24904 Federal Register / Vol. 60, No. 90 / Wednesday, May 10, 1995 / Notices

Agreement (the Guaranteed Amount) is

the Receivership Date Accumulated

Book Value of the GIC, which is

$1,034,447.59, less the sum of GIC

Proceeds (cash proceeds actually

received by the Plan from Confederation

Life or any other entity making payment

with respect to Confederation Life’s

obligations under the terms of the GIC,

or from the sale or transfer of the GIC

to unrelated third parties) and Advances

under the Agreement as described

below, plus interest on the net of the

foregoing amount after the Receivership

date at the Contract Rate of 7.15 percent.

The Advances: On the monthly

occasions when the Employer, as Plan

administrator, would otherwise request

a withdrawal from the GIC to fund

Withdrawal Events with respect to

Account balances invested in the GIC,

the Employer will instead notify the

Trustee of the requested withdrawal

amount. The Trustee will then

determine whether it can satisfy the

withdrawal request by using the assets

in the G.I. Fund other than the GIC. If

the Trustee determines that the funds

available from the G.I. Fund are

insufficient to honor the withdrawal

request, the Trustee will determine the

amount of additional funds necessary to

honor the withdrawal request, and the

Employer will make an Advance in that

amount to the Plan. Valuation of the

Account balances invested in the GIC

for purposes of the Advances will be

based on the Guaranteed Amount as

described above.

Final Advance: The Agreement

provides for a final Advance after the

completion of the Receivership. After

the Trustee has determined that the Plan

will not receive any further proceeds

from Confederation Life or its

successors with respect to the GIC, the

Employer shall make a final Advance to

the Plan in the amount necessary to

enable the Plan’s recovery of the

Guaranteed Amount. In the event the

Receivership extends beyond the year

2000, the Employer will make the final

Advance on the first business day in the

year 2001 in the amount required on

such date to enable the Plan to recover

the Guaranteed Amount.

The Repayments: The Agreement

provides that the Repayments of the

Advances are restricted to the principal

amounts of the Advances, and the Plan

will pay no interest and will incur no

expenses with respect to the Advances.

The Repayments may be made only

from the GIC Proceeds received by the

Plan. No other Plan assets will be

available for the Repayments. If the GIC

Proceeds are not sufficient to repay fully

the Advances, the Agreement provides

that the Employer will have no recourse

against the Plan, or against any

participants or beneficiaries of the Plan,

for the unpaid amount. To the extent the

Plan receives GIC Proceeds in excess of

the total amount of the Advances, such

additional amounts will be retained by

the Plan and allocated among the

Accounts invested in the G.I. Fund.

6. In summary, the applicant

represents that the proposed transaction

satisfies the criteria of section 408(a) of

the Act for the following reasons: (1)

The Advances enable the Plan to resume

the full funding of the Withdrawal

Events; (2) The Advances will protect

the Plan’s investment in the GIC and

will ensure that the Plan will recover all

amounts due under the terms of the GIC;

(3) The Plan will pay no interest or

incur any expenses with respect to the

Advances; (4) Repayment of the

Advances will be made only from GIC

Proceeds and no other Plan assets will

be involved in the transactions; (5)

Repayment of the Advances will be

waived to the extent the Plan recoups

less from the GIC Payors than the total

amount of the Advances; and (6) In the

event the Plan receives GIC Proceeds in

excess of the Guaranteed Amount, such

amounts will be retained by the Plan

and allocated among the Accounts.

FOR FURTHER INFORMATION CONTACT:

Ronald Willett of the Department (202)

219–8881. (This is not a toll-free

number.)

General Information

The attention of interested persons is

directed to the following:

(1) The fact that a transaction is the

subject of an exemption under section

408(a) of the Act and/or section

4975(c)(2) of the Code does not relieve

a fiduciary or other party in interest of

disqualified person from certain other

provisions of the Act and/or the Code,

including any prohibited transaction

provisions to which the exemption does

not apply and the general fiduciary

responsibility provisions of section 404

of the Act, which among other things

require a fiduciary to discharge his

duties respecting the plan solely in the

interest of the participants and

beneficiaries of the plan and in a

prudent fashion in accordance with

section 404(a)(1)(b) of the act; nor does

it affect the requirement of section

401(a) of the Code that the plan must

operate for the exclusive benefit of the

employees of the employer maintaining

the plan and their beneficiaries;

(2) Before an exemption may be

granted under section 408(a) of the Act

and/or section 4975(c)(2) of the Code,

the Department must find that the

exemption is administratively feasible,

in the interests of the plan and of its

participants and beneficiaries and

protective of the rights of participants

and beneficiaries of the plan;

(3) The proposed exemptions, if

granted, will be supplemental to, and

not in derogation of, any other

provisions of the Act and/or the Code,

including statutory or administrative

exemptions and transitional rules.

Furthermore, the fact that a transaction

is subject to an administrative or

statutory exemption is not dispositive of

whether the transaction is in fact a

prohibited transaction; and

(4) The proposed exemptions, if

granted, will be subject to the express

condition that the material facts and

representations contained in each

application are true and complete, and

that each application accurately

describes all material terms of the

transaction which is the subject of the

exemption.

Ivan Strasfel,

Director of Exemption Determinations,

Pension and Welfare Benefits Administration,

U.S. Department of Labor.

[FR Doc. 95–11536 Filed 5–9–95; 8:45 am]

BILLING CODE 4510–29–P

UNITED STATES NUCLEAR

REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to

Facility Operating LicensesInvolving

No Significant Hazards Considerations

I. Background

Pursuant to Public Law 97-415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 97-415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from April 17,

1995, through April 28, 1995. The last

biweekly notice was published on April

26, 1995.
Federal Register / Vol. 60, No. 90 / Wednesday, May 10, 1995 / Notices 24905

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 June 9, 1995, the licensee may file

a request for a hearing with respect to

issuance of the amendment to the

subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public
24906 Federal Register / Vol. 60, No. 90 / Wednesday, May 10, 1995 / Notices

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: April 5,

1995

Description of amendment request:

The licensee proposes to revise

Technical Specification (TS) 3/4.9,

Refueling Operations, to be consistent

with NUREG-1431, Standard Technical

Specifications, Westinghouse Plants,

and to relocate the applicable sections

from the TS that do not meet the

Commission’s screening criteria for

retention.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

This change does not involve a significant

hazards consideration for the following

reasons:

The proposed amendment does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes will have no

significant impact on the safety, reliability, or

operation of fuel handling equipment or

activities. These changes will simplify the

Technical Specifications and implement the

recommendations of the Commission’s Final

Policy Statement on Technical Specification

Improvements based upon the assumptions

and analyses contained in the bases of

NUREG-1431. Those elements that involve

relocations to plant procedures are

administrative in nature and do not involve

any modifications to plant equipment or

operation. 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 changes do not introduce

any new equipment or require existing

equipment to operate to perform a function

different from that previously evaluated in

the Final Safety Analysis Report or Technical

Specifications. The changes are consistent

with the new Standard Techical

Specification and assumptions contained in

NUREG-1431 and in the Commission’s Final

Policy Statement on Technical Specification

Improvements. Therefore, the proposed

changes would not increase the possibility of

a new or different type of accident from any

accident previously evaluated.

3. The proposed amendment does not

involve a significant reduction in the margin

of safety.

The proposed changes do not affect any of

the parameters which relate to the margin of

safety as described in the [Bases] of the

Technical Specifications or the Final Safety

Analysis Report. Accordingly, NRC

Acceptance Limits are not affected by these

changes. For those specifications being

relocated to other plant documents, these

changes are purely administrative. 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: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh,

North Carolina 27605

Attorney for licensee: R. E. Jones,

General Counsel, Carolina Power &

Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602

NRC Project Director: David B.

Matthews

Commonwealth Edison Company,

Docket Nos. 50-237 and 50-249,

Dresden Nuclear Power Station, Units 2

and 3, Grundy County, Illinois, Docket

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

Nuclear Power Station, Units 1 and 2,

Rock Island County, Illinois

Date of application for amendment

request: September 15, 1992, as

supplemented April 21, 1995

Description of amendment request: As

a result of findings by a Diagnostic

Evaluation Team inspection performed

by the NRC staff at the Dresden Nuclear

Power Station in 1987, Commonwealth

Edison Company (ComEd, the licensee)

made a decision that both the Dresden

Nuclear Power Station and sister site

Quad Cities Nuclear Power Station,

needed attention focused on the existing

custom Technical Specifications (TSs).

The licensee made the decision to

initiate a Technical Specification

Upgrade Program (TSUP) for both

Dresden and Quad Cities. The licensee

evaluated the current TSs for both

Dresden and Quad Cities against the

Standard Technical Specifications

(STSs) contained in NUREG-0123,

‘‘Standard Technical Specifications

General Electric Plants BWR/4.’’ The

licensee’s evaluation identified

numerous potential improvements such

as clarifying requirements, changing TSs

to make them more understandable and

to eliminate interpretation, and deleting

requirements that are no longer

considered current with industry

practice. As a result of the evaluation,

ComEd has elected to upgrade both the

Dresden and Quad Cities TSs to the

STSs contained in NUREG-0123.

The TSUP for Dresden and Quad

Cities is not a complete adaption of the

STSs. The TSUP focuses on (1)

integrating additional information such

as equipment operability requirements

during shutdown conditions, (2)

clarifying requirements such as limiting

conditions for operations and action

statements utilizing STS terminology,

(3) deleting superseded requirements

and modifications to the TSs based on

the licensee’s responses to Generic

Letters (GLs), and (4) relocating specific

items to more appropriate TS locations.

The application dated September 15,

1992, as supplemented April 21, 1995,

proposed to upgrade only Sections 2.0

(Safety Limits and Limiting Safety

System Settings), 3/4.11 (Power

Distribution Limits), and 3/4.12 (Special

Test Exceptions) of the Dresden and

Quad Cities TS.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the
Federal Register / Vol. 60, No. 90 / Wednesday, May 10, 1995 / Notices 24907

issue of no significant hazards

consideration, which is presented

below:

Section 2.0

The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated because:

The proposed changes to Specifications 1/

2.1 and 1/2.2 to delete the present

Applicability and Objective sections

represent administrative changes to format

and presentation of material. The proposed

changes provide the user with a format that

will allow better access to needed

information and provides concise Safety

Limit, Limiting Safety System Settings,

Applicability and Action requirements. The

additions of Applicability and Action

requirements represent clarification of

intended requirements that do not presently

state all required conditions of operability or

provide clearly stated Action statements if

the requirements are not met. The combining

of the two sections and added requirements

follow STS guidelines that are in use at many

operating BWRs with similar design and

operating configurations as Dresden and

Quad Cities Stations. Operability

requirements for Safety Limits have been

chosen to reflect only those Operational

Modes where the Safety Limits apply.

Operability requirements for Limiting Safety

System Settings are already stated in other

sections of the Technical Specifications, thus

reference to the appropriate operability

requirement is made rather than repeating

the requirement in the Limiting Safety

System Setting Specification.

Deletion of the Power Transient Safety

Limit does not impact any safety analyses.

The safety analyses assume the Reactor

Protection System (RPS) operates as designed

and the reactor scrams when the neutron flux

exceeds the limiting safety system setting.

The proposed Technical Specifications will

continue to provide a highly reliable system

to operate as assumed in the safety analyses.

Therefore, this change does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The reactor water level low scram setpoint

is changed (for Quad Cities) to be consistent

with other reactor water level setpoints in the

Technical Specifications and the STS. The

setpoint is equivalent to the current

requirement but is expressed as the reactor

water level above the top of active fuel.

The scram discharge volume scram level is

converted for Dresden Unit 2 and Unit 3 to

gallons to be consistent with the Quad Cities

Units. The proposed setpoints are consistent

with the current specifications. The change

in the units does not represent a change in

the physical setpoint.

The proposed change to delete the APRM

Downnscale Scram trip function for Quad

Cities has been evaluated by Commonwealth

Edison and General Electric and previously

approved for Dresden Station. The events of

concern with respect to the APRM/IRM

companion trip are the Control Rod Drop

Accident and the low power Rod Withdrawal

Error. The FSAR and reload safety analyses

do not credit this scram function in the

termination of either of these events. Since

this scram function is not credited in the

termination of these events, the elimination

of this scram function has no adverse effect

on previously evaluated accidents.

The change to the low condenser vacuum

scram setpoint from 23 inches Hg to 21

inches of Hg is consistent with an identical

change made to Quad Cities Units 1 and 2.

The low condenser vacuum scram is an

anticipatory scram and is not credited in any

transient analysis. Thus the reduction in the

setpoint will not affect any transient analysis.

The proposed changes do not alter the

intent of existing setpoints or accident

assumptions and follow existing

requirements at other operating BWRs for

operability and Action statements. Therefore,

the proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes do not create the

possibility of a new or different kind of

accident from any previously evaluated

because:

The proposed administrative changes to

the format and arrangement of material do

not affect technical requirements or

assumptions of any potential accident and;

therefore, cannot create the possibility of a

new or different kind of accident from any

previously evaluated.

The proposed addition of Applicability

and Action requirements enhance the

understanding and usability of the Technical

Specifications and thus represent an

improvement over present specifications.

New requirements are modeled after those in

use at operating BWRs and do not represent

requirements that will adversely affect

potential accident analyses or assumptions.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any previously

evaluated.

Deletion of the Power Transient Safety

Limit does not involve a change in the design

or operation of any systems assumed to

operate in the safety analyses. Therefore, this

change does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

The change in the units for the Reactor

Water Level scram function do not change

any physical plant setpoints. The setpoint

will remain the same but will be expressed

as the level above the top of active fuel. The

change does not create the possibility of a

new or different kind of accident.

The conversion of the Scram Discharge

Volume scram setpoint from inches to

gallons does not alter any physical plant

setpoints. The setpoint will remain the same

but will be expressed in gallons rather than

inches. The change will provide consistency

between Dresden and Quad Cities.

The deletion of the APRM Downscale

Scram Trip Function does not introduce any

new accident. The limiting accidents, Control

Rod Drop, Rod Withdrawal Error, in the

operating region of transition between the

Startup and Run Operational Modes are well

understood and are evaluated in FSAR and

reload analyses. Other control rod initiated

events which are less limiting in this region

are subsets of the low power Rod Withdrawal

Error event and are bounded by it and the

design basis Control Rod Drop Accident.

