Federal Register / Vol. 60, No. 31 / Wednesday, February 15, 1995 / Notices 8741

prepare an environmental impact

statement for the proposed amendment.

For further details with respect to this

proposed action, see the licensee’s letter

dated October 15, 1993, as

supplemented by letters dated April 15,

and November 10, 1994, which are

available for public inspection at the

Commissin’s Public Document Room,

the Gelman Building, 2120 L Street,

NW., Washington, DC, and at the local

public document room located at the

Judge George W. Armstrong Library, 220

S. Commerce Street, Natchez,

Mississippi 39120.

Dated at Rockville, Maryland this 9th day

of February 1995.

For the Nuclear Regulatory Commission.

James R. Hall,

Senior Project Manager, Project Directorate

IV–I, Division of Reactor Projects III/IV, Office

of Nuclear Reactor Regulation.

[FR Doc. 95–3772 Filed 2–14–95; 8:45 am]

BILLING CODE 7590–01–M

Applications and Amendments to

Facility Operating Licenses Involving

No Significant Hazards

Considerations; Biweekly Notice

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 January 20,

1995, through February 3, 1995. The last

biweekly notice was published on

February 1, 1995 (60 FR 6296).

Notice of Consideration of Issuance of

Amendments to Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a

proposed determination that the

following amendment requests involve

no significant hazards consideration.

Under the Commission’s regulations in

10 CFR 50.92, this means that operation

of the facility in accordance with the

proposed amendment would not (1)

involve a significant increase in the

probability or consequences of an

accident previously evaluated; or (2)

create the possibility of a new or

different kind of accident from any

accident previously evaluated; or (3)

involve a significant reduction in a

margin of safety. The basis for this

proposed determination for each

amendment request is shown below.

The Commission is seeking public

comments on this proposed

determination. Any comments received

within 30 days after the date of

publication of this notice will be

considered in making any final

determination.

Normally, the Commission will not

issue the amendment until the

expiration of the 30-day notice period.

However, should circumstances change

during the notice period such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received before

action is taken. Should the Commission

take this action, it will publish in the

Federal Register a notice of issuance

and provide for opportunity for a

hearing after issuance. The Commission

expects that the need to take this action

will occur very infrequently.

Written comments may be submitted

by mail to the 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 March 17, 1995, the licensee may

file a request for a hearing with respect

to issuance of the amendment to the

subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) The nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the
8742 Federal Register / Vol. 60, No. 31 / Wednesday, February 15, 1995 / Notices

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.

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: January

19, 1995.

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS)

Surveillance Requirement 4.0.3 and its

associated bases to provide for a delay

period of up to 24 hours in which to

perform a surveillance which has been

discovered not to have been performed

within its specified frequency. This

change would adopt the requirements of

NUREG–1431, ‘‘Standard Technical

Specifications, Westinghouse Plants.’’

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.

The proposed change will reduce the

requirement to unnecessarily

manipulate and challenge plant systems

and equipment. The most probable

result of performing a surveillance

during the delay period will be to verify

its conformance with Technical

Specification requirements. Since this

change does not affect plant design,

operation, or the manner in which

testing is performed, the consequences

of accident scenarios postulated in the

Final Safety Analysis Report will not

increase. Therefore, 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.

The proposed change does not

introduce any new equipment, nor does

it require existing systems to perform a

different type of function than they are

currently designed to perform.

Therefore, the proposed change does 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.

The margin of safety is neither

described or prescribed for this

specification. The proposed change

simply provides additional time to

perform a surveillance and verify that

the operability of equipment is in

conformance with the Technical

Specification requirements. Therefore,

the proposed 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: 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: William H.

Bateman.

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:

December 23, 1994.

Description of amendment request:

The proposed amendments would

increase the allowable enrichment of

new fuel stored in the new fuel storage

vault (NFSV), revise the enrichment

description of fuel in the reactor core,

and include references to documents

previously approved by the staff in the
Federal Register / Vol. 60, No. 31 / Wednesday, February 15, 1995 / Notices 8743

Technical Specifications that provide

analytical methods used to determine

core operating limits.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

A.1. The proposed change does not

involve a significant increase in the

probability of occurrence or

consequences of any accident

previously evaluated.

The Updated Final Safety Analysis

Report (UFSAR) does not consider any

accidents involving the NFSV. The Fuel

Handling Accidents that are analyzed

(Section 15.7.4) include dropping of a

spent fuel assembly onto the spent fuel

pool floor and breaking of all fuel rods,

and dropping of a fuel assembly inside

containment onto the top of the core.

The proposed change to increase the

NFSV fuel enrichment limit from 4.0 to

4.65 weight percent U–235 does not

affect any of the initiators or precursors

of any accident previously evaluated.

The proposed change will not increase

the likelihood that a transient initiating

event will occur because transients are

initiated by equipment malfunction

and/or catastrophic system failure.

Since the proposed change does not

involve the introduction of new or

redesigned plant equipment, failure

mechanisms are not affected. As a

result, the probability of occurrence of

accidents previously evaluated is not

significantly increased.

A new criticality analysis for the

proposed change to increase the NFSV

fuel enrichment limit from 4.0 to 4.65

weight percent U–235 was performed

for the NFSV. It was determined that

even in worst case conditions the

acceptance criteria was met since the

maximum Keff was determined to be

well below the 0.95 limit with a 95/95

probability/confidence level. The

consequences of any accident, including

a fuel handling accident involving the

NFSV, are not significantly increased.

A.2. The proposed change does not

create the possibility of a new or

different kind of accident from any

previously analyzed.

The proposed change to the Technical

Specifications does 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 analysis. 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.

A.3. The proposed change does not

involve a significant reduction in a

margin of safety.

Plant safety margins are established

through LCOs, limiting safety system

settings, and safety limits specified in

the Technical Specifications. There will

be no changes to either the physical

design of the plant or to any of these

settings and limits as a result of

increasing the NFSV fuel enrichment

limit. The change does not involve a

significant increase in the probability of

occurrence or consequences of any

accident previously evaluated or create

the possibility of a new or different kind

of accident from any previously

analyzed. Additionally, the revised

criticality analysis demonstrates that the

maximum Keff under all postulated

conditions remains below the

acceptance value of 0.95. Therefore, the

change will not result in a significant

reduction in a margin of safety.

B.1. The proposed change does not

involve a significant increase in the

probability of occurrence or

consequences of any accident

previously evaluated.

The proposed change to increase the

reactor core fuel enrichment range

discussed in the Design Features section

of Technical Specifications from

‘‘between 2.2 to 4.0’’ to ‘‘up to 4.65’’

weight percent U–235 is administrative

in nature and does not affect any of the

initiators or precursors of any accident

previously evaluated. The proposed

change will not increase the likelihood

that a transient initiating event will

occur because transients are initiated by

equipment malfunction and/or

catastrophic system failure. Since the

proposed change does not involve the

introduction of new or redesigned plant

equipment, failure mechanisms are not

affected. As a result, the probability of

occurrence of accidents previously

evaluated is not significantly increased.

The fuel enrichment limit of each core

is determined by the core specific

design and is determined to be

acceptable with respect to the accident

analysis by the reload analysis and is

not impacted by the value specified in

the description in the Design Features

section of Technical Specifications. This

value is only provided as the highest

expected core fuel enrichment in the

Design Features section discussion of

the reactor core. This change is

administrative in nature and does not

affect the consequences of any accident

previously evaluated.

B.2. The proposed change does not

create the possibility of a new or

different kind of accident from any

previously analyzed.

The proposed change in the reactor

core fuel enrichment description

contained in the Design Features section

of Technical Specifications does 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 analysis. No safety related

equipment or function will be altered as

a result of the proposed change. Also,

the procedures governing normal plant

operation and recovery from an accident

are not changed by the proposed

Technical Specification change. Since

no new failure modes or mechanisms

are added by the proposed change, the

possibility of a new or different kind of

accident is not created.

B.3. The proposed change does not

involve a significant reduction in a

margin of safety.

Plant safety margins are established

through LCOs, limiting safety system

settings, and safety limits specified in

the Technical Specifications. There will

be no changes to either the physical

design of the plant or to any of these

settings and limits as a result of

increasing reactor core fuel enrichment

value given in the Design Features

section of Technical Specifications. The

change does not involve a significant

increase in the probability of occurrence

or consequences of any accident

previously evaluated or create the

possibility of a new or different kind of

accident from any previously analyzed.

Based on the above discussion, the

ability to safely shutdown the operating

unit and mitigate the consequences of

all accidents previously evaluated will

be maintained. Therefore, the margin of

safety is not significantly affected.

C.1. The proposed change does not

involve a significant increase in the

probability of occurrence or

consequences of any accident

previously evaluated.

The proposed change to add three

documents to the list of documents that

provide the analytical methods to

determine core operating limits is

administrative in nature and does not

affect any of the initiators or precursors

of any accident previously evaluated.

The proposed change will not increase

the likelihood that a transient initiating

event will occur because transients are

initiated by equipment malfunction
8744 Federal Register / Vol. 60, No. 31 / Wednesday, February 15, 1995 / Notices

and/or catastrophic system failure.

Since the proposed change does not

involve the introduction of new or

redesigned plant equipment, failure

mechanisms are not affected.

The documents have been previously

reviewed and approved by the NRC and

it was determined that they provide an

acceptable means to determine core

operating limits. As a result, the

probability of occurrence of accidents

previously evaluated is not significantly

increased. Since the documents provide

NRC approved methodologies for

determining core operating limits, the

addition of the documents to Technical

Specifications or use of the documents

to determine core operating limits will

not significantly increase the

consequences of any accident

previously evaluated.

C.2. The proposed change does not

create the possibility of a new or

different kind of accident from any

previously analyzed.

The proposed change to add three

documents to the list of documents that

provide the analytical methods to

determine core operating limits is

administrative in nature and does 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 analysis. 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.

C.3. The proposed change does not

involve a significant reduction in a

margin of safety.

Plant safety margins are established

through LCOs, limiting safety system

settings, and safety limits specified in

the Technical Specifications. There will

be no changes to either the physical

design of the plant or to any of these

settings and limits as a result of adding

references to the new documents. The

ability to 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.

Consolidated Edison Company of New

York, Docket No. 50–247, Indian Point

Nuclear Generating Unit No. 2,

Westchester County, New York Date of

amendment request: September 19,

1994.

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) Section

4.4.A.3 to reference the testing

frequency requirements of 10 CFR Part

50, Appendix J, and to state that NRC

approved exemptions to the applicable

regulatory requirements are permitted.

This proposed administrative revision

simply deletes the paraphrased language

and directly references Appendix J.

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 will provide a

one-time exemption from the 10 CFR

[Part] 50, Appendix J Section III.D.1.(a)

leak rate test schedule requirement. This

change will allow for a one-time test

interval for Type A Integrated Leak Rate

Tests (ILRTs) of approximately 70

months.

Leak rate testing is not an initiating

event in any accident, therefore this

proposed 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. The history

at IP–2 [Indian Point 2] demonstrates

that Type B and C Local Leak Rate Tests

(LLRTs) have consistently detected any

excessive local leakages.

Administrative controls govern the

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, an LLRT is performed to

ensure acceptable leakage levels,

following any LLRT 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 IP–2

has been in operation it has never failed

a Type A ILRT. Therefore, a one-time

exemption increasing the interval for

performing an ILRT should not result in

a significant decrease in the confidence

in the leak tightness of the containment

structure.

The proposed change also revises

Technical Specification 4.4.A.3 to

reference the testing frequency

requirements of 10 CFR [Part] 50,

Appendix J, and to state that NRC

approved exemptions to the applicable

regulatory requirements are permitted.

The current language of TS 4.4.A.3

paraphrases the requirements of Section

III.D.1.(a) 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.1.(a) will require a

submittal from Consolidated Edison

under 10 CFR 50.12 and subsequent

review and approval by the NRC prior

to implementation. The proposed

change is stated generically to avoid the

need for further TS changes if different

exemptions are approved in the future.

The proposed change, in itself, does

not affect reactor operations or accident

analysis and has no radiological

consequences. The change provides

clarification so that future Technical

Specifications changes will not be

necessary to correspond to applicable

NRC approved exemptions from the

requirements of Appendix J.

Therefore, the proposed change does

not involve a significant increase in the

probability or consequences of any

accident previously evaluated.

Criterion 2—Does Not Create the

Possibility of a New or Different Kind of

Accident from any Previously

Evaluated.

The proposed exemption request does

not affect normal plant operations or

configuration, nor does it affect leak rate

test methods. The proposed change

allows a one-time test interval of
Federal Register / Vol. 60, No. 31 / Wednesday, February 15, 1995 / Notices 8745

approximately 70 months for the ILRT.

Given the test history of IP–2 of 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 provides clarification to a

specification that paraphrases a codified

requirement.

Since the proposed change would not

change the design, configuration or

method of operation of the plant, it

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 ILRTs is to ensure that the release of

radioactive materials will be restricted

to those leak paths and leak rates

assumed in accident analyses. The

relaxed schedule for ILRTs does not

allow for relaxation of Type B and C

LLRTs. Therefore, methods for detecting

local containment leak paths and leak

rates are unaffected by this proposed

change. Given that the test history for

ILRTs shows no failure during plant life,

a one-time increase of the test interval

does not lead to a significant probability

of creating a new leakage path or

increased leakage rates, and the margin

of safety inherent in existing accident

analyses is maintained.

The proposed Technical Specification

change is administrative and clarifies

the relationship between the

requirements of TS 4.4.A.3, Appendix J

and any approved exemptions to

Appendix J. It does not, in itself, change

a safety limit, an LCO [limiting

condition for operation], or a

surveillance requirement on equipment

required to operate the plant. The NRC

will directly approve any proposed

change or exemption to [Section]

III.D.1.(a) of Appendix J prior to

implementation.

Therefore, this change does not

involve a significant reduction in the

margin of safety.

Based on the Safety Analysis, it is

concluded that: (1) The proposed

change does not constitute a significant

hazards consideration as defined by 10

CFR 50.92 and (2) there is reasonable

assurance that the health and safety of

the public will not be endangered by the

proposed change. Moreover, because

this action does not involve a significant

hazards consideration, it will also not

result in a condition which significantly

alters the impact of the station on the

environment as described in the NRC

Final Environmental Statement.

Although the licensee has included an

evaluation of a proposed exemption to

10 CFR part 50, Appendix J

requirements in the above

determination of no significant hazards

consideration, only the part related to

the amendment is pertinent to this

notice of proposed amendment. The

exemption request will be considered as

a separate matter on its own merits. 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: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10610.

Attorney for licensee: Brent L.

Brandenburg, Esq., 4 Irving Place, New

York, New York 10003.

NRC Project Director: Ledyard B.

