CFR50 Appendix J Type ‘A’ test and is

monitored and controlled via Technical

Specification 6.8.4.a. TS 6.8.4.a establishes a

program to monitor and control leakage from

systems located outside containment that

could contain highly radioactive fluids

during a serious transient or accident. This

program applies to the ECCS systems and

RCIC affected by the proposed change and

ensures that leakage into secondary

containment via packing, flanges, seals, etc.,

is controlled. Leakage from these systems,

plus the Scram Discharge Volume, Reactor

Water Clean-up, and PASS, has been found

to be very low, and well below the 5 gpm

limit established for these systems. Current

leakage for Unit 1 is 0.14 gpm and for Unit

2, 0.043 gpm. The proposed change is not

expected to contribute to higher levels of

system leakage. Any leakage from these

systems is processed via Standby Gas

Treatment and the radwaste system to

maintain ALARA and comply with

regulatory guidance. The closed systems are

maintained filled, so that a supply of water

exists on both sides of the isolation valves.

While suppression pool leakage is a

function of closed system leakage for the

subject penetrations, a review of Type ‘C’ test

data for the subject CIVs showed that the

valves have had low leakage rates during

previous tests. This leakage is on the order

of 0.6 gpm, per unit. Proposed testing of the

valves under Section XI and the current

requirements of the Generic Letter 8910

program will ensure valve operability.

Therefore, leakage past the CIV and out of

the closed system is expected to be low and

in keeping with the design basis for the

suppression pool. However, the capability

does exist to make-up water to the

suppression pool from the Condensate

Storage Tank or Spray Pond if necessary.

Existing Emergency Operating Procedures

require actions if suppression pool level is

less than 22 feet or greater than 24 feet. Thus,

the level of the suppression pool is ensured,

independent of the current CIV Type ‘C’

testing requirement.

The proposed change to the scope of Type

‘C’ testing for the subject valves maintains

the existing barriers to primary containment

leakage, and ensures that the suppression

pool level is assured for 30 days during all

design basis, post-accident modes of

operation. Therefore, the plant response to

the design basis events is unchanged, and the

proposal does not create the possibility of a

new or different kind of accident from any

accident previously evaluated.

III. This change does not involve a

significant reduction in a margin of safety.

As discussed in questions I and II, the

proposed change does not alter the plant

response to existing accident scenarios, and

does not introduce new or different

scenarios. So the margin of safety from a

design basis accident standpoint is

maintained.

Historically, the leakage rate through the

subject valves has been determined under the

Type ‘C’ testing program. This leakage rate

has been found to be very low, and is

currently on the order of 0.6 gpm.

Quantifying leakage past the CIVs has been

used to ensure that the suppression pool

level is assured for 30 days post-accident.

Under the proposed change, this leakage rate

will not be quantified. This is acceptable

since leakage from the suppression pool is in

reality a function of closed system leakage,

not solely CIV leakage. Closed system leakage

is monitored and controlled by an existing

Technical Specification program. Closed

system leakage has been found to be very low

on both units, and is currently a small

fraction of a gallon per minute compared

with a 5 gpm allowable. Therefore, leakage

past the CIV and out of the closed system is

expected to be low and in keeping with the

design basis for the suppression pool.

However, the capability does exist, and is

proceduralized, to make-up water to the

suppression pool from the Condensate

Storage Tank or Spray Pond if necessary.

Thus the current capability to maintain

adequate suppression pool level for 30 days

postaccident is assured under the proposed

change.

Therefore the proposed change to the scope

of Type ‘C’ testing for the subject 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,

Pennsylvania 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 Nos. 50–387 and 50–

388 Susquehanna Steam Electric

Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of amendment request: March

31, 1995.

Description of amendment request:

These amendments would modify the

surveillance requirement for reactor

coolant system pH analysis in section

4.4.4 of the Technical Specifications

(TS) for each unit. Also, they would

clarify in the TS that the pH analysis

would be taken at least every 72 hours

whenever reactor coolant conductivity

exceeds 1.0 μmho/cm.

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:

I. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The pH limits on reactor coolant are not

affected by this change. The pH will be

measured whenever it is theoretically

possible for it to be outside the Tech Spec

[Technical Specifications] limits of <5.6 or

>8.6 (i.e., whenever the conductivity is

greater than 1.0 μmho/cm). Because of the

theoretical relationship between pH and

conductivity as shown in Attachment A [see

application dated March 31, 1995, for this

reference], it is possible to establish pH limits

on the reactor coolant by limiting the

conductivity. As shown in this figure, the pH

must be >5.6 and <8.6 if the conductivity is

less than or equal to 1.0 μmho/cm.

Attachment A was taken from Regulatory

Guide 1.56 Revision 1, July 1978

‘‘Maintenance of Water Purity in Boiling

Water Reactors’’. As noted in both FSAR final

safety analysis report and Technical

Specification Bases, the pH and conductivity

limits for OPERATIONAL CONDITION 1 are

consistent with this theoretical relationship.

The Bases for Section 3/4.4.4 of the Tech

Specs [Technical Specifications] contains

[contain] the following statement: ‘‘When the

conductivity is within limits, the pH,
Federal Register / Vol. 60, No. 80 / Wednesday, April 26, 1995 / Notices 20523

chlorides and other impurities affecting

conductivity must also be within their

acceptable limits[’’].

Since the conductivity is measured by grab

sampling at least every 72 hours to verify that

it is within limits, this will also verify that

pH is within limits every 72 hours. If the

conductivity should exceed 1.0 μmho/cm, pH

measurements will be made to determine if

the Tech Spec [Technical Specifications] pH

limits have been exceeded. It should also be

noted that inline conductivity

instrumentation is very stable and reliable

and is used to continuously monitor the

reactor coolant per Tech Spec [Technical

Specifications] requirements, with

instrumentation connected to redundant

sources (reactor water cleanup influent and

reactor recirculation loop). Therefore, the

proposed change will not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

II. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

As stated above, the pH limits on reactor

coolant are not affected by this change. Since

the conductivity is monitored continuously,

to verify that it is within limits, this will also

verify that pH is within limits. If the

conductivity should exceed 1.0 μmho/cm, pH

measurements will be made to determine if

the Tech Spec [Technical Specifications] pH

limits have been exceeded. Therefore, the

incorporation of this change will not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

III. Involve a significant reduction in a

margin of safety.

The in-line conductivity instrumentation

has been determined to be very stable and

reliable in its use to continuously monitor

the reactor coolant per Tech Spec [Technical

Specifications] requirements. To maintain

this reliability, this instrumentation is

connected to redundant sources (reactor

water cleanup influent and reactor

recirculation loop). Based on this continuous

monitoring of reactor coolant conductivity, as

provided by this instrumentation, the

incorporation of this change will have no

impact on current safety margins, nor will it

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: Osterhout Free Library,

Reference Department, 71 South

Franklin Street, Wilkes-Barre,

Pennsylvania 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 Nos. 50–387 and 50–

388, Susquehanna Steam Electric

Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of amendment request: March

31, 1995.

Description of amendment request:

These amendments would delete from

the Technical Specifications of each

unit, the operational condition

restriction in Surveillance Requirement

4.8.1.1.2.d.7 which requires that 24-

hour emergency diesel generator testing

be performed with at least one unit in

operational condition 4 or 5 (cold

shutdown or refueling).

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 changes do not:

I. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change to permit the 24 hour

testing of the emergency diesel generators

during power operation does not increase the

chances for a previously analyzed accident to

occur. The function of the EDGs [emergency

diesel generators] is to supply emergency

power in the event of a loss of offsite power.

As stated above [,] the diesel generator being

tested has been determined to remain

operable and available to supply the

emergency loads within the required times.

In addition, the three remaining EDGs will be

operable during this test. Operations

[Operation] of an EDGs [EDG] is not a

precursor to any accident. If, however, an

offsite disturbance were to occur that affected

the operability of the DG [emergency diesel

generator] being tested, the remaining EDGs

are capable of feeding the loads necessary for

safe shutdown of the plant. In summary, the

proposed change does not adversely affect

the performance or the ability of the diesel

generators to perform their intended safety

function. Therefore, the proposed change

will not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

II. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed change to the 24 hour

surveillance requirement will not affect the

operation of any safety system or alter its

response to any previously evaluated

accident. The diesel generator will

automatically transfer from test mode if

necessary to supply emergency loads in the

required time. The test mode is used for the

monthly surveillances of these diesel

generators, resulting in no new plant

operating modes being introduced. In the

event the EDG fails the functional test[,] it

will be declared inoperable and the actions

required for an inoperable diesel will be

performed. The remaining three EDGs will be

operable and are capable of feeding the loads

necessary for safe shutdown of the plant.

Therefore, the incorporation of this change

will not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

III. Involve a significant reduction in a

margin of safety.

Changing the EDG test timing results is no

reduction in the safety margin as defined in

the design basis. Because loss of an EDG is

not expected as a result of LOOP [loss of

offsite power] or LOCA/LOOP [loss-ofcoolant accident with a loss of offsite power]

during the 24 hour test, SSES [Susquehanna

Steam Electric Station] remains within its

design basis. In fact, because the test EDG

loads the ESS [engineered safety system] bus

8.5 seconds earlier than the non-test EDGs

during LOCA with LOOP, plant response is

actually improved. Risk of operation during

the 24 hour EDG test is certainly less than

during the current 84 hour allowed outage

time (AOT) because both the impact of the

initiating events evaluated (EDG in test is not

actually failed) and the frequency of the

limiting plant condition (loss of two EDGs)

are less. No increase in frequency or impact

of design basis events, and no reduction in

the safety margin occurs during the 24 hour

EDG test. Therefore, the incorporation of this

change will have no impact on current safety

margins, nor will it involve a significant

reduction in the margin to [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,

Pennsylvania 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: March

31, 1995.

Description of amendment request:

This amendment would change

Susquehanna Unit 2 Technical

Specifications (TS) by incorporating the

General Electric (GE) NRC approved

methodology for GE–12 type lead use

fuel assemblies (NEDE–24011–P–A–10)

into the list of references in Section

6.9.3.2. The licensee plans to insert four

of these fuel assemblies into the Unit 2

core during the fall of 1995. The

addition of the reference to the TS

would allow the use of the GE

methodology to document that all
20524 Federal Register / Vol. 60, No. 80 / Wednesday, April 26, 1995 / Notices

applicable requirements of the safety

analysis are met by the assemblies.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The proposed change does not:

I. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Incorporation of this proposed change of

adding reference NEDE–24011–P–A–10,

‘‘General Electric Standard Application for

Reactor Fuel’’ to the list of references in [the]

Unit 2 Technical Specifications will allow

the use of the GE methodology to calculate

the operating limits for the four GE Lead Use

Assemblies which are of a different

mechanical design than the Siemens 9X9 fuel

[currently installed in the reactor core]. This

NRC approved methodology will be

referenced as the approved methodology in

showing that all applicable safety limits of

the safety analysis are met by the four GE–

12 LUAs. Results of incorporating this

change will not significantly increase the

probability or the consequences of an

accident previously evaluated.

II. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

As stated above, the incorporation of this

change will allow the use of the GE

methodology to be referenced as the

approved methodology to show that all

applicable limits of the safety analysis are

met by the four GE–12 LUAs. Therefore, the

incorporation of this change will not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

III. Involve a significant reduction in a

margin of safety.

The use of the GE methodology will not

result in a change in safety margin, but will

ensure that the safety margin is maintained

with the insertion of the four GE LUAs of the

GE–12 type in Unit 2 Cycle 8. Therefore, the

incorporation of these changes will have no

impact on current safety margins, nor will

they involve a significant reduction in the

margin to [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,

Pennsylvania 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.

Philadelphia Electric Company, Docket

Nos. 50–352 and 50–353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of amendment request:

December 2, 1994.

Description of amendment request:

The proposed change to Limerick

Generating Station, Units 1 and 2,

Technical Specifications (TS) relocates

the TS Fire Protection Requirements to

Licensee controlled documents

consistent with NRC Generic Letter (GL)

86–10 ‘‘Implementation of Fire

Protection Requirements,’’ and GL 88–

12, ‘‘Removal of Fire Protection

Requirements from 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 Technical Specifications

changes do not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

The proposed changes are administrative

in nature and are consistent with NRC GL

86–10 and GL 88–12. Removal of Fire

Protection Program (FPP) requirements does

not affect any fire protection equipment nor

plant equipment important to safety, or

involve any physical modifications to plant

structures, systems or components, and

therefore is not associated with an accident

initiator or accident mitigator and can not

affect the probability of occurrence of an

accident or increase the consequences of an

accident. The licensee controlled Technical

Requirements Manual (TRM) containing the

relocated requirements will be maintained in

accordance with TS Section 6.0.

‘‘Administrative Controls’’ and subject to

review in accordance with 10 CFR 50.59.

Since future changes to the FPP (i.e.,

Updated Final Safety Analysis Report and

the TRM) will be evaluated per 10 CFR 50.59,

no increase (significant or insignificant) in

the probability or consequences of an

accident previously evaluated will be

allowed. Therefore, these changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed TS changes do not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The changes will not alter the plant

configuration (no new or different type of

equipment will be installed) or create

changes in methods governing normal plant

operation that will introduce new failure

modes. These changes will not impose

different requirements and proper control of

information will be maintained. These

changes will not alter assumptions made in

the safety analysis and licensing basis.

Therefore, these changes do not create the

possibility of a new or different kind of

accident previously evaluated.

3. The proposed TS changes do not involve

a significant reduction in a margin of safety.

The proposed changes are administrative

in nature and are consistent with NRC GL

86–10 and GL 88–12. The changes will not

reduce the margin of safety since they have

no impact on any safety analysis assumptions

or sequence of events used in any accident

analysis. In addition, any future changes to

the FPP will be evaluated per the

requirements of 10 CFR 50.59. Therefore, the

changes do not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania

19464.

Attorney for licensee: J. W. Durham,

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

Counsel, Philadelphia Electric

Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101.

NRC Project Director: John F. Stolz.

Philadelphia Electric Company, Docket

Nos. 50–352 and 50–353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of amendment request: January

27, 1995.

Description of amendment request:

The proposed change to Limerick

Generating Station (LGS) Units 1 and 2

Technical Specifications (TS) will

eliminate the TS active safety function

designation of eight (i.e., four per unit)

Drywell Chilled Water System (DCWS)

valves.

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 Technical Specifications

changes do not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

The proposed changes will eliminate the

TS active safety function designation of eight

(i.e., four per unit) DCWS valves. The DCWS

motor operated valves (MOVs) are designated

in TS as Primary Containment Isolation

Valves (PCIVs), during operational

conditions (OPCONS) 1, 2, and 3, which

mitigate the consequences of design basis

accidents. The proposed changes will

prohibit the subject DCWS valves from

opening during OPCONs 1, 2, and 3, thereby,
Federal Register / Vol. 60, No. 80 / Wednesday, April 26, 1995 / Notices 20525

eliminating the active safety function, and

maintaining a passive safety function. The

postulated accidents which require the

Primary Containment to act as a barrier in

order to mitigate the release of radioactivity

described in the LGS Updated Final Safety

Analysis Review [Report] (UFSAR) Section

15, are not affected by these changes.

Therefore, the previously evaluated

postulated on-site and off-site radiological

effects of these accidents will not change.

The DCWS valves will be prohibited from

opening during OPCONs 1, 2, and 3 by

physical changes made to the electrical

control circuitry and administrative controls.

Therefore, the probability of the valves to fail

in the open position will diminish, and the

required Primary Containment isolation

safety function will be maintained.

Therefore, these proposed changes will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed TS changes do not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes remove the affected

automatic isolation relays from the DCWS

MOVs’ circuitry. These changes eliminate

any postulated relay failure effects on the

associated control circuits and electrical

power supplies. The proposed changes do

not introduce any new accident initiators or

any new valve failure modes not previously

evaluated.

Therefore, these changes will not create the

possibility of a new or different kind of

accident from any accidents previously

evaluated.

3. The proposed TS changes do not involve

a significant reduction in a margin of safety.

The proposed changes will prohibit the

opening of the DCWS valves which provide

backup cooling from RECW [reactor

enclosure cooling water] during OPCONs 1,

2, and 3. The RECW System is not the normal

DCWS cooling alignment, is not required as

a backup safety related drywell cooling

system, and this backup alignment is not an

automatic function. The proposed changes do

not affect the function or operation of DCWS,

and since the proposed changes and

administrative controls ensure the valves will

remain closed during OPCONs 1, 2, and 3,

the capability for Primary Containment

isolation is not affected. Therefore, the

changes 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: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania

19464.

Attorney for licensee: J. W. Durham,

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

Counsel, Philadelphia Electric

Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101.

NRC Project Director: John F. Stolz.

Philadelphia Electric Company, Docket

Nos. 50–352 and 50–353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of amendment request: February

22, 1995.

Description of amendment request:

The proposed change to Limerick

Generating Station (LGS) Units 1 and 2

Technical Specifications (TS) revises

various TS Surveillance Requirements

to clarify the Emergency Diesel

Generator acceptable steady state

voltage range.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed Technical Specifications

changes do not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

The proposed Emergency Diesel Generator

steady state voltage range of 4280 [plus or

minus] 120 volts provides voltages through

the 4160V and 480V distribution systems

which are within the operating range of the

connected equipment and power system

components. Therefore, the reduced steady

state voltage range will not cause the

malfunction of any equipment or affect the

operation of any equipment in a manner

which would increase the probability of

occurrence of an accident previously

evaluated in the [Safety Analysis Report]

SAR.

Reducing the Emergency Diesel Generator

steady state voltage range in the Technical

Specifications maintains the capability of the

Emergency Diesel Generator to start and

attain rated voltage and frequency within 10

seconds and to accept the engineered

safeguard loads in the required time in order

to mitigate the consequences of an accident.

The Emergency Diesel Generator automatic

voltage regulator setting is calibrated to

within a range of 4266.5 volts to 4308.5 volts.

A review of results from recent monthly

Emergency Diesel Generator Surveillance

Tests has confirmed that the voltage

regulators currently maintain the Emergency

Diesel Generator steady state voltage within

the 4280 [plus or minus] 120 volt range to be

included in the Technical Specifications.

Establishing, via Technical Specification

surveillance requirements and administrative

limits within Station Surveillance Test

Procedures, that the Emergency Diesel

Generator voltage regulator is maintaining

the steady state voltage within a narrower

range (within the existing range) provides

increased assurance that connected

equipment required to mitigate the

consequences of an accident will operate as

required.

Therefore, the proposed TS changes do not

involve an increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed TS changes do not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

Reducing the Emergency Diesel Generator

steady state voltage range in the Technical

Specifications to a range of 4280 [plus or

minus] 120 volts does not create any new

accident initiators or affect any existing

accident initiators such that a different type

of accident than previously evaluated could

result. The function and operation of the

Emergency Diesel Generators and their

connected loads are not changed in a manner

which would create the possibility of an

accident of a different type than any

previously evaluated.

Therefore, the proposed TS changes do not

create the possibility of a new or different

kind of accident from any previously

evaluated.

3. The proposed TS changes do not involve

a significant reduction in a margin of safety.

Reducing the Emergency Diesel Generator

steady state voltage range in the Technical

Specifications to a range of 4280 [plus or

minus] 120 volts does not reduce the margin

of safety. The reduced range provides

increased assurance that the equipment

powered by the Emergency Diesel Generators

will start and operate as designed in order to

perform their design basis functions.

Therefore, the proposed TS changes do not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania

19464.

Attorney for licensee: J. W. Durham,

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

Counsel, Philadelphia Electric

Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101.

NRC Project Director: John F. Stolz.

Philadelphia Electric Company, Docket

Nos. 50–352 and 50–353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of amendment request: March 1,

1995.

Description of amendment request:

The proposed changes will clarify the

concentrations of calibration gas

required to calibrate the Hydrogen and

Oxygen Analyzers, and support the

requirements of Limerick Generating

Station (LGS) Transient Response

Implementation Plant (TRIP) T–102,

‘‘Primary Containment Control Bases.’’

Basis for proposed no significant

hazards consideration determination:
20526 Federal Register / Vol. 60, No. 80 / Wednesday, April 26, 1995 / Notices

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed TS changes do not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed TS [Technical Specification]

changes remove calibration of the H2/O2

Analyzers using zero volume percent

hydrogen (H2) and 100% bottled Nitrogen

(N2). A calibration gas containing zero

volume percent H2 and 100% bottled N2 is

not required for calibration of the analyzers

to the required accuracy. Calibration of the

H2/O2 Analyzers is done in accordance with

the manufacturer’s instructions. The

proposed TS changes also revise the span gas

concentration from 5% to 7% to support the

requirements of TRIP T–120. The H2/O2

Analyzers provide indication of the

concentrations of combustible gases in the

primary containment and provide

annunciation when combustible gas

concentrations reach unacceptable levels.

Failure of the analyzers is not an accident

initiator. The analyzers do not connect to the

reactor coolant pressure boundary; therefore,

they do not increase the probability of a

LOCA [loss-of-coolant accident]. The

proposed TS changes do not involve any

design changes to analyzers. Therefore, these

TS changes will not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

2. The proposed TS changes do not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The H2/O2 Analyzers provide indication

and alarms for H2 and O2 concentrations in

containment. No physical or design changes

to the analyzers are being made by these TS

changes. During normal operations, the

potential for an explosive atmosphere is

negligible due to the absence of H2 sources.

For Post-LOCA, conditions the levels of H2

and O2 in containment have already been

evaluated in LGS UFSAR [Updated Final

Safety Analysis Report] Section 6.2.5. No

physical or design changes which could

introduce a new analyzer failure mode are

being made. The failure modes of the

analyzers are evaluated in UFSAR Table 6.2–

21. Therefore, these TS changes will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed TS changes do not involve

a significant reduction in a margin of safety.