General Electric has indicated that, for

reactivity insertion mechanisms at very low

power, the only effect of the deletion of the

APRM downscale scram would be that the

initial power level could be a few percent

lower which would not have a significant

effect on the severity of the event. In

addition, proper overlap between the IRMs

and APRMs is not affected since the

calibration requirements are not being

changed.

The change in the low condenser vacuum

scram function will not create the possibility

of a new or different kind of accident because

the function is not recognized in any of the

transient analysis. The low condenser

vacuum scram function is an anticipatory

scram.

The proposed changes do not involve a

significant reduction in the margin of safety

because:

The proposed administrative changes to

format, arrangement of material, clarification

of requirements and other non-technical

changes do not affect any safety aspects of

the plant and as such can not involve a

significant reduction in the margin of safety.

The proposed Applicability statements

require availability of Safety Limits and

Limiting Safety System Settings when

required to perform their respective

functions. Proposed Actions for Safety Limits

allow only 2 hours to be in Hot Shutdown

and then reference Specification 6.4 to

ensure that proper reports are made and

restart is prohibited until approved by the

NRC. These provisions help ensure that

present margins are not significantly

reduced.

Deletion of the Power Transient Safety

Limit does not impact the margin assumed in

the safety analyses. The safety analyses

assume the RPS operates as designed and the

reactor scrams when the neutron flux

exceeds the limiting safety system setting.

The margins assumed in the design of the

RPS and in the safety and transient analyses

calculations have not been revised.

Therefore, this change does not involve a

significant reduction in the margin of safety.

The change in units to the Reactor Water

Level scram setpoint and the Scram

Discharge Volume scram setpoint do not

involve a significant reduction in the margin

of safety because the changes do not

represent a change in the physical setpoints.

The reduction in the Low Condenser

Vacuum scram setpoint does not represent a

reduction in the margin of safety because the

scram is not credited in any transient

analysis.

The APRM Downscale Scram Trip

Function is not credited in the termination of

any FSAR or reload safety analysis event. As

such, the elimination of this scram function

has no effect on any margin of safety.

Section 3/4.11

The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated because:

In general, the proposed changes represent

the conversion of current requirements to a
24908 Federal Register / Vol. 60, No. 90 / Wednesday, May 10, 1995 / Notices

more generic format, or the addition of

requirements which are based on the current

safety analysis. Implementation of these

changes will provide increased reliability of

equipment assumed to operate in the current

safety analysis, or provide continued

assurance that specified parameters remain

within their acceptance limits, and as such,

will not significantly increase the probability

or consequences of a previously evaluated

accident.

Some of the proposed changes represent

minor curtailments of the current

requirements which are based on generic

guidance or previously approved provisions

for other stations. These proposed changes

are consistent with the current safety

analyses and have been previously

determined to represent sufficient

requirements for the assurance of reliability

of equipment assumed to operate in the

safety analysis, or provide continued

assurance that specified parameters remain

within their acceptance limits. As such, these

changes will not significantly increase the

probability or consequences of a previously

evaluated accident.

The Generic Changes to the technical

specifications involve administrative changes

to format and arrangement of the material. As

such, these changes cannot involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The current specifications require the

reactor to be placed in cold shutdown when

a thermal limit was exceeded and not

restored within the allotted 2 hours, but the

proposed specifications require the reactor to

be less than 25% of rated thermal power if

this condition occurred. The change

eliminates a shutdown and requires the

power level to be reduced to the point that

the limits are no longer applicable.

Therefore, the change will not increase the

probability or consequences of an accident.

Create the possibility of a new or different

kind of accident from any previously

evaluated because:

In general, the proposed changes represent

the conversion of current requirements to a

more generic format, or the addition of

requirements which are based on the current

safety analysis. Others represent minor

curtailments of the current requirements

which are based on generic guidance or

previously approved provisions for other

stations. These changes do not involve

revisions to the design of the station. Some

of the changes may involve revision in the

operation of the stations; however, these

changes provide additional restrictions

which are in accordance with the current

safety analyses, or are to provide for

additional testing or surveillance which will

not introduce new failure mechanisms

beyond those already considered in the

current safety analyses. Therefore, these

changes will not create the possibility of a

new or different kind of accident from any

accident previously evaluated.

Since the Generic Changes proposed to the

technical specifications are administrative in

nature, they cannot create the possibility of

a new or different kind of accident from any

previously evaluated.

The requirement to reduce thermal power

to less than 25% of rated thermal power

rather than place the reactor in cold

shutdown will not create a new or different

kind of accident because the thermal limits

are not required in operational mode 1 when

thermal power is less than 25% of rated

power.

Involve a significant reduction in the

margin of safety because:

In general, the proposed changes represent

the conversion of current requirements to a

more generic format, or the addition of

requirements which are based on the current

safety analysis. Others represent minor

curtailments of the current requirements

which are based on generic guidance or

previously approved provisions for other

stations. Some of the latter individual items

may introduce minor reductions in the

margin of safety when compared to the

current requirements. However, other

individual changes are the adoption of new

requirements which will provide significant

enhancement of the reliability of the

equipment assumed to operate in the safety

analysis, or provide enhanced assurance that

specified parameters remain within their

acceptance limits. These enhancements

compensate for the individual minor

reductions, such that taken together, the

proposed changes will not significantly

reduce the margin of safety.

The Generic Changes proposed in this

amendment request are administrative in

nature and, as such, do not involve a

reduction in the margin of safety.

Section 3/4.12

The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated because:

The proposed Specification 3/4.12 is a new

section which will provide the user with a

format that will allow better access to needed

information and provide concise

Applicability and Action requirements. The

additions of Applicability and Action

requirements represent classification of

intended requirements that do not presently

state all required conditions of operability or

provide clearly stated Action statements if

the requirements are not met. The combining

of the two sections and the added

requirements follow Standard Technical

Specifications (STS) guidelines that are in

use at many operating BWRs with similar

design and operating configurations as

Dresden and Quad Cities Stations.

The proposed Section 3/4.12 involves the

relocation of present requirements into one

section identical to STS provisions. The

changes also implement the Applicability

and Action provisions of the STS and later

operating BWR plants that have been

evaluated and found acceptable for use at

Dresden and Quad Cities. Present

Surveillance Requirements are replaced,

where applicable, with proven STS

guidelines that are being used at plants with

a system similar to that at Dresden and Quad

Cities. The changes in the present

Surveillance Requirements add testing

requirements that are not presently in the

Dresden and Quad Cities technical

specifications. The proposed changes do not

affect accident assumptions other than a

minor increase in the initial power level

(approximately 0.2% to 1%) and as such, do

not involve a significant increase in the

probability of an accident previously

evaluated. The proposed specifications add

additional requirements to specifications

currently contained in the Technical

Specifications. Since the proposed changes to

the Technical Specifications implement

requirements that have been demonstrated to

provide acceptable operability provisions at

other facilities with a design similar to that

at Dresden and Quad Cities, the proposed

changes do not significantly increase the

consequences of an accident previously

evaluated.

The proposed changes do not create the

possibility of a new or different kind of

accident from any previously evaluated

because:

The proposed administrative changes to

the format and arrangement of material do

not affect technical requirements or

assumptions of any potential accident and;

therefore, cannot create the possibility of a

new or different kind of accident from any

previously evaluated.

The proposed addition of Applicability

and Action requirements enhance the

understanding and usability of the Technical

Specifications and thus represent an

improvement over present specifications.

New requirements are modeled after those in

use at operating BWRs and do not represent

requirements that will adversely affect

potential accident analyses or assumptions.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any previously

evaluated.

The proposed changes do not involve a

significant reduction in the margin of safety

because:

The proposed administrative changes to

format, arrangement of material, clarification

of requirements and other non technical

changes do not affect any safety aspects of

the plant and as such can not involve a

significant reduction in the margin of safety.

In addition, the commission has provided

guidance concerning the application of

standards for determining whether

significant hazards consideration exists by

providing certain examples (51 FR 7751) of

amendments that are considered not likely to

involve significant hazards considerations.

Commonwealth Edison has reviewed the

proposed changes against these examples and

believes that the proposed changes fall

within the scope of example (ii) ‘‘a change

that constitutes an additional limitation,

restriction, or control not presently included

in the technical specifications’’.

The proposed amendment does not involve

a significant relaxation of the criteria used to

establish safety limits, a significant relaxation

of the bases for the limiting safety system

settings or a significant relaxation of the

bases for the limiting conditions for

operations. Therefore, based on the guidance

provided in the Federal Register and the

criteria established in 10 CFR 50.92(c), the

proposed change does not constitute a

significant hazards consideration.

The NRC staff has reviewed the

licensee’s analysis and, based on this
Federal Register / Vol. 60, No. 90 / Wednesday, May 10, 1995 / Notices 24909

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: for Dresden, Morris Area

Public Library District, 604 Liberty

Street, Morris, Illinois 60450; for Quad

Cities, Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois 61021

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60603

NRC Project Director: Robert A. Capra

Commonwealth Edison Company,

Docket Nos. 50-237 and 50-249,

Dresden Nuclear Power Station, Units 2

and 3, Grundy County, IllinoisDocket

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

Nuclear Power Station, Units 1 and 2,

Rock Island County, Illinois

Date of application for amendment

request: December 15, 1993, as

supplemented by letter dated April 21,

1995

Description of amendment request: As

a result of findings by a Diagnostic

Evaluation Team inspection performed

by the NRC staff at the Dresden Nuclear

Power Station in 1987, Commonwealth

Edison Company (ComEd, the licensee)

made a decision that both the Dresden

Nuclear Power Station and sister site

Quad Cities Nuclear Power Station,

needed attention focused on the existing

custom Technical Specifications (TSs)

used.

The licensee made the decision to

initiate a Technical Specification

Upgrade Program (TSUP) for both

Dresden and Quad Cities. The licensee

evaluated the current TSs for both

Dresden and Quad Cities against the

Standard Technical Specifications

(STSs) contained in NUREG-0123,

‘‘Standard Technical Specifications

General Electric Plants BWR/4.’’ The

licensee’s evaluation identified

numerous potential improvements such

as clarifying requirements, changing TSs

to make them more understandable and

to eliminate interpretation, and deleting

requirements that are no longer

considered current with industry

practice. As a result of the evaluation,

ComEd has elected to upgrade both the

Dresden and Quad Cities TSs to the

STSs contained in NUREG-0123.

The TSUP for Dresden and Quad

Cities is not a complete adaption of the

STSs. The TSUP focuses on (1)

integrating additional information such

as equipment operability requirements

during shutdown conditions, (2)

clarifying requirements such as limiting

conditions for operations and action

statements utilizing STS terminology,

(3) deleting superseded requirements

and modifications to the TSs based on

the licensee’s responses to Generic

Letters (GLs), and (4) relocating specific

items to more appropriate TS locations.

The December 15, 1993, and April 21,

1995, applications proposed to upgrade

only Section 5.0 (Design Features) of the

Dresden and Quad Cities TSs.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated because:

In general, the proposed amendment

represents the conversion of current

requirements to a more generic format, or the

addition of requirements which are based on

the current safety analysis. Implementation

of these changes will provide continued

assurance that specified [parameters remain]

within their acceptance limits, and as such,

will not significantly increase the probability

or consequences of a previously evaluated

accident. Some of the proposed changes to

the current Technical Specifications (CTS)

represent minor curtailments of the current

requirements which are based on generic

guidance or previously approved provisions

for other stations. The proposed amendment

for current Dresden and Quad Cities Station’s

Technical Specifications Section 5.0

represent a minor relaxation of the current

requirements, and is based on BWR-STS

(NUREG-0123) guidelines or later operating

BWR plant’s NRC accepted changes. The

proposed changes are consistent with the

current safety analyses and have been

previously determined to represent sufficient

requirements for the assurance and reliability

of equipment assumed to operate in the

safety analysis. Any deviations from CTS or

STS requirements do not significantly

increase the probability or consequences of

any previously evaluated accidents for

Dresden or Quad Cities Stations.

Details describing the plant’s design are

presented in TSUP Section 5.0. There are no

Limiting Conditions for Operation (LCO) or

Surveillance Requirements (SR)

encompassed within TSUP Section 5.0. This

information is administrative in nature and

consistent to the UFSAR; therefore, the

probability of any accident previously

evaluated is not increased by the proposed

amendment.

Create the possibility of a new or different

kind of accident from any previously

evaluated because:

In general, the proposed amendment

represents the conversion of current

requirements to a more generic format, or the

addition of requirements which are based on

the current safety analysis. Others represent

minor relaxations of the current requirements

which are based on generic guidance or

previously approved provisions for other

stations. These changes do not involve

revisions to the design of the station. The

proposed changes are administrative in

nature and do not involve a revision in the

operation of the station. As such, there are no

changes to the current safety analysis.

Therefore, the proposed changes will not

introduce new failure mechanisms beyond

those already considered in the current safety

analyses.

The proposed amendment for Dresden and

Quad Cities Station’s Technical

Specifications Section 5.0 is based on BWRSTS guidelines or later operating BWR

plants’ NRC accepted changes. The proposed

amendment has been reviewed for

acceptability at the Dresden or Quad Cities

Nuclear Power Stations considering

similarity of system or component design

versus the BWR-STS or later operating BWRs.

Any deviations from CTS or BWR-STS

requirements do not create the possibility of

a new or different kind of accident

previously evaluated for Dresden and Quad

Cities Stations. No new modes of operation

are introduced by the proposed changes. The

proposed changes maintain at least the

present level of operability, and in some

cases are more conservative. Therefore, the

proposed changes do not create the

possibility of a new or different kind of

accident from any previously evaluated.

Involve a significant reduction in the

margin of safety because:

In general, the proposed amendment

represents the conversion of current

requirements to a more generic format, or the

addition of requirements which are based on

the current safety analysis. Others represent

minor curtailments of the current

requirements which are based on generic

guidance or previously approved provisions

for other stations. The proposed amendment

to Technical Specification Section 5.0

implements present requirements, or the

intent of present requirements in accordance

with the guidelines set forth in the STS. Any

deviations from CTS or BWR-STS

requirements do not significantly reduce the

margin of safety for Dresden or Quad Cities

Stations. These changes do not involve

revisions to the design of the station. The

proposed changes are administrative in

nature and do not involve a revision in the

operation of the station. As such, there are no

changes to the current safety analysis.