Marsh

Duke Power Company, et al., Docket

Nos. 50–413 and 50–414, Catawba

Nuclear Station, Units 1 and 2, York

County, South Carolina

Date of amendment request: October

31, 1994

Description of amendment request:

The requested amendments would

remove the stroke times for the steam

generator power operated relief valves

(PORVs) from Technical Specification

(TS) Tables 3.6–2a and 3.6–2b. The

PORVs are part of the main steam vent

to atmosphere system. The PORV

actuators have difficulty developing

enough closing thrust to adequately

overcome all of the friction loads within

the valves; therefore, difficulty exists in

consistently meeting the present 5-

second closing stroke time requirement.

The licensee requests the proposed

change on the basis that the PORVs do

not receive an actual containment

isolation signal; therefore, it is justified

to remove the stroke times from TS

Tables 3.6–2a and 3.6–2b.

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:

In 48 FR 14870, the Commission has

set forth examples of amendments that

are considered not likely to involve

significant hazards considerations.

Example (vi) describes a change which

either may result in some increase to the

probability or consequences of a

previously-analyzed accident or may

reduce in some way a safety margin, but

where the results of the change are

clearly within all acceptable criteria

with respect to the system or component

specified in the Standard Review Plan.

In this case, the proposed amendment is

similar to example (vi) in that it

removes the required isolation time of

the steam generator PORVs from TS

Tables 3.6–2a and 3.6–2b; however, no

adverse impact upon accident analyses

is created as a result.

Criterion 1

The requested amendments will not

involve a significant increase in the

probability or consequences of an

accident previously evaluated. The

effects of the delays in isolation times

on the various transients affected have

been analyzed and found to be

acceptable. Since these valves do not

receive a containment isolation signal,

and no credit is taken for operation of

these valves in the dose analysis for a

containment isolation function, a

maximum stroke time does not apply for

containment isolation.

Criterion 2

The requested amendments will not

create the possibility of a new or

different kind of accident from any

accident previously evaluated. SV

PORV closure (provided the valves are

not already closed at the start of the

transient) is a response to a transient

already in progress. The possibility of a

spurious SV PORV opening will not be

affected by the requested amendments.

No equipment or component

reconfiguration will occur as a result of

this change. Finally, no changes to plant

procedures are being made which

would affect any accident causal

mechanisms.

Criterion 3

The requested amendments will not

involve a significant reduction in a

margin of safety. The isolation times

which are applicable to these valves are

specified in TS Table 3.3–5, Engineered

Safety Features Response Times. The

effects of the isolation of these valves

were evaluated based on their ESF

function, not a containment isolation

function, and determined to be

acceptable.

Based upon the preceding analyses,

Duke Power Company concludes that

the requested amendments 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 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the
8746 Federal Register / Vol. 60, No. 31 / Wednesday, February 15, 1995 / Notices

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: York County Library, 138 East

Black Street, Rock Hill, South Carolina

29730

Attorney for licensee: Mr. Albert Carr,

Duke Power Company, 422 South

Church Street, Charlotte, North Carolina

28242

NRC Project Director: Herbert N.

Berkow

Duke Power Company, Docket Nos. 50–

369 and 50–370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg

County, North Carolina

Date of amendment request: June 13,

1994, as supplemented August 15, 1994.

Description of amendment request:

The proposed changes would increase

the initial fuel enrichment limit from a

current maximum of 4.0 weight % to

4.75 weight % and establish new

loading patterns for new and irradiated

fuel in the spent fuel pool to

accommodate this increase. These

changes would also increase the

efficiency of fuel storage cell use in the

spent fuel pools and provide additional

flexibility to the reload design efforts at

Duke Power Company, while at the

same time maintaining sufficient

criticality safety margin and decay heat

removal capabilities.

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.

There is no increase in the probability

or consequences of an accident in the

new fuel vault since the only credible

accidents for this area are criticality

accidents and it has been shown that

calculated, worst case Keff for this area

is ≤0.95 under all conditions.

There is no increase in the probability

of a fuel drop accident in the Spent Fuel

Storage Pool since the mass of an

assembly will not be affected by the

increase in fuel enrichment. The

likelihood of other accidents, previously

evaluated and described in Section 9.1.2

of the FSAR [Final Safety Analysis

Report], is also not affected by the

proposed changes. In fact, it could be

postulated that since the increase in fuel

enrichment will allow for extended fuel

cycles, there will be a decrease in fuel

movement and the probability of an

accident may likewise be decreased.

There is also no increase in the

consequences of a fuel drop accident in

the Spent Fuel Pool since the fission

product inventory of individual fuel

assemblies will not change significantly

as a result of increased initial

enrichment. In addition, no change to

safety related systems is being made.

Therefore, the consequences of a fuel

rupture accident remain unchanged.

Also, it has been shown that keff is

≤0.95, under all conditions therefore,

the consequences of a criticality

accident remain unchanged as well.

2. The proposed changes do not create

the possibility of a new or different kind

of accident from any accident

previously evaluated.

The proposed changes do not create

the possibility of a new or different kind

of accident since fuel handling

accidents (fuel drop and misplacement)

are not new or different kinds of

accidents. Fuel handling accidents are

already discussed in the FSAR for fuel

with enrichments up to 4.1 weight %.

As described in Section VI.9 of

Attachment IV, additional analyses have

been performed for fuel with

enrichment up to 4.75 weight %. Worst

case misloading accidents associated

with the new loading patterns were

evaluated. For all possible misloading

accidents the negative reactivity

provided by soluble boron maintains keff

≤0.95. of safety.

3. The proposed changes do not

involve a significant reduction in the

margin of safety.

The proposed change does not

involve a significant reduction in the

margin of safety since, in all cases, a keff

≤0.95 is being maintained. Criticality

analyses have been performed which

show that the new fuel storage vault

will remain subcritical under a variety

of moderation conditions, from fully

flooded to optimum moderation. As

discussed above, the Spent Fuel Pool

will remain sufficiently subcritical

during any fuel misplacement accident.

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: Atkins Library, University of

North Carolina, Charlotte (UNCC

Station), North Carolina 28223.

Attorney for licensee: Mr. Albert Carr,

Duke Power Company, 422 South

Church Street, Charlotte, North Carolina

28242.

NRC Project Director: Herbert N.

Berkow.

Duke Power Company, Docket Nos. 50–

269, 50–270 and 50–287, Oconee

Nuclear Station, Units 1, 2 and 3,

Oconee County, South Carolina

Date of amendment request:

November 11, 1994, as supplemented

January 30, 1995.

Description of amendment request:

The amendments would revise the

Technical Specifications Design

Features section to establish restricted

loading patterns and associated burnup

criteria for placing fuel in the Oconee

Spent Fuel Pools. These changes are

necessary to address two new fuel

designs which have increased initial

fuel enrichment and therefore cannot be

stored in the spent fuel pools under

existing 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:

Standard 1. The proposed

amendments will not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

Each accident analysis addressed in

the Oconee Final Safety Analysis Report

(FSAR) has been examined with respect

to changes in Cycle 15 parameters to

determine the effect of the Cycle 16

reload and to ensure that the acceptance

criteria of the FSAR safety analyses

remain satisfied. The transient

evaluation of Cycle 16 is considered to

be bounded by previously accepted

analyses. Section 7 of the Reload Report

addresses ‘‘Accident and Transient

Analysis’’ for this core reload.

There is no increase in the probability

or consequences of an accident due to

the spent fuel storage restrictions

proposed in this amendment request. It

has been shown that the calculated,

worst case keff for this area is [less than

or equal to] 0.95 under all conditions.

There is no increase in the probability

of a fuel drop accident in the SFP [spent

fuel pool] since the mass of the new

assemblies is not significantly different

from the mass of the old assemblies. The

likelihood of other accidents, previously

evaluated and described in the FSAR, is

also not affected by the proposed

changes. In fact, it could be postulated

that since the increase in fuel

enrichment will allow for extended fuel

cycle lengths, there will be a decrease in

fuel movement and the probability of an

accident may actually be reduced. There

is also no increase in the consequences

of a fuel rod drop accident in the SFP

since the fission product inventory of
Federal Register / Vol. 60, No. 31 / Wednesday, February 15, 1995 / Notices 8747

individual fuel assemblies will not

change significantly as a result of

increasing the initial enrichment. In

addition, no change to safety related

systems is being made. Therefore, the

consequences of a fuel rupture accident

remain unchanged. In addition, it has

been shown that keff is [less than or

equal to] 0.95 under all conditions.

Therefore, the consequences of a

criticality accident in the SFP remain

unchanged as well. The above analysis

ensures that the proposed reload

amendment request 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.

The analyses performed in support of

this reload are in accordance with the

NRC approved methods delineated in

Specification 6.9.2. The predicted

operating characteristics of Oconee 3

Cycle 16 are similar to previously

licensed designs. The Mark B10T and

Mark B11 fuel assembly designs remain

mechanically compatible with all fuel

handling equipment. Therefore, no new

or different kind of fuel handling

accident is created by the proposed

amendment request.

Section 15.11 of the Oconee FSAR

states that the refueling boron

concentration is maintained such that a

criticality accident during refueling is

not considered credible. The proposed

amendment request continues to assure

that a criticality accident in the SFP or

during refueling is not credible. The

double contingency principle discussed

in ANSI N–16.1–1975 and the April

1978 NRC letter allows credit for soluble

boron under other abnormal or accident

conditions, since only a single accident

need be considered at one time. Thus,

by requiring a minimum boron

concentration in the SFP, a criticality

accident caused by violating the SFP

storage restrictions is not considered

credible. Therefore, the proposed

amendment request does 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 the

margin of safety.

The Oconee 3 Cycle 16 design was

performed using the NRC approved

methods given in Specification 6.9.2.

The safety limits for Oconee 3 Cycle 16

are unchanged from previous cycles.

The limits and margins summarized in

the Oconee 3 Cycle 16 Reload Report are

well within the allowable limits and

requirements, and reflect no reductions

to any margins of safety.

The proposed change does not

involve a significant reduction in the

margin of safety related to SFP

criticality. In all cases, a keff [less than

or equal to] 0.95 is maintained.

Criticality analyses have been

performed which show that the SFP will

remain sufficiently subcritical during

any fuel misplacement accident. In

summary the proposed changes do not

involve a significant reduction in the

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Oconee County Library, 501

West South Broad Street, Walhalla,

South Carolina 29691.

Attorney for licensee: J. Michael

McGarry, III, Winston and Strawn, 1200

17th Street, NW., Washington, DC

20036.

NRC Project Director: Herbert N.

Berkow.

Niagara Mohawk Power Corporation,

Docket No. 50–410, Nine Mile Point

Nuclear Station, Unit 2, Oswego County,

New York

Date of amendment request: January

6, 1995

Description of amendment request:

The proposed amendment would revise

Technical Specifications (TSs) 3/4.8.1.1,

‘‘AC Sources-Operating,’’ and 3/4.8.1.2,

‘‘AC Sources-Shutdown,’’ to (1) revise

the minimum quantity of fuel oil

required in the day tanks and the

storage tanks, (2) add specific actions to

be taken if the storage tank levels fall

below minimum requirements, (3)

revise and relocate to the associated

Bases the fuel oil sampling and testing

criteria, and (4) add specific actions to

be taken if the fuel oil properties do not

meet specified limits. The proposed

amendment would also revise TS 6.8.4,

‘‘Programs,’’ to add a requirement for a

diesel fuel oil testing program. The

licensee stated that the proposed

changes are consistent with the NRC’s

Improved Standard Technical

Specifications (NUREG–1434).

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The operation of Nine Mile Point Unit

2, in accordance with the proposed

amendment, will not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The diesel generators are not initiators

or precursors to an accident previously

evaluated. The diesel generators are

required to provide onsite power to safe

shutdown loads as assumed in the

accident analysis. Therefore, the

proposed changes to the diesel generator

fuel oil specifications cannot

significantly affect the probability of a

previously evaluated accident.

The proposed change to the minimum

required diesel generator fuel oil levels

is based on updated calculations of fuel

consumption rates. Because the updated

calculations assume a lower

consumption rate, the new minimum

fuel oil levels are lower but still assure

that a seven-day fuel oil capacity is

available. Accordingly, the proposed

change has no effect on the operation of

the diesel generator. The proposed

change to allow 48-hours to restore

diesel generator fuel oil to the minimum

required level does not affect short-term

diesel generator operability and is

acceptable based on the remaining fuel

oil capacity (>6 days), initiating the

process for procuring additional fuel

and the low probability of an event

requiring a diesel generator during this

interval. Also, the proposed allowance

of a limited time to restore diesel fuel

oil properties to required limits will not

affect the short-term operability of the

diesel generator. Even with minor

degradation of the fuel oil properties,

the diesels will start and perform their

intended function. Relocation of the

testing requirements to the bases and

adding a description of the Diesel Fuel

Oil Testing Program to the

Administrative Control section are

administrative changes. The diesel fuel

oil will continue to be sampled and

tested in a manner to assure its quality.

In summary, the changes will not

adversely affect the performance or the

ability of the diesel generators to

perform their intended function.

Therefore, the proposed changes will

not significantly increase the

consequences of an accident previously

evaluated.

The operation of Nine Mile Point Unit

2, in accordance with the proposed

amendment, will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes will revise the

minimum required diesel generator fuel

oil levels and requirements associated

with diesel generator fuel oil properties.
8748 Federal Register / Vol. 60, No. 31 / Wednesday, February 15, 1995 / Notices

The changes do not introduce any new

accident precursors and do not involve

any alterations to plant configurations

which could initiate a new or different

kind of accident. The proposed changes

do not affect the short-term operability

of the diesel generator. In addition, the

operability of the diesel generators is

assured by periodic testing and

preventive maintenance. Therefore, the

proposed changes will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The operation of Nine Mile Point Unit

2, in accordance with the proposed

amendment, will not involve a

significant reduction in a margin of

safety.

Safety margins are established

through safety analyses. These analyses

assume that at least one diesel generator

will start and load whenever offsite

power is lost. The proposed change to

the minimum required diesel generator

fuel oil levels is based on updated

calculations of fuel consumption rates.

The updated calculations use the

guidance delineated in Regulatory

Guide 1.137 which is based on timedependent loads of the diesel-generators

during design basis events. Calculations

based on time dependent loads result in

new minimum fuel oil levels which are

lower. This change has no effect on the

operation of the diesel generator or on

a margin of safety. The allowance of a

limited time to restore the fuel oil

levels, or to analyze and restore fuel oil

properties to required limits, is justified

since the short term operability of the

diesel generators is not affected.

Relocation of the fuel oil testing

requirements to the Bases does not

affect the quality of the fuel oil. The

10CFR50.59 process will assure that

future changes to the Bases will

maintain the current margins of safety,

and that the diesel fuel oil will continue

to be sampled and tested in such a

manner as to assure its quality. Adding

a description of the Diesel Fuel Oil

Testing Program to the Administrative

Control section of Technical

Specifications are administrative.

Therefore, the diesel generator will

continue to operate as analyzed and

there will not be a significant reduction

in a margin of safety.

The proposed changes are further

justified in that they are consistent with

the requirements of the Improved

Standard Technical Specifications

(NUREG–1434).