These TS changes will clarify statements in

the LGS UFSAR and TS concerning

calibrated ranges of the analyzers. The

change of the span gas from 5% to 7% falls

within conditions previously analyzed. The

Bases for TS 3/4.3.7.5 and 3/4.6.6 require

operable H2/O2 Analyzers to ensure the

analyzers will be available for monitoring,

assessing and controlling H2 and O2 in

containment following a LOCA. These TS

changes do not adversely affect operability of

the analyzers or their availability for use

during Post-LOCA conditions; therefore, the

margin of safety is not reduced.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania

19464.

Attorney for licensee: J. W. Durham,

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

Counsel, Philadelphia Electric

Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101

NRC Project Director: John F. Stolz.

Public Service Electric & Gas Company,

Docket No. 50–354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of amendment request:

December 15, 1994.

Description of amendment request: In

accordance with 10CFR50.90, PSE&G

proposes to remove Technical

Specification Surveillance Requirement

4.8.1.1.2.h.1, and utilize plantcontrolled programs to govern diesel

generator maintenance. To ensure

procedural consistency and reduce the

impact of this change on Hope Creek

procedures, the remaining Surveillance

Requirements of Technical Specification

4.8.1.1.2.h are not renumbered.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Will not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

The proposed change is consistent with the

improved Standard Technical Specifications

(NUREG–1433) and does not result in any

changes to the existing plant design. The

Hope Creek preventative maintenance

program will utilize diesel generator

performance history, engineering analyses

and manufacturer’s recommendations as

appropriate for determining diesel generator

inspection requirements. Since the changes

do not impact the ability of the diesel

generators and the AC electrical power

sources to perform their function, the

changes do not result in a significant increase

in the consequences of any accident

previously evaluated. The diesel generators

will continue to function as designed.

Therefore, the proposed change will not

impact the probability of occurrence of any

accident previously evaluated.

2. Will not create the possibility of a new

or different kind of accident from any

accident previously evaluated.

This request does not result in any change

to the plant design nor does it involve a

significant change in current plant operation.

The diesel generators will be inspected

utilizing diesel generator operating history,

engineering analyses and manufacturer’s

recommendations as appropriate, and the

remaining surveillance requirements will not

be changed. As a result, no new failure

modes will be introduced, and the proposed

changes will not create the possibility of a

new or different kind of accident from any

accident previously evaluated.

3. Will not involve a significant reduction

in a margin of safety. The proposed request

does not adversely impact the reliability of

the diesel generators. As stated above, the

diesel generator operating history,

engineering analyses and the manufacturer’s

recommendations will be utilized as

appropriate to perform the diesel generator

inspections. In addition, the diesel generators

will continue to perform their design

functions. This request does not involve an

adverse impact on diesel generator operation

or reliability. Since the diesel generator

function is not affected by the proposed

changes, this request 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: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey

08070.

Attorney for licensee: M. J.

Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW.,

Washington, DC 20005–3502.

NRC Project Director: John F. Stolz.

Tennessee Valley Authority, Docket No.

50–327, Sequoyah Nuclear Plant, Unit 1,

Hamilton County, Tennessee

Date of amendment request: April 6,

1995 (TS 95–09).

Description of amendment request:

The proposed change would revise

Operating Condition 2.C.(25) to extend

the ice condenser Surveillance 4.6.5.1.d

to October 1, 1995, to coincide with the

Unit 1 Cycle 7 refueling outage.

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:

TVA has evaluated the proposed technical

specification (TS) change and has determined

that it does not represent a significant

hazards consideration based on criteria

established in 10 CFR 50.92(c). Operation of

Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:
Federal Register / Vol. 60, No. 80 / Wednesday, April 26, 1995 / Notices 20527

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change is temporary and

allows a one-time extension of the ice

condenser Surveillance Requirement

4.6.5.1.d for Cycle 7 to allow surveillance

testing to coincide with the seventh refueling

outage. The proposed surveillance interval

extension will not cause a significant

reduction in system reliability nor affect the

ability of the system to perform the design

function. Current monitoring of plant

conditions and continuation of the

surveillance testing required during normal

plant operation will continue to be

performed to ensure conformance with TS

operability requirements. Therefore, this

change does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

Extending the surveillance interval for the

performance of ice condenser testing will not

create the possibility of a new or different

kind of accident. No change is required to

any system configuration, plant equipment,

or analyses.

3. Involve a significant reduction in a

margin of safety.

The safety limits assumed in the accident

analyses and the design function of the

equipment required to mitigate the

consequences of postulated accidents will

not be changed significantly. Existing

analysis indicates that the potential

reduction in ice weight resulting from the

proposed extension will continue to maintain

the maximum containment accident pressure

below 12 pounds per square inch gauge. The

ice condenser will continue to support

accident mitigation functions although some

Row 1 baskets could drop slightly below the

required 993-pound analysis limit. Therefore,

the plant will be maintained with acceptable

ice weights for accident mitigation and the

proposed extension will not significantly

reduce the margin of safety.

The NRC has reviewed the licensee’s

analysis and, based on this review, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402.

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H,

Knoxville, Tennessee 37902.

NRC Project Director: Frederick J.

Hebdon.

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of amendment request: April 6,

1995 (TS 95–11).

Description of amendment request:

The proposed change would relocate the

constant numerical value found in the

overtemperature delta temperature and

overpower delta temperature equations

of Technical Specifications Table 2.2–1

and place them in the Core Operating

Limits Report (COLR). This would be

accomplished by revising notes 1 and 2

in Table 2.2–1 to state that the values

are located in the COLR. The values of

the constants, however, would not be

changed. Also, the ‘‘Overtemperature

and Overpower Delta Temperature

Setpoint Parameter Values for

Specification 2.2.1’’ would be added to

the list of core operating limits specified

in Section 6.9.1.14 that are required to

be included in the COLR. In addition, a

reference to WCAP–8745–P–A, ‘‘Design

Bases for the Thermal Overpower deltaT and Thermal Overtemperature delta-T

Trip Functions,’’ would be added to

Section 6.9.1.14.a.

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:

TVA has evaluated the proposed technical

specification (TS) change and has determined

that it does not represent a significant

hazards consideration based on criteria

established in 10 CFR 50.92(c). Operation of

Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed changes will allow changes

to the constant numerical values for the

overtemperature delta temperature

(OT[delta-]T) and overpower delta

temperature (OP[delta-]T) equations in

accordance with the 10 CFR 50.59

requirements. This revision does not revise

these constants, but relocates them to the

core operating limits report (COLR) that is

governed by the 10 CFR 50.59 requirements.

The addition of the lag compensator

functions for measured [delta-]T and average

temperature in these equations does not alter

the setpoint because this lag function has a

value of unity. Therefore, the proposed

revision does not alter plant functions or

setpoints, but does allow for a more timely

revision process for parameters that may

require changes due to specific fuel cycle

requirements or updated plant analyses. The

use of the lag functions and revisions to the

constant numerical values will be maintained

within the safety analysis for the plant by the

10 CFR 50.59 process requirements. The

probability of an accident is not increased

because the plant functions are not altered by

the proposed revision and future changes

will be in accordance with 10 CFR 50.59.

Additionally, the consequences of an

accident are not increased because the

mitigation functions of the OT[delta-]T and

OP[delta-]T functions are not changed and

revisions to the equations that derive the

setpoints will be processed under the

requirements of the 10 CFR 50.59 program.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

The proposed revision will not change

plant functions and future revisions will

continue to be controlled in accordance with

the 10 CFR 50.59 requirements. The addition

of the lag functions does not create a new

accident potential because these functions

have already been considered in the analysis

as shown in NUREG 1431. Therefore, the

possibility of a new or different kind of

accident is not created by the proposed

revision.

3. Involve a significant reduction in a

margin of safety.

Plant parameters are not altered by the

proposed revision and the OT[delta-]T and

OP[delta-]T functions will not reflect a

change in setpoint generation or value. The

proposed change will allow revision of the

constant numerical values and use of the lag

compensator functions in accordance with

the 10 CFR 50.59 provisions to ensure the

design basis of the plant is maintained. This

revision does not result in changes that

reduce the margin of safety because the

OT[delta-]T and OP[delta-]T functions

remain unchanged and future revisions to

these functions will be performed in

accordance with 10 CFR 50.59.

The NRC has reviewed the licensee’s

analysis and, based on this review, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402.

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H,

Knoxville, Tennessee 37902.

NRC Project Director: Frederick J.

Hebdon.

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of amendment request: April 6,

1995 (TS 94–15).

Description of amendment request:

The proposed change would modify the

Technical Specifications associated

with the Post Accident Sampling (PAS)

system by deleting License Condition

2.C.(23)F for Unit 1 and 2.C.(16)g for

Unit 2 that require operation of the PAS

system in accordance with referenced

letters no later than startup from the

second refueling outage. The submittal

also includes a revised description of

operation of the PAS system for

insertion into the Updated Final Safety
20528 Federal Register / Vol. 60, No. 80 / Wednesday, April 26, 1995 / Notices

Analysis Report for staff approval. This

information supersedes the information

contained in the letters referenced in the

License Conditions listed above and

would be maintained in accordance

with the 10 CFR 50.59 process.

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:

TVA has evaluated the proposed technical

specification (TS) change and has determined

that it does not represent a significant

hazards consideration based on criteria

established in 10 CFR 50.92(c). Operation of

Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change involves the deletion

of license conditions that authorized TVA to

operate SQN’s postaccident sampling (PAS)

system. TVA proposed change establishes

programmatic control of SQN’s PAS system

under SQN TS 6.8.4.e and the SQN Final

Safety Analysis Report. Any future changes

to SQN’s PAS Program would be governed by

the 10 CFR 50.59 process. PAS and analysis

will continue at SQN through grab sample

acquisition and laboratory analysis and will

continue to meet the PAS objectives in

NUREG–0737, Item II.B.3 and Regulatory

Guide 1.97, Revisions 2. Accordingly, the

proposed change does not affect the

probability or consequences of an accident.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

The proposed change involves

improvements in the operational reliability of

SQN’s PAS system by using more reliable

laboratory analysis methods, reducing

sampling personnel radiation dose, and

incorporating practical methods for sample

acquisition and analysis. Because the

proposed change involves license conditions

and sampling methods that are utilized for

postaccident recovery, the potential for an

unanalyzed accident is not created.

Consequently, no new failure modes are

introduced.

3. Involve a significant reduction in a

margin of safety.

Plant safety margins are established

through limiting conditions of operation,

limiting safety system settings, and safety

limits specified in the TSs. As a result of the

proposed amendment, there will be no

change to either the physical design of the

plant or to any of these settings and limits.

The proposed changes do not affect the safe

operation of SQN. Therefore, there are no

changes to any of the margins of safety.

The NRC has reviewed the licensee’s

analysis and, based on this review, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402.

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H,

Knoxville, Tennessee 37902.

NRC Project Director: Frederick J.

Hebdon.

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of amendment request: April 6,

1995 (TS 94–18).