Therefore, the proposed changes will not

introduce new failure mechanisms beyond

those already considered in the current safety

analyses. Therefore, because the proposed

changes are administrative in nature, do not

involve a revision in the operation of the

station and maintains the current design

requirements specified in the UFSAR, 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.
24910 Federal Register / Vol. 60, No. 90 / Wednesday, May 10, 1995 / Notices

Local Public Document Room

location: For Dresden, Morris Area

Public Library District, 604 Liberty

Street, Morris, Illinois 60450; for Quad

Cities, Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois 61021

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60603

NRC Project Director: Robert A. Capra

Consumers Power Company, Docket

No. 50-255, Palisades Plant, Van Buren

County, Michigan

Date of amendment request:

December 13, 1994

Description of amendment request:

The proposed amendment would revise

the Palisades’ technical specifications

(TSs) to add a high thermal performance

(HTP) departure from nucleate boiling

correlation to Safety Limit 2.1. The HTP

correlation is used for the high thermal

performance fuel loaded during recent

fuel cycles.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change to the TS adds the

HTP critical heat flux correlation to the

Safety Limit - Reactor Core Section 2.1. The

HTP correlation is an NRC approved

methodology for a Departure from Nucleate

Boiling (DNB) Correlation for high thermal

performance (HTP) fuel as is used at

Palisades. The HTP correlation is an

extension of the currently approved ANFP

correlation. There are no associated changes

in plant operation. Palisades fuel loaded in

cycle 9 and later meet the requirements of the

HTP correlation. Therefore, operation of the

facility in accordance with the proposed TS

would not result in a significant increase in

the probability or consequences of an

accident previously evaluated.

2. Create the possibility of a new or

different kind of accident from any

previously evaluated.

The HTP correlation will allow for more

accurate DNB predictions within the

applicable operating conditions for fuels with

the HTP design used at Palisades. There are

no changes in plant operation. Therefore

operation of the facility in accordance with

the proposed TS would not create the

possibility of a new or different kind of

accident from any previously evaluated.

3. Involve a significant reduction in a

margin of safety.

As stated previously, the HTP correlation

will allow for more accurate DNB predictions

within the applicable operating conditions

for fuel with the HTP design. There are no

associated changes in plant operation.

Therefore, operation of the facility in

accordance with the proposed TS would 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: Van Wylen Library, Hope

College, Holland, Michigan 49423.

Attorney for licensee: Judd L. Bacon,

Esquire, Consumers Power Company,

212 West Michigan Avenue, Jackson,

Michigan 49201

NRC Project Director: Cynthia A.

Carpenter, Acting

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: January

18, 1995

Description of amendment request:

The proposed amendments would

relocate the requirements for the seismic

instrumentation, meteorological

instrumentation, and loose-part

detection system from the Technical

Specifications to the Selected Licensee

Commitment (SCL) Manual. This will

allow future changes to these controls to

be performed under the provisions of 10

CFR 50.59. No changes are being made

to the technical content of the affected

Technical Specification pages.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Criterion 1

The requested amendments will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated. Relocation of the

affected TS sections to the SLC Manual will

have no effect on the probability of any

accident occurring. In addition, the

consequences of an accident will not be

impacted since the above instrumentation

will continue to be utilized in the same

manner as before. No impact on the plant

response to accidents will be created.

Criterion 2

The requested amendments will not create

the possibility of a new or different kind of

accident from any accident previously

evaluated. No new accident causal

mechanisms will be created as a result of

relocating the affected TS requirements to the

SLC Manual. Plant operation will not be

affected by the proposed amendments and no

new failure modes will be created.

Criterion 3

The requested amendments will not

involve a significant reduction in a margin of

safety. No impact upon any plant safety

margins will be created. Relocation of the

affected TS requirements to the SLC Manual

is consistent with the content of the

Westinghouse RSTS [Revised Standard

Technical Specifications], as the NRC did not

require technical specification controls for

the affected instrumentation in the RSTS.

The proposed amendments are consistent

with the NRC philosophy of encouraging

utilities to propose amendments that are

consistent with the content of the RSTS.

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

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

Florida Power and Light Company, et

al., Docket Nos. 50-335 and 50-389, St.

Lucie Plant, Unit Nos. 1 and 2, St. Lucie

County, Florida

Date of amendment request: April 3,

1995

Description of amendment request:

The amendments will incorporate lineitem TS improvements to Specifications

3/4.8.1 ‘‘Electrical Power Systems-A.C.

Sources,’’ and 4.8.1.2.2 ‘‘Electrical

Power Systems-Shutdown.’’ The

proposed changes are consistent with

recommendations for Emergency Diesel

Generator (EDG) Surveillance

Requirements in NUREG-1366, and

regulatory guidance provided in Generic

Letter (GL) 93-05 and GL 94-01. This

proposal also contains FPL’s

commitment to implement a

maintenance program for monitoring

and maintaining EDG performance for

both St. Lucie Units consistent with 10

CFR 50.65 and the guidance of

Regulatory Guide 1.160.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) Operation of the facility in accordance

with the proposed amendment would not
Federal Register / Vol. 60, No. 90 / Wednesday, May 10, 1995 / Notices 24911

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The license amendments proposed for St.

Lucie Units 1 and 2 will incorporate lineitem Technical Specification (TS)

improvements for Emergency Diesel

Generators (EDG) pursuant to guidance

provided in Generic Letters (GL) 93-05 and

94-01. The EDGs are not accident initiators,

the proposed TS changes do not involve any

assumptions relative to accident initiators in

the plant safety analyses, and therefore the

proposed amendments will not impact the

probability of occurrence for accidents

previously analyzed.

The EDG line-item TS improvements

associated with GL 93-05 are based on

recommendations designed to remove

unwarranted requirements for testing during

power operation and other factors that are

counter-productive to safety in terms of

equipment degradation and availability.

These recommendations resulted from a

comprehensive study of industry-wide EDG

surveillance requirements and subsequent

findings reported by the NRC in NUREG1366. The proposed amendments are

consistent with the GL 93-05 guidance for

implementing such recommendations.

Similarly, GL 94-01 provides guidance for

a line-item TS improvement that will remove

accelerated testing requirements from the TS

provided that the licensee commits to a

maintenance program for monitoring and

maintaining EDG performance that includes

the applicable provisions of the maintenance

rule (10 CFR 50.65). Such a program will

further assure EDG availability. Since the

availability of EDGs is assumed in certain

success paths for mitigating analyzed

accidents, an improvement in EDG

availability will enhance accident mitigation

capabilities.

Therefore, operation of the facility in

accordance with the proposed amendments

would not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

(2) Operation of the facility in accordance

with the proposed amendment would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed amendments incorporate

line-item TS improvements to EDG

surveillance testing requirements, and will

not change the physical plant or the modes

of plant operation defined in the Facility

License. The changes do not involve the

addition or modification of equipment, nor

do they alter the design or methods of

operation of plant systems. Plant

configurations that are prohibited by TS will

not be created by the amendments. Therefore,

operation of the facility in accordance with

the proposed amendment would not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

(3) Operation of the facility in accordance

with the proposed amendment would not

involve a significant reduction in a margin of

safety.

The proposed amendments are designed to

improve EDG availability by eliminating

unwarranted surveillance testing. The

presently specified surveillance intervals are

not changed. The proposed changes do not

otherwise alter the basis for any technical

specification that is related to the

establishment of, or the maintenance of a

nuclear safety margin. Therefore, operation of

the facility in accordance with the proposed

amendment would not involve a significant

reduction in a margin of safety.

Based on the above discussion and the

supporting Evaluation of Technical

Specification changes, FPL has determined

that the proposed license amendment

involves no significant hazards

consideration.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort

Pierce, Florida 34954-9003

Attorney for licensee: J. R. Newman,

Esquire, Morgan, Lewis & Bockius, 1800

M Street, NW., Washington, DC 20036

NRC Project Director: David B.

Matthews, Director

Maine Yankee Atomic Power Company,

Docket No. 50-309, Maine Yankee

Atomic Power Station, Lincoln County,

Maine

Date of amendment request: March 7,

1995

Description of amendment request:

The proposed amendment would add an

Exception to Technical Specifications

(TS) 3.6.A and 3.6.C. The Exception

would permit reduced component

cooling water flow for short periods of

time, while component cooling water

heat exchangers are shifted.

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

staff’s review is presented below:

1. The proposed change does not

involve a significant increase in the

probability or consequences of an

accident previously evaluated.

Plant experience shows that the

component cooling water heat

exchangers can be shifted in a few

minutes; well within the time limit for

Remedial Action under this TS 3.6.A or

C, or TS 3.0.A. Thus, the proposed

change does not involve a significant

increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed change does not

create the possibility of a new or

different kind of accident from any

accident previously evaluated.

The proposed change does not affect

equipment reliability when such

equipment is required to be operable.

Existing TS 3.6 and its Remedial Action

statement govern the plant

circumstances under which cooling

water subsystems are required, and

specify the maximum time such

subsystems may be unavailable. The

proposed change does affects neither

operating requirements nor the time

limit on restoring system operability.

3. The proposed change does not

involve a significant reduction in a

margin of safety.

The proposed change does not

significantly alter the availability or

condition of the cooling water

subsystems and, therefore, does not alter

the accident analysis or its associated

conclusions. The proposed change

would permit flow in one component

cooling water train to be reduced below

that required for operation of the

emergency core cooling systems in the

recirculation mode, for a short period of

time. The amount of time that flow is

reduced is small, and full flow

operation can be easily restored within

the time required for design heat load

removal. Thus, there is no significant

reduction in a margin of safety.

Based on this review, it appears that

the three standards of 10 CFR 50.92(c)

are satisfied. Therefore, the NRC staff

proposes to determine that this

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Wiscasset Public Library, High

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

04578

Attorney for licensee: Mary Ann

Lynch, Esquire, Maine Yankee Atomic

Power Company, 329 Bath Road,

Brunswick, ME 04011

NRC Project Director: Phillip F.

McKee

Northeast Nuclear Energy Company

(NNECO), Docket No. 50-245, Millstone

Nuclear Power Station, Unit 1, New

London County, Connecticut

Date of amendment request: April 18,

1995

Description of amendment request:

The proposed amendment would allow

the use of the ANSI/ANS 5.1-1979 decay

heat model for post-loss of coolant

accident containment cooling analysis.

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
24912 Federal Register / Vol. 60, No. 90 / Wednesday, May 10, 1995 / Notices

consideration, which is presented

below:

NNECO has reviewed the proposed change

in accordance with 10CFR50.92 and

concluded that the change does not involve

a significant hazards consideration (SHC).

The basis for this conclusion is that the three

criteria of 10CFR50.92(c) are not

compromised. The proposed change does not

involve an SHC because the change would

not:

1. Involve a significant increase in the

probability or consequences of an accident

previously analyzed.

The change to the decay heat model used

to determine post-accident conditions cannot

affect the probability of any accident. No

changes to plant operation or design would

occur due to the new analysis.

The new model cannot directly affect the

consequences of an accident, since it is the

tool used to predict the temperature effects

of the postulated accident. However, using

the ANSI/ANS 5.1-1979 model could change

the anticipated actions necessary to respond

to an event. Changing the response action

could possibly affect the consequences of an

accident. This model change will not have

such an effect. Operator actions to throttle

LPCI [low pressure coolant injection], CS

[core spray], or ESW [emergency service

water] pump flow are taken based upon

observed conditions, not predetermined data

points from the analysis.

Operability of the emergency core cooling

systems (ECCS) can be shown for

temperatures that are higher than those

predicted by the containment cooling

analysis.

Therefore, the utilization of the ANSI/ANS

5.1-1979 decay heat model does not involve

a significant increase in the probability or

consequences of a previously evaluated

accident.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

The proposed license amendment only

revises the predicted temperature that result

from a postulated accident. There is no

change to the design or operation of any

system or component. Since this change only

deals with the post-accident effects of

currently analyzed accidents, there is no

possibility of creating a new or different kind

of accident.

3. Involve a significant reduction in the

margin of safety.

The early design documentation stated that

the ECCS components were designed for

post-accident torus temperatures of 203°F. As

this issue evolved, NNECO performed

operability determinations which showed

that peak temperatures of 209°F were

acceptable. Utilizing a more accurate decay

heat model which results in lower predicted

peak temperatures demonstrates the

acceptability of the plant design. Therefore,

replacing the May-Witt decay heat model

with the ANSI/ANS 5.1-1979 model does not

result in a decrease in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: 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: March

29, 1995

Description of amendment request:

The proposed amendment changes

Technical Specifications to revise

peaking factor penalties based on NRC

approved methods.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The proposed changes do not involve an

SHC because the changes would not:

1. Involve a significant increase in the

probability or consequences of an accident

previously analyzed.

The proposed changes to the action

statements of Sections 3.2.2.1 and 3.2.2.2 are

purely administrative and therefore they do

not adversely affect the probability or

consequences of an accident previously

analyzed. The proposed changes to

Surveillance Requirements 4.2.2.1.2.e,

4.2.2.1.4.e, 4.2.2.2.2.e and 4.2.2.2.4.e and

Section 6.9.1.6.b are based on the NRC

approved methodology for calculating the

penalty to be applied to FQM(Z). The margin

for the FQRTP limit is still maintained by the

proposed changes. In addition, the penalty is

included in the COLR [Core Operating Limits

Report] which will be maintained and

controlled per the requirements of

10CFR50.59. Therefore, the proposed

changes do not increase 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.

The proposed changes to the Action

Statement of Sections 3.2.2.1 and 3.2.2.2 are

purely administrative and therefore, they do

not create the possibility of a new or different

kind of accident from any previously

analyzed. The proposed changes to

Surveillance Requirements 4.2.2.1.2.e,

4.2.2.1.4.e, 4.2.2.2.2.e, and 4.2.2.2.4.e and

Section 6.9.1.6.b do not create a malfunction

that is different from those previously

evaluated. The changes do not involve

positioning reactivity systems or plant

components into any new configuration or

sequence not previously analyzed. Therefore,

the changes will not create the possibility of

a new or different kind of accident from any

other previously analyzed.