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston & Strawn,

1400 L Street, NW., Washington, DC

20005–3502.

NRC Project Director: Ledyard B.

Marsh.

Niagara Mohawk Power Corporation,

Docket No. 50–410, Nine Mile Point

Nuclear Station, Unit 2, Oswego County,

New York

Date of amendment request: January

6, 1995.

Description of amendment request:

The proposed amendment would revise

Technical Specifications (TSs) 3/4.3.7.5,

‘‘Accident Monitoring Instrumentation,’’

and TS 3/4.4.2, ‘‘Safety/Relief Valves.’’

TS 3/4.3.7.5 would be revised to delete

certain instruments not classified as

Category 1 (Type A or non-Type A) as

defined in Regulatory Guide 1.97 and to

delete the requirement that accident

monitoring instrumentation be operable

in Operational Condition 3. The

ACTIONS of TS Table 3.3.7.5–1 would

be revised to allow 30 days to restore

one inoperable channel and 7 days to

restore two inoperable channels. TS

3.3.7.5 would be revised to add an

exception to the requirements of TS

3.0.4. In addition, editorial changes

would be made to TS Tables 3.3.7.5–1

and 4.3.7.5–1 for consistency and

clarity.

The proposed amendment would also

revise TS 3/4.4.2 to remove

requirements related to safety/relief

valve acoustic monitors to be consistent

with the proposed changes to TS Tables

3.3.7.5–1 and 4.3.7.5–1.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The operation of NMP2 [Nine Mile

Point Nuclear Station Unit 2] in

accordance with the proposed

amendment, will not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

PAM [Post-Accident Monitoring]

instruments are used to help guide

operator response to postulated

accidents. Thus, the status or operability

of PAM instrumentation does not affect

the probability of previously analyzed

accidents. The non-Category 1 PAM

instruments being removed from the

Technical Specifications do not meet

any of the Commission’s screening

criteria and are not of controlling

importance to safety or necessary to

obviate the possibility of an abnormal

situation or event giving rise to an

immediate threat to public health and

safety. The operability of critical

parameters necessary to assure proper

response to previously analyzed

accidents (i.e., Category 1 instruments)

is still controlled by the Technical

Specifications. Thus, deleting nonCategory 1 instruments will not increase

the consequences of any accident

previously evaluated.

PAM instruments are related to the

diagnosis and preplanned actions

required to mitigate DBAs [Design Basis

Accidents] assumed to occur in

Operational Conditions 1 and 2. A DBA

during Operational Condition 3 is

extremely unlikely. The requirement to

maintain the Reactor Water Level,

Suppression Pool Water Level and

Drywell High Range Radiation Monitor

instrumentation operable in Operational

Condition 3 will be deleted. Because

Suppression Pool Water Level

indication will no longer be required in

Operational Condition 3, its ACTION

requirement was revised to delete the

requirement to place the plant in COLD

SHUTDOWN, Operational Condition 4.

This is consistent with ITS [Improved

Standard Technical Specifications]

which requires that the plant be brought

to an operational condition in which the

LCO [Limiting Condition for Operation]

does not apply if a required action

cannot be met. Therefore, deleting the

requirement that PAM instruments be

operable during Operational Condition

3 and changing the ACTION

requirement for Suppression Pool Water

Level Monitoring does not affect the

probability or consequences of an

accident.

The passive nature of the Category 1

PAM instruments (i.e., those

instruments that initiate no critical

automatic action) and the alternate

means available to obtain the required

information assure an acceptable level

of safety is maintained during operation

with instrument channels out of service.

Since an acceptable level of safety is

maintained with inoperable channels,

plant startup or operation with

inoperable channels will not alter plant

response to analyzed accidents. Thus,

the proposed changes to the required

ACTIONS and the proposed exemption

to Specification 3.0.4 will not increase

the consequences of analyzed events.

The proposed changes to the

requirements for PCIV [Primary
Federal Register / Vol. 60, No. 31 / Wednesday, February 15, 1995 / Notices 8749

Containment Isolation Valve] indication

are consistent with the proposed

required ACTIONS. Position indication

will still be required for each operable

PCIV and penetrations without adequate

PCIV indication status will be isolated,

thus assuring containment integrity in

the event of an accident. Deletion of the

‘‘Minimum Required Actions’’ column

in Table 3.3.7.5–1 is consistent with the

proposed ACTIONS for LCO 3.3.7.5,

since compensatory actions are based on

compliance with the ‘‘Required Number

of Channels.’’ Deleting the ‘‘Applicable

Operating Conditions’’ column is

consistent with the proposed changes

and other NMP2 Technical

Specifications sections. Finally,

referencing Specification 4.0.5 is an

administrative change which does not

alter any existing surveillance

requirements for the safety relief valves.

In aggregate, the proposed changes do

not affect the plant in a way that could

directly contribute to causing or

mitigating the effects of an accident.

Therefore, the operation of NMP2, in

accordance with the proposed

amendment, will not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The operation of NMP2, in

accordance with the proposed

amendment, will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes do not

represent a physical change to the plant

as described in the NMP2 USAR

[Updated Safety Analysis Report]. The

proposed changes do not modify any

plant equipment and the initial

conditions used for the design basis

accident analysis are still valid. Thus,

no potential initiating events are created

which would cause any new or different

kinds of accidents. PAM

instrumentation is used to guide

operator response during postulated

accidents. Those PAM instruments

considered of controlling importance to

safety are retained in the Technical

Specifications. Thus, plant response to

previously analyzed events is not

altered so as to create any new or

different kinds of accidents. Therefore,

operation of Nine Mile Point Unit 2 in

accordance with the proposed change

will not create the possibility of a new

or different kind of accident from any

previously assessed.

The operation of NMP2, in

accordance with the proposed

amendment, will not involve a

significant reduction in a margin of

safety.

The non-Category 1 PAM instruments

being removed from the Technical

Specifications do not meet any of the

Commission’s screening criteria. That is,

the instruments being proposed for

removal are not of controlling

importance to safety or necessary to

obviate the possibility of an abnormal

situation or event giving rise to an

immediate threat to public health and

safety. Thus, they are not critical to any

margin of safety.

PAM instruments are related to the

diagnosis and preplanned actions

required to mitigate DBAs assumed to

occur in Operational Conditions 1 and

2. A DBA during Operational Condition

3 is extremely unlikely. The

requirement to maintain the Reactor

Water Level, Suppression Pool Water

Level and Drywell High Range

Radiation Monitor instrumentation

operable in Operational Condition 3

will be deleted. Because Suppression

Pool Water Level indication will no

longer be required in Operational

Condition 3, its ACTION requirement

was revised to delete the requirement to

place the plant in COLD SHUTDOWN,

Operational Condition 4. This is

consistent with the ITS, which requires

that the plant be brought to an

operational condition in which the LCO

does not apply if a required action

cannot be met. Therefore, deleting the

requirement that PAM instruments be

operable during Operational Condition

3 and changing the ACTION

requirement for Suppression Pool Water

Level Monitoring does not significantly

reduce a margin of safety.

Since the Category 1 PAM

instruments are passive in nature (i.e.,

no critical automatic action is assumed

to occur from these instruments) and

alternate means exist to obtain the

required information, an acceptable

level of safety is assured when

instrument channels are out of service.

Also, the probability of an event

requiring PAM instrumentation is low.

Continued operation with one channel

out of service, and limited plant

operation with two channels out of

service, does not compromise plant

safety margins. An acceptable level of

safety is maintained during plant

startups and operation with instrument

channels out of service. Thus, the

proposed changes to the required

ACTIONS and the proposed exemption

to Specification 3.0.4 will not

significantly reduce a margin of safety.

The proposed changes to PCIV

indication will assure correct

implementation of the ACTIONS

discussed above. Isolating the flow path

associated with one or two inoperable

PCIV indication channels is

conservative since the subject valve will

be positioned as required to assure

primary containment integrity. The

remaining editorial changes are

administrative in nature and by

definition do not affect safety margins.

Deleting the ‘‘Minimum Operable

Channels’’ and ‘‘Applicable Operating

Conditions’’ columns is consistent with

the proposed changes. Finally,

referencing the requirements of

Specification 4.0.5 is an administrative

change and by definition does not

reduce the margin of safety.

Therefore, the operation of NMP2 in

accordance with the proposed change

will not involve a significant reduction

in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston & Strawn,

1400 L Street, NW., Washington, DC

20005–3502.

NRC Project Director: Ledyard B.

Marsh.

Northeast Nuclear Energy Company, et

al., Docket No. 50–423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of amendment request:

December 9, 1994.

Description of amendment request:

The proposed changes incorporate NRC

recommendations contained in Generic

Letter 93–05 related to the diesel

generator (DG) surveillance

requirements and other DG surveillance

requirements related to the cold starts.

The proposed changes to the DG

operability testing surveillance

requirements are consistent with the

intent of GL 93–05 however vary in

some particulars, because of

circumstances specific to Millstone 3.

The proposed changes will modify the

requirement for the DG operability

testing when the other DG is inoperable,

delete the requirement for DG

operability testing when one or both

offsite AC sources are inoperable,

eliminate fast loading of DGs except for

the 18-month test, and modify the hot

restart test from the 24-hour loaded test

run for the DGs.

Basis for proposed no significant

hazards consideration determination:
8750 Federal Register / Vol. 60, No. 31 / Wednesday, February 15, 1995 / Notices

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration (SHC), which is presented

below:

* * * The proposed changes do not

involve a SHC because the changes

would not:

1. Involve a significant increase in the

probability or consequences of an

accident previously analyzed.

The proposed changes revise the

action requirements regarding

operability testing of a non-affected DG

when the other DG is inoperable, delete

the requirement for operability testing of

the DGs when one or both offsite AC

sources are inoperable and eliminate the

fast loading of DGs except for the 18-

month test. These changes will improve

DGs performance by reducing the

number of unnecessary quick starts and

by requiring more appropriate testing of

the DGs when there is a potential for

common mode failure. The proposed

change, to revise the method of

verifying DG hot restart capability after

a 24-hour run without loading the DG

with LOP/SI [loss of offsite power/safety

injection] load, meets an intent of

Regulatory Guide 1.108, Position

C.2.a.5, which states the purpose of the

test as to ‘‘demonstrate functional

capability at full load temperature

conditions.’’ Functional capability of

the DG can be adequately demonstrated

by manually or automatically restarting

the DG within five minutes after a 24-

hour test run without loading it with

LOP/SI loads, provided that a full load

temperature condition is maintained

prior to restart. The proposed DG restart

method does not reduce the

effectiveness of the test. The proposed

revisions of the DG surveillance

requirements will not increase the

probability of an accident and it will not

change the response of the DG to a LOP

as described in the Millstone Unit No.

3 FSAR. Since the plant response to an

accident will not change, there is no

change in the potential for an increase

in the consequences of an accident

previously analyzed.

2. Create the possibility of a new or

different kind of an accident previously

evaluated.

The proposed changes of the DG

surveillance requirements and

operability testing requirements do not

affect the operation or response of any

plant equipment or introduce any new

failure mechanisms. The proposed

changes do not affect the test results and

the DGs will be verified to be operable

and their response to a loss of voltage

will be unchanged. The plant

equipment will respond per the design

and analyses and there will not be a

malfunction of a new or any type

introduced by the revision to the DG

surveillance requirements. As such, the

changes do not create the possibility of

a new or different kind of accident

previously evaluated.

3. Involve a significant reduction in

the margin of safety.

The bases of Technical Specification

3/4.8, ‘‘Electrical Power Systems,’’ state

that the operability of the AC and DC

power systems and associated

distribution systems ensure that

sufficient power will be available to

supply the safety-related equipment

required for safe shut down and

mitigation and control of accident

conditions. The bases also state that the

surveillance requirements for

determining the operability of the DGs

are in accordance with the

recommendations of Regulatory Guide

1.108, Revision 1. The revisions of the

surveillance requirements establishes

tests that will continue to verify that the

DGs are operable and the testing will

still meet the intent of Regulatory Guide

1.108, Revision 1. Operable DGs ensure

that the assumptions in the bases of the

Technical Specifications are not affected

and ensure that the margin of safety is

not reduced. Therefore, the assumptions

in the bases of the technical

specifications are not affected and these

changes do not result in 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.

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.

Northeast Nuclear Energy Company, et

al., Docket No. 50–423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of amendment request:

December 14, 1994.

Description of amendment request:

The proposed amendment would revise

the Millstone Unit No. 3 Technical

Specifications by:

1. Increasing the upper bound of the

overall containment integrated leakage

rate required by Technical Specification

3.6.1.2.a from 0.3 wt. % per day to 0.65

wt. % per day of the containment air per

24 hours at design basis pressure.

2. Revising Technical Specification

4.6.6.1.d.3 by providing more margin

with respect to the drawdown time for

secondary containment vacuum.

3. Revising Bases Section 3/4.7.9 to

reflect the above changes.

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 (SHC), which is presented

below:

* * * 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 evaluated.

* * * There is a reasonable assurance

that the modified criteria for the

negative pressure in the secondary

containment boundary proposed via the

proposed change (i.e., a negative

pressure of 0.1 inches in one minute

and a negative pressure of 0.4 inches

within the next two minutes), can be

accomplished in the prescribed time.

Extension of the time allowed to

achieve the final drawdown of

secondary containment from 120

seconds to 180 seconds (these times

include the diesel generator start and

load time of approximately 11 seconds)

will have a negligible impact on heating

and cooling. Plant experience has

shown that heatup and cooldown of

thick-walled concrete structures, such

as the Millstone Unit No. 3 auxiliary

building, is a relatively slow process.

Also, natural convection within the

auxiliary building tends to stabilize

temperatures. Following an accident

signal, ventilation equipment is

restarted promptly. Therefore, heatup or

cooldown, during short periods while

ventilation fans and/or heaters are

inactive, is insignificant and can be

neglected.

The proposed change to reinstate the

containment integrated leakage rate at

the design basis pressure from 0.3 wt %

per day to 0.65 wt % per day has been

evaluated to determine the impact to the

Appendix J requirements for Type A, B

and C Testing. In addition, the

radiological consequence evaluation

also addressed the increase in La (i.e.,

from 0.3 wt % per day to 0.65 wt % per

day).

On October 12, 1993, Millstone Unit

No. 3 successfully conducted the second
Federal Register / Vol. 60, No. 31 / Wednesday, February 15, 1995 / Notices 8751

Type A test in the first 10-year service

period. Test results indicated that the

‘‘As-Found’’ and ‘‘As-Left’’ ILRTs

[integrated leakage rate tests] passed the

technical specification acceptance

criteria. The ‘‘As-Found’’ value was

0.1327 weight percent per day and the

‘‘As-Left’’ value was 0.1313 weight

percent per day. These values represent

27.2% and 26.9% of the technical

specification criterion of 0.4875 wt %

per day (0.75 La), based on La equal to

0.65 wt % per day, respectively. In

addition, as of October 9, 1993, the total

Type B and C ‘‘As-Found’’ and ‘‘AsLeft’’ leakage results were 0.099 wt %

per day, and 0.084 wt % per day,

respectively. These values represent

approximately 25.3% and 21.5% of the

technical specification limit of 0.39%

wt % per day (0.6 La), based on La equal

of 0.65 wt % per day, respectively.