Description of amendment request:

The proposed change would revise

Surveillance Requirement (SR) 4.0.5 by

replacing the current Inservice

Inspection program requirements with

the requirements stated in the Standard

Technical Specifications (NUREG–

1431). As a result, SR 4.0.5 would more

clearly specify the inservice inspection

(ISI) program requirements and the

inservice testing (IST) program

requirements of the American Society of

Mechanical Engineers (ASME) Code

Class 1, 2, and 3 components. The

licensee also proposed deletion of

Technical Specification 3/4.4.10,

‘‘Structural Integrity ASME Code Class

1, 2 and 3 Components,’’ and its related

Bases information.

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:

TVA has evaluated the proposed technical

specification (TS) change and has determined

that it does not represent a significant

hazards consideration based on criteria

established in 10 CFR 50.92(c). Operation of

Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Operation of the facility in accordance

with the proposed amendment would not

involve any increase in the probability of

occurrence or consequences of an accident

previously evaluated. The Inservice

Inspection and Testing Programs, pursuant to

10 CFR 50.55a are described in the TSs. The

proposed amendment, in accordance with

NUREG–1431 and draft NUREG–1482

permits relief from an American Society of

Mechanical Engineers (ASME) code

requirement in the interim between the time

of submittal of a relief request and NRC

approval of the relief. The changes being

proposed do not affect assumptions

contained in plant safety analyses or change

the physical design and/or operation of the

plant, nor do they affect TSs that preserve

safety analysis assumptions. Any relief from

the approved ASME Section XI code

requirements that is implemented prior to

NRC review and approval will require

evaluation under the 10 CFR 50.59 process to

determine that no TS changes or unreviewed

safety questions exist. This evaluation

process will ensure that the impact of any

code relief is thoroughly evaluated and that

the structures, systems, and components

remain in conformance with assumptions

made in the safety analysis. The proposed

change to delete SQN TS 3/4.4.10, Structural

Integrity, does not affect plant safety analyses

or change the physical design or operation of

the plant. The proposed amendment

relocates the structural integrity requirements

under the existing TS Surveillance

Requirement (SR) 4.0.5 to allow these

requirements to be governed and controlled

within the inservice inspection (ISI) program.

Therefore, operation of the facility in

accordance with the proposed amendment

would not affect the probability or

consequences of an accident previously

evaluated.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

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 Inservice

Inspection and Testing Programs, pursuant to

10 CFR 50.55a are described in the TSs. The

proposed amendment, in accordance with

NUREG–1433 and draft NUREG–1482,

permits relief from an ASME code

requirement in the interim between the time

of submittal of a relief request and NRC

approval of the relief. The changes being

proposed will not change the physical plant

or the modes of operation defined in the

Facility License. The changes do not involve

the addition or modification of equipment

nor do they alter the design or operation of

plant systems. Any relief from the approved

ASME Section XI code requirements that is

implemented prior to NRC review and

approval will require evaluation under the 10

CFR 50.59 process to determine that no TS

changes or unreviewed safety questions exist.

This evaluation process will ensure that the

impact of the code relief is thoroughly

evaluated and that the structures, systems,

and components remain in conformance with

assumptions made in the safety analysis. The

proposed change to delete SQN TS 3/4.4.10

does not affect plant safety analyses or

change the physical design or operation of

the plant. The proposed amendment

relocates the structural integrity requirements

under the existing TS SR 4.0.5 to allow these

requirements to be governed and controlled

within the ISI program. 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. Involve a significant reduction in a

margin of safety.

Operation of the facility in accordance

with the proposed amendment would not

involve any reduction in a margin of safety.
Federal Register / Vol. 60, No. 80 / Wednesday, April 26, 1995 / Notices 20529

The Inservice Inspection and Testing

Programs, pursuant to 10 CFR 50.55a, are

required by the SQN TSs. The proposed

amendment, in accordance with NUREG–

1431 and draft NUREG–1482 permits relief

from an ASME code requirement in the

interim between the time of submittal of a

relief request and NRC approval of the relief.

Any relief from the ASME Section XI code

is required to be evaluated under the 10 CFR

50.59 process to determine that no TS

changes or unreviewed safety questions exist.

This evaluation process will ensure that code

relief does not affect the ability of structures,

systems, or components to perform their

design function, affect compliance with any

TS requirements or reduce the margin of

safety. The proposed change to delete SQN

TS 3/4.4.10 does not affect plant safety

analyses or change the physical design or

operation of the plant. The proposed

amendment relocates the structural integrity

requirements under the existing TS SR 4.0.5

to allow these requirements to be governed

and controlled within SQN’s ISI program.

Therefore, operation of the facility in

accordance with the proposed amendment

would not involve a reduction in the margin

of safety.

The NRC has reviewed the licensee’s

analysis and, based on this review, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402.

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H,

Knoxville, Tennessee 37902.

NRC Project Director: Frederick J.

Hebdon.

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of amendment request: April 6,

1995 (TS 95–06).

Description of amendment request:

The proposed change would delete the

technical specification requirement that

limits and controls loads traveling over

the spent fuel pool (Specification 3.9.7),

the graph that relates the Load Carried

Over the Shield to the Allowable Height

Above the Shield Surface (Figure 3.9–1),

the crane interlocks and physical stops

surveillance requirements

(Specifications 4.9.7.1 and 4.9.7.2), and

the related Bases information. These

requirements would be relocated to

administratively controlled procedures.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

TVA has evaluated the proposed technical

specification (TS) change and has determined

that it does not represent a significant

hazards consideration based on criteria

established in 10 CFR 50.92(c). Operation of

Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed TS change involves the

relocation of a requirement that does not

pertain to limitations or conditions of reactor

operation or to equipment to mitigate design

basis accidents or transients. SQN is

proposing to relocate this TS based on NRC’s

final policy statement on TS improvement

(58 FR 39132, dated July 22, 1993). Based on

this criteria, the spent fuel pit (SFP) crane

travel is not important to operational safety

and may be relocated to administratively

controlled procedures. By placing the crane

travel requirements in administratively

controlled procedures, adequate controls will

remain in place to prevent heavy loads from

traveling over fuel assemblies in the SFP. The

administratively controlled procedure that

controls the by-passing of the interlocks and

physical stops is subject to the requirements

of TS 6.5.1A. Therefore, the relocation of this

TS will not involve an increase in the

probability or consequences of an accident

previously evaluated.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

The proposed change involves relocating

TS requirements to another administratively

controlled document. No modifications to the

plant are involved. Additionally, there are no

changes to the operation of the plant or

equipment proposed. Based on this, the

relocation of this TS will not create the

possibility of a new or different kind of

accident from any previously analyzed.

3. Involve a significant reduction in a

margin of safety.

The proposed change involves the

relocation of TS requirements to

administratively controlled procedures. The

relocation of this requirement is based on the

criteria endorsed in the Commission’s Policy

Statement on TS improvements as it pertains

to 10 CFR 50.36. Additionally, this change

does not alter the basic design and safety

analysis requirements, as discussed in the

Updated Final Safety Analysis Report.

Therefore, the deletion of this TS will not

involve a reduction in the margin of safety.

The NRC has reviewed the licensee’s

analysis and, based on this review, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402.

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H,

Knoxville, Tennessee 37902.

NRC Project Director: Frederick J.

Hebdon.

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of amendment request: April 6,

1995 (TS 94–19).

Description of amendment request:

The proposed change would revise the

Action statement for Technical

Specification 3.8.1.1 by inserting a new

Action a, relabeling and modifying

existing Action a to become Action b,

adding a footnote referenced to Action

b, renumbering the subsequent action

statements, and adding information to

the Bases that amplifies the action

statements. The proposed new Action a

would no longer address required

actions for diesel generator testing. It

would require that, should one of the

AC electrical power sources listed be

inoperable, then operability of the

remaining offsite AC circuit be

demonstrated by performing

Surveillance Requirement 4.8.1.1.1.a

within 1 hour and at least once per 8

hours thereafter. If two offsite circuits

cannot be restored within 72 hours, the

unit must be placed in hot standby

within the next 6 hours and in cold

shutdown within the following 30

hours.

The proposed change to Action b

would address the testing requirements

should a diesel generator become

inoperable. It would require testing of

operable diesel generators if the

inoperability of the affected diesel

generator has the potential to be the

result of a common cause failure. A

footnote would clarify that the common

cause determination must be completed

regardless of how long the diesel

generator inoperability persists or

Surveillance 4.8.1.1.2.a.4 must be

completed to verify diesel generator

operability. The proposed change to the

Bases would provide clear guidance on

the use of common cause failure

determinations to eliminate unnecessary

diesel generator testing and would

define when testing is required to verify

diesel generator operability.

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:

TVA has evaluated the proposed technical

specification (TS) change and has determined
20530 Federal Register / Vol. 60, No. 80 / Wednesday, April 26, 1995 / Notices

that it does not represent a significant

hazards consideration based on criteria

established in 10 CFR 50.92(c). Operation of

Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed revisions do not alter the

plant features or operating practices. This

revision will reduce unnecessary starts of the

diesel generators (D/Gs) when a common

cause failure is not involved or for an

inoperable offsite circuit. This change will

not affect the accident mitigation capabilities

of the D/Gs, but should improve the

reliability by reducing the wear and tear

associated with starting the D/Gs. The D/Gs

are not the source of a postulated accident

and because this change does not alter plant

functions or operating practices the

probability of an accident is not increased.

The D/G’s operability will continue to be

verified for conditions that indicate a

potential common-cause failure to ensure

accident mitigation capabilities are not

affected. Therefore, this revision will

continue to provide actions that will support

alternating-current (ac) electrical power

source safety functions without unnecessary

degradation of the D/Gs and will not increase

the consequences of an accident.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

The D/Gs are not the source of accidents

and the proposed revision will not alter plant

functions or actions by more appropriately

limiting the conditions when a D/G must be

verified operable. Therefore, the possibility

of a new or different accident is not created

3. Involve a significant reduction in a

margin of safety.

This revision does not alter plant functions

that provide the margin of safety. The

reduction of D/G testing will only be allowed

for situations where the operable D/Gs are

not affected by the conditions resulting in the

ac power source inoperability. This reduced

testing should improve D/G reliability for

accident mitigation functions and further

ensure the margin of safety provided by the

D/Gs. Therefore, the margin of safety is not

reduced by the proposed revision to limit

unnecessary D/G starts.

The NRC has reviewed the licensee’s

analysis and, based on this review, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402.

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H,

Knoxville, Tennessee 37902.

NRC Project Director: Frederick J.

Hebdon.

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of amendment request: April 6,

1995 (TS 95–02).