3. Involve a significant reduction in the

margin of safety.

The proposed changes to the action

statements of Sections 3.2.2.1 and 3.2.2.2 are

purely administrative and therefore they will

not reduce the margin of safety. The

proposed changes to Surveillance

Requirements 4.2.2.1.2.e, 4.2.2.1.4.e,

4.2.2.2.2.e and 4.2.2.2.4.e and Section

6.9.1.6.b do not reduce the margin to the

FQRTP limit. The approved methods more

distinctly evaluate the expected changes to

FQM than previously existed. 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

PECO Energy Company, Public Service

Electric and Gas Company, Delmarva

Power and Light Company, and

Atlantic City Electric Company, Docket

No. 50-277, Peach Bottom Atomic

Power Station, Unit No. 2, York County,

Pennsylvania

Date of application for amendment:

March 30, 1995

Description of amendment request:

The proposed change would revise

Technical Specifications Section

4.7.D.1.b.(1) by adding a footnote to

exempt the High Pressure Coolant

Injection [HPCI] motor-operated valve

MO-2-23-015 from quarterly stoke

testing requirements until refueling

outage 2RO11.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or
Federal Register / Vol. 60, No. 90 / Wednesday, May 10, 1995 / Notices 24913

consequences of an accident previously

evaluated.

The proposed change does not serve as an

initiator or contributor to any accidents

previously evaluated. It does not decrease the

effectiveness of equipment relied upon to

mitigate previously evaluated accidents. A

calculation was performed and it has been

determined the leakage through the valve’s

packing will be within the allowable limits

of containment leakage (La). While

positioning the valve in the backseated

position does increase its stroke time, it has

been calculated and demonstrated that the

valve will close within the TS time limit of

20 seconds.

Therefore, the proposed change does not

involve a significant increase in the

probability or consequence of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

The proposed change does not serve as an

initiator or contributor to any of the accidents

previously evaluated. The proposed change

does not introduce any new modes of plant

operation.

Implementation of the proposed changes

will not affect the design function or

configuration of any component or introduce

any new operating scenarios or failure modes

or accident initiation. It does not impair or

prevent safety systems from performing their

safety function.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any previously

evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed change does not serve as an

initiator or contributor to any accidents

evaluated in the [Safety Analysis Report]

SAR. It has no impact on any safety analysis

assumptions. Exempting the HPCI valve MO2-23-015 from quarterly stroke testing until

2RO11 does not impact its reliability or affect

its ability to perform its intended safety

function. The change does not adversely

affect the assumptions or sequence of events

used in any accident analysis.

Therefore, the proposed change does not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Education

Building, Walnut Street and

Commonwealth Avenue, Box 1601,

Harrisburg, Pennsylvania 17105.

Attorney for licensee: J. W. Durham,

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

Counsel, PECO Energy Company, 2301

Market Street, Philadelphia,

Pennsylvania 19101

NRC Project Director: John F. Stolz

PECO Energy Company, Public Service

Electric and Gas Company, Delmarva

Power and Light Company, and

Atlantic City Electric Company,

Dockets Nos. 50-277 and 50-278, Peach

Bottom Atomic Power Station, Units

Nos. 2 and 3, York County,

Pennsylvania

Date of application for amendments:

March 16, 1995

Description of amendment request:

This amendment would change the

existing requirements for the Source

Range Monitors (SRM) while the plant

is in the refueling condition to

requirements based on the Improved

Technical Specifications in NUREG1433, ‘‘Standard Technical Specification

General Electric Plants, BWR/4.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of any accident previously

evaluated.

The proposed changes to the SRM

requirements will not increase the

probability or consequences of an accident

previously evaluated. The SRMs are not

assumed to function during any UFSAR

[Updated Final Safety Analysis Report]

design basis accident or transient analysis.

This TS change will not alter any safety

limits which ensure the integrity of fuel

barriers, and will not result in any increase

to onsite or offsite dose. Additionally,

continued availability of the SRMs in the

refuel mode is ensured through additional

testing requirements being added by this TS

change. The changes to the SRM

requirements will not alter the operation of

equipment assumed to be available for the

mitigation of accidents or transients.

The proposed changes are based on

NUREG-1433, ‘‘Standard Technical

Specifications General Electric Plants, BWR/

4,’’ and are consistent with the PECO Energy

submittal of September 29, 1994, requesting

an overall conversion, based on NUREG1433. The overall conversion to the ITS

[Improved Technical Specifications]

included both technically justified deviations

from the NUREG, and technically justified

changes from the PBAPS current TS.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

The proposed changes to the SRM

requirements will not create the possibility of

a new or different type of accident from any

previously evaluated. The SRMs are not

assumed to function during any analyzed

UFSAR design basis accident or transient

analysis. Additionally, the changes will not

involve any changes to plant systems,

structures or components (SCCs) which

could act as new accident initiators.

Implementation of the proposed changes will

effect the manner in which these SCCs are

tested; however, TS requirements that govern

routine testing and verification of plant

components and variables are not assumed to

be initiators of any analyzed event.

3. The proposed change does not result in

a significant reduction in the margin of

safety.

No margins of safety are reduced as a result

of the proposed TS changes. No safety limits

will be changed as a result of this TS change.

The proposed change does not involve a

reduction in the margin of safety because

SRMs are not credited in any safety analysis.

At least one SRM will remain operable

during rod withdrawal during core

alterations and rod withdrawal will not occur

if no SRMs are operable. Excessive reactivity

additions will be quickly identified and

mitigated by the Intermediate Range

Monitors and associated rod blocks. The

Average Power Range Monitor Flux scram,

and not any SRM function, is credited for

mitigating a rod withdrawal or reactivity

addition accident.

Use of a spiral offload or reload pattern

will provide assurance that the SRM will be

in the optimum position for monitoring

changes in neutron flux levels during core

alternations.

The changes proposed in this TS change do

not introduce any hardware changes, and

will not alter the intended operation of plant

structures, systems or components utilized in

the mitigation of accidents or transients.

Additionally, these changes will not

introduce any new failure modes of plant

equipment not previously evaluated.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Education

Building, Walnut Street and

Commonwealth Avenue, Box 1601,

Harrisburg, Pennsylvania 17105.

Attorney for licensee: J. W. Durham,

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

Counsel, PECO Energy Company, 2301

Market Street, Philadelphia,

Pennsylvania 19101

NRC Project Director: John F. Stolz

PECO Energy Company, Public Service

Electric and Gas Company, Delmarva

Power and Light Company, and

Atlantic City Electric Company,

Dockets Nos. 50-277 and 50-278, Peach

Bottom Atomic Power Station, Units

Nos. 2 and 3, York County,

Pennsylvania

Date of application for amendments:

March 22, 1995
24914 Federal Register / Vol. 60, No. 90 / Wednesday, May 10, 1995 / Notices

Description of amendment request:

The amendment would revise Note (1)

for Technical Specifications Tables 3.7.2

through 3.7.4 by reducing the Local

Leak Rate Test (LLRT) hold time

duration from one hour to 20 minutes.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed change does not serve as an

initiator or contributor to any accidents

previously evaluated. It does not decrease the

effectiveness of equipment relied upon to

mitigate previously evaluated accidents. The

change does not involve any physical

changes to any plant systems, structures, or

components.

Therefore, the proposed change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed changed does not create

the possibility of a new or different kind of

accident from any previously evaluated.

The proposed change does not serve as an

initiator or contributor to any of the accidents

previously evaluated. The proposed change

does not introduce any new modes of plant

operation.

Implementation of the proposed changes

will not affect the design function or

configuration of any component or introduce

any new operating scenarios or failure modes

or accident initiation. It does not impair or

prevent safety systems from performing their

safety function.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any previously

evaluated.

3. The proposed changes do not involve a

significant reduction in a margin of safety.

The proposed change does not serve as an

initiator or contributor to any accidents

evaluated in the SAR [Safety Analysis

Report]. It has no impact on any safety

analysis assumptions. Changing the LLRT

duration hold time from one hour to 20

minutes does not impact equipment

reliability. The change does not adversely

affect the assumptions or sequence of events

used in any accident analysis. Therefore, the

propose change does not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Education

Building, Walnut Street and

Commonwealth Avenue, Box 1601,

Harrisburg, Pennsylvania 17105.

Attorney for licensee: J. W. Durham,

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

Counsel, PECO Energy Company, 2301

Market Street, Philadelphia,

Pennsylvania 19101

NRC Project Director: John F. Stolz

Philadelphia Electric Company, Docket

Nos. 50-352 and 50-353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of amendment request:

November 21, 1994, as supplemented by

letter dated April 6, 1995

Description of amendment request:

The proposed amendment would make

changes affecting the Administrative

Controls Section of the Technical

Specifications (TSs). The areas proposed

to be changed are: 1) NEEDS [Nuclear

Effectiveness and efficiency Design

Study] Organization Title Changes, 2)

Minimum Shift Crew Composition, 3)

Delete Independent Techincal Review

Section from TS, 4) Delete NRB [Nuclear

Review Board] Review Section from TS,

and 5) Delete NRB Audit Section from

TS.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed Technical Specifications

changes do not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

The proposed TS changes to revise the

organization position titles, PORC [Plant

Operations Review Committee] composition

description, and eliminate the Assistant

Superintendent - Operations position do not

involve any physical modifications to plant

structures, systems, or components (SSC), or

the manner in which these SSC are operated,

maintained, modified, tested, or inspected.

The proposed changes to position titles will

not change the requirements for the

qualifications and training of personnel in

any management or supervisory position.

Personnel will continue to meet the guidance

specified in ANSI/ANS 3.1-1978 as required

by Technical Specification 6.3.1. The

probability of occurrence of an accident is

based in part on: the training and

qualifications of the personnel filling key

plant management and supervisory positions;

clear lines of authority, responsibility and

communication; and, adequate management

and corporate oversight of plant performance

and activities. The proposed TS changes do

not change any of these management and

organizational elements.

Allowing the Plant Manager to designate

appropriately qualified, trained and

experienced members of the LGS [Limerick

Generating Station] staff as members of the

PORC, as proposed, will not degrade the

effectiveness of the PORC. The qualifications,

training and experience level of the PORC

will meet the requirements listed in ANSI/

ANS 3.1-1978, and the required PORC

quorum (including the use of alternates) will

not be affected.

Elimination of the position of Assistant

Superintendent - Operations eliminates a

level of supervision between the Plant

Manager and the Shift Managers. The Shift

Managers, who hold SRO licenses, will

report directly to the Senior Manager -

Operations. Other organizational changes

within the Operations group (i.e.,

establishment of the positions of Manager -

Operations Services and Manager -

Operations Support) will ensure that the

Senior Manager - Operations has sufficient

time to properly supervise and monitor onshift performance. The Senior Manager

-Operations and/or an Operations Manager

will be required to hold a Senior Reactor

Operator (SRO) license. Individuals filling

these positions will satisfy the applicable

training, qualifications, and experience

requirements of ANSI/ANS 3.1-1978.

The consequences of an accident could be

affected by the qualifications and training of

plant management and supervisory

personnel. However, the proposed changes

do not change the qualifications and training

of personnel in any management or

supervisory position. Personnel will continue

to meet the criteria specified in ANSI/ANS

3.1-1978 as required by TS 6.3.1.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed TS changes to increase the

minimum shift crew composition do not

involve any physical changes to plant SSC.

The probability of the occurrence of an

accident is based in part on the operating

crew and their ability to safely operate the

plant. The increase in the minimum on-shift

crew composition and the associated changes

improves the capability of the on-shift crew

to safely operate the plant and SSC, thereby

reducing the probability of a situation that

could result in an accident. The increase in

the minimum on-shift crew composition will

improve the manner in which the SSC are

operated, maintained, tested, and inspected.

The consequences of an accident could be

affected by an operating error. However, the

proposed TS changes increase the number of

licensed operators required to be on-shift,

and therefore, increase the capability of the

on-shift crew to properly operate the facility

and to implement the appropriate emergency

procedures to reduce the consequences of an

accident.

The proposed changes will also delete

redundant and/or relocate existing

independent technical review and, Nuclear

Review Board review and audit requirements

from TS that are and/or will be contained in

the LGS UFSAR [Updated Final Safety

Analysis Report]. Removal of redundant/

relocation of existing requirements does not

affect any equipment important to safety, or

involve any physical modifications to plant

SSC, therefore, is not associated with an

accident initiator or accident mitigator and
Federal Register / Vol. 60, No. 90 / Wednesday, May 10, 1995 / Notices 24915

can not affect the probability of occurrence

of an accident or increase the consequences

of an accident. The licensee controlled

UFSAR containing the requirements will be

maintained using the provisions of 10 CFR

50.59, or 10 CFR 50.54(a), as appropriate, and

are subject to the change control process in

the Administrative Controls Section (6.0) of

the Technical Specifications. Since future

changes to related licensee-controlled

documents will be evaluated per 10 CFR

50.59 or 10 CFR 50.54(a), no increase

(significant or insignificant) in the

probability or consequences of an accident

previously evaluated will be allowed.

Therefore, these proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed TS changes do not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed TS changes to revise the

organization position titles, PORC

composition description, and eliminate the

Assistant Superintendent - Operations

position do not involve any physical

modifications to plant structures, systems, or

components (SSC), or the manner in which

these SSC are operated, maintained,

modified, tested, or inspected. The proposed

changes to position titles will not change the

requirements for the qualifications and

training of personnel in any management or

supervisory position. Personnel will continue

to meet the guidance specified in ANSI/ANS

3.1-1978 as required by Technical

Specification 6.3.1. Therefore, these

proposed TS changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes to the on-shift crew

composition can not create the possibility of

a new or different type of accident than

previously evaluated in the SAR since

implementation of the changes will not

involve any physical changes to the plant

SSC. The increase in the minimum on-shift

crew composition increases the ability of the

operating crew to ensure that the SSC are

properly operated, maintained, tested and

inspected. An increase in the required

number of licensed operators on each shift

improves the ability of the crew to

adequately operate the facility, to respond to

accident conditions, and to implement

applicable plant procedures. Therefore, these

proposed TS changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes will also delete

redundant and/or relocate existing

independent technical review and, Nuclear

Review Board review and audit requirements

from TS that are and/or will be contained in

the UFSAR. The changes will not alter the

plant configuration (no new or different type

of equipment will be installed) or create

changes in methods governing normal plant

operation that will introduce new failure

modes. These changes will not impose

different requirements and proper control of

information will be maintained. These

changes will not alter assumptions made in

the safety analysis and licensing basis.