Correspondingly, the 1993 Type A, B,

and C test results indicate that the ‘‘AsFound’’ and ‘‘As-Left’’ result in each test

case was below the existing Technical

Specification limit of 0.3 wt % per day.

This further demonstrates the overall

leakage integrity of the containment and

its boundaries.

Based on the relatively low ‘‘As-Left’’

ILRT leakage rate (i.e., 0.1313 wt % per

day is well below the existing technical

specification limit of 0.225 wt % per

day (0.75 La), based on La equal to 0.3

wt % per day), which represents the

overall containment integrated leakage

rate for the containment prior to startup, there is reasonable assurance that

containment integrity will be

maintained below the allowable leakage

rate limit of 0.65 wt % per day. In

addition, the total Type B and C ‘‘AsLeft’’ leakage result of 0.084 wt % per

day (this is well below the existing

technical specification limit of 0.18 wt

% per day (0.6 La), based on La equal to

0.3 wt % per day), provides further

assurance that leakage, based on

individual penetration, will be

maintained within sufficient margin of

the leakage limits.

Because the last Type A, B, and C

tests were performed under the

technical specification limit of 0.65 wt

% per day, the proposed change to

restore La to 0.65 wt % per day has no

impact to these systems from a leakage

allowance perspective. As indicated

above, the previous test results met the

technical specification leakage limits

(based on 0.65 wt % per day) within

sufficient margin and, therefore, would

not present any challenge to these

leakage limits.

NNECO has evaluated the proposed

changes to Surveillance Requirement

4.6.6.1.d.3 that increase the time to

draw a final required negative pressure

as measured at the 24’-6’’ elevation of

the auxiliary building in conjunction

with the proposed change to reinstate

the containment integrated leakage rate

of 0.65 wt % per day to determine the

impact on the offsite doses following a

LOCA. The calculated radiological

doses are, in most cases, less than the

previously calculated doses (i.e., EAB

[exclusion area boundary] and LPZ

[low-population zone] doses) and are

within the 10CFR100 limits. Previously,

the EAB thyroid and whole body doses

as documented in the November 4,

1993, submittal were calculated to be

141 REM and 9.4 REM respectively,

while the previously docketed (i.e., the

November 4, 1993, submittal) LPZ doses

to the thyroid and whole body were

calculated to be 29.8 REM and 1.7 REM

respectively. Utilizing the revised

application of containment recirculation

spray DF, the EAB thyroid and whole

body doses were calculated to be 61

REM and 16.7 REM, respectively, and

the LPZ thyroid and whole body doses

were calculated to be 10.9 REM and 2.8

REM respectively. The assumptions

used in the above radiological dose

calculations are provided in Attachment

1. It is noted that a LOCA at Millstone

Unit No. 3 is also one of the bounding

accidents for the Millstone Unit No. 3

control room, Millstone Unit No. 2

control room, and the Millstone

Technical Support Center habitability

analysis. Therefore, the doses for these

areas were recalculated and are

presented in the Safety Assessment

section above. The Millstone Unit No. 1

control room and the Emergency

Operating Facility doses are bounded by

the Millstone Unit No. 1 LOCA

calculations.

The Millstone Unit Nos. 2 and 3

control rooms and Millstone Technical

Support Center doses were not

recalculated in 1993 (i.e., November 4,

1993, submittal) since EAB/LPZ doses

proved that the releases were less than

the 1990 submittal. In summary, all

control room and Technical Support

Center doses are within the guidelines

of GDC 19. Therefore, the proposed

changes do not result in an increase in

consequences of an accident (i.e., a

LOCA) previously analyzed.

The proposed changes to Bases

Section 3/4.6.6 do not have any safety

impact since they only reflect the

changes proposed to Surveillance

Requirement 4.6.6.1.d.3.

2. Create the possibility of a new or

different kind of accident from any

accident previously evaluated.

The proposed changes do not

compromise the ability of the SLCRS

[supplementary leak collection and

release system] and ABFS [auxiliary

building filter system] to mitigate the

consequences of an accident. The

proposed changes do not make any

physical or operational changes to

existing plant structures, systems or

components. The proposed changes do

not introduce any new or unique

operational modes or accident

precursors. Therefore, the proposed

changes do not create the possibility of

a new or different kind of accident from

any accident previously evaluated.

3. Involve a significant reduction in a

margin of safety.

NNECO has evaluated the proposed

changes to Surveillance Requirement

4.6.6.1.d.3 that increase the time to

draw a final required negative pressure

as measured at the 24’–6’’ elevation of

the auxiliary building in conjunction

with the proposed change to reinstate

the containment integrated leakage rate

of 0.65 wt % per day to determine the

impact on the offsite doses following a

LOCA. The calculated radiological

doses are, in most cases, less than the

previously calculated doses and these

doses are within the 10CFR100 limits.

All control rooms and technical support

center doses are within the guidelines of

GDC 19. Therefore, the proposed

changes do not involve a significant

reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Learning Resource Center,

Three Rivers Community-Technical

College, Thames Valley Campus, 574

New London Turnpike, Norwich, CT

06360.

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.

Northeast Nuclear Energy Company, et

al., Docket No. 50–423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of amendment request:

December 23, 1994.

Description of amendment request:

The proposed amendment would

change the acceptance criteria for the

peak transient generator voltage from

4784 volts to 5000 volts during full load

rejection tests of the diesel generator

(DG), and delete the 10-year

surveillance requirement to perform a
8752 Federal Register / Vol. 60, No. 31 / Wednesday, February 15, 1995 / Notices

110% pressure test of the DG fuel oil

system.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration (SHC), which is presented

below:

* * * The proposed changes do not

involve a SHC because the changes

would not:

1. Involve a significant increase in the

probability or consequences of an

accident previously analyzed.

DG Full-Load Rejection Test

NNECO is proposing to modify

Surveillance Requirement 4.8.1.1.2.g.3

of the Millstone Unit No. 3 Technical

Specifications by changing the

acceptable transient voltage to 5000

volts from 4784 volts. This change will

permit the DG full load rejection tests to

be performed at realistic plant

conditions using a power factor that will

envelope the calculated power factor

during the worst kW loading conditions.

The transient voltage of 5000 volts is

within the normal design limits of the

DGs.

The proposed change does not alter

the intent of the surveillance, does not

involve any physical changes to the

plant, does not alter the way any

structure, system, or component

functions, and does not modify the

manner in which the plant is operated.

As such, the proposed change to

Surveillance Requirement 4.8.1.1.2.g.3

will not degrade the capability of the

DGs to perform their intended safety

function, and will not reduce the

availability of the DGs. Actually, the

proposed change will increase the

effectiveness of the full load rejection

tests, because the DGs will be tested in

a configuration that is closer to the

design basis conditions.

Pressure Test of the DG Fuel Oil System

The DG fuel oil system is classified as

an ASME Code Class 3 system in

accordance with the guidance of

Regulatory Guide 1.26, ‘‘Quality Group

Classification and Standards for

Water-, Steam-, and Radioactive-waste

Components of Nuclear Power Plants.’’

Surveillance Requirement 4.0.5 requires

the testing of ASME Class 1, 2, and 3

components in accordance with Section

XI of the ASME Code. Surveillance

Requirement 4.8.1.1.2.i.2 is redundant

to the ASME Section XI pressure test

requirements of Surveillance

Requirement 4.0.5. Additionally, the DG

fuel oil tank cannot be tested in the

configuration required by Surveillance

Requirement 4.8.1.1.2.i.2, because the

tanks are vented to the atmosphere and

the vent cannot be isolated. Therefore,

NNECO is proposing to delete

Surveillance Requirement 4.8.1.1.2.i.2.

The proposed change does not modify

the manner in which the DGs respond

to an accident. Also, the proposed

change does not reduce the reliability of

the DGs.

Conclusion

Based on the above, the proposed

changes to Surveillance Requirements

4.8.1.1.2.g.3 and 4.8.1.1.2.i.2 of the

Millstone Unit No. 3 Technical

Specifications do not involve a

significant increase in the probability or

consequences of an accident previously

analyzed.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

DG Full-Load Rejection Test

The DGs are required to operate in

response to a loss of offsite power. Their

failure cannot initiate an accident.

Additionally, the proposed change to

Surveillance Requirement 4.8.1.1.2.g.3

does not affect the operation or response

of any plant structure, system, or

component, and it does not introduce

any new failure mechanisms.

Pressure Test of the DG Fuel Oil System

The proposed change to Surveillance

Requirement 4.8.1.1.2.i.2 does not affect

the design or function of the DG fuel oil

system. Failure of the DG fuel oil system

would not initiate an accident.

Conclusion

Based on the above, the proposed

changes to Surveillance Requirements

4.8.1.1.2.g.3 and 4.8.1.1.2.i.2 of the

Millstone Unit No. 3 Technical

Specifications will not create the

possibility of a new or different kind of

accident from any previously evaluated.

3. Involve a significant reduction in

the margin of safety.

DG Full-Load Rejection Test

NNECO is proposing to modify

Surveillance Requirement 4.8.1.1.2.g.3

of the Millstone Unit No. 3 Technical

Specifications by changing the

acceptable transient voltage to 5000

volts from 4784 volts. The intent of the

proposal is to permit the DG full load

rejection tests to be conducted at

conditions which simulate design basis

conditions.

The proposed change does not alter

the intent of the surveillance, does not

involve any physical changes to the

plant, does not alter the way any

structure, system, or component

functions, and does not modify the

manner in which the plant is operated.

As such, the proposed change to

Surveillance Requirement 4.8.1.1.2.g.3

will not degrade the ability of the DGs

to perform their intended safety

function, and will not reduce the

availability of the DGs.

The bases of Technical Specification

3/4.8, ‘‘Electrical Power Systems,’’ state

that the operability of the AC and DC

power systems and associated

distribution systems ensure that

sufficient power will be available to

supply the safety related equipment

required for safe shutdown and for the

mitigation of transients. The proposed

change to the surveillance requirement

will increase the effectiveness of the full

load rejection tests.

This will ensure the operability of the

DGs. Operable DGs ensure that the

assumptions for the bases of the

Millstone Unit No. 3 Technical

Specifications are not affected.

Pressure Test of the DG Fuel Oil System

NNECO is proposing to delete

Surveillance Requirement 4.8.1.1.2.i.2

from the Millstone Unit No. 3 Technical

Specifications. This surveillance

requirement is redundant to the

requirements of Surveillance

Requirement 4.0.5 which invokes ASME

Section XI. Additionally, the fuel oil

system cannot be tested to the

requirements of Surveillance

Requirement 4.8.1.1.2.i.2 because the

DG fuel oil tanks are vented to the

atmosphere and this vent path cannot be

isolated.

Millstone Unit No. 3 will include the

DG fuel oil system pressure test as an

augmented inspection within the

Inservice Inspection program.

Inspections will be performed in

compliance with the requirement of the

1983 Edition of ASME Section XI, Table

IWD–2500–1, ‘‘Test and Examination

Categories.’’ Testing (i.e., a system

hydrostatic test) in accordance with

ASME Section XI will provide

equivalent assurance of tank and piping

integrity.

Conclusion

Based on the above, the proposed

changes to Surveillance Requirements

4.8.1.1.2.g.3 and 4.8.1.1.2.i.2 of the

Millstone Unit No. 3 Technical

Specifications do not involve a

significant reduction in the margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the
Federal Register / Vol. 60, No. 31 / Wednesday, February 15, 1995 / Notices 8753

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.

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.

Northeast Nuclear Energy Company, et

al., Docket No. 50–423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of amendment request: January

18, 1995.

Description of amendment request:

The proposed changes to the technical

specifications will increase the

minimum required boron concentration

in the boric acid tank (BAT) from 6300

ppm to 6600 ppm.

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 (SHC), which is presented

below:

* * * 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 evaluated.

The change affects the minimum

required boron concentration in the

BAT. Changes in the tank’s boron

concentration will not affect the

probability of any plant accident.

An increase in the minimum BAT

concentration of 6600 ppm was

recommended by Westinghouse based

on their Cycle 6 BORDER evaluation.

The BORDER evaluation conservatively

determines the ability to maintain

shutdown margin when the plant is

taken from an initial operating

condition of Mode 1 or 2 to a final

condition of Mode 5 or 6 using an

assumed minimum BAT concentration.

Therefore, the ability to maintain

shutdown margin is assured and the

change will not adversely affect the

consequences of any plant accident.

2. Create the possibility of a new or

different kind of accident from any

Previously Analyzed.

The change conservatively increases

the minimum required boron

concentration in the BAT from 6300

ppm to 6600 ppm. There is no impact

on the operability of plant systems or

equipment. Therefore, the change does

not create a malfunction that is different

from those previously evaluated.

3. Involve a significant reduction in

the margin of safety.

The proposed increase in the

minimum boron concentration in the

BAT provides conservatism in the

calculated shutdown margin for

Millstone Unit No. 3. The change does

not adversely affect any equipment

credited in the safety analysis. Also, the

change does not adversely affect the

probability or consequences of any plant

accident, including the calculated PCT

[peak clad temperature] or offsite doses.

Therefore, there is no impact on the

margin of safety as specified in the

Technical Specifications.

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.

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.

Northern States Power Company,

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

Island Nuclear Generating Plant, Unit

Nos. 1 and 2, Goodhue County,

Minnesota

Date of amendment requests: January

10, 1995.

Description of amendment requests:

The proposed amendments would

revise the Prairie Island Event

Monitoring Instrumentation Technical

Specifications and associated Bases to

conform to Standard Technical

Specifications for post-accident

monitoring.

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[s] will

not involve a significant increase in the

probability or consequences of an

accident previously evaluated.

The primary purpose of post accident

monitoring instrumentation is to display

plant variables that provide information

to the control room operators during

accident situations. Plant

instrumentation was evaluated for

importance for this function when

Regulatory Guide 1.97

[’’Instrumentation for Light Water

Cooled Nuclear Power Plants to Assess

Plant Conditions During and Following

an Accident’’] classifications were

determined. The Prairie Island

Regulatory Guide 1.97 classification of

instruments was previously approved

by the NRC on October 18, 1985. This

amendment request proposes to base

Prairie Island Technical Specifications

on the results of the Regulatory Guide

1.97 evaluation in accordance with the

guidance of the industry standard.

Revising the allowed outage time for

these instruments will not significantly

increase the probability or consequences

of an accident since these instruments

do not initiate automatic actions, there

are available backup indications and the

probability of an event requiring these

instruments to be operable is very low.

Therefore, the probability or

consequences of an accident previously

evaluated are not affected by any of the

proposed amendments.

2. The proposed amendment[s] will

not create the possibility of a new or

different kind of accident from any

accident previously analyzed.