Description of amendment request:

The proposed change would add

Limiting Condition for Operation 3.0.6

to allow equipment that has been

removed from service or declared

inoperable to be returned to service

under administrative control in order to

perform testing required to demonstrate

operability. It would be applicable for

operability testing of the inoperable

equipment or other equipment that

requires the operability feature to be in

service in order to perform the test. A

related change to the Bases would

provide amplifying explanation on the

use of this new provision. In addition,

a proposed change to Action 18 of Table

3.3–3, ‘‘Engineered Safety Feature

Actuation System Instrumentation,’’

would clarify the time interval that an

instrument channel may be in the

bypass condition. For those instruments

that reference Action 18, the change

would allow the bypass for 6 hours.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

TVA has evaluated the proposed technical

specification (TS) change and has determined

that it does not represent a significant

hazards consideration based on criteria

established in 10 CFR 50.92(c). Operation of

Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The addition of the exception to TSs 3.0.1

and 3.0.2 and the definition for the time to

place a channel in bypass will not change

plant equipment or the operating practices at

SQN. The exception will allow testing to be

performed with inoperable equipment

returned to service under administrative

controls, but will not change functions. The

function will be available from other

redundant channels during the brief

durations that the new provision would be

utilized. The specified time interval to

achieve a bypass condition will clarify the

implementation of the action requirement

with the affected functions remaining

available through the redundant operable

channels. This clarification does not change

the intent of the action but does set the

previously undefined time interval.

The proposed change affects actions

associated with the actuation of functions to

mitigate accidents and are not the source of

an accident. Therefore, the probability of an

accident is not increased. The affected

functions provide accident mitigation

functions and the proposed revisions serve to

ensure equipment can be maintained in

required conditions within acceptable time

intervals and administrative controls. The

brief periods utilized for the TSs 3.0.1 and

3.0.2 exception will not significantly affect

the accident mitigation capabilities because

of the availability of redundant equipment. In

addition, the benefit of performing

operability testing to return equipment

permanently to service or to maintain the

operability of other equipment outweighs the

slight reduction in safety function actuation

redundancy. Therefore, the proposed change

will not significantly increase the

consequences of an accident.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

The proposed changes affect functions

utilized to mitigate an accident and are not

the source of an accident. The exception

provides reasonable flexibility to maintain

equipment operability and the bypass time

interval reduces the potential for damage to

safety related equipment. Because plant

functions are not changed as a result of this

request the possibility of a new or different

kind of accident is not created.

3. Involve a significant reduction in a

margin of safety.

The proposed change does not alter

setpoints or operating considerations that

maintain the margin of safety for SQN. These

changes provide needed flexibility to perform

TS required testing and clarifications for

implementing action requirements. These

changes will slightly affect the redundancy of

the affected safety functions but provide

greater benefit for maintaining equipment in

an operable condition. Therefore, the margin

of safety provided by the affected equipment

has not changed and the proposed change

will not result in a reduction.

The NRC has reviewed the licensee’s

analysis and, based on this review, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402.

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H,

Knoxville, Tennessee 37902.

NRC Project Director: Frederick J.

Hebdon.

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of amendment request: April 6,

1995 (TS 94–04).

Description of amendment request:

The proposed amendment would

change the power range neutron flux

channel calibration frequency
Federal Register / Vol. 60, No. 80 / Wednesday, April 26, 1995 / Notices 20531

surveillance requirement from monthly

to every 31 effective full power days and

delay the requirement to perform the

surveillance for 96 hours after reaching

15 percent power. A proposed change to

the Bases would provide amplifying

information.

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:

TVA has evaluated the proposed technical

specification (TS) change and has determined

that it does not represent a significant

hazards consideration based on criteria

established in 10 CFR 50.92(c). Operation of

Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The likelihood that an accident will occur

is neither increased or decreased by this TS

change, which only affects when the first

surveillance is performed following an

outage and changes the frequency of

performance of the surveillance. Before startup following refueling outage, the power

range high trip setpoint is set below 85

percent power, typically 60 percent, for

conservatism. The power range low trip

setpoint is set at 22 percent power, TS

requires the setpoint to be less than or equal

to 25 percent power. These settings are in

addition to the conservatism built into startup following a refueling outage. Therefore,

delaying the first performance for 96 hours

will not impact on the operation of the plant

since the setpoints are set conservatively.

Also, the change of the frequency to every 31

effective full power days (EFPD) only delays

the surveillance when the plant is operated

at reduced power. During operation at

reduced power changes in the neutron flux

are also reduced. Therefore, changing the

frequency from monthly to every 31 EFPD

allows slow changes in neutron flux during

the fuel cycle to be more accurately detected

and evaluated.

This TS change will not impact the

function or method of operation of plant

equipment. Thus, there is not a significant

increase in the probability of a previously

analyzed accident due to this change. No

systems, equipment, or components are

affected by the proposed change. Thus, the

consequences of a malfunction of equipment

important to safety previously evaluated in

the Updated Final Safety Analysis Report are

not increased by this change.

The proposed changes provide TS

improvements that ensure the system

operates within the bounds of SQN’s

accident analysis as contained in the Final

Safety Analysis Report (FSAR) and only

affects when a surveillance is performed.

This change has no impact on accident

initiators and does not involve a physical

modification to the plant. Accordingly, the

proposed changes do not involve an increase

in the probability or consequences of an

accident previously evaluated.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

This revision will not change any plant

equipment, system configurations, or

accident assumptions. This change will more

accurately monitor changes in the condition

of the core.

Fuel burn-up is necessary to change the

relationship between the incore axial power

and the excore detectors response. At

reduced levels the effectiveness of the

monitoring activity is reduced. Therefore,

changing the frequency to 31 EFPD allows

slow changes in neutron flux during the fuel

cycle to be more accurately detected and

evaluated. Delaying the first performance of

the surveillance requirement, until 96 hours

after reaching 15 percent rated thermal

power, will allow the unit to be in a more

stable condition. Therefore, this change will

not affect the safety function of any

components and will create the possibility of

a new or different kind of accident.

3. Involve a significant reduction in a

margin of safety.

The proposed changes provide TS

improvements for SQN’s power range

monitoring system that ensure the system

operates within the bounds of SQN’s

accident analysis as contained in the FSAR

since only the time interval between

performances of the surveillance is being

extended. This change does not involve a

physical modification to SQN’s power range

monitoring system. Accordingly, the margin

of safety has not been reduced.

The NRC has reviewed the licensee’s

analysis and, based on this review, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402.

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H,

Knoxville, Tennessee 37902.

NRC Project Director: Frederick J.

Hebdon.

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of amendment request: April 6,

1995 (TS 95–08).

Description of amendment request:

The proposed change would (1) change

the core alteration definition to limit the

term to reactor vessel internal activities

that could have an affect on core

reactivity, (2) change the quadrant

power tilt ratio definition to eliminate

the conflict in the definition of the term

and its use in Surveillance Requirement

4.2.4.2, and (3) revise the Unit 1

Operational Modes parameters in Table

1.1 to be consistent with the description

in Table 1.1 for Unit 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:

TVA has evaluated the proposed technical

specification (TS) change and has determined

that it does not represent a significant

hazards consideration based on criteria

established in 10 CFR 50.92(c). Operation of

Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated. The proposed changes

provide TS improvements that ensure the

plant operates within the bounds of SQN’s

accident analysis as contained in the Final

Safety Analysis Report (FSAR) and only

affects the definitions and does not have any

affect on any work performed. The change to

core alteration is to clarify those components

that may result in reactivity changes. The

change will not effect movement of fuel or

components that effect reactivity, therefore, a

fuel handling accident will not be effected.

The change in the definition of quadrant

power tilt ratio (QPTR) allows the alternate

method of determining QPTR to be utilized.

The current TS surveillance requirement (SR)

and bases allow alternate means for

determining QPTR, therefore, revising the

definition will have no effect on any

accident. The revision to the mode

parameters is administrative in nature,

therefore it will have no effect on any

accident. This change has no impact on

accident initiators and does not involve a

physical modification to the plant.

Accordingly, the proposed changes do not

involve an increase in the probability or

consequences of an accident previously

evaluated.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

This revision will not change any plant

equipment, system configurations, or

accident assumptions. This change will

better define the associated parameters and

will eliminate potential ambiguity and

confusion. The change in the definition of

core alteration allows components that do

not affect reactivity to be moved within the

reactor vessel. The change in the definition

will not effect the monitoring of QPTR with

one channel inoperable. The core will be

monitored in accordance with the SRs.

Therefore, this change will not affect the

safety function of any components and will

not create the possibility of a new or different

kind of accident.

3. Involve a significant reduction in a

margin of safety.

The proposed changes provide

improvements for SQN’s TS. This change

does not involve a physical modification to

the plant nor change the methods of

monitoring plant parameters. Accordingly,

the margin of safety has not been reduced.
20532 Federal Register / Vol. 60, No. 80 / Wednesday, April 26, 1995 / Notices

The NRC has reviewed the licensee’s

analysis and, based on this review, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402.

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H,

Knoxville, Tennessee 37902.

NRC Project Director: Frederick J.

Hebdon.

Wolf Creek Nuclear Operating

Corporation, Docket No. 50–482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request: March

24, 1995.

Description of amendment request:

This amendment request proposes to

revise Technical Specification (TS) 1.7,

‘‘Containment Integrity,’’ TS 3/4.6.1,

‘‘Containment Integrity,’’ TS 3/4.6.3,

‘‘Containment Isolation Valves,’’ and the

associated Bases. These proposed

changes will relocate TS Table 3.6–1,

‘‘Containment Isolation Valves,’’ to Wolf

Creek Generating Station procedures.

This proposed change is in accordance

with the guidance provided in Generic

Letter 91–08, ‘‘Removal of Component

Lists from Technical Specifications,’’

dated May 6, 1991.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes simplify the

technical specifications, meet the regulatory

requirements for control of containment

isolation, and are consistent with the

guidelines of GL 91–08. The procedural

details of Technical Specification Table 3.6–

1 have not been changed, but only relocated

to a different controlling document. The

proposed changes are administrative in

nature, should result in improved

administrative practices, and do not affect

plant operations.

The probability of occurrence of a

previously evaluated accident is not

increased because this change does not

introduce any new potential accident

initiating conditions. The consequences of an

accident previously evaluated is not

increased because the ability of [the]

containment to restrict the release of any

fission product radioactivity to the

environment will not be degraded by this

change.

(2) The proposed change does not create

the possibility of a new or different kind of

accident from any previously evaluated.

The proposed changes are administrative

in nature, do not result in physical

alterations or changes to the operation of the

plant, and cause no change in the method by

which any safety-related system performs its

function. Therefore, this 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 the margin of

safety.

The administrative change to relocate

Technical Specification Table 3.6–1 to

appropriate plant procedures does not alter

the basic regulatory requirements for

containment isolation and will not adversely

affect containment isolation capability for

credible accident scenarios. Adequate control

of the content of the table is assured by

existing plant procedures.

The proposed relocation of Technical

Specification Table 3.6–1 does not alter the

requirements for containment isolation valve

operability currently in the technical

specifications. The LCO [limiting condition

for operation] and Surveillance Requirements

would be retained in the revised technical

specifications. Therefore, the proposed

change will not affect the meaning,

application, and function of the current

technical specification requirements for the

valves in Table 3.6–1.

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

locations: Emporia State University,

William Allen White Library, 1200

Commercial Street, Emporia, Kansas

66801 and Washburn University School

of Law Library, Topeka, Kansas 66621.

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts and Trowbridge,

2300 N Street, N.W., Washington, D.C.

20037.

NRC Project Director: William H.

Bateman.