Therefore, these changes will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The proposed TS changes do not involve

a significant reduction in a margin of safety.

The proposed TS changes to revise the

organization position titles, PORC

composition description, and eliminate the

Assistant Superintendent - Operations

position, do not reduce the margin of safety

because positions with equivalent authority

and responsibility are established and the

new positions have equivalent requirements

for education, experience and training.

Allowing the Plant Manager to designate

appropriately qualified, trained and

experienced members of the LGS staff as

members of the PORC will not degrade the

effectiveness of the PORC because the

qualifications, training and experience level

of the PORC will meet the requirements

listed in ANSI/ANS 3.1-1978 and the

required PORC quorum (including the use of

alternates) will not be affected. Elimination

of the position of Assistant Superintendent -

Operations eliminates a level of supervision

between the Plant Manager and the Shift

Managers. If the Senior Manager - Operations

does not hold an SRO license, then an

Operations Manager must hold an SRO

license. This individual will 1) be qualified

to fill the Senior Manager - Operations

position, 2) have the same management

authority over the licensed operators as the

Senior Manager - Operations, and 3) by being

designated by Administrative procedures

assures that there is always an individual

holding a current SRO license in one of the

Operations management positions. Other

organizational changes (i.e., establishment of

the positions of Manager - Operations

Services and Manager - Operations Support),

will ensure that the Senior Manager

-Operations has sufficient time to properly

supervise and monitor on-shift performance.

Therefore, these changes do not involve a

significant reduction in a margin of safety.

The proposed changes to the on-shift crew

composition increases the number of

licensed SROs per shift to be one (1) above

the minimum number required by the

regulations. Additionally, the title changes

are consistent with the organization and

reporting relationships discussed in the

regulation and the LGS Updated Final Safety

Analysis Report (UFSAR). The Shift Manager

holds a SRO license for both units and is

assigned responsibility for overall plant

operation at all times when there is fuel in

any unit. The other SROs on the shift report

to the Shift Manager and at least one (1) of

the SRO licensed individuals is in the Main

Control Room when either unit is in an

operating mode other than cold shutdown or

refuel. The increase in the minimum on-shift

crew composition and the associated changes

improves the capability of the on-shift crew

to safely operate the plant and SSC.

Therefore, these changes do not involve a

significant reduction in a margin of safety.

The proposed changes will also delete

redundant and/or relocate existing

independent technical review and, Nuclear

Review Board review and audit requirements

from TS that are and/or will be contained in

the LGS UFSAR. The changes will not reduce

the margin of safety since they have no

impact on any safety analysis assumptions.

In addition, any future changes to the UFSAR

will be evaluated per the requirements of 10

CFR 50.59 or 10 CFR 50.54(a), as appropriate.

Therefore, these changes will not involve a

significant reduction in a margin of safety.

The existing requirement for NRC review

and approval of revisions, in accordance with

10 CFR 50.90, to these TS details and

requirements proposed for relocation, does

not have a specific margin of safety upon

which to evaluate. However, since the

proposed changes to delete redundant and/or

relocate requirements are consistent with the

BWR Standard Technical Specifications

(NUREG-1433) and the four criteria set forth

in the NRC ‘‘Final Policy Statement on

Technical Specifications Improvements for

Nuclear Power Reactors,’’ and since the

change controls for proposed relocated

details and requirements provide an

equivalent level of regulatory authority,

revising the TS to reflect the approved level

of detail and requirements ensures no

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: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania

19464.

Attorney for licensee: J. W. Durham,

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

Counsel, Philadelphia Electric

Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101

NRC Project Director: John F. Stolz

Power Authority of the State of New

York, Docket No. 50-333, James A.

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of amendment request: February

22, 1995

Description of amendment request:

The proposed changes to the James A.

Fitzpatrick Technical Specifications

establish operability and surveillance

requirements for the Reactor Vessel

Overfill Protection Instrumentation that

initiates feedwater pump turbine trips,

and a main turbine trip, on high reactor

vessel water level.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated because:
24916 Federal Register / Vol. 60, No. 90 / Wednesday, May 10, 1995 / Notices

The proposed changes involve the addition

of new operability and surveillance

requirements to the Technical Specification

regarding the current high reactor water level

trip feature for the feedwater pump turbines

and main turbine. The changes do not

introduce any new modes of plant operation,

make any physical changes, or alter any

operational setpoints associated with the

plants instrumentation and controls. Further,

the Fitzpatrick UFSAR [Updated Final Safety

Analysis Report], Section 14.5.9, for the

Feedwater Controller Failure operational

transient does not take credit for the

automatic high reactor vessel water level trip

of the feedwater pump turbines. The

Fitzpatrick UFSAR analysis (Section 14.5.9),

for the Feedwater Controller Failure

operational transient assumes an automatic

high reactor vessel water level trip of the

main turbine. Incorporating these

requirements into the Technical

Specifications provides additional assurance

that a trip feature described in the UFSAR

remains functional. For these reasons the

changes do not increase the probability or

consequences of an accident previously

evaluated.

2. Create the possibility of a new or

different kind of accident from those

previously evaluated because:

The proposed changes do not introduce

any new accident initiators or failure

mechanisms since the changes do not

introduce any new modes of plant operation,

make any physical changes, or alter any

operational setpoints. Accordingly, the

changes do not create the possibility of a new

or different kind of accident from those

previously evaluated.

3. Involve a significant reduction in the

margin of safety because:

The proposed changes establish operability

and surveillance requirements for the design

feature that trips the feedwater pump

turbines and main turbine on high reactor

vessel water level. The requirements will

assure the continued operability of a trip

function that is designed to initiate protective

measures in the event of excessive feedwater

flow. Tripping the feedwater pump turbines

and main turbine on high reactor vessel

water level, precludes potential adverse

safety implications associated with a reactor

overfill condition. Accordingly, the proposed

changes will enhance the plant safety margin.

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: Mr. Charles M.

Pratt, 1633 Broadway, New York, New

York 10019.

NRC Project Director: Ledyard B.

Marsh

Power Authority of the State of New

York, Docket No. 50-333, James A.

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of amendment request: March 2,

1995

Description of amendment request:

The proposed changes to the James A.

Fitzpatrick Technical Specifications

extend the surveillance test intervals for

the snubber systems to support 24

month operating cycles.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed changes increase the interval

between snubber functional tests. These

changes are consistent with the guidance

provided in Generic Letter 91-04. These

changes do not involve any physical changes

to the plant, nor do they alter the way

snubbers function. The type of testing and

the actions taken if a snubber fails a

functional test remain the same. The review

of the snubber installation and maintenance

records will continue to ensure that the

snubbers service life is not exceeded prior to

the next scheduled review. The proposed

changes to bases 4.0 and 4.6 clarify that the

snubber functional testing interval is

consistent with the length of the operating

cycle. Therefore, the proposed changes do

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed changes increase the interval

between snubber functional tests. These

changes are consistent with the guidance

provided in Generic Letter 91-04. The

proposed changes do not change the ability

of the snubbers to provide dynamic load

support during a design basis accident. Past

operating experience indicates that the

snubber program at the FitzPatrick plant

adequately identifies snubber failures. No

changes are proposed to the type of testing

performed only to the surveillance interval

length. The proposed changes do not modify

the design or operation of plant equipment,

therefore, no new or different failure modes

are introduced. The Technical Specification

for snubber testing is self-corrective. If any

snubber fails a functional test, Technical

Specifications require additional testing of a

10% sample of that type of snubber until no

more failures are found. The functional test

criteria remains unchanged and ensures a

95% confidence level that at least 90% of the

snubbers are operable. The proposed changes

to bases 4.0 and 4.6 clarify that the snubber

functional testing interval is consistent with

the length of the operating cycle. 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.

The proposed changes increase the interval

between snubber functional tests. These

changes are consistent with the guidance

provided in Generic Letter 91-04. The

proposed changes do not alter the

configuration of the snubbers nor change the

manner in which the snubbers function.

Operation of the facility remains unchanged

by the proposed changes. An evaluation of

past equipment performance indicates that

snubber operability is not time dependent.

The proposed changes to bases 4.0 and 4.6

clarify that the snubber functional testing

interval is consistent with the length of the

operating cycle. Therefore, a longer

surveillance test interval will not degrade

snubber performance and 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: Mr. Charles M.

Pratt, 1633 Broadway, New York, New

York 10019.

NRC Project Director: Ledyard B.

Marsh

Power Authority of the State of New

York, Docket No. 50-333, James A.

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of amendment request: April 12,

1995

Description of amendment request:

The proposed changes to the James A.

FitzPatrick Technical Specifications

extend the surveillance test intervals for

the nuclear steam supply system to

support 24 month operator cycles.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed changes extend the

surveillance test intervals for nuclear steam

supply system components. These changes

are consistent with the guidance provided in

Generic Letter 91-04. The proposed changes

do not involve any modification to the plant,

nor do they alter equipment functions. Online testing will provide a redundant and

early means of demonstrating system
Federal Register / Vol. 60, No. 90 / Wednesday, May 10, 1995 / Notices 24917

operability. Based on past results, SRV

[safety/relief valve] mechanical performance

has been good. No SRV setpoint changes are

involved in this application. The proposed

change to bases section 4.6 clarifies that the

nuclear steam supply system surveillance

testing interval is consistent with the length

of the operating cycle. Therefore, the

proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed changes extend the

surveillance test intervals for nuclear steam

supply system components. These changes

are consistent with the guidance provided in

Generic Letter 91-04. The proposed changes

do not affect the way in which the nuclear

steam supply system operates nor alter the

type of surveillance testing performed. SRV

drift analyses indicate that SRV drift with a

3% tolerance would be acceptable for (i.e.,

bounded by) a 24 to 30 month interval.

Leaking or partially open SRVs are detected

by the acoustic monitoring system. Since the

proposed changes do not modify the design

or equipment of the plant, no new failure

modes are introduced. The proposed change

to bases section 4.6 clarifies that the nuclear

steam supply system surveillance testing

interval is consistent with the length of the

operating cycle. 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.

The proposed changes extend the

surveillance test intervals for nuclear steam

supply system components. These changes

are consistent with the guidance provided in

Generic Letter 91-04. The proposed changes

do not alter the configuration of the nuclear

steam supply system nor change the manner

in which the system functions. Operation of

the facility remains unchanged by the

proposed changes. An evaluation of past

equipment performance indicates that SRV

mechanical performance has been good. In

addition, SRV drift has been analyzed to be

within the allowable tolerance for the

extended surveillance interval. The proposed

change to bases section 4.6 clarifies that the

nuclear steam supply system surveillance

testing interval is consistent with the length

of the operating cycle. 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 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: Mr. Charles M.

Pratt, 1633 Broadway, New York, New

York 10019.

NRC Project Director: Ledyard B.

Marsh

Power Authority of The State of New

York, Docket No. 50-286, Indian Point

Nuclear Generating Unit No. 3,

Westchester County, New York

Date of amendment request: March 3,

1995, as supplemented April 12, 1995

Description of amendment request:

The licensee commenced operating on a

24-month fuel cycle, instead of the

previous 18-month fuel cycle, with

cycle 9. Fuel cycle 9 started in August

1992; however, the licensee shut down

the facility in February 1993 for a

performance improvement outage.

Although a firm restart date has not yet

been established, restart is expected in

the spring of 1995. In order to

accommodate operation on a 24-month

cycle after the facility restarts, the

licensee requested an amendment to the

Technical Specifications (TSs) to

incorporate the indicating instrument

calibration frequency changes listed

below:

(1) The licensee proposed changing

the calibration frequency for the

containment water level monitor

instrumentation (specified in TS Table

4.1-1) to accommodate operation on a

24-month cycle.

(2) The licensee proposed changing

the calibration frequency for the

auxiliary feedwater (AFW) flow rate

instrumentation (specified in TS Table

4.1-1) to accommodate operation on a

24-month cycle.

(3) The licensee proposed changing

the calibration frequency for the

containment building ambient

temperature sensors (specified in TS

Table 4.1-1) to accommodate operation

on a 24-month cycle.

(4) The licensee proposed changing

the calibration frequency for the seismic

monitoring instrumentation (specified

in TS Table 4.10-2) to accommodate

operation on a 24-month cycle.

In addition, the licensee proposed

adding a new surveillance requirement

to TS Table 4.1-1 for testing the core exit

thermocouples.

These proposed changes follow the

guidance provided in Generic Letter 91-

04, ‘‘Changes in Technical Specification

Surveillance Intervals to Accommodate

a 24-Month Fuel Cycle,’’ as applicable.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Consistent with the criteria of 10 CFR

50.92, the enclosed application is judged to

involve no significant hazards based on the

following information:

(1) Does the proposed license amendment

involve a significant increase in the

probability or consequences of any accident

previously evaluated?

Response:

The proposed changes do not involve a

significant increase in the probability or

consequences of any accident previously

evaluated. The proposed changes extend the

calibration frequency (to 24 months) for the:

• containment temperature channels,

• containment water level monitoring

system channels,

• seismic instrumentation channels, and

• auxiliary feedwater flow rate channels.

These changes are being made to

accommodate a 24 month operating cycle.

The proposed changes in the calibration

frequencies do not involve any plant

hardware changes, nor do they change the

way the systems function.

Extension of the calibration and

surveillance test intervals in question were

evaluated and the results documented in

[New York Power Authority (NYPA) Report

No. IP3-RPT-MULT-00424, ‘‘Indicating

Instruments Surveillance Test Extensions,’’

May 1993]. An Instrument Drift Analysis for

the indicating instruments [NYPA Report No.

IP3-RPT-MULT-00407, ‘‘Instrument Drift

Analysis for Indicating Loops,’’ April 1993]

was performed to evaluate past and future

instrument drift. The results of these

evaluations and analyses indicate that the

calibrations in question can safely be

extended to accommodate the 24 month

operating cycle.