The license amendment request

proposes to add instruments to the

Technical Specifications which have

been previously determined to be

important for post accident monitoring,

and to remove instruments from

Technical Specifications which have

been previously determined to be less

important for post accident monitoring.

This amendment ensures the control

room operators are provided with the

instrumentation required to properly

manage an accident situation.

Therefore, based on the above

considerations, the possibility of a new

or different kind of accident from any

accident previously evaluated would

not be created.

3. The proposed amendment[s] will

not involve a significant reduction in

the margin of safety.

The post accident monitoring

functions do not initiate any automatic

actions. The instrumentation to be

added to the Event Monitoring

Instrumentation Table was previously

recognized through the Regulatory

Guide 1.97 evaluation process as

important for post accident monitoring

and would be relied upon if there were

an event without this license

amendment. Instrumentation to be

removed from Technical Specifications

was previously recognized to be less
8754 Federal Register / Vol. 60, No. 31 / Wednesday, February 15, 1995 / Notices

important and would not be relied upon

very much in an event. Overall, with the

trade-off of adding and deleting

instrumentation, the margin of safety

will not be significantly affected.

The proposed license amendment will

increase the allowed outage time for

most of the instruments. Again, these

instruments do not provide automatic

actions, they provide indications for

monitoring post accident conditions. All

of the instruments have backup or

corroborating indications which could

be relied upon if the Technical

Specifications instruments were

inoperable. Also, an event requiring use

of these instruments has a very low

probability. For these reasons the

proposed changes in allowed outage

time will not result in a significant

reduction in the margin of safety.

For these same reasons, the proposed

changes in radiation instrument

surveillance requirements will not

significantly reduce the margin of

safety.

Overall, a significant reduction in the

margin of safety would not result from

this license 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 requests involve no

significant hazards consideration.

Local Public Document Room

location: Minneapolis Public Library,

Technology and Science Department,

300 Nicollet Mall, Minneapolis,

Minnesota 55401.

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts, and Trowbridge,

2300 N Street, NW, Washington, DC

20037.

NRC Project Director: John N.

Hannon.

North Atlantic Energy Service

Corporation, Docket No. 50–443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request: January

25, 1995.

Description of amendment request:

The proposed Technical Specification

change would replace a specific

requirement for the frequency of Type A

tests with a general requirement to

perform Type A tests. The proposed

amendment would change Surveillance

Requirement 4.6.1.2.a. Specifically, the

change would require the performance

of Type A tests (overall containment

integrated leak rate tests (ILRTs)) at

intervals as specified in 10 CFR 50,

Appendix J, instead of on a specific

schedule for performance of ILRTs of

‘‘40 plus or minus 10 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. The NRC staff has

reviewed the licensee’s analysis against

the standards of 10 CFR 50.92(c). The

NRC staff’s review is presented below.

A. The change does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated (10 CFR 50.92(c)(1)) because

the proposed change merely replaces a

prescriptive schedule for performing

ILRTs with a requirement to conduct the

ILRTs on a schedule consistent with the

Commission’s regulations. The change

does not alter the methodology,

frequency, or acceptance criteria for

ILRTs, does not affect the design basis

of the containment, and does not change

the post-accident response of the

containment.

B. The change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated (10 CFR 50.92(c)(2)) because

the change does not affect the manner

by which the facility is operated and

does not make any changes to existing

plant structures, systems, or

components. The proposed change

merely replaces a prescriptive schedule

for performing ILRTs with a

requirement to conduct the ILRTs on a

schedule consistent with the

Commission’s regulations.

C. The change does not involve a

significant reduction in a margin of

safety (10 CFR 50.92(c)(3)) because the

proposed change does not affect the

manner by which the facility is operated

or involve changes to equipment or

features which affect the operational

characteristics of the facility.

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: Exeter Public Library, 47 Front

Street, Exeter, NH 03833.

Attorney for licensee: Thomas Dignan,

Esquire, Ropes & Gray, One

International Place, Boston, MA 02110–

2624.

NRC Project Director: Phillip F.

McKee.

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:

December 7, 1994.

Description of amendments request:

The amendments would provide a

permanent voltage-based steam

generator tube repair criteria for both

units. This criteria is based on the

guidance contained in the NRC

Proposed Generic Communication

(Generic Letter 94–XX), ‘‘Voltage-Based

Repair Criteria for the Repair of

Westinghouse Steam Generator Tubes

Affected by Outside Diameter Stress

Corrosion Cracking,’’ that was issued for

public comment in the Federal Register

(59 FR 41520) on August 12, 1994. The

licensee’s submittal also includes

responses to and identifies exceptions

taken to the draft Generic Letter. The

significant exceptions are: (1) The

requirement to reinspect all tubes if

bobbin probe wear exceeds 15%; (2) the

1×10¥2 limit on the calculated

conditional burst probability; and (3)

the need to pull additional steam

generator tubes to evaluate the current

condition of the steam generator tubes.

In addition, the operational leakage

requirement for Unit 2 will be modified

to reduce the total allowable primary-tosecondary leakage for any steam

generator from 500 gallons per day (gpd)

to 150 gallons per day.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) Operation of Farley units in

accordance with the proposed license

amendment does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

Testing of model boiler specimens for

free standing tubes at room temperature

conditions shows burst pressures as

high as approximately 5000 psi for

indications of outer diameter stress

corrosion cracking with voltage

measurements as high as 26.5 volts.

Burst testing performed on pulled tubes

with up to 7.5 volt indications show

burst pressures in excess of 5900 psi at

room temperature. As stated earlier,

tube burst criteria are inherently

satisfied during normal operating

conditions by the presence of the tube

support plate. Furthermore, correcting

for the effects of temperature on

material properties and minimum

strength levels (as the burst testing was
Federal Register / Vol. 60, No. 31 / Wednesday, February 15, 1995 / Notices 8755

done at room temperature), tube burst

capability significantly exceeds the R.G.

[Regulatory Guide] 1.121 criterion

requiring the maintenance of a margin

of 1.43 times the steam line break

pressure differential on tube burst if

through-wall cracks are present without

regard to the presence of the tube

support plate. Considering the existing

data base, this criterion is satisfied with

bobbin coil indications with signal

amplitudes over twice the 2.0 volt

voltage-based repair criteria, regardless

of the indicated depth measurement.

This structural limit is based on a lower

95% confidence level limit of the data.

The 2.0 volt criterion provides an

extremely conservative margin of safety

to the structural limit considering

expected growth rates of outside

diameter stress corrosion cracking at

Farley. Alternate crack morphologies

can correspond to a voltage so that a

unique crack length is not defined by a

burst pressure to voltage correlation.

However, relative to expected leakage

during normal operating conditions, no

field leakage has been reported from

tubes with indications with a voltage

level of under 7.7 volts for 3/4 inch tube

which correlates to 10 volts for 7/8 inch

tubing (as compared to the 2.0 volt

proposed voltage-based tube repair

limit). Thus, the proposed amendment

does not involve a significant increase

in the probability or consequences of an

accident.

Relative to the expected leakage

during accidents (sic) condition

loadings, the accidents that are affected

by primary-to-secondary leakage and

steam release to the environment are

Loss of External Electrical Load and/or

Turbine Trip, Loss of All AC Power to

Station Auxiliaries, Major Secondary

System Pipe Failure, Steam Generator

Tube Rupture, Reactor Coolant Pump

Locked Rotor, and Rupture of a Control

Rod Drive Mechanism Housing. Of

these, the Major Secondary System Pipe

Failure is the most limiting for Farley in

considering the potential for off-site

doses. The offsite doses analyses for the

other events which model primary-tosecondary leakage and steam releases

from the secondary side to the

environment assume that the secondary

side remains intact. The steam generator

tubes are not subjected to a sustained

increase in differential pressure, as is

the case following a steam line break

event. This increase in differential

pressure is responsible for the

postulated increase in leakage and

associated offsite doses following a

steam line break event. In addition, the

steam line break event results in a

bypass of containment for steam

generator leakage. Upon implementation

of the voltage-based repair criteria, it

must be verified that the expected

distributions of cracking indications at

the tube support plate intersections are

such that primary-to-secondary leakage

would result in site boundary dose

within the current licensing basis. Data

indicate that a threshold voltage of 2.8

volts could result in through-wall cracks

long enough to leak at steam line break

conditions. Applications of the

proposed repair criteria requires that the

current distribution of a number of

indications versus voltage be obtained

during the refueling outages. The

current voltage is then combined with

the rate of change in voltage

measurement and a voltage

measurement uncertainty to establish an

end of cycle voltage distribution and,

thus, leak rate during steam line break

pressure differential. The leak rate

during a steam line break is further

increased by a factor related to the

probability of detection of the flaws. If

it is found that the potential steam line

break leakage for degraded intersections

planned to be left in service coupled

with the reduced specific activity levels

allowed result in radiological

consequences outside the current

licensing basis, then additional tubes

will be plugged or repaired to reduce

steam line break leakage potential to

within the acceptance limit. Thus, the

consequences of the most limiting

design basis accident are constrained to

present licensing basis limits.

(2) The proposed license amendment

does not create the possibility of a new

or different kind of accident from any

accident previously evaluated.

Implementation of the proposed

voltage-based tube support plate

elevation steam generator tube repair

criteria does not introduce any

significant changes to the plant design

basis. Use of the criteria does not

provide a mechanism that could result

in an accident outside of the region of

the tube support plate elevations.

Neither a single or multiple tube rupture

event would be expected in steam

generator in which the repair criteria

have been applied during all plant

conditions. The bobbin probe signal

amplitude repair criteria are established

such that operational leakage or

excessive leakage during a postulate

steam line break condition is not

anticipated. Southern Nuclear has

previously implemented a maximum

leakage limit of 140/150 gpd (Unit 1/

Unit 2) per steam generator. The R.G.

1.121 criterion for establishing

operational leakage limits that require

plant shutdown are based upon leakbefore-break considerations to detect a

free span crack before potential tube

rupture. The 140/150 gpd limit provides

for leakage detection and plant

shutdown in the event of the occurrence

of an unexpected single crack resulting

in leakage that is associated with the

longest permissible crack length. R.G.

1.121 acceptance criteria for

establishing operating leakage limits are

based on leak-before-break

considerations such that plant

shutdown is initiated if the leakage

associated with the longest permissible

crack is exceeded. The longest

permissible crack is the length that

provides a factor safety of 1.43 against

bursting at steam line break pressure

differential. A voltage amplitude of

approximately 9 volts for typical outside

diameter stress corrosion cracking

corresponds to meeting this tube burst

requirement at the 95% prediction

interval on the burst correlation.

Alternate crack morphologies can

correspond to a voltage so that a unique

crack length is not defined by the burst

pressure versus voltage correlation.

Consequently, typical burst pressure

versus throughwall crack length

correlations is used below to define the

‘‘longest permissible crack’’ for

evaluating operating leakage limits.

The single through-wall crack lengths

that results in tube burst at 1.43 times

steam line break pressure differential

and steam line break conditions are

about 0.53 inch and 0.84 inch,

respectively. Normal leakage for these

crack lengths would range from about

0.4 gallons per minute to 4.5 gallons per

minute, respectively, while lower 95%

confidence level leak rates would range

from about 0.06 gallons per minute to

0.6 gallons per minute, respectively.

An operating leak rate of 140/150 gpd

per steam generator has been

implemented. This leakage limit

provides for detection of 0.4 inch long

cracks at nominal leak rates and 0.6

inch long cracks at the lower 95%

confidence level leak rates. Thus, the

140/150 gpd limit provides for plant

shutdown prior to reaching critical

crack lengths for steam line break

conditions at leak rates less than 95%

confidence level and for three times

normal operating pressure differential at

less than nominal leak rates.

Considering the above, the

implementation of voltage-based

plugging criteria will not create

possibility of a new or different kind of

accident from any previously evaluated.

(3) The proposed license amendment

does not involve a significant reduction

in margin of safety.

The use of the voltage-based tube

support plate elevation repair criteria is

demonstrated to maintain steam
8756 Federal Register / Vol. 60, No. 31 / Wednesday, February 15, 1995 / Notices

generator tube integrity commensurate

with the requirements of R.G. 1.121.

R.G. 1.121 describes a method

acceptable to the NRC staff for meeting

GDCs [General Design Criteria] 2, 14, 15,

31, and 32 by reducing the probability

of the consequences of steam generator

tube rupture. This is accomplished by

determining the limiting conditions of

degradation of steam generator tubing,

as established by inservice inspection,

for which tubes with unacceptable

cracking should be removed from

service. Upon implementation of the

criteria, even under the worst case

conditions, the occurrence of outside

diameter stress corrosion cracking at the

tube support plant elevations is not

expected to lead to a steam generator

tube rupture event during normal or

faulted plant conditions. The most

limiting effect would be a possible

increase in leakage during a steam line

break event. Excessive leakage during a

steam line break event, however, is

precluded by verifying that, once the

criteria are applied, the expected end of

cycle distribution of crack indications at

the tube support plate elevations would

result in minimal, and acceptable

primary to secondary leakage during the

event and, hence, help to demonstrate

radiological conditions are less than an

appropriate fraction of the 10 CFR 100

guideline.

The margin to burst for the tubes

using the voltage-based repair criteria is

comparable to that currently provided

by existing technical specifications.

In addressing the combined effects of

LOCA [loss-of-coolant accident] + SSE

[safe shutdown earthquake] on the

steam generator component (as required

by GDC 2), it has been determined that

tube collapse may occur in the steam

generators at some plants. This is the

case as the tube support plates may

become deformed as a result of lateral

loads at the wedge supports at the

periphery of the plate due to either the

LOCA rarefaction wave and/or SSE

loadings. Then, the resulting pressure

differential on the deformed tubes may

cause some of the tubes to collapse.

There are two issues associated with

steam generator tube collapse. First, the

collapse of steam generator tubing

reduces the RCS [reactor coolant

system] flow area through the tubes. The

reduction in flow area increases the

resistance to flow of steam from the core

during a LOCA which, in turn, may

potentially increase Peak Clad

Temperature (PCT). Second, there is a

potential the partial through-wall cracks

in tubes could progress to through-wall

cracks during tube deformation or

collapse or that short through-wall

indications would leak at significantly

higher leak rates than included in the

leak rate assessments.

Consequently, a detailed leak-beforebreak analysis was performed and it was

concluded that the leak-before-break

methodology (as permitted by GDC 4) is

applicable to the Farley reactor coolant

system primary loops and, thus, the

probability of breaks in the primary loop

piping is sufficiently low that they need

not be considered in the structural

design basis of the plant. Excluding

breaks in the RCS primary loops, the

LOCA loads from the large branch line

breaks were analyzed at Farley and were

found to be of insufficient magnitude to

result in steam generator tube collapse

or significant deformation.