Previously Published Notices of

Consideration of Issuance of

Amendments to Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing

The following notices were previously

published as separate individual

notices. The notice content was the

same as above. They were published as

individual notices either because time

did not allow the Commission to wait

for this biweekly notice or because the

action involved exigent circumstances.

They are repeated here because the

biweekly notice lists all amendments

issued or proposed to be issued

involving no significant hazards

consideration.

For details, see the individual notice

in the Federal Register on the day and

page cited. This notice does not extend

the notice period of the original notice.

Entergy Operations, Inc., Docket No. 50–

382, Waterford Steam Electric Station,

Unit 3, St. Charles Parish, Louisiana

Date of amendment request: April 4,

1995, as supplemented by letter dated

April 5, 1995.

Description of amendment request:

The proposed amendment would

change the technical specifications on

moderator temperature coefficient. The

proposed change constitutes a one time

deviation not to perform the two-thirds

end-of-cycle moderator temperature

coefficient test for Cycle 7.

Date of individual notice in the

Federal Register: April 11, 1995 (60 FR

18432).

Expiration date of individual notice:

May 11, 1995.

Local Public Document Room

location: University of New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, Louisiana 70122.

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
Federal Register / Vol. 60, No. 80 / Wednesday, April 26, 1995 / Notices 20533

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the applications for

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

the Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, the Gelman Building, 2120 L

Street, NW., Washington, DC, and at the

local public document rooms for the

particular facilities involved.

Baltimore Gas and Electric Company,

Docket No. 50–318, Calvert Cliffs

Nuclear Power Plant, Unit No. 2, Calvert

County, Maryland

Date of application for amendment:

February 24, 1995.

Brief description of amendment: The

amendment revises Technical

Specification Section 4.6.1.2.a, Primary

Containment/Containment Leakage, to

reference 10 CFR Part 50, Appendix J,

as modified by approved exemptions.

Date of issuance: April 10, 1995.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 183.

Facility Operating License No. DPR–

69: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 8, 1995 (60 FR 12789).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 10, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Calvert County Library, Prince

Frederick, Maryland 20678.

Carolina Power & Light Company, et al.,

Docket Nos. 50–325 and 50–324,

Brunswick Steam Electric Plant, Units 1

and 2, Brunswick County, North

Carolina

Date of application for amendments:

September 30, 1994, as supplemented

on March 24, 1995.

Brief Description of amendments: The

proposed change will revise Technical

Specification requirements to eliminate

the reactor scram and isolation

functions of the Main Steam Line

Radiation Monitors. The March 24,

1995, supplement provided clarifying

information only and did not affect the

NRC’s determination of no significant

hazards considerations.

Date of issuance: March 31, 1995.

Effective date: March 31, 1995.

Amendment No.: 160.

Facility Operating License Nos. DPR–

71 and DPR–62. Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: November 9, 1994 (59 FR

55867). The March 24, 1995, submittal

provided clarifying information only

and did not affect the no significant

hazards consideration as published in

the Federal Register.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated March 31, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of North Carolina at

Wilmington, William Madison Randall

Library, 601 S. College Road,

Wilmington, North Carolina 28403–

3297.

Carolina Power & Light Company,

Docket No. 50–261, H. B. Robinson

Steam Electric Plant, Unit No. 2,

Darlington County, South Carolina

Date of application for amendment:

July 22, 1994, as supplemented March 6,

1995.

Brief description of amendment: The

amendments change the Technical

Specifications to implement a

performance based assessment program,

including corresponding organizational

and functional changes. Specifically, the

changes affect the independent

assessment of plant activity and the

independent review function, the

independent assessment of plant

activity and the Independent Safety

Engineering Group. These functions will

be performed by the Nuclear

Assessment Section (NAS). The NAS’s

fundamental role will be to: (1) Assist

plant management in the early

identification of issues that may prevent

the plant from achieving quality, and (2)

ensure effective correction of

deficiencies.

Date of issuance: March 31, 1995.

Effective date: March 31, 1995.

Amendment No.: 160.

Facility Operating License No. DPR–

23. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: August 31, 1994 (59 FR

45017). The March 6, 1995, submittal

added Radiation Protection to the list of

assessments in TS 6.5.5.2 and reworded

Section 6.5.4.4, but did not change the

no significant hazards consideration

determination as published in the

Federal Register.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated March 31, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Hartsville Memorial Library,

147 West College Avenue, Hartsville,

South Carolina 29550.

Carolina Power & Light Company,

Docket No. 50–261, H. B. Robinson

Steam Electric Plant, Unit No. 2,

Darlington County, South Carolina

Date of application for amendment:

August 23, 1994, as supplemented

March 2, 1995.

Brief description of amendment: The

amendment increases the trip voltage

settings of the degraded grid voltage

relays which are shown in TS Table

3.5–1, Engineering Safety Feature

System Initiation Instrument Setting

Limits, Item 6b.

Date of issuance: April 14, 1995.

Effective date: April 14, 1995.

Amendment No.: 161.

Facility Operating License No. DPR–

23. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: March 2, 1995 (60 FR 11692).

The licensee’s March 2, 1995, submittal

provided clarifying information that did

not affect the no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 14, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Hartsville Memorial Library,

147 West College Avenue, Hartsville,

South Carolina 29550.

Commonwealth Edison Company,

Docket Nos. 50–373 and 50–374, LaSalle

County Station, Units 1 and 2, LaSalle

County, Illinois

Date of application for amendments:

June 9, 1994.

Brief description of amendments: The

amendments implement a partial

application of the Generic Electric

ARTS (Average Power Range Monitor

(APRM)/Rod Block Monitor (RBM)/

Technical Specification) Improvement

Program. Four new ARTS thermal limits

replace the existing flow-referenced

APRM trip setpoint setdown

requirements and the Minimum Critical

Power Ratio (MCPR) Kf factor.

Date of issuance: April 13, 1995.

Effective date: Immediately to be

implemented within 60 days.

Amendment Nos.: 103 and 88.

Facility Operating License Nos. NPF–

11 and NPF–18. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: August 3, 1994 (59 FR 39582).
20534 Federal Register / Vol. 60, No. 80 / Wednesday, April 26, 1995 / Notices

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated April 13, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Jacobs Memorial Library,

Illinois Valley Community College,

Oglesby, Illinois 61348.

Detroit Edison Company, Docket No.

50–341, Fermi–2, Monroe County,

Michigan

Date of application for amendment:

July 29, 1993, as supplemented October

8, 1993.

Brief description of amendment: The

amendment revises the Technical

Specification ACTION STATEMENTS

related to operability of the control

room emergency filtration system. A

portion of the amendment request was

denied. A separate Notice of Denial of

Amendment has been sent to the

Federal Register for publication.

Date of issuance: March 31, 1995.

Effective date: March 31, 1995, with

full implementation within 45 days.

Amendment No.: 103.

Facility Operating License No. NPF–

43. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: August 18, 1993 (58 FR

43925). The October 8, 1993, letter

provided clarifying information within

the scope of the original submittal and

did not change the staff’s initial no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated March 31, 1995.

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–

369 and 50–370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg

County, North Carolina

Date of application for amendments:

January 18, 1995.

Brief description of amendments: The

amendments revise TS Table 4.3–3 to

allow the analog channel operational

test interval for radiation monitoring

instrumentation to be increased from

monthly to quarterly, and are consistent

with the guidance in Generic Letter 93–

05, ‘‘Line-Item Technical Specifications

Improvements to Reduce Surveillance

Requirements for Testing During Power

Operation.’’

Date of issuance: April 3, 1995.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment Nos.: 154 and 136.

Facility Operating License Nos. NPF–

9 and NPF–17. Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: March 1, 1995 (60 FR 11132).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated April 3, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Atkins Library, University of

North Carolina, Charlotte (UNCC

Station), North Carolina 28223.

Entergy Operations, Inc., Docket No. 50–

368, Arkansas Nuclear One, Unit No. 2,

Pope County, Arkansas

Date of application for amendment:

July 22, 1993.

Brief description of amendment: The

amendment revised operability

requirements for the Reactor Protection

System and the Engineered Safety

Features Acturation System.

Date of issuance: April 3, 1995.

Effective date: April 3, 1995.

Amendment No.: 159.

Facility Operating License No. NPF–6.

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: September 1, 1993 (58 FR

46229).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 3, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801.

Florida Power and Light Company,

Docket Nos. 50–250 and 50–251, Turkey

Point Plant, Units 3 and 4, Dade County,

Florida

Date of application for amendments:

July 19, 1994, resubmitted October 20,

1994, and supplemented February 20,

1995.

Brief description of amendments:

These amendments relate to the

maximum allowable reactor thermal

power operation with inoperable main

steam safety valves.

Date of issuance: April 11, 1995.

Effective date: April 11, 1995.

Amendment Nos.: 172 and 166.

Facility Operating Licenses Nos. DPR–

31 and DPR–41. Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: November 23, 1994 (59 FR

60380).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated April 11, 1995.

The February 20, 1995 submittal

provided additional information that

did not change the staff’s proposed no

significant hazards consideration

determination.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Florida International

University, University Park, Miami,

Florida 33199.

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

Duane Arnold Energy Center, Linn

County, Iowa

Date of application for amendment:

October 28, 1994.

Brief description of amendment: The

amendment revised the license by

deleting the ‘‘Plan for the Integrated

Scheduling of Plant Modifications for

the Duane Arnold Energy Center,’’ as a

condition of the license.

Date of issuance: April 3, 1995.

Effective date: April 3, 1995.

Amendment No.: 208.

Facility Operating License No. DPR–

49. Amendment revised the license.

Date of initial notice in Federal

Register: March 1, 1995 (60 FR 11134).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 3, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Cedar Rapids Public Library,

500 First Street SE., Cedar Rapids, Iowa

52401.

Maine Yankee Atomic Power Company,

Docket No. 50–309, Maine Yankee

Atomic Power Station, Lincoln County,

Maine

Date of application for amendment:

December 6, 1994, as supplemented on

February 27, and April 4, 1995.

Brief description of amendment: The

amendment revises the Technical

Specifications to allow the use of the

Combustion Engineering sleeving

process for repairing steam generator

tubes. (The current requirement

specifies that degraded steam generator

tubes be repaired by plugging.)

Date of issuance: April 14, 1995.

Effective date: As of the date of

issuance, to be implemented within 30

days.

Amendment No.: 149.

Facility Operating License No. DPR–

36. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 18, 1995 (60 FR 3673).

The February 27, and April 4, 1995,
Federal Register / Vol. 60, No. 80 / Wednesday, April 26, 1995 / Notices 20535

letters provided clarifying information

that did not change the initial proposed

no significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 14, 1995.

No significant hazards consideration

comments received: Yes.

Comments were provided by letter

dated February 17, 1995, from the

Maine State Nuclear Safety Inspector,

Office of Nuclear Safety, Division of

Health Engineering, Department of

Human Services. The NRC staff

responded to his comments in its letter

dated March 15, 1995.

Local Public Document Room

location: Wiscasset Public Library, High

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

04578.

Northeast Nuclear Energy Company, et

al., Docket No. 50–423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of application for amendment:

September 30, 1994, as supplemented

February 13, 1995.