For containment temperature, auxiliary

feedwater flow and seismic instrumentation,

past instrument drift has generally been

within acceptable limits. Some drift

exceeding the calibration tolerance did occur

for the triaxial time-history accelographs, but

on-line testing should ensure that instrument

drift over the longer cycle does not degrade

system performance. For containment water

level systems (except containment building

level), new electronic transmitters were

recently installed. Due to the lack of data, an

instrument drift analysis was not performed.

However, the new containment water level

transmitters improved the overall channel

accuracy.

Future instrument drift was predicted and

used to update existing loop accuracy

calculations, with the following results. (1)

For the containment temperature channels,

the loop accuracy calculations were revised

to incorporate the larger channel

uncertainties. Postulated drift over 30

months should have a negligible effect on the

EOPs [Emergency Operating Procedures] and

plant shutdown. (2) For the containment

system sump water levels, future drift is not

a concern because the containment building

water level is used post accident. The larger

uncertainties can safely be accommodated by

changing the EOP setpoint for transfer to cold

leg recirculation. (3) For the seismic

instrumentation, past drift was negligible,

and future drift is not expected to be cycle

length dependent. (4) For the auxiliary
24918 Federal Register / Vol. 60, No. 90 / Wednesday, May 10, 1995 / Notices

feedwater flow rate channels, the larger

uncertainties can be safely accommodated by

changing the EOP setting for the minimum

AFW flow required for heat removal.

For the containment temperature and

seismic instrumentation, on-line testing

provides added assurance that the

instrumentation is functioning as required.

[For the core exit thermocouples, adding a

requirement to conduct testing every 18

months will serve to ensure system

operability. This new testing requirement

does not change the way the plant operates

or involve hardware modifications.]

(2) Does the proposed license amendment

create the possibility of a new or different

kind of accident from any previously

evaluated?

Response:

The proposed changes do not create the

possibility of a new or different kind of

accident from any previously evaluated. The

proposed changes extend the calibration

frequency (to 24 months) for the:

• containment temperature channels,

• containment water level monitoring

system channels,

• seismic instrumentation channels, and

• auxiliary feedwater flow rate channels.

These changes are being made to

accommodate a 24 month operating cycle.

The proposed changes in the calibration

frequencies do not involve any plant

hardware changes, nor do they change the

way the systems function.

Extension of the calibration and

surveillance test intervals in question were

evaluated and the results documented in

[same as Question (1)]. An Instrument Drift

Analysis for the indicating instruments [same

as Question (1)] was performed to evaluate

past and future instrument drift. The results

of these evaluations and analyses indicate

that the calibrations in question can safely be

extended to accommodate the 24 month

operating cycle. For the containment

temperature and seismic instrumentation, online testing provides added assurance that

the instrumentation is functioning as

required.

[For the core exit thermocouples, adding a

requirement to conduct testing every 18

months will serve to ensure system

operability. This new testing requirement

does not change the way the plant operates

or involve hardware modifications.]

(3) Does the proposed amendment involve

a significant reduction in a margin of safety?

Response:

The proposed changes do not involve a

significant reduction in a margin of safety.

The proposed changes extend the calibration

frequency (to 24 months) for the:

• containment temperature channels,

• containment water level monitoring

system channels,

• seismic instrumentation channels, and

• auxiliary feedwater flow rate channels.

These changes are being made to

accommodate a 24 month operating cycle.

The proposed changes in the calibration

frequencies do not involve any plant

hardware changes, nor do they change the

way the systems function.

For containment temperature, auxiliary

feedwater flow and seismic instrumentation,

past instrument drift has generally been

within acceptable limits. Some drift

exceeding the calibration tolerance did occur

for the triaxial time-history accelographs, but

on-line testing should ensure that instrument

drift over the longer cycle does not degrade

system performance. For containment water

level systems (except containment building

level), new electronic transmitters were

recently installed. Due to the lack of data, an

instrument drift analysis was not performed.

However, the new containment water level

transmitters improved the overall channel

accuracy.

[For the core exit thermocouples, adding a

requirement to conduct testing every 18

months will serve to ensure system

operability. This new testing requirement

does not change the way the plant operates

or involve hardware modifications.]

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 10601.

Attorney for licensee: Mr. Charles M.

Pratt, 10 Columbus Circle, New York,

New York 10019.

NRC Project Director: Ledyard B.

Marsh

Public Service Electric & Gas Company,

Docket Nos. 50-272 and 50-311, Salem

Nuclear Generating Station, Unit Nos. 1

and 2, Salem County, New Jersey

Date of amendment request: March

30, 1995

Description of amendment request:

The proposed change to the Technical

Specifications eliminates the defined

term CONTROLLED LEAKAGE,

removes Controlled Leakage flow from

the Reactor Coolant System Operational

Leakage Limiting Condition for

Operation (LCO), and establishes a new

Seal Injection Flow LCO.

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. Do not involve a significant increase in

the probability or consequence of an accident

previously evaluated.

Changing the Technical Specification to

limit seal injection flow instead of seal

leakoff flow does not affect the probability of

any accident previously evaluated.

Maintaining adequate Emergency Core

Cooling System (ECCS) flow during Loss of

Coolant Accident (LOCA) ensures that the

consequences of these accidents are

unaffected. The existing Technical

Specification allows seal injection throttle

valve positioning that could result in seal

injection flow path resistance values below

those used in the Salem ECCS hydraulic flow

analyses. Reduced line resistances could

result in inadequate ECCS flow to the reactor

core. Revising the Technical Specification to

limit RCP seal injection flow ensures that the

accident analysis assumptions are

maintained, and the previously evaluated

accident consequences remain unchanged.

Therefore, it may be concluded that the

proposed changes do not increase the

probability or consequences of an accident

previously evaluated.

2. 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

hardware modifications or result in any

functional changes to system operation. RCP

seal injection flow is used as a limiting

parameter in-place of RCP seal leakoff flow.

Since design requirements continue to be

met and the RCS pressure boundary is not

challenged, no new failure mode is created.

Thus, an accident different from any already

evaluated is not created by this change.

Therefore, it may be concluded that the

proposed changes do not create the

possibility of a new or different kind of

accident from any previously evaluated.

3. Do not involve a significant reduction in

a margin of safety.

The proposed changes do not alter the

manner in which Safety Limits or Limiting

Safety System Setpoints are determined.

Controlled Leakage (RCP seal leakoff)is

removed from the Reactor Coolant System

Leakage Limiting Condition for Operation

(LCO), and a new seal injection LCO is

established. The new LCO continues to limit

seal injection flow during accident

conditions. The limiting parameter is

changed from RCP seal leakoff flow to RCP

seal injection flow. These changes ensure

that the accident analysis assumptions and

existing margins of safety are maintained.

The seal injection flow specification limit is

not applicable in Mode 4 and lower, because

high seal injection flow is less critical due to

lower Reactor Coolant System (RCS) pressure

and decay heat removal requirements in

these modes. Reactor coolant pump seal

injection flow must be limited in Modes 1,

2, and 3 to ensure adequate Emergency Core

Cooling System Flow.

Therefore, it may be concluded that 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: Salem Free Public library, 112

West Broadway, Salem, New Jersey

08079

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston and
Federal Register / Vol. 60, No. 90 / Wednesday, May 10, 1995 / Notices 24919

Strawn, 1400 L Street, NW, Washington,

DC 20005-3502

NRC Project Director: John F. Stolz

Tennessee Valley Authority, Docket

Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of amendment request: April 6,

1995 (TS 95-05)

Description of amendment request:

The proposed change would (1) replace

the reference to Table 3.6-2 from

Definition 1.7.a.2 for Containment

Integrity with a phrase that will allow

the valves to be opened under

administrative control; (2) replace the

reference to Table 3.6-2 from

Surveillance Requirement 4.6.1.1 with a

phrase that will allow the valves to be

opened under administrative control; (3)

delete the reference to Table 3.6-1 from

Technical Specification 3.6.1.2; (4)

delete Table 3.6-1, ‘‘Bypass Leakage

Paths to the Auxiliary Building --

Secondary Containment Bypass Leakage

Paths;’’ (5) revise Specification 3.6.3 to

delete the reference to Table 3.6-2, add

a footnote that discusses the opening of

penetrations intermittently, add the

phrase to take exception to the

containment vacuum isolation valves,

and add an action statement to indicate

that Specification 3.0.4 does not apply

to the specification; (6) delete

Surveillance Requirement 4.6.3.1; (7)

delete references to Table 3.6-2 in

Specifications 4.6.3.2 and 4.6.3.3 and

additional wording added to indicate

that the specifications apply to

automatic containment isolation valves;

(8) delete Table 3.6-2, ‘‘Containment

Isolation Valves’’ and add a note to the

page indicated that the information has

been intentionally deleted; (9) revise

Specification 3.8.3.1 to specify that the

Limiting Condition for Operation

applies to primary and backup

containment penetration conductor

overcurrent protective devices

associated with each containment

electrical penetration shall be operable,

add a phrase to indicate that the scope

of these protective devices excludes

those circuits for which credible fault

currents would not exceed the electrical

penetration design rating, and delete the

phrase that references appropriate plant

instructions in the action statement; (10)

delete the phrase that references

appropriate plant procedures from

Specification 4.8.3.1; (11) delete the

phrase from SR 4.8.3.1.a.3 that indicates

that a complete listing of all fuses to be

verified in accordance with the

requirement will be maintained in

appropriate plant instructions; (12)

replace the phrase ‘‘appropriate plant

instructions based on’’ with

‘‘procedures prepared in conjunction

with’’ in SR 4.8.3.1.b; (13) replace the

reference to Table 3.8-2 in Specification

3.8.3.2 with a phrase that indicates that

the Requirement is applicable to valves

used in safety systems; (14) delete Table

3.8-2, ‘‘Motor Operated Valves Thermal

Overload Protection,’’ and replace it

with a note that indicates that the pages

are intentionally blank; and (15)

incorporate appropriate changes to the

Bases to reflect these 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, which is presented

below:

TVA has evaluated the proposed technical

specification (TS) change and has determined

that it does not represent a significant

hazards consideration based on criteria

established in 10 CFR 50.92(c). Operation of

Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The removal of the component listings

from the SQN TSs will not create an increase

in the probability or consequences of any

accident previously evaluated. Although no

longer in the TSs, the components listed in

Tables 3.6-1, 3.6-2, and 3.8-2 will be

contained in administratively controlled

documents. This equipment must be tested at

the required intervals and each unit’s action

statements must still be adhered to. These

procedures are revised and approved in

accordance with requirements of TS Section

6.5.1A. This review process also requires an

evaluation based on 10 CFR 50.59

requirements. As indicated in GL 91-08, this

is adequate control for changes to these

components lists.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

The removal of the component lists from

the TSs does not modify safety-related

equipment or systems, nor does it change any

safety-related setpoints used to prevent or

mitigate previously analyzed accidents. The

component lists are presently located in

separate documents that are subject to the

requirements of 10 CFR 50.59. Also, the

limiting condition of operation requirements

remain in effect and appropriate actions will

be taken if any limits are exceeded.

Therefore, the proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Involve a significant reduction in a

margin of safety.

The margin of safety is not affected by the

removal of the previously discussed

component lists from the TS. Appropriate

measures presently exist to control the

setpoint of the components listed. Any

changes to these setpoints are controlled by

the SQN design change process that is subject

to the requirements of 10 CFR 50.59 in which

the reduction of the present margin of safety

is addressed. The proposed amendment

continues to require operation within the set

values for these components, and appropriate

actions to be taken when or if the limits are

exceeded. Based on these controls, this

amendment will not involve a reduction in

a margin of safety.

The NRC has reviewed the licensee’s

analysis and, based on thisreview, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H,

Knoxville, Tennessee 37902

NRC Project Director: Frederick J.

Hebdon

The Cleveland Electric Illuminating

Company, Centerior Service Company,

Duquesne Light Company, Ohio Edison

Company, Pennsylvania Power

Company, Toledo Edison Company,

Docket No. 50-440, Perry Nuclear

Power Plant, Unit No. 1, Lake County,

Ohio

Date of amendment request: March

24, 1995

Description of amendment request:

The licensee has requested a one-time

extension of the performance intervals

for certain Technical Specification

Surveillance Requirements (SR).

Affected SRs include penetration leak

rate testing, valve operability testing,

instrument calibration, response time

testing, and logic system functional

tests. The proposed changes are

requested to support refueling outage 5

scheduled to begin no later than

February 15, 1996.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration which is presented below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed TS change requests a onetime extension of the surveillance intervals

related to: a) RPS Instrumentation

calibration, LSFTs, and response time testing;

b) Isolation Actuation System

Instrumentation calibration, LSFTs, and

response time testing; c) ECCS Actuation

Instrumentation calibration, LSFTs, and

response time testing; d) Control Rod Block

Instrumentation calibration and LSFTs; e)

Remote Shutdown Instrumentation and

Controls calibration and operability testing; f)
24920 Federal Register / Vol. 60, No. 90 / Wednesday, May 10, 1995 / Notices

Accident Monitoring Instrumentation

calibration; g) Plant Systems Instrumentation

calibration and LSFTs; h) Primary

Containment automatic valve actuation; i)

Reactor Coolant System Pressure Isolation

Valve (PIV) testing; j) system automatic

initiation testing; and, k) Emergency Diesel

Generator inspection and testing.

Also proposed is the re-establishment of

the baseline for the ‘‘N times 18 months’’

cumulative surveillance interval for response

time testing.