Regardless of whether or not leakbefore-break is applied to the primary

loop piping at Farley, any flow area

reduction is expected to be minimal

(much less than 1%) and PCT margin is

available to account for this potential

effect. Based on analyses’ results, no

tubes near wedge locations are expected

to collapse or deform to the degree that

secondary to primary in-leakage would

be increased over current expected

levels. For all other steam generator

tubes, the possibility of secondary-toprimary leakage in the event of a LOCA

+ SSE event is not significant. In

actuality, the amount of secondary-toprimary leakage in the event of a LOCA

+ SSE is expected to be less than that

previously allowed, i.e., 500 gpd per

steam generator. Furthermore,

secondary-to-primary in-leakage would

be less than primary-to-secondary

leakage for the same pressure

differential since the cracks would tend

to tighten under a secondary-to-primary

pressure differential. Also, the presence

of the tube support plate is expected to

reduce the amount of in-leakage.

Addressing the R.G. 1.83

considerations, implementation of the

tube repair criteria is supplemented by

100% inspection requirements at the

tube support plate elevations having

outside diameter stress corrosion

cracking indications, reduced operating

leakage limits, eddy current inspection

guidelines to provide consistency in

voltage normalization, and rotating

pancake coil inspection requirements

for the larger indications left in service

to characterize the principle degradation

mechanism as outside diameter stress

corrosion cracking.

As noted previously, implementation

of the tube support plate elevation

repair criteria will decrease the number

of tubes that must be taken out of

service with tube plugs or repaired. The

installation of steam generator tube

plugs or tube sleeves would reduce the

RCS flow margin, thus implementation

of the voltage-based repair criteria will

maintain the margin of flow that would

otherwise be reduced through increased

tube plugging or sleeving.

Considering the above, it is concluded

that the proposed change does not result

in a significant reduction in margin with

respect to plant safety as defined in the

Final Safety Analysis Report or any

bases of the plant Technical

Specifications.

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: Houston-Love Memorial

Library, 212 W. Burdeshaw Street, Post

Office Box 1369, Dothan, Alabama

36302

Attorney for licensee: M. Stanford

Blanton, Esq., Balch and Bingham, Post

Office Box 306, 1710 Sixth Avenue

North, Birmingham, Alabama 35201

NRC Project Director: William H.

Bateman.

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: January

9, 1995.

Description of amendments request:

The requested changes to the Technical

Specifications (TS) would implement

the recommended changes from Generic

Letter 93–05, ‘‘Line Item Technical

Specification Improvements to Reduce

Surveillance Requirements for Testing

During Power Operation.’’ Specifically,

the amendments would implement TS

changes corresponding to the following

GL 93–05 line-item improvement issues:

Control Rod Movement Test for

Pressurized Water Reactors, Radiation

Monitors, Surveillance of Boron

Concentration in the Accumulator/

Safety Injection/Core Flood Tank,

Containment Spray System, Hydrogen

Recombiner, and Special Test

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 changes do not

involve a significant increase in the

probability or consequences of an

accident previously evaluated. The

proposed changes do not involve any

change to the configuration or method
Federal Register / Vol. 60, No. 31 / Wednesday, February 15, 1995 / Notices 8757

of operation of any plant equipment

used to mitigate the consequences of an

accident. The changes to the

surveillance requirements will result in

an overall improvement in plant safety

by reducing the likelihood of plant trips

and subsequent challenges to safety

systems, decreasing equipment

degradation due to excessive testing,

reducing radiation exposure to plant

personnel, increasing the availability of

safety related equipment, and

eliminating an unnecessary burden on

plant personnel. Therefore, the

proposed changes do 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. The proposed

changes do not involve any change to

the configuration or method of

operation of any plant equipment used

to mitigate the consequences of an

accident. The relaxation of surveillance

tests curtails the excessive amount of

testing that increases wear on the

equipment and reduces the likelihood of

plant trips and subsequent challenges to

safety systems. The relaxation also

increases the availability of safety

related equipment. Accordingly, no new

failure modes have been defined for any

plant system or component important to

safety nor has any new limiting failure

been identified as a result of the

proposed changes. 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 changes do not

involve a significant reduction in a

margin of safety. The proposed changes

eliminate an unnecessary burden

without compromising protection for

public health and safety. The proposed

changes were generically analyzed by

the NRC as part of a comprehensive

study and presented in NUREG–1366

‘‘Improvement to Technical

specifications (sic) Surveillance

Requirements.’’ The NRC concluded

that while some testing at power is

essential to verify equipment and

system operability, safety can be

improved, equipment degradation

decreased, and unnecessary personnel

burden relaxed by reducing the amount

of testing at power. SNC has analyzed

plant operations and made a

comparison with the criteria stated in

NUREG–1366 for the line-item

improvements contained in this request

and has found the NUREG–1366 basis to

be consistent with the Farley design and

operation experience. Therefore, 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: Houston-Love Memorial

Library, 212 W. Burdeshaw Street, Post

Office Box 1369, Dothan, Alabama

36302.

Attorney for licensee: M. Stanford

Blanton, Esq., Balch and Bingham, Post

Office Box 306, 1710 Sixth Avenue

North, Birmingham, Alabama 35201.

NRC Project Director: William H.

Bateman.

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50–346, Davis-Besse Nuclear Power

Station, Unit No. 1, Ottawa County,

Ohio

Date of amendment request:

December 6, 1994.

Description of amendment request:

The proposed change to Technical

Specification 3/4.1.3.2 will delete

Surveillance Requirement (SR) 4.1.3.2.2,

that presently requires, every 31 days,

the movement of at least 2% of its

height for each Axial Power Shaping

Rod not fully withdrawn. The proposed

amendment would also change the

surveillance intervals for the following

Technical Specifications (TS) in

accordance with the guidance of

Generic Letter 93–05, ‘‘Line Item

Technical Specifications Improvements

to Reduce Surveillance Requirements

For Testing During Power Operation,’’

and NUREG–1366, ‘‘Improvements to

Technical Specifications Surveillance

Requirements:’’

1. TS 4.1.3.2 for the Movable Control

Assemblies ‘‘Group Height—Safety and

Regulating Rod Groups,’’ will relax

testing requirements from at least once

every 31 days to every 92 days.

2. TS 4.4.6.2, for ‘‘Operational

Leakage,’’ relaxes the requirement to

leakage test RCS pressure isolation

valves prior to MODE 2 whenever the

plant has been in COLD SHUTDOWN

for 72 hours to whenever the plant has

been in COLD SHUTDOWN for 7 days.

3. SR 4.5.2.c.2 for TS 4.5.2, ‘‘ECCS

Subsystems—Tavg equal to or greater

than 280° F,’’ relaxes the inspection

requirements for ensuring no debris in

containment from ‘‘at the completion of

each containment entry’’ to ‘‘at least

once daily.’’

4. TS 4.6.2.1.d, for the ‘‘Containment

Spray System,’’ relaxes the SR to

perform an air or smoke flow test

through the spray header and nozzles

from once per 5 years to once per 10

years.

5. TS 4.10.4.2 for ‘‘Special Test

Exceptions Shutdown Margin’’ relaxes

the SR interval for testing rod insertion

capability prior to reducing shutdown

margin from 24 hours to 7 days.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

NRC has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) The proposed changes do not

involve a significant increase in the

probability or consequences of an

accident previously evaluated.

The change does not involve a

significant increase in the probability of

an accident previously evaluated nor

does it involve a significant increase in

the consequences of an accident

previously evaluated because no change

is being made to any accident initiator

and no accident conditions or

assumptions used in evaluating the

radiological consequences of an

accident are changed. Relaxation of

surveillance requirements is in

accordance with GL 93–05, NUREG–

1366, and is compatible with plant

operating experience. Deletion of SR

4.1.3.2 is consistent with NUREG–1430,

‘‘Improved Standard Technical

Specifications for B&W Plants.’’ No

credit is taken in any accident analysis

or mitigation requirements for the Axial

Power Shaping Rod Group.

(2) The proposed changes do not

create the possibility of a new or

different kind of accident from any

accident previously evaluated.

The proposed changes do not create

the possibility of any new or different

kind of accident from any accident

previously evaluated because no new

accident initiators or assumptions are

introduced by these proposed changes.

Relaxation of SRs as discussed in GL

93–05 was evaluated as reducing

equipment degradation with no increase

in safety consequences consistent with

the maintenance of plant specific

reliability of the equipment and systems

affected. Deletion of the SR to move the

Axial Power Shaping Rod Group does

not affect the requirement to verify rod

position, and there is no credit taken for

movement of these rods to mitigate an

accident.

(3) The proposed changes do not

result in a significant reduction in the

margin of safety.
8758 Federal Register / Vol. 60, No. 31 / Wednesday, February 15, 1995 / Notices

The changes do not involve a

significant reduction in the margin of

safety, because the proposed changes

affect only surveillance requirements,

do not affect the function of the

components and systems involved, and

do not decrease the estimated

equipment or system reliability.

Based on the NRC staff analysis, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of Toledo Library,

Documents Department, 2801 Bancroft

Avenue, Toledo, Ohio 43606.

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, N.W.,

Washington, DC 20037.

NRC Project Director: Leif J.

Norrholm.

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50–346, Davis-Besse Nuclear Power

Station, Unit No. 1, Ottawa County,

Ohio

Date of amendment request:

December 6, 1994.

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) 4.0.5,

‘‘Applicability’’ and its associated

Bases; TS 3/4.1.2.3, ‘‘Reactivity Control

Systems—Makeup Pump—Shutdown;

TS 3/4.1.2.4, ‘‘Reactivity Control

Systems—Makeup Pump—Operating;

TS 3/4.1.2.6, Reactivity Control

Systems—Boric Acid Pump—

Shutdown; and TS 3/4.1.2.7, ‘‘Reactivity

Control System—Boric Acid Pumps—

Operating.’’ The proposed change

would replace the specific monthly

surveillance requirements associated

with the makeup pumps and boric acid

pumps with a surveillance requirement

referencing TS 4.0.5, which references

Section XI of the American Society of

Mechanical Engineers Boiler and

Pressure Vessel Code for quarterly

pump testing requirements. The

proposed change to TS 4.0.5 and its

associated Bases would revise the

requirement regarding the NRC’s

approval of relief requests to be in

accordance with the NRC Staff’s

recommendation contained in NUREG–

1482, ‘‘Guidelines for Inservice Testing

at Nuclear Power Plants.’’ Additionally,

TS 4.0.5.a.2 which describes historical

requirements for inservice inspection

and testing would be deleted and TS

4.0.5.a.1 would be renumbered as TS

4.0.5.a.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

NRC Staff has performed an 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.

Operation of the Davis-Besse Nuclear

Power Station, Unit No. 1, in

accordance with these changes, would

not involve a significant increase in the

probability of an accident previously

evaluated because no accident initiators,

conditions, or assumptions are affected

by the proposed changes to replace the

specific monthly surveillance

requirements for the makeup and boric

acid pumps with surveillance

requirements referencing TS 4.0.5

(ASME Boiler and Pressure Vessel Code

Section XI requirements) and to delete

wording regarding NRC approval of

relief requests. The changes do not

involve a significant increase in the

consequences of an accident previously

evaluated, because no accident

conditions or assumptions are affected

that would increase the radiological

consequences of a previously evaluated

accident.

(2) The proposed changes do not

create the possibility of a new or

different kind of accident from any

accident previously evaluated.

The proposed changes do not result in

any new accident initiators nor do they

alter any accident scenarios. The

changes do not create the possibility of

a different kind of accident from any

accident previously evaluated, because

the surveillance requirements for the

makeup and boric acid pumps only

affect the testing of existing

components, systems, and functions,

and do not introduce any new

requirements.

(3) The proposed changes do not

result in a significant reduction in the

margin of safety.

The proposed changes do not reduce

or adversely affect the capabilities or

reliability of any plant structures,

systems or components. Relaxation of

the surveillance testing interval for the

boric acid and makeup pumps and

modifying the testing requirements is

consistent with previous NRC guidance.

Based on this NRC staff evaluation, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of Toledo Library,

Documents Department, 2801 Bancroft

Avenue, Toledo, Ohio 43606.

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, N.W.,

Washington, DC 20037.

NRC Project Director: Leif J.

Norrholm.

Union Electric Company, Docket No.

50–483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of amendment request: January

13, 1995.

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) by

relocating Tables 3.3–2, ‘‘Reactor Trip

System Instrumentation Response

Times,’’ and 3.3–5, ‘‘Engineered Safety

Features Response Times,’’ to FSAR

Chapter 16, Section 16.3. The Bases

discussion specific to Table 3.3–5

would also be relocated to FSAR

Section 16.3.

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 revision does not

involve a significant hazards

consideration because operation of

Callaway Plant with this change would

not:

(1) Involve a significant increase in

the probability or consequences of an

accident previously evaluated.

Overall protection system

performance will remain within the

bounds of the accident analyses

documented in FSAR Chapter 15,

WCAP–10961–P, and WCAP–11883

since no changes to the response times

or measurement interval are proposed.

The RTS and ESFAS will continue to

function in a manner consistent with

the above analysis assumptions and the

plant design basis. As such, there will

be no degradation in the performance of

nor an increase in the number of

challenges to equipment assumed to

function during an accident situation.

These Technical Specification

revisions do not involve any hardware

changes nor do they affect the

probability of any event initiators. There

will be no change to normal plant

operating parameters or accident

mitigation capabilities. Therefore, there

will be no increase in the probability or

consequences of any accident occurring

due to these changes.

(2) Create the possibility of a new or

different kind of accident from any

previously evaluated.
Federal Register / Vol. 60, No. 31 / Wednesday, February 15, 1995 / Notices 8759

As discussed above, there are no

hardware changes associated with these

Technical Specification revisions nor

are there any changes in the method by

which any safety-related plant system

performs its safety function. The normal

manner of plant operation is unaffected.

No new accident scenarios, transient

precursors, failure mechanisms, or

limiting single failures are introduced as

a result of these changes. There will be

no adverse effect or challenges imposed

on any safety-related system as a result

of these changes. Therefore, the

possibility of a new or different type of

accident is not created.

(3) Involve a significant reduction in

a margin of safety.

No response time changes are

proposed in this amendment

application; only the document where

these limits are listed will be changed.

There will be no effect on the manner

in which safety limits or limiting safety

system settings are determined nor will

there be any effect on those plant

systems necessary to assure the

accomplishment of protection functions.

There will be no impact on DNBR

limits, FQ, F-delta-H, LOCA PCT, peak

local power density, or any other margin

of safety.

Based upon the preceding

information, it has been determined that

the proposed changes to the Technical

Specifications do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated, create the possibility of a

new or different kind of accident from

any accident previously evaluated, or

involve a significant reduction in a

margin of safety. Therefore, it is

concluded that the proposed changes

meet the requirements of 10CFR50.92(C)

[sic] and 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 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: Callaway County Public

Library, 710 Court Street, Fulton,

Missouri 65251.

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, N.W.,

Washington, DC 20037.

NRC Project Director: Leif J.

Norrholm.

Vermont Yankee Nuclear Power

Corporation, Docket No. 50–271,

Vermont Yankee Nuclear Power Station,

Vernon, Vermont

Date of amendment request:

December 8, 1994.