Brief description of amendment: The

amendment revises the Technical

Specifications to (1) clarify the

definition of core alterations, (2) change

the verbiage in the Limiting Condition

For Operation (LCO) addressing the

refueling operations, (3) make changes

of surveillance requirements involving

source range instrumentation, and (4)

change the LCO regarding the Residual

Heat Removal and coolant circulation

water levels to be consistent with the

guidance provided in NUREG–1431,

Standard Technical Specifications for

Westinghouse plants.

Date of issuance: April 12, 1995.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 107.

Facility Operating License No. NPF–

49. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: November 9, 1994 (59 FR

55877). The February 13, 1995, letter

provided clarifying information that did

not change the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 12, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

Location: Learning Resources Center,

Three Rivers Community-Technical

College, Thames Valley Campus, 574

New London Turnpike, Norwich, CT

06360.

Northeast Nuclear Energy Company, et

al., Docket No. 50–423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of application for amendment:

January 26, 1993, as supplemented

August 4, 1993.

Brief description of amendment: The

amendment changes the Technical

Specifications governing electrical

power systems, AC and DC power

sources, and onsite power distribution

for shutdown conditions (modes 5 and

6).

Date of issuance: April 12, 1995.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 108.

Facility Operating License No. NPF–

49. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: May 12, 1993 (58 FR 28058).

The August 4, 1993, submittal provided

clarifying information that did not

change the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 12, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, Thames Valley Campus, 574

New London Turnpike, Norwich, CT

06360.

North Atlantic Energy Service

Corporation, Docket No. 50–443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request: April 23,

1993.

Description of amendment request:

The amendment revises the Appendix A

Technical Specifications to allow longer

surveillance test intervals and allowed

outage times for the reactor protection

system and the engineered safety

features actuation system. Also, the

amendment removes the requirement to

perform the reactor trip system analog

channel operational test on a staggered

basis.

Date of issuance: April 10, 1995.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 36.

Facility Operating License No. NPF–

86. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 4, 1993 (58 FR 41507).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 10, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Exeter Public Library, 47 Front

Street, Exeter, NH 03833.

Pacific Gas and Electric Company,

Docket Nos. 50–275 and 50–323, Diablo

Canyon Nuclear Power Plant, Unit Nos.

1 and 2, San Luis Obispo County,

California

Date of application for amendments:

February 16, 1994 (Reference LAR 94–

03)

Brief description of amendments: The

amendments revise TS 4.6.1.2,

‘‘Containment Integrity,’’ to allow a

more flexible schedule for testing the

primary containment integrated leakage

rate.

Date of issuance: April 11, 1995.

Effective date: April 11, 1995, to be

implemented within 30 days of

issuance.

Amendment Nos.: Unit 1—

Amendment No. 99; Unit 2—

Amendment No. 98

Facility Operating License Nos. DPR–

80 and DPR–82. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: March 30, 1994 (59 FR 14893).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated April 11, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: California Polytechnic State

University, Robert E. Kennedy Library,

Government Documents and Maps

Department, San Luis Obispo, California

93407.

Pacific Gas and Electric Company,

Docket Nos. 50–275 and 50–323, Diablo

Canyon Nuclear Power Plant, Unit Nos.

1 and 2, San Luis Obispo County,

California

Date of application for amendments:

August 17, 1994.

Brief description of amendments: The

amendments revise the combined

technical specifications (TS) to change

TS 3/4.4.9.1, ‘‘Reactor Coolant System—

Pressure/Temperature Limits,’’ Figures

3.4–2, ‘‘Reactor Coolant System Heatup

Limitations—Applicable Up to 8 EFPY,’’

and 3.4–3, ‘‘Reactor Coolant System

Cooldown Limitations—Applicable Up

to 8 EFPY,’’ to extend the applicability

up to 12 effective full-power years

(EFPYs). TS 3/4/4/9/3, ‘‘Overpressure

Protection Systems,’’ is revised to
20536 Federal Register / Vol. 60, No. 80 / Wednesday, April 26, 1995 / Notices

specify a new low-temperature

overprotection (LTOP) system actuation

pressure setpoint. The associated Bases

were also appropriately revised.

Additionally, TS 3/4.1.2.2, ‘‘Flow

Paths—Operating;’’ TS 3/4.1.2.4,

‘‘Charging Pumps—Operating;’’ TS 3/

4.4.1.3, ‘‘Hot Shutdown;’’ TS 3/4.4.1.4.1,

‘‘Cold Shutdown—Loops Filled;’’ TS 3/

4.4.9.3, ‘‘Overpressure Protection

Systems;’’ and TS 3/4.5.3, ‘‘ECCS

Subsystems—Tavg Less than 350 Degrees

F,’’ are revised to specify a new LTOP

system enable temperature.

Date of issuance: April 13, 1995.

Effective date: April 13, 1995.

Amendment Nos.: Unit 1—

Amendment No. 100; Unit 2—

Amendment No. 99.

Facility Operating License Nos. DPR–

80 and DPR–82. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: October 12, 1994 (59 FR

51622).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated April 13, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: California Polytechnic State

University, Robert E. Kennedy Library,

Government Documents and Maps

Department, San Luis Obispo, California

93407.

Power Authority of the State of New

York, Docket No. 50–333, James A.

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of application for amendment:

June 13, 1994.

Brief description of amendment: The

amendment removes License Condition

2.E from the Facility Operating License.

License Condition 2.E incorporated the

requirements of U.S. Department of

Interior publication ‘‘Environmental

Criteria for Electric Transmission

Systems’’—1970, which applies to the

construction cleanup, restoration, and

maintenance of transmission lines. The

NRC staff has determined that removing

this condition from the Facility

Operating License has no bearing on

plant safety or the health and safety of

the public, and is therefore acceptable.

Date of issuance: March 31, 1995.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 224.

Facility Operating License No. DPR–

59. Amendment revised the Facility

Operating License.

Date of initial notice in Federal

Register: March 1, 1995 (60 FR 11140).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated March 31, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Southern Nuclear Operating Company,

Inc., Docket Nos. 50–348 and 50–364,

Joseph M. Farley Nuclear Plant, Units 1

and 2, Houston County, Alabama

Date of amendments request: June 10,

1995.

Brief Description of amendments: The

amendments allow modifications to be

made for both units to relocate the lower

level steam generator water level taps

during the upcoming refueling outages.

These modifications affect the Technical

Specifications associated with the

reactor trip system and engineered

safety feature actuation system

setpoints.

Date of issuance: April 7, 1995.

Effective date: As of the date of

issuance to be implemented within 30

days

Amendment Nos.: 114 and 105.

Facility Operating License Nos. NPF–

2 and NPF–8. Amendments revise the

Technical Specifications.

Date of initial notice in Federal

Register: March 6, 1995 (60 FR 12253).

Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated April 7, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Houston-Love Memorial

Library, 212 W. Burdeshaw Street, Post

Office Box 1369, Dothan, Alabama

36302.

Southern Nuclear Operating Company,

Inc., Joseph M. Farley Nuclear Plant,

Unit 2, Houston County, Alabama

Date of application for amendment:

December 7, 1994, as supplemented

February 14 and March 20, 1995.

Brief description of amendment: The

December 7, 1994, submittal requested a

permanent change to the Technical

Specifications for both units related to

steam generator tube support plate

voltage-based repair criteria in

accordance with the draft Generic Letter

on this issue.

Date of issuance: April 7, 1995.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 106.

Facility Operating License No. NPF–8.

Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: February 15, 1995 (60 FR

8754). The February 14 and March 20,

1995, letters provided clarifying

information that did not change the

scope of the original December 7, 1994,

application and the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 7, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Houston-Love Memorial

Library, 212 W. Burdeshaw Street, Post

Office Box 1369, Dothan, Alabama

36302.

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of application for amendments:

December 16, 1994; supplemented

February 10, 1995 (TS 94–07).

Brief description of amendments: The

amendments revise the technical

specifications to reduce the high reactor

power level setpoints when one or more

main steam safety valves are inoperable

and incorporate related changes.

Date of issuance: April 4, 1995.

Effective date: April 4, 1995.

Amendment Nos.: 196 and 187.

Facility Operating License Nos. DPR–

77 and DPR–79. Amendments revise the

technical specifications.

Date of initial notice in Federal

Register: March 1, 1995 (60 FR 11140).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 4, 1995.

No significant hazards consideration

comments received: None.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402.

TU Electric Company, Docket Nos. 50–

445 and 50–446, Comanche Peak Steam

Electric Station, Unit Nos. 1 and 2,

Somervell County, Texas

Date of amendment request: February

14, 1994 (Reference LAR 94–002, TXX–

94008), as supplemented by letters

dated May 17, 1994 (Reference TXX–

94142), and April 3, 1995 (TXX–95098).

Brief description of amendments: The

amendments revised Technical

Specification (TS) 3/4.2.4, ‘‘Quadrant

Power Tilt Ratio,’’ by replacing the

existing TS and associated Bases

concerning the quadrant power tilt ratio

with a TS consistent with the improved

Standard Technical Specifications

(NUREG-1431).
Federal Register / Vol. 60, No. 80 / Wednesday, April 26, 1995 / Notices 20537

Date of issuance: April 4, 1995.

Effective date: April 4, 1995.

Amendment Nos.: Unit 1—

Amendment No. 36; Unit 2—

Amendment No. 22.

Facility Operating License Nos. NPF–

87 and NPF–89. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: July 20, 1994 (59 FR 37087).

The additional information contained

in the supplemental letter dated April 3,

1995, was clarifying in nature and thus,

within the scope of the initial notice

and did not affect the staff’s proposed

no significant hazards consideration

determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated April 4, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of Texas at

Arlington Library, Government

Publications/Maps, 702 College, P.O.

Box 19497, Arlington, TX 76019.

TU Electric Company, Docket Nos. 50–

445 and 50–446, Comanche Peak Steam

Electric Station, Unit Nos. 1 and 2,

Somervell County, Texas

Date of amendment request: February

28, 1995 (Reference LAR 95–01).

Brief description of amendments:

These amendments replace Surveillance

Requirement (SR) 4.6.2.1b of Technical

Specification (TS) 3/4.6.2,

‘‘Depressurization and Cooling

Systems—Containment Spray System,’’

with the corresponding SR from

NUREG–1431. Bases Section 3/4.6.2.1

‘‘Containment Spray System’’ has also

been revised to expand the detail

consistent with the corresponding Bases

from NUREG–1431. The SR, and its

associated Bases, for confirming the

performance of the containment spray

pumps are changed by replacing the

specific pump head and flow values

with the general requirement that the

pumps provide the required head at the

flow test point while the specific

required values are moved to the

Comanche Peak Steam Electric Station

Technical Requirements Manual.

Date of issuance: April 6, 1995.

Effective date: April 6, 1995.

Amendment Nos.: Unit 1—

Amendment No. 37; Unit 2—

Amendment No. 23.

Facility Operating License Nos. NPF–

87 and NPF–89. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: March 6, 1995 (60 FR 12255).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated April 6, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of Texas at

Arlington Library, Government

Publications/Maps, 702 College, P.O.