The discussion in the License Amendment

Request demonstrates the following:

i) Rosemount transmitter calibration period

extension is acceptable based on Rosemount

D8900126, Revision A which supported

extension of the calibration interval from 18

months to 30 months based on the reduction

in the drift allowance;

ii) Extrapolation of plant specific

calibration data is acceptable in supporting

the extension of other calibration

surveillance intervals to RFO-5;

iii) LSFT interval extension is acceptable

based on the NRC Safety Evaluation Report

(Peach Bottom Atomic Power Plant, Units 2

and 3, dated August 2, 1993) which

supported extension of the interval for LSFT

from 18 to 24 months. This was based on the

small probability of relay or contact failure

relative to mechanical component failure

probability and, therefore, the increase in

LSFT interval represented no significant

change in the overall safety system

unavailability;

iv) Response time testing interval

extension for Isolation Actuation and ECCS

Actuation instrumentation channels is

acceptable based on the BWR Owners Group

(BWROG) Licensing Topical Report NEDO32291 (January 1994) which provided the

necessary justification for elimination of

response time testing and, therefore, provides

a suitable argument for extending the interval

for a short period of time. The NRC approved

the use of NEDO-32291 as a basis for License

Amendment Requests, with additional

conditions specified, in a letter to the

BWROG in December 1994.

v) Response time testing interval extension

for RPS Instrumentation channels is

acceptable because: i) there are redundant

sensors that can initiate the scram function;

ii) one-out-of-two redundancy exists in every

individual instrument channel within each

trip function; iii) several redundant and

diverse instrument channels are provided

which can detect and generate a scram signal;

iv) the failure probability is a small fraction

of the total control rod insertion (scram)

failure probability; v) failure of

instrumentation in the sluggish mode is a

small fraction of its overall failure modes;

and iv) NRC Safety Evalution Report dated

August 2, 1993 (Peach Bottom Atomic Power

Station, Units 2 and 3 docket) has previously

provided approval for extension of the RPS

response time testing surveillance interval

from 18 to 24 months.

vi) Response time testing interval

extension for the Main Steam Line isolation

is acceptable because i) redundancy and

diversity exist in individual instrument

channels within a trip function; ii)

instrumentation response time is a small

fraction of the overall response time of the

actuating device; iii) instrumentation failure

probability is a very small portion of the total

MSIV failure probability; and, iv) failure of

instrumentation in the sluggish responding

mode is a small fraction of its overall failure

modes.

vii) Containment Isolation Valve leakage

determination and actuation interval

extension is acceptable based on: i)

redundancy provided in the design of the

penetrations; ii) the periodic testing of the

valves during power operation; and, iii) the

short period of time the interval is being

extended.

viii) Reactor Coolant System PIVs have

exhibited low as-found leak rates as

measured during the last refueling outage;

there is substantial margin available for the

PIVs from the as-left leakage to the allowed

TS leakage; the requested extension of the

surveillance interval is small; and the

conclusion of NUREG-1463, ‘‘Regulatory

Analysis for the Resolution of Generic Safety

Issue 105: Interfacing System Loss-of-Coolant

Accident in Light Water Reactors’’ (July

1993), and the confirmation of the PNPP

Individual Plant Examination that the

ISLOCA (for which PIVs are provided to

prevent) is not a risk concern to BWRs or

PNPP.

ix) System initiation and actuation testing

interval is acceptable based on the periodic

testing of components during power

operation and the short period of time the

interval is being extended.

x) Emergency Diesel Generator testing

interval extension is acceptable based on: i)

the past testing results which support

extension for the short period of time; ii) the

testing that is done during power operation;

and, iii) the short period of time the interval

is being extended.

xi) The re-establishment of the baseline for

the ‘‘N times 18 months’’ cumulative

surveillance interval for response time testing

is acceptable in that the extension of the

cumulative interval would not be for more

than the individual extension requested and

justified herein.

Therefore, from the above it is shown that

the proposed change will not significantly

increase the probability of an accident

previously evaluated.

2. The proposed change would not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed TS change requests a onetime extension of the surveillance intervals

for instrument calibration, instrument

channel LSFT and response time testing,

containment isolation valve leakage

determination and actuation, PIV leak rate

determination, system actuation testing, and

diesel generator inspection and testing. The

proposed changes do not necessitate a

physical alteration to the plant (no new or

different type of equipment will be installed).

The requested extension durations are small

as compared to the overall interval allowed

by TS; drift data supports extension of the

calibration intervals; NRC and industry

evaluations support extension of LSFT;

industry evaluations and redundancy in

system design support extension of response

time testing; past testing and periodic testing

provides confidence of no effect on

equipment availability by extending the

confidence of no effect on equipment

availability by extending the surveillance

interval. Therefore, the change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

In addition, the requested re-establishment

of the baseline at RFO-5 for the ‘‘N time 18

months’’ cumulative surveillance interval for

response time testing is acceptable in that the

cumulative surveillance interval will not be

extended by more than that which is

proposed for individual response time tests

during RFO-5. The individual response time

test surveillance interval extensions have

been justified herein. The justification for

individual response time test surveillance

interval extensions applies to the cumulative

surveillance interval extension which is

requested and will be granted by allowing the

re-establishment of the baseline of the ‘‘N

times 18 months’’ surveillance interval to the

response time testing dates for those response

time tests to be performed during RFO-5. The

proposed changes do not necessitate a

physical alteration to the plant (no new or

different type of equipment will be installed).

Therefore, the change does not create the

possibility of a new or different kind of

accident.

3. The proposed change will not involve a

significant reduction in the margin of safety.

The proposed TS change requests a onetime extension of the surveillance intervals

for instrument calibration, instrument

channel LSFT, and response time testing,

containment isolation valve leakage

determination and actuation, PIV leak rate

determination, system actuation testing, and

diesel generator inspection and testing. The

proposed changes do not necessitate a

physical alteration to the plant (no new or

different type of equipment will be installed).

In that the requested extension durations are

small as compared to the overall interval

allowed by TS, drift data supports extension

of the calibration intervals, NRC and industry

evaluations support extension of LSFT,

industry evaluations and redundancy in

system design support extension of response

time testing, past testing and periodic testing

provides confidence of no effect on

equipment availability by extending the

surveillance interval, the change does not

involve a significant reduction in the margin

of safety.

In addition, the requested re-establishment

of the baseline at RFO-5 for the ‘‘N times 18

months’’ cumulative surveillance interval for

response time testing is acceptable in that the

cumulative surveillance interval will not be

extended by more than that which is

proposed for individual response time tests

during RFO-5. The individual response time

test surveillance interval extensions have

been justified herein. The justification for

individual response time test surveillance

interval extensions applies to the cumulative

surveillance interval extension which is

requested and will be granted by allowing the

re-establishment of the baseline of the ‘‘N

times 18 months’’ surveillance interval to the

response time testing dates for those response
Federal Register / Vol. 60, No. 90 / Wednesday, May 10, 1995 / Notices 24921

time tests to be performed during RFO-5. The

proposed changes do not necessitate a

physical alteration to the plant (no new or

different type of equipment will be installed).

Therefore, the 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: Perry Public Library, 3753

Main Street, Perry, Ohio 44081

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts & Trowbridge,

2300 N Street, NW., Washington, DC

20037

NRC Project Director: Gail H. Marcus

The Cleveland Electric Illuminating

Company, Centerior Service Company,

Duquesne Light Company, Ohio Edison

Company, Pennsylvania Power

Company, Toledo Edison Company,

Docket No. 50-440, Perry Nuclear

Power Plant, Unit No. 1, Lake County,

Ohio

Date of amendment request: April 3,

1995

Description of amendment request:

The proposed amendment would add

new programmatic requirements

governing radiological effluent into the

Administrative Controls section of the

Technical Specifications in accordance

with Generic Letter 89-01,

‘‘Implementation of Programmatic

Controls for Radiological Effluent

Technical Specifications in the

Administrative Controls Section of

Technical Specifications and the

Relocation of Procedural Details of

RETS to the Offsite Dose Calculation

Manual or to the Process Control

Program.’’

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration which is presented below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes are administrative

in nature and alter only the format and

location of programmatic controls and

procedural details relative to radioactive

effluent, radiological environmental

monitoring, solid radioactive wastes, and

associated reporting requirements.

Compliance with applicable regulatory

requirements will continue to be maintained.

In addition, the proposed changes do not

alter the conditions or assumptions in any of

the Updated Safety Analysis Report (USAR)

accident analyses. Since the USAR accident

analyses remain bounding, the radiological

consequences previously evaluated are not

adversely affected by the proposed changes.

Therefore, it can be concluded that the

proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

The proposed changes do not involve any

changes to the configuration or method of

operation of any plant equipment.

Accordingly, no new failure modes have

been defined for any plant system or

component important to safety nor has any

new limiting single failure been identified as

a result of the proposed changes. Also, there

will be no change in types or increase in the

amounts of any radioactive effluent released

offsite. Therefore, it can be concluded that

the proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The proposed change does not involve

a significant reduction in the margin of

safety.

The proposed changes do not involve any

actual change in the methodology used in the

control of radioactive effluents, solid

radioactive wastes, or radiological

environmental monitoring. These changes are

considered administrative in nature, provide

for the relocation of procedural details

outside the Technical Specifications, and add

appropriate administrative controls in the

Technical Specifications to provide

continued assurance of compliance with

applicable regulatory requirements. These

proposed changes also comply with the

guidance contained in Generic Letter 89-01.

Therefore, it can be concluded that 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: Perry Public Library, 3753

Main Street, Perry, Ohio 44081

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts & Trowbridge,

2300 N Street, NW., Washington, DC

20037

NRC Project Director: Gail H. Marcus

Union Electric Company, Docket No.

50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of amendment request: February

24, 1995

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS)

Surveillance Requirement 4.6.1.7.4 and

its associated Bases to delete the

quarterly verification of the measured

leakage rate for containment mini-purge

supply and exhaust isolation valves.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

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.

The proposed revision to the T/S will not

adversely impact plant safety since the

requirement to perform the quarterly

surveillance will still be implemented to

verify valve leakage and seal degradation.

The mini-purge valves will still perform their

intended safety function to close within 5

seconds after receipt of an isolation signal.

2) Create the possibility of a new or

different kind of accident from any

previously evaluated.

There are no design changes being made

that would create a new type of accident or

malfunction and the method and manner of

plant operation remain unchanged. Deletion

of the individual leakage rate for these valves

does not affect the severity of any accident

previously evaluated. The consequences of a

valve failure or malfunction are not increased

by the removal of the acceptance criteria,

leakage rate will still be measured on a

quarterly basis as is currently done to

determine if the seals are degrading.

3) Involve a significant reduction in a

margin of safety.

There are no changes being made to the

safety limits or safety system settings that

would adversely impact plant safety. The

valves will still be surveilled on a quarterly

basis to verify leakage and seal degradation

to assure gross failure will not occur and that

containment integrity is maintained.

Based on the above discussions, it has been

determined that the requested Technical

Specification change does not involve a

significant increase in the probability or

consequences of an accident or create the

possibility of a new or different kind of

accident or condition over previous

evaluations; or involve a significant

reduction in a margin of safety. Therefore,

the requested license amendment does 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.
24922 Federal Register / Vol. 60, No. 90 / Wednesday, May 10, 1995 / Notices

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, N.W.,

Washington, DC 20037

NRC Project Director: Gail H. Marcus

Union Electric Company, Docket No.

50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of amendment request: April 17,

1995

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) Table 2.2-

1 and associated Bases to reduce

repeated alarms and partial reactor trips

related to the C-4 control system

interlock and the Overpower Delta-T

(OP[delta]T) reactor trip setpoint.

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 Final Safety

Analyses Report (FSAR) Chapter 15, WCAP10961-P for Category 1 plants such as

Callaway, and WCAP-11883 since no

hardware changes are proposed.

The OP[delta]T reactor trip function

provides protection against excessive power

(fuel rod integrity protection within the fuel

temperature design basis). No credit is taken

for the OP[delta]T trip in the Chapter 15

licensing basis accident analyses. The

[delta]T trip function is credited in nonlicensing basis analyses of various steamline

breaks.

The OP[delta]T trip will continue to

function in a manner consistent with the

plant design basis. There will be no change

to the OP[delta]T safety analysis limit listed

in FSAR Table 15.0-4. Therefore, there will

be no degradation in the performance of or

an increase in the number of challenges to

equipment assumed to function during an

accident situation.

The reactor trip system response time, as

defined in the Technical Specifications, will

be unaffected.

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, these

changes will not increase the probability or

consequences of an accident or malfunction.

2) Create the possibility of a new or

different kind of accident from any

previously evaluated.

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 safetyrelated plant system performs its safety

function. Revisions to the OP[delta]T values

for K4 and K6 will require scaling changes for

summing amplifier cards (NSA cards) in the

7300 Process Protection System. These

scaling changes are straightforward and

similar in nature to those performed to

implement OL Amendments 72 and 84

associated with the implementation of

relaxed axial offset control (RAOC) and a

revised OT[delta]T f1([delta]I) penalty

function. These scaling changes will not

affect the normal manner of plant operation.

There will be a reduction in the incidence of

C-4 alarms and partial reactor trips. There

will be less of a need to reduce power during

on-line surveillance testing.

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 kind of accident is not created.

3) Involve a significant reduction in a

margin of safety.

There will be no change to the Overpower

[delta]T safety analysis limit listed in FSAR

Table 15.0-4. Available setpoint calculation

margin will be used to increase the K4 value,

reflected as a new bias on a summing

amplifier card in each of the four protection

loops. This will also require corresponding

decreases in the OP[delta]T Total Allowance

and Allowable Value in Technical

Specification Table 2.2-1. Available margin

in the OP[delta]T trip protection function

will be used to decrease the K6 value,

reflected as a new gain on a summing

amplifier card in each of the four protection

loops.

As discussed above, the response time of

the OP[delta]T reactor trip function will

remain unchanged.

It has been confirmed that the Z and S

terms currently listed in Table 2.2-1 for the

OP[delta]T trip function will remain

conservative. The change in K4 will result in

a decrease in the Total Allowance and

Allowable Value for OP[delta]T; however,

this does not affect any margin of safety since

the safety analysis limit, which preserves the

overpower safety margin, is unchanged.

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 the

overpower limit, 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.

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: Gail H. Marcus

Vermont Yankee Nuclear Power

Corporation, Docket No. 50-271,

Vermont Yankee Nuclear Power

Station, Vernon, Vermont

Date of amendment request: October

28, 1994

Description of amendment request:

The proposed amendment would

remove the Neutron Monitoring System

(NMS) and Control Rod Position

instrumentation from the Vermont

Yankee Technical Specifications for

post-accident monitoring.

Administrative changes are also

proposed.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change to remove the

NMS and Control Rod Position

instrumentation from the Technical

Specifications for post-accident monitoring is

consistent with NRC requirements

concerning this instrumentation.