Description of amendment request:

The proposed amendment would

change Standby Gas Treatment Power

Supply Requirements during refueling

operations.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

SGTS [Standby Gas Treatment System]

DURING REFUELING OPERATIONS

(Specification 3.7.B.1, 3.7.B.3)

1. The proposed amendment will not

involve a significant increase in the

probability or consequences of an

accident previously evaluated. The

Standby Gas Treatment System (SGTS)

is not the initiator of any accident.

SGTS may be required to operate for a

design basis loss of coolant accident or

for a refueling accident in order to

mitigate the consequences of said

accident by providing a filtered exhaust

path to minimize the potential release of

radioactive material to the environs. The

proposed amendment does not reduce

or change the operational requirements

for the SGTS for an accident. The

proposed amendment now clearly

defines the operability requirements

during refueling conditions. The

proposed amendment further requires

the availability of a second auxiliary

power supply in the event that an

Emergency Diesel Generator (EDG) is

out of service during refueling

operations, not currently required. We

conclude, therefore, that the proposed

amendment does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed amendment will not

create the possibility of a new or

different kind of accident from any

accident previously evaluated. The

SGTS is not an accident initiator,

therefore, the proposed amendment will

not create the possibility of a new or

different kind of accident from any

accident previously evaluated.

3. The proposed amendment will not

involve a significant reduction in a

margin of safety. The proposed

amendment requires the availability of a

second auxiliary power supply in the

event that an EDG is out of service

during refueling operations, not

currently required. Maintaining

availability of a specific reliable

auxiliary electrical power source as an

alternative to an EDG in this mode

provides assurance that SGTS can, if

required, be operated without placing

undue constraints on EDG availability

and represents an enhancement that

increases a margin of safety. We

conclude, therefore, that the proposed

amendment does not involve a

significant reduction in a margin of

safety.

Based on the above discussion, we

have determined that this change does

not constitute a significant hazards

consideration as defined in

10CFR50.92(c).

LABORATORY CARBON SAMPLE

ANALYSIS (Specification 3.7.B.2.b)

1. The proposed amendment will not

involve a significant increase in the

probability or consequences of an

accident previously evaluated. The

Standby Gas Treatment System (SGTS)

is not the initiator of any accident.

SGTS may be required to operate for a

design basis loss of coolant accident or

for a refueling accident in order to

mitigate the consequences of said

accident by providing a filtered exhaust

path to minimize the potential release of

radioactive material to the environs. The

proposed amendment does not reduce

or change the operational requirements

for the SGTS for an accident. The

proposed amendment now clearly

defines the operability requirements

during the interval between sample

removal and completion of laboratory

analysis.

2. The proposed amendment will not

create the possibility of a new or

different kind of accident from any

accident previously evaluated. The

SGTS is not an accident initiator,

therefore, the proposed amendment will

not create the possibility of a new or

different kind of accident from any

accident previously evaluated.

3. The proposed amendment will not

involve a significant reduction in a

margin of safety. The proposed change

does not reduce the requirements or

acceptance criteria for sampling, testing

or analysis. The proposed change only

incorporates into the specification an

existing clarification which addresses

the determination of operability during

the time between sample removal and

completion of laboratory analysis. The

change provides an explicit time limit

consistent with current regulatory

criteria for completion of analyses.

Based on the above discussion, we

have determined that this change does

not constitute a significant hazards
8760 Federal Register / Vol. 60, No. 31 / Wednesday, February 15, 1995 / Notices

consideration as defined in

10CFR50.92(c).

TORUS VENT MODE (Specification 4.7

B.2.c)

1. The proposed amendment will not

involve a significant increase in the

probability or consequences of an

accident previously evaluated. The

Standby Gas Treatment System (SGTS)

is not the initiator of any accident.

SGTS may be required to operate for a

design basis loss of coolant accident or

for a refueling accident in order to

mitigate the consequences of said

accident by providing a filtered exhaust

path to minimize the potential release of

radioactive material to the environs. The

proposed amendment does not reduce

or change the operational requirements

for the SGTS for an accident.

2. The proposed amendment will not

create the possibility of a new or

different kind of accident from any

accident previously evaluated. The

SGTS is not an accident initiator,

therefore, the proposed amendment will

not create the possibility of a new or

different kind of accident from any

accident previously evaluated.

3. The proposed amendment will not

involve a significant reduction in a

margin of safety. The proposed change

will incorporate into the specification

an existing clarification. Use of the

SGTS filters during Torus venting

results in an insignificant flow through

the filters. Further, maintaining

humidity control prevents any adsorber

degradation. Past sample testing on a six

month calendar interval when 720

hours operating time has not

accumulated has shown no detectable

impact.

Based on the above discussion, we

have determined that this change does

not constitute a significant hazards

consideration as defined in

10CFR50.92(c).

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: Brooks Memorial Library, 224

Main Street, Brattleboro, Vermont

05301.

Attorney for licensee: John A. Ritsher,

Esquire, Ropes and Gray, One

International Place, Boston,

Massachusetts 02110–2624.

NRC Project Director: Walter R.

Butler.

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.

Power Authority of The State of New

York, Docket No. 50–286, Indian Point

Nuclear Generating Unit No. 3,

Westchester County, New York

Date of application for amendment:

November 19, 1994.

Brief description of amendment: The

proposed amendment would revise

Section 3.10.8 and the associated Bases

of the Indian Point Nuclear Generating

Unit No. 3 Technical Specifications.

Specifically, the proposed revision

would reduce the maximum allowable

control rod drop time from 2.4 to 1.8

seconds. The change would remove, for

testing purposes, the allowance for a

seismic event (0.6 seconds), which had

been integral to the 2.4 second safety

analysis basis. Since a seismic event

cannot be simulated during the rod drop

time test, the more conservative testing

acceptance criteria value of 1.8 seconds

is needed to ensure that the plant is

within its design basis. This proposed

revision will support control rod testing

which is required during startup from

the current outage.

Date of publication of individual

notice in Federal Register: January 20,

1995 (60 FR 4203).

Expiration date of individual notice:

February 21, 1995.

Local Public Document Room

location: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10610.

Notice of Issuance of Amendments to

Facility Operating Licenses

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application

complies with the standards and

requirements of the Atomic Energy Act

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

Commission’s rules and regulations.

The Commission has made appropriate

findings as required by the Act and the

Commission’s rules and regulations in

10 CFR Chapter I, which are set forth in

the license amendment.

Notice of Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for A Hearing in

connection with these actions was

published in the Federal Register as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the applications for

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

the Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, the Gelman Building, 2120 L

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

local public document rooms for the

particular facilities involved.

Arizona Public Service Company, et al.,

Docket Nos. STN 50–528, STN 50–529,

and STN 50–530, Palo Verde Nuclear

Generating Station, Units 1, 2, and 3,

Maricopa County, Arizona

Date of application for amendments:

November 30, 1994.

Brief description of amendments:

These amendments relocate Table 3.3–

2, ‘‘Reactor Protective Instrumentation

Response Times,’’ and Table 3.3–5,

‘‘Engineered Safety Features Response

Times,’’ of TS 3/4.3.1 and 3/4.3.2,

respectively, to the Palo Verde Updated

Final Safety Analysis Report (UFSAR)

in accordance with the guidance

provided in Generic Letter 93–08. In

addition, the amendments make

administrative changes to two previous

TS amendment requests to maintain

consistency with the deletion of Tables

3.3–2 and 3.3–5. The amendments also

delete an obsolete footnote on page 3/4

3–17 of the Palo Verde Unit 2’s TS.
Federal Register / Vol. 60, No. 31 / Wednesday, February 15, 1995 / Notices 8761

Date of issuance: February 3, 1995.

Effective date: February 3, 1995.

Amendment Nos.: 88, 75 and 59.

Facility Operating License Nos. NPF–

41, NPF–51, and NPF–74: The

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 4, 1995 (60 FR 496)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated February 3, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona

85004.

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:

November 16, 1994.

Brief description of amendments: The

proposed amendments change the

Technical Specifications to revise the

wording for the containment integrated

leakage rate testing in Section 3/4.6.1.2

to make it consistent with the

requirements of the BWR–4 Improved

Standard Technical Specifications

(NUREG–1433).

Date of issuance: January 26, 1995.

Effective date: January 26, 1995.

Amendment Nos.: 173 and 204.

Facility Operating License Nos. DPR–

71 and DPR–62.

Date of initial notice in Federal

Register: December 21, 1994 (59 FR

65810).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated January 26,

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.

Commonwealth Edison Company,

Docket Nos. STN 50–454 and STN 50–

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

Ogle County, Illinois Docket Nos. STN

50–456 and STN 50–457, Braidwood

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

Illinois

Date of application for amendments:

June 13, 1994, as supplemented on

October 7, 1994.

Brief description of amendments: The

amendments revise the administrative

controls in Section 6 of the technical

specifications (TS). The changes

include: (1) a change to the submittal

frequency of the Radiological Effluent

Release Report from semiannually to

annually; (2) changes to the Shift

Technical Advisor (STA) description;

(3) a clarification of the Shift Engineer

responsibilities; and (4) several editorial

changes.

Date of issuance: February 2, 1995.

Effective date: February 2, 1995.

Amendment Nos.: 69, 69, 59 and 59.

Facility Operating License Nos. NPF–

37, NPF–66, NPF–72 and NPF–77: The

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 26, 1994 (59 FR

53839).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated February 2,

1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: For Byron, the Byron Public

Library, 109 N. Franklin, P.O. Box 434,

Byron, Illinois 61010; for Braidwood,

the Wilmington Township Public

Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481.

Connecticut Yankee Atomic Power

Company, Docket No. 50–213, Haddam

Neck Plant, Middlesex County,

Connecticut

Date of application for amendment:

May 17, 1993 as supplemented October

12, 1994.

Brief description of amendment: The

amendment replaces License Condition

2.C.4, relating to the implementation

and maintenance of the approved Fire

Protection Program, in its entirety with

a new License Condition. In

conjunction, with this change, and in

accordance with GL 86–10, Technical

Specification provisions related to the

Fire Protection Program are being

deleted and placed in the Updated Final

Safety Analysis Report.

Date of Issuance: February 1, 1995.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 179.

Facility Operating License No. DPR–

61. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 7, 1993 (58 FR 36432).

The October 12, 1994, letter provided

clarifying information that did not

change the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of this amendment is contained in a

Safety Evaluation dated February 1,

1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Russell Library, 123 Broad

Street, Middletown, CT 06457.

Duke Power Company, et al., Docket

Nos. 50–413 and 50–414, Catawba

Nuclear Station, Units 1 and 2, York

County, South Carolina Date of

application for amendments: August 25,

1994, as supplemented November 16,

1994.

Brief description of amendments: The

amendments revise Technical

Specification Table 3.3–4, by revising

the ‘‘Trip Setpoint’’ and ‘‘Allowable

Value’’ for the 4 kV bus undervoltage

grid degraded voltage relays and the

‘‘Allowable Value’’ for the 4 kV

undervoltage loss of voltage/loss of

offsite power relays. This revision was

submitted in response to a concern

identified by the licensee in their SelfInitiated Technical Audit and during

the electrical distribution system

functional inspection team findings.

Date of issuance: January 20, 1995.

Effective date: To be implemented

within 30 days from the date of

issuance.

Amendment Nos.: 127 and 121.

Facility Operating License Nos. NPF–

35 and NPF–52: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: October 12, 1994 (59 FR

51619).

The November 16, 1994, letter

provided clarifying information that did

not change the scope of the August 25,

1994, application and the initial

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated January 20,

1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: York County Library, 138 East

Black Street, Rock Hill, South Carolina

29730.

Florida Power and Light Company,

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

No. 1, St. Lucie County, Florida

Date of application for amendment:

July 28, 1994.

Brief description of amendment: This

amendment revises Technical

Specifications 3/4.4.13 to incorporate

Low Temperature Overpressure

Protection requirements similar to those

recommended by the NRC staff via

Generic Letter 90–06.
8762 Federal Register / Vol. 60, No. 31 / Wednesday, February 15, 1995 / Notices

Date of Issuance: January 27, 1995.

Effective Date: January 27, 1995.

Amendment No.: 132.

Facility Operating License No. DPR–

67: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 17, 1994 (59 FR

42341).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated January 27,

1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort

Pierce, Florida 34954–9003.

Georgia Power Company, Oglethorpe

Power Corporation, Municipal Electric

Authority of Georgia, City of Dalton,

Georgia, Docket Nos. 50–424 and 50–

425, Vogtle Electric Generating Plant,

Units 1 and 2, Burke County, Georgia

Date of application for amendments:

February 3, 1994.

Brief description of amendments: The

amendments relocate the requirements

of Technical Specification 3/4.7.10,

Area Temperature Monitoring, to

section 16.3 of the VEGP Final Safety

Analysis Report (FSAR). With this

relocation to the FSAR, GPC plans to

clarify the basis for areas to be

monitored and modify these

surveillance requirements. This change

is in accordance with NUREG–1431,

‘‘Standard Technical Specifications,

Westinghouse Plants.’’

Date of issuance: January 23, 1995.

Effective date: To be implemented

within 30 days from the date of

issuance.

Amendment Nos.: 83 and 61.

Facility Operating License Nos. NPF–

68 and NPF–81: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: September 2, 1994 (59 FR

45735).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated January 23,

1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Burke County Library, 412

Fourth Street, Waynesboro, Georgia

30830.

GPU Nuclear Corporation, et al., Docket

No. 50–289, Three Mile Island Nuclear

Station, Unit No. 1, Dauphin County,

Pennsylvania

Date of application for amendment:

May 13, 1992.

Brief description of amendment: The

amendment changes the allowable

primary-to-secondary leakage rate, as

specified in License Condition 2.c.(8)2,

from 0.1 gallons per minute (gpm) to 0.2

gpm.

Date of Issuance: January 31, 1995.

Effective date: January 31, 1995.

Amendment No.: 193.

Facility Operating License No. DPR–

50. Amendment revises a License

Condition.

Date of initial notice in Federal

Register: October 14, 1992 (57 FR

47137).

The Commission’s related evaluation

of this amendment is contained in a

Safety Evaluation dated January 31,

1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Government Publications

Section, State Library of Pennsylvania,

Walnut Street and Commonwealth

Avenue, Box 1601, Harrisburg,

Pennsylvania 17105.

Nebraska Public Power District, Docket

No. 50–298, Cooper Nuclear Station,

Nemaha County, Nebraska

Date of amendment request:

September 26, 1994.

Brief description of amendment: The

amendment revised Technical

Specification 3.5.C.1 and 3.5.C.4 to

increase the minimum pressure at

which the high pressure coolant

injection system is required to be

operable from 113 psig to 150 psig.

Date of issuance: January 25, 1995.

Effective date: January 25, 1995.

Amendment No.: 166.

Facility Operating License No. DPR–

46. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 26, 1994 (59 FR

53841). The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

January 25, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Auburn Public Library, 118

15th Street, Auburn, Nebraska 68305.