Box 19497, Arlington, TX 76019.

TU Electric Company, Docket Nos. 50–

445 and 50–446, Comanche Peak Steam

Electric Station, Unit Nos. 1 and 2,

Somervell County, Texas

Date of amendment request: April 22,

1994 (LAR 94–011, TXX–94116).

Brief description of amendments:

These amendments changed

Surveillance Requirement 4.7.1.2 of

Technical Specification 3/4.7.1.2

‘‘Auxiliary Feedwater System,’’ for the

operational test frequency of the motor

driven and turbine driven pumps from

‘‘at least once per 31 days on a

STAGGERED TEST BASIS’’ to ‘‘at least

once per 92 days on a STAGGERED

TEST BASIS.’’ This change is consistent

with ASME Section XI requirements

and Generic Letter 93–05, ‘‘Line-Item

Technical Specifications Improvements

to Reduce Surveillance Requirements

for Testing During Power Operations.’’

Date of issuance: April 7, 1995.

Effective date: April 7, 1995.

Amendment Nos.: Unit 1—

Amendment No. 38; Unit 2—

Amendment No. 24.

Facility Operating License Nos. NPF–

87 and NPF–89. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: August 3, 1994 (59 FR 39598).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated April 7, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of Texas at

Arlington Library, Government

Publications/Maps, 702 College, P.O.

Box 19497, Arlington, TX 76019.

Union Electric Company, Docket No.

50–483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application for amendment:

December 9, 1994, as supplemented on

January 27, 1995.

Brief description of amendment: The

amendment revises the Technical

Specification Surveillance Requirement

4.6.1.2.a and its associated Bases. The

change defers the requirement to

perform the Type A Containment

Integrated Leak Rate Test until Refuel 8

(October 1996), in conjunction with the

exemption to 10 CFR Part 50, Appendix

J.

Date of issuance: April 5, 1995.

Effective date: April 5, 1995.

Amendment No.: 98.

Facility Operating License No. NPF–

30. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: March 1, 1995 (60 FR 11141).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 5, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Callaway County Public

Library, 710 Court Street, Fulton,

Missouri 65251.

Wolf Creek Nuclear Operating

Corporation, Docket No. 50–482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request:

December 12, 1994.

Brief description of amendment: The

amendment clarifies the surveillance

requirements for verifying the correct

required position for the valves in the

auxiliary feedwater system.

Date of issuance: April 3, 1995.

Effective date: April 3, 1995, to be

implemented within 30 days of

issuance.

Amendment No.: 85.

Facility Operating License No. NPF–

42. The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: January 18, 1995 (60 FR 3677).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 3, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

locations: Emporia State University,

William Allen White Library, 1200

Commercial Street, Emporia, Kansas

66801 and Washburn University School

of Law Library, Topeka, Kansas 66621.

Wisconsin Public Service Corporation,

Docket No. 50–305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of application for amendment:

April 11, 1994, as supplemented on

November 30, and December 22, 1994,

and March 3, 1995.

Brief description of amendment: The

amendment revises Kewaunee Nuclear

Power Plant (KNPP) Technical

Specification (TS) 3.1.f, ‘‘Minimum

Conditions for Criticality,’’ and its

associated basis, by specifying that the

moderator temperature coefficient

(MTC) shall be no greater than 5.0 pcm/

°F when at or below 60% rated thermal

power and shall be zero or negative

when above 60% rated thermal power.

Additionally, the MTC shall be no less
20538 Federal Register / Vol. 60, No. 80 / Wednesday, April 26, 1995 / Notices

negative than ¥8 pcm/°F for 95% of the

cycle time at full power. The

amendment also incorporates required

actions to be implemented, if the MTC

specification is not met.

Date of issuance: April 3, 1995.

Effective date: April 3, 1995.

Amendment No.: 117.

Facility Operating License No. DPR–

43. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: September 28, 1994 (59 FR

49442).

The November 30, and December 22,

1994, and March 3, 1995, submittals,

provided clarifying information and

expanded the basis portion of the TS,

but did not change the initial no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 3, 1995.

No significant hazards consideration

comments received: None.

Local Public Document Room

location: University of Wisconsin

Library Learning Center, 2420 Nicolet

Drive, Green Bay, Wisconsin 54301.

Notice of Issuance of Amendments to

Facility Operating Licenses and Final

Determination of No Significant

Hazards Consideration and

Opportunity for a Hearing (Exigent

Public Announcement or Emergency

Circumstances)

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application for the

amendment complies with the

standards and requirements of the

Atomic Energy Act of 1954, as amended

(the Act), and the Commission’s rules

and regulations. The Commission has

made appropriate findings as required

by the Act and the Commission’s rules

and regulations in 10 CFR Chapter I,

which are set forth in the license

amendment.

Because of exigent or emergency

circumstances associated with the date

the amendment was needed, there was

not time for the Commission to publish,

for public comment before issuance, its

usual 30-day Notice of Consideration of

Issuance of Amendment, Proposed No

Significant Hazards Consideration

Determination, and Opportunity for a

Hearing.

For exigent circumstances, the

Commission has either issued a Federal

Register notice providing opportunity

for public comment or has used local

media to provide notice to the public in

the area surrounding a licensee’s facility

of the licensee’s application and of the

Commission’s proposed determination

of no significant hazards consideration.

The Commission has provided a

reasonable opportunity for the public to

comment, using its best efforts to make

available to the public means of

communication for the public to

respond quickly, and in the case of

telephone comments, the comments

have been recorded or transcribed as

appropriate and the licensee has been

informed of the public comments.

In circumstances where failure to act

in a timely way would have resulted, for

example, in derating or shutdown of a

nuclear power plant or in prevention of

either resumption of operation or of

increase in power output up to the

plant’s licensed power level, the

Commission may not have had an

opportunity to provide for public

comment on its no significant hazards

consideration determination. In such

case, the license amendment has been

issued without opportunity for

comment. If there has been some time

for public comment but less than 30

days, the Commission may provide an

opportunity for public comment. If

comments have been requested, it is so

stated. In either event, the State has

been consulted by telephone whenever

possible.

Under its regulations, the Commission

may issue and make an amendment

immediately effective, notwithstanding

the pendency before it of a request for

a hearing from any person, in advance

of the holding and completion of any

required hearing, where it has

determined that no significant hazards

consideration is involved.

The Commission has applied the

standards of 10 CFR 50.92 and has made

a final determination that the

amendment involves no significant

hazards consideration. The basis for this

determination is contained in the

documents related to this action.

Accordingly, the amendments have

been issued and made effective as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the application for

amendment, (2) the amendment to

Facility Operating License, and (3) the

Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment, as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, the Gelman Building, 2120 L

Street, NW., Washington, DC, and at the

local public document room for the

particular facility involved.

The Commission is also offering an

opportunity for a hearing with respect to

the issuance of the amendment. By May

26, 1995, the licensee may file a request

for a hearing with respect to issuance of

the amendment to the subject facility

operating license and any person whose

interest may be affected by this

proceeding and who wishes to

participate as a party in the proceeding

must file a written request for a hearing

and a petition for leave to intervene.

Requests for a hearing and a petition for

leave to intervene shall be filed in

accordance with the Commission’s

‘‘Rules of Practice for Domestic

Licensing Proceedings’’ in 10 CFR Part

2. Interested persons should consult a

current copy of 10 CFR 2.714 which is

available at the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington, DC and

at the local public document room for

the particular facility involved. If a

request for a hearing or petition for

leave to intervene is filed by the above

date, the Commission or an Atomic

Safety and Licensing Board, designated

by the Commission or by the Chairman

of the Atomic Safety and Licensing

Board Panel, will rule on the request

and/or petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) The nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.
Federal Register / Vol. 60, No. 80 / Wednesday, April 26, 1995 / Notices 20539

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses. Since the Commission has

made a final determination that the

amendment involves no significant

hazards consideration, if a hearing is

requested, it will not stay the

effectiveness of the amendment. Any

hearing held would take place while the

amendment is in effect.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

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

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

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

1–(800) 342–6700). The Western Union

operator should be given Datagram

Identification Number N1023 and the

following message addressed to (Project

Director): petitioner’s name and

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555, and to the attorney for the

licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

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).

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:

March 29, 1995.

Brief description of amendments: The

amendments change the Technical

Specifications by revising the

periodicity of the channel functional

test of the turbine driven auxiliary

feedwater pump from quarterly to each

refueling outage.

Date of issuance: April 14, 1995.

Effective date: April 14, 1995.

Amendment Nos.: 161 and 149.

Facility Operating License Nos. DPR–

39 and DPR–48. The amendments

revised the Technical Specifications.

Public comments requested as to

proposed no significant hazards

consideration: No.

The Commission’s related evaluation

of the amendments, finding of

emergency circumstances, and final

determination of no significant hazards

consideration are contained in a Safety

Evaluation dated April 14, 1995.

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60690.

Local Public Document Room

location: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois

60085.

NRC Project Director: Robert A. Capra.

Dated at Rockville, Maryland, this 19th day

of April 1995.

For the Nuclear Regulatory Commission.

Elinor G. Adensam,

Acting Director, Division of Reactor Projects—

III/IV, Office of Nuclear Reactor Regulation.

[FR Doc. 95–10127 Filed 4–25–95; 8:45 am]

BILLING CODE 7590–01–P

Regulatory Guide; Issuance,

Availability

The Nuclear Regulatory Commission

has issued a revision to a guide in its

Regulatory Guide Series. This series has

been developed to describe and make

available to the public such information

as methods acceptable to the NRC staff

for implementing specific parts of the

Commission’s regulations, techniques

used by the staff in evaluating specific

problems or postulated accidents, and

data needed by the staff in its review of

applications for permits and licenses.

Revision 3 of Regulatory Guide 1.118,

‘‘Periodic Testing of Electric Power and

Protection Systems,’’ describes a

method acceptable to the NRC staff for

complying with the Commission’s

regulations with respect to the periodic

testing of the electric power and

protection systems.

Comments and suggestions in

connection with items for inclusion in

guides currently being developed or

improvements in all published guides

are encouraged at any time. Written

comments may be submitted to the

Rules Review and Directives Branch,

Division of Freedom of Information and

Publications Services, Office of

Administration, U.S. Nuclear Regulatory

Commission, Washington, DC 20555.

Regulatory guides are available for

inspection at the Commission’s Public

Document Room, 2120 L Street, NW.,

Washington, DC. Copies of issued

guides may be purchased from the

Government Printing Office at the

current GPO price. Information on

current GPO prices may be obtained by

contracting the Superintendent of

Documents, U.S. Government Printing

Office, Post Office Box 37082,

Washington, DC 20013–7082, telephone

(202) 512–2249. Issued guides may also

be purchased from the National

Technical Information Service on a

standing order basis. Details on this

service may be obtained by writing

NTIS, 5285 Port Royal Road,

Springfield, VA 22161.

(5 U.S.C. 552(a))

Dated at Rockville, Maryland, this 12th day

of April 1995.