Nebraska Public Power District, Docket

No. 50–298, Cooper Nuclear Station,

Nemaha County, Nebraska

Date of amendment request:

December 22, 1994.

Brief description of amendment: The

amendment revised Technical

Specification 1.0.J, definition of limiting

conditions for operation, consistent

with the guidance provided in NRC

Generic Letter 87–09, ‘‘Sections 3.0 and

4.0 of the Standard Technical

Specifications on the Applicability of

Limiting Conditions for Operation and

Surveillance Requirements.’’

Date of issuance: February 3, 1995.

Effective date: February 3, 1995.

Amendment No.: 168.

Facility Operating License No. DPR–

46. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 3, 1995 (60 FR 153).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated February 3,

1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Auburn Public Library, 118

15th Street, Auburn, Nebraska 68305.

Niagara Mohawk Power Corporation,

Docket No. 50–220, Nine Mile Point

Nuclear Station Unit No. 1, Oswego

County, New York.

Date of application for amendment:

July 21, 1994.

Brief description of amendment: The

amendment revises Technical

Specifications 2.2.2, 3.2.8, 4.2.8, and the

associated Bases to reduce the number

of reactor head safety valves required

operable from 16 valves to 9 valves. The

setpoints of the valve groups are

unchanged by this amendment. The

amendment requires testing of the safety

valves in accordance with the approved

NMP–1 Inservice Test Program.

Date of issuance: January 25, 1995.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 152.

Facility Operating License No. DPR–

63: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: August 31, 1994 (59 FR

45027).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated January 25,

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.

Niagara Mohawk Power Corporation,

Docket No. 50–410, Nine Mile Point

Nuclear Station, Unit 2, Oswego

County, New York.

Date of application for amendment:

October 28, 1994.

Brief description of amendment: The

amendment revises Technical
Federal Register / Vol. 60, No. 31 / Wednesday, February 15, 1995 / Notices 8763

Specification (TS) 1.7, ‘‘CORE

ALTERATION,’’ to state that movement

or replacement of incore

instrumentation is not considered to be

a CORE ALTERATION and that

movement of control rods is not

considered a CORE ALTERATION

provided there are no fuel assemblies in

the associated core cell. This

amendment includes changes to TS 3/

4.9.3, ‘‘Control Rod Position,’’ and

associated Bases to be consistent with

the revision to TS 1.7. TS 3/4.9.3 is

being revised to require that all control

rods be inserted only during loading of

fuel assemblies into the core rather than

during CORE ALTERATIONS. These

changes are consistent with the NRC’s,

‘‘Improved Standard Technical

Specifications,’’ (NUREG–1434).

This amendment also revises Item

1.i.3) of TS Tables 3.3.2–1 and 4.3.2.1–

1 to delete the requirement for Reactor

Water Cleanup isolation due to

actuation of the Standby Liquid Control

System (SLCS) in OPERATIONAL

CONDITION 5. License Amendment No.

48 issued on September 30, 1993,

deleted the requirement for the SLCS to

be OPERABLE in OPERATIONAL

CONDITION 5; however, due to an

oversight, Item 1.i.3) and associated

notations were not deleted from TS

Tables 3.3.2–1 and 4.3.2.1–1 as part of

License Amendment No. 48. This

amendment corrects that oversight.

Date of issuance: January 20, 1995.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 61.

Facility Operating License No. NPF–

69: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: November 23, 1994 (59 FR

60382).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated January 20,

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.

Niagara Mohawk Power Corporation,

Docket No. 50–410, Nine Mile Point

Nuclear Station, Unit 2, Oswego County,

New York

Date of application for amendment:

November 14, 1994.

Brief description of amendment: The

amendment revises Technical

Specification 4.5.1.e.2.e) to reduce the

leak rate test pressure for the Automatic

Depressurization System (ADS) nitrogen

receiving tanks from 385 psig to 365

psig.

Date of issuance: January 31, 1995.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 62.

Facility Operating License No. NPF–

69: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: December 21, 1994 (59 FR

65817).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated January 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.

North Atlantic Energy Service

Corporation, Docket No. 50–443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request: January

14, 1994, as modified by letter dated

October 17, 1994.

Description of amendment request:

The amendment revises the Appendix A

Technical Specifications (TS) to specify

the composition of the Station

Operation Review Committee (SORC)

based on experience and expertise vice

organizational position, to implement a

Station Qualified Reviewer Program

(SQRP), and to revise the time within

which the Nuclear Safety Audit Review

Committee (NSARC) must issue reports

and minutes.

The amendment also incorporated a

number of editorial changes to delete

certain items that are no longer

applicable; remove inconsistencies

involving the names of systems,

equipment and NSARC function,

composition, and use of alternates; and

correct the value for the reactor coolant

system volume. Other editorial changes

have been incorporated for document

format consistency. The amendment

affects the following: TS Sections 1.31,

3.3.3.6, 3.4.1.2, 4.6.3.2, 3.7.1.2, 3/4 10.6,

5.4.2, 6.3.1, 6.4, 6.7, and 6.8.1.4, and

Table 4.3–1.

Date of issuance: January 26, 1995.

Effective date: January 26, 1995.

Amendment No.: 34.

Facility Operating License No. NPF–

86. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: May 25, 1994 (59 FR 27057)

The licensee’s letter dated October 17,

1994, provided clarification and minor

revision to the application but does 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

January 26, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Exeter Public Library, 47 Front

Street, Exeter, NH 03833.

Northeast Nuclear Energy Company, et

al., Docket No. 50–423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of application for amendment:

July 22, 1994.

Brief description of amendment: The

amendment revises the Technical

Specifications to incorporate a different

setpoint and transient methodology for

determining the maximum allowable

power range neutron flux setpoint.

These changes allow Millstone Unit 3 to

operate with a reduced number of main

steam-line safety valves at a reduced

power level, as determined by the high

flux setpoint.

Date of issuance: January 31, 1995.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 102.

Facility Operating License No. NPF–

49. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: September 14, 1994 (59 FR

47171).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated January 31,

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.

Pennsylvania Power and Light

Company, Docket Nos. 50–387 and 50–

388 Susquehanna Steam Electric

Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of application for amendments:

October 25, 1994.

Brief description of amendments:

These amendments add to the

Susquehanna, Units 1 and 2, Technical

Specifications, isolation signals to Table

3.6.3–1 for the containment isolation

valves on the sample lines for the

containment radiation monitoring and
8764 Federal Register / Vol. 60, No. 31 / Wednesday, February 15, 1995 / Notices

wetwell sample lines. This change is

based on the licensee’s design change

for installation of a new CRM and

wetwell sample system.

Date of issuance: January 31, 1995.

Effective date: January 31, 1995.

Amendment Nos.: 141 and 111.

Facility Operating License Nos. NPF–

14 and NPF–22. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: December 7, 1994 (59 FR

63126). The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

January 31, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Osterhout Free Library,

Reference Department, 71 South

Franklin Street, Wilkes-Barre,

Pennsylvania 18701.

Philadelphia Electric Company, Docket

No. 50–352, Limerick Generating

Station, Unit 1, Montgomery County,

Pennsylvania.

Date of application for amendment:

June 10, 1994, as supplemented by letter

dated December 19, 1994.

Brief description of amendment: This

amendment involves a one-time change

affecting the Allowed Outage Time

(AOT) for the Emergency Sevice Water

(ESW) system, Residual Heat Removal

Service Water (RHRSW) System, the

Suppression Pool Cooling, the

Suppression Pool Spray, and Low

Pressure Coolant Injection modes of the

Residual Heat Removal System, and

Core Spray System to be extended from

3 and 7 days to 14 days during the Unit

2 refueling outage scheduled to begin in

January 1995. This proposed extended

AOT allows adequate time to install

isolation valves and cross-ties on the

ESW and RHRSW Systems to facilitate

future inspections or maintenance.

Date of issuance: January 27, 1995.

Effective date: January 27, 1995.

Amendment No. 86.

Facility Operating License No. NPF–

39. This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: July 20, 1994 (59 FR 37077).

The December 19, 1994 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 January 27,

1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania

19464.

Philadelphia Electric Company, Docket

No. 50–353, Limerick Generating

Station, Unit 2, Montgomery County,

Pennsylvania

Date of application for amendment:

June 30, 1994.

Brief description of amendment: This

amendment removes the controls for a

remote shutdown system control valve

and the primary containment isolation

valves from TS Tables 3.3.7.4–1 and

3.6.3–1 respectively, as a result of

eliminating the steam condensing mode

of the Residual Heat Removal system.

Date of issuance: January 27, 1995.

Effective date: January 27, 1995.

Amendment No. 47.

Facility Operating License No. NPF–

85. This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: August 17, 1994 (59 FR

42343).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated January 27,

1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania

19464.

Philadelphia Electric Company, Docket

No. 50–353, Limerick Generating

Station, Unit 2, Montgomery County,

Pennsylvania

Date of application for amendment:

August 27, 1993, supplemented by letter

dated November 17, 1993.

Brief description of amendment: The

amendment allows an expanded

operating domain for the Limerick

Generating Station (LGS), Unit 2,

resulting from the implementation of

the Average Power Range Monitor—Rod

Block Monitor Technical Specifications/

Maximum Extended Load Line Limit

Analysis. These improvements are a

prerequisite for Power Rerate Program

implementation at Limerick Generating

Station, Unit 2.

Date of issuance: January 27, 1995.

Effective date: January 27, 1995.

Amendment No. 48.

Facility Operating License No. NPF–

85. The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: October 13, 1993 (58 FR

52992). The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

January 27, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania

19464.

Philadelphia Electric Company, Docket

Nos. 50–352 and 50–353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of application for amendments:

August 22, 1994.

Brief description of amendments:

These amendments revise TS 3.1.5,

‘‘Standby Liquid Control System,’’ to

remove the requirement for the standby

liquid control system to be operable in

OPERATIONAL CONDITION 5,

Refueling, when any control rod is

withdrawn and the TS definition of

CORE ALTERATION to exclude control

rod movement in a control cell that

contains no fuel assemblies.

Date of issuance: January 27, 1995.

Effective date: January 27, 1995.

Amendment Nos. 87/49.

Facility Operating License Nos. NPF–

39 and NPF–85. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: November 9, 1994 (59 FR

55881).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated January 27,

1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania

19464.

Power Authority of the State of New

York, Docket No. 50–333, James A.

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of application for amendment:

September 28, 1993.

Brief description of amendment: The

amendment revised Technical

Specification (TS) Section 4.11.D to

change the surveillance requirements

for the Emergency Service Water System

pumps. The change added pump flow

rate requirements and tests the pumps

in accordance with the licensee’s

Inservice Testing Program. The

respective TS Bases were also revised.

Date of issuance: January 30, 1995.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 223.

Facility Operating License No. DPR–

59: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: November 24, 1993 (58 FR

62156).
Federal Register / Vol. 60, No. 31 / Wednesday, February 15, 1995 / Notices 8765

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated January 30,

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.

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:

January 3, 1995 (TS 95–01).

Brief description of amendments: The

amendments add a permissive statement

to Surveillance Requirement 4.9.7.1 that

will allow the auxiliary building bridge

crane interlocks and physical stops to be

defeated during implementation of the

spent fuel pool storage capacity increase

modification.

Date of issuance: January 24, 1995.

Effective date: January 24, 1995.

Amendment Nos.: 194 and 185.

Facility Operating License Nos. DPR–

77 and DPR–79: Amendments revise the

technical specifications.

Date of initial notice in Federal

Register: January 9, 1995 (60 FR 2404)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated January 24, 1995.

No significant hazards consideration

comments received: None.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402.

Notice of Issuance of Amendments to

Facility Operating Licenses and Final

Determination of No Significant

Hazards Consideration amd

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

March 17, 1995, the licensee may file a

request for a hearing with respect to

issuance of the amendment to the

subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted
8766 Federal Register / Vol. 60, No. 31 / Wednesday, February 15, 1995 / Notices

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses. Since the Commission has

made a final determination that the

amendment involves no significant

hazards consideration, if a hearing is

requested, it will not stay the

effectiveness of the amendment. Any

hearing held would take place while the

amendment is in effect.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555, 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).

Nebraska Public Power District, Docket

No. 50–298, Cooper Nuclear Station,

Nemaha County, Nebraska

Date of amendment request: July 26,

1994, as supplemented by letters dated

December 27, 1994, and January 27,

1995.

Brief description of amendment: The

amendment changed the Technical

Specification Section 3/4.12.A to allow

for increased flow capacity of the

control room emergency filter system.

By increasing the maximum allowed

makeup capacity of this system,

additional margin is provided for the

positive pressurization of the control

room envelope.

Date of issuance: January 27, 1995.

Effective date: January 27, 1995.

Amendment No.: 167.

Facility Operating License No. DPR–

46. Amendment revised the Technical

Specifications.

Public comments requested as to

proposed no significant hazards

consideration: No.

The Commission’s related evaluation

of the amendment, finding of emergency

circumstances, and final determination

of no significant hazards consideration

are contained in a Safety Evaluation

dated

Local Public Document Room

location: Auburn Public Library, 118

15th Street, Auburn, Nebraska 68305.

Attorney for licensee: Mr. G.D.

Watson, Nebraska Public Power District,

Post Office Box 499, Columbus,

Nebraska 68602–0499.

NRC Project Director: William D.

Beckner.

Dated at Rockville, Maryland, this 8th day

of February 1995.

For the Nuclear Regulatory Commission.

Elinor G. Adensam,

Deputy Director, Division of Reactor

Projects—III/IV, Office of Nuclear Reactor

Regulation.

[FR Doc. 95–3629 Filed 2–14–95; 8:45 am]

BILLING CODE 7590–01–P

SECURITIES AND EXCHANGE

COMMISSION

[Rel. No. IC–20893; 811–3095]

Pacific American Fund; Notice of

Application

February 9, 1995.

AGENCY: Securities and Exchange

Commission (‘‘SEC’’).

ACTION: Notice of application for

deregistration under the Investment

Company Act of 1940 (the ‘‘Act’’).

APPLICANT: Pacific American Fund.

RELEVANT ACT SECTION: Section 8(f).

SUMMARY OF APPLICATION: Applicant

seeks an order declaring that it has

ceased to be an investment company.

FILING DATES: The application on Form

N–8F was filed on January 11, 1995.

HEARING OR NOTIFICATION OF HEARING: An

order granting the application will be

issued unless the SEC orders a hearing.

Interested persons may request a

hearing by writing to the SEC’s

Secretary and serving applicant with a

copy of the request, personally or by

mail. Hearing requests should be

received by the SEC by 5:30 p.m. on

March 6, 1995, and should be

accompanied by proof of service on the

applicant, in the form of an affidavit or,

for lawyers, a certificate of service.

Hearing requests should state the nature

of the writer’s interest, the reason for the

request, and the issues contested.

Persons who wish to be notified of a