Wide Range Neutron Flux (NMS

instrumentation) is presently included in the

[boiling water reactor] BWR Standard

Technical Specifications, but the NRC has

recently determined [letter, USNRC to

VYNPC, dated April 29, 1993] that this

instrumentation need not meet R.G. 1.97

Category 1 criteria and that licensees may

request the removal of this instrumentation

from their post-accident monitoring

Technical Specifications. Control Rod

Position instrumentation is considered R.G.

1.97 Category 3 which is required to meet the

least stringent design and qualification

criteria as specified in this regulatory guide.

Testing, calibration and maintenance of

this instrumentation will continue to assure

operability of instrumentation. The portions

of the NMS and the Control Rod Position

instrumentation systems to be removed from

the post-accident monitoring Technical

Specifications do not perform any automatic

control or trip function. In addition, this

instrumentation does not provide

information that is required to permit the

control room operator to take manual actions

that are required for safety systems to

accomplish their safety functions for design

basis accident events.
Federal Register / Vol. 60, No. 90 / Wednesday, May 10, 1995 / Notices 24923

At a BWR, when all control rods are

inserted, these control rods cannot be

withdrawn without deliberate operator

action. The proposed change does not result

in any system hardware modification or new

plant configuration. The requested change to

post-accident monitoring instrumentation

does not impact any [Final Safety Analysis

Report] FSAR safety analysis involving the

NMS or Control Rod Position System. These

monitoring functions are not contributors to

the initiation of accidents.

The administrative changes to correct a

typographical error and instrument ranges

will have no effect on plant hardware, plant

design, safety limit setting or plant system

operation and therefore, do not modify or

add any initiating parameters that would

significantly increase the probability or

consequences of any previously analyzed

accident.

Therefore, it is concluded that there is not

a significant increase in the probability or

consequences of an accident previously

evaluated.

2. The function of the instrumentation to

be removed from the Technical

Specifications is for monitoring only. These

indications are not necessary for operators to

accomplish any safety functions.

The proposed change does not involve any

change in hardware, Technical Specification

setpoints, plant operation, redundancy,

protective function or design basis of the

plant. There is no impact on any existing

safety analysis or safety design limits. NMS

and Control Rod Position monitoring

functions do not initiate nuclear system

parameter variations which are considered

potential initiating causes of threats to the

fuel and the nuclear system process barrier.

As discussed above, the proposed

administrative change only corrects a

typographical error concerning equipment

identification numbers and listed instrument

ranges. This change does not affect any

equipment and they do not involve any

potential initiating events that would create

any new or different kind of accident.

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 change to remove the

NMS and Control Rod Position

instrumentation from the Technical

Specifications for post-accident monitoring

does not affect any existing safety margins.

The original NMS design basis for BWRs

never required a post-accident neutron

monitoring function since there are no design

basis accidents that rely on operator action to

control reactor power. This is also true for

Control Rod Position monitoring.

Existing Technical Specifications

requirements for automatic trip functions are

unaffected. Failure of the indication of

reactor power from the NMS or the Control

Rod Position System does not preclude the

ability of the reactor operator to determine

reactor power levels. Alternate indications

are available to ascertain reactor power.

These include reactor coolant boron

concentrations, flux levels from the

Traversing Incore Probe (TIP) System and the

status of plant parameters which are linked

to reactor power. In addition, alternate means

of determining reactor power have been

incorporated into the Emergency Operating

Procedures (EOPs).

Operation, testing and maintenance of this

instrumentation will remain the same.

System functions are the same. Post-accident

functional design criteria as described in

[BWR Owners Group Topical Report NEDO31558-A, dated March 29, 1993], and

approved by the NRC are satisfied by present

equipment installed at VY. NMS

instrumentation is still included in the

Technical Specifications for the [Reactor

Protection System] RPS. Control Rod Position

instrumentation does not perform any safety

function.

As discussed above, the proposed

administrative changes do not affect any

equipment involved in potential initiating

events or safety limits.

Based upon the above, it is concluded that

the proposed change does not involve a

significant reduction in a margin of safety.

Based upon the above, we conclude that

the proposed 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, VT 05301

Attorney for licensee: John A. Ritsher,

Esquire, Ropes and Gray, One

International Place, Boston, MA 02110-

2624

NRC Project Director: Phillip F.

McKee

Virginia Electric and Power Company,

Docket Nos. 50-338 and 50-339, North

Anna Power Station, Units No. 1 and

No. 2, Louisa County, Virginia

Date of amendment request: March

30, 1995

Description of amendment request:

The licensee is requesting temporary

changes to Technical Specifications (TS)

3.7.3.1, ‘‘Component Cooling Water

Subsystem - Operating,’’ and 3.7.4.1,

‘‘Service Water System - Operating,’’ for

NA-1&2. The proposed TS changes will

allow one of the two service water loops

to be isolated from the component

cooling water heat exchangers during

power operation in order to refurbish

the isolated service water headers.

NA-1&2 is currently pursuing

refurbishment of the 18-inch, 20-inch

and 24-inch diameter service water

supply and return lines to/from the NA1 and NA-2 component cooling heat

exchangers (CCHXs). Refurbishment of

this piping presents a challenge in that

it is not possible to isolate and plug or

blank the section to be worked in a 7-

day time period. The purpose of the

proposed change is to request temporary

changes to the existing servicewater

(SW) and component cooling water (CC)

TS to permit orderly and efficient

conduct of the pipe refurbishment

project during two-unit power

operation. Specifically, the licensee is

proposing to temporarily change TS

3.7.4.1 ‘‘Service Water System -

Operating’’ to allow operation of the SW

system with one independent source of

SW to/from the NA-1 and NA-2 CCHXs

for two periods of up to 49 days each.

This proposed change also allows the

automatic closure feature of the SW

valves to/from the CCHXs to be defeated

during the 49-day periods. In addition,

the licensee proposes to temporarily

change TS 3.7.3.1 ‘‘Component Cooling

Water Subsystem - Operating’’ with a

footnote which considers the CC

subsystems OPERABLE with only one

independent source of SW provided to/

from the CCHXs during these 49-day

periods. Further, the proposed change

would allow that during operation with

only one SW header available to/from

the CCHXs, the provisions of

Specification 3.0.4 would not be

applicable provided two SW loops are

capable of providing cooling for the

other operable plant components.

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:

Specifically, operation of North Anna

Power Station in accordance with the

proposed Technical Specifications

changes will not:

Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The piping refurbishment project and the

proposed temporary changes to the SW and

CC Technical Specifications have been

evaluated to assess their impact on the

normal operation of the SW and CC systems

and to ensure that the design basis safety

functions of each system are preserved. The

SW system is required to function during all

normal and emergency operating conditions.

During normal plant operation, the SW

system provides cooling water to the CCHXs,

charging pump coolers, instrument air

compressor coolers, and control room chiller

condensors of both units. During the two 49-

day periods, one header will [operate] with

its 24-inch piping to/from the CCHXs

temporarily blanked. To avoid operation of

the SW pump at abnormal conditions (low

flow) on this ‘‘partially deadlocked’’ header,

a temporary cross-connect will be installed to

by-pass the CCHXs.
24924 Federal Register / Vol. 60, No. 90 / Wednesday, May 10, 1995 / Notices

SW system operation with the crossconnect installed was evaluated for design

basis accident (DBA) conditions. The DBA

condition for the SW system is a loss-ofcoolant accident on one unit with

simultaneous loss-of-offsite-power to both

units. A SW system hydraulic analysis has

been performed to verify that adequate flow

is provided to the containment recirculation

spray heat exchangers (RSHXs) with the

temporary cross-connect installed and

throttled open assuming the occurrence of

the most limiting single failure. Therefore,

there is no increase in probability or

consequences of the DBA condition.

Utilizing only one SW header to supply

flow to the CCHXs has the potential to affect

the reliability of the CC system and all of the

equipment cooled by CC. The activities to be

performed during the refurbishment project

and the various system alignments required

have been evaluated using the Individual

Plant Examination (IPE) Probabilistic Safety

Assessment (PSA) model for North Anna

Power Station. This model is used in a

manner that is generally consistent with the

Nuclear Energy Institute (NEI)/Electric Power

Research Institute (EPRI) draft PSA

Applications Guide (Revision H). The effect

on the PSA model is a slight increase in the

frequency of reactor trips and an increase in

the probability of RHR failure.

The increased frequency of reactor trips is

due to the decreased reliability of the CC

system to supply cooling to the reactor

coolant pump (RCP) motors. When only one

SW header is available to the CCHXs, the

increased frequency of losing this single

header can be conservatively estimated by

combining the failure probability of both SW

pumps (approximately 1.5E-4 based on IPE

PSA data). Also considered was the

frequency of pipe rupture anywhere in the

single available header. When the single SW

header fails to supply cooling to the CCHXs,

the CC system will heatup causing

inadequate cooling for sustained operation of

the RCPs. Tripping these pumps results in a

reactor trip. The second SW header can be

expected to supply other equipment with

cooling. A sensitivity analysis shows the

increase in CDF as a result of the increased

reactor trip frequency to be less than 1E-8 per

year.

The CC system is also included in the PSA

model as a support system for RHR cooling.

The RHR system is used to reduce reactor

coolant system temperatures from 350°F (hot

shutdown) to 140°F (cold shutdown). The

only accident initiator that requires the unit

to be cooled down and placed on RHR

cooling are sequences which are initiated

with a steam generator tube rupture. (Note

that, for the North Anna plant design, RHR

is separate from the safety injection system

and the low head safety injection pumps.)

The increased probability for the loss of RHR

when only one SW header is available to the

CCHXs is estimated using fault tree analysis

and is dominated by the failure of both SW

pumps. The probability for the loss of both

SW pumps aligned to the CCHXs is estimated

to be 1.5E-4. The effect of this increase in

RHR failure probability was determined by

adding this probability to the top single event

in the RHR function and recalculating the

new CDF. The resulting increase in CDF as

a result of RHR system failure following a

steam generator tube rupture is less than 1E8 per year.

The CC system is further included in the

PSA model as part of the loss of RCP seal

cooling as an initiating event and as a loss

of function during other initiating event

scenarios. The effect on the probability for a

loss of RCP seal cooling due to losing CC

cooling to the RCP thermal barriers is

negligible due to the high reliability of the

charging system to provide seal injection.

The total effect of this pipe refurbishment

project was estimated by a sensitivity

analysis combining both the change in the

reactor trip initiating event frequency and the

increased failure probability of RHR resulting

in less than a 1E-6 per year increase in CDF.

Since this project will not affect the

containment systems, there would not be any

significant change in off-site dose, except that

resulting directly from the increase in CDF.

These minor increases in CDF and off-site

dose are less than what is defined as risk

significant in the NEI/EPRI draft PSA

Applications Guide.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed temporary Technical

Specifications changes do not affect the basic

method of operation of the SW or CC

systems. The purpose of the proposed

changes is to permit extended operation of

the CC system with one independent source

of SW cooling. During the project, there will

be a significant time period when all the

CCHXs are aligned to one SW loop, the

possibility of an interruption of SW supply

to the heat exchangers during a DBA is

eliminated by defeating the closure of the 24-

inch SW isolation MOVs to the CCHXs on a

SI/CDA signal. Both SW headers will be

available for equipment required for safe

shutdown of the units (i.e., RSHXs, charging

pumps, and CR/ESGR chillers). The SW pipe

repair activities and the installation/removal

of the SW cross-connect piping do not create

the possibility for a malfunction of

equipment different than previously

evaluated. Therefore, implementation of the

restoration project and approval of the

proposed Technical Specifications changes

will not introduce any new accident

initiators nor affect the performance of

accident mitigation systems.

3. Involve a significant reduction in a

margin of safety.

The proposed changes to the schedule only

provide operational flexibility to perform the

required SW pipe refurbishment. The

Technical Specifications continue to require

the SW and CC systems to remain functional

during the period with a single SW supply

to the CCHXs. As stated in item (1) above, the

SW system is fully capable of performing its

DBA function during the course of the pipe

refurbishment project with the proposed

Technical Specification changes in place.

The effect of this pipe refurbishment project

on CC system reliability was estimated by a

sensitivity analysis combining both the

change in the reactor trip initiating event

frequency and the increased failure

probability of RHR resulting in less than a

1E-6 per year increase in CDF. Since this

project will not affect the containment

systems, there would not be any significant

change in off-site dose, except that resulting

directly from the increase in CDF. These

minor increases in CDF and off-site dose are

less than what is defined as risk significant

in the NEI/EPRI draft PSA Applications

Guide. Therefore, there is not a significant

reduction in margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: The Alderman Library, Special

Collections Department, University of

Virginia, Charlottesville, Virginia 22903-

2498.

Attorney for licensee: Michael W.

Maupin, Esq., Hunton and Williams,

Riverfront Plaza, East Tower, 951 E.

Byrd Street, Richmond, Virginia 23219.

NRC Project Director: David B.

Matthews

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.

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:

March 31, 1995

Brief description of amendments: The

proposed amendments would provide

an exception to Technical Specification

(TS) 3.0.4. TS 3.0.4 allows entry of a

unit into another operational condition

only if the conditions of the Limiting

Conditions for Operation (LCOs) are met

without reliance on TS action

statements. The exception requested by
Federal Register / Vol. 60, No. 90 / Wednesday, May 10, 1995 / Notices 24925

the licensee would allow a change in a

unit’s operational condition in a specific

situation in which the unit’s LCO

concerning the minimum number of

operable offsite power circuits is not

fully satisfied. Specifically, the

exception would allow an operational

mode change of a unit if the second unit

is in Operational Condition 4 or 5 (i.e.,

cold shutdown or refueling) and one of

the second unit’s offsite power circuits

is inoperable.

Date of publication of individual

notice in Federal Register: April 13,

1995 (60 FR 18860)

Expiration date of individual notice:

May 15, 1995

Local Public Document Room

location: The University of North

Carolina at Wilmington, William

Madison Randall Library, 601 S. College

Road, Wilmington, North Carolina

28403-3297

Duquesne Light Company, et al., Docket

No. 50-412, Beaver Valley Power

Station, Unit No. 2, Shippingport,

Pennsylvania

Date of amendment request: April 10,

1995, as supplemented April 12, 1995

Brief description of amendment

request: The proposed amendment

would revise Technical Specification

(TS) 4.6.2.2.d to delete the reference to

the specific test acceptance criteria for

the Containment Recirculation Spray

Pumps and replace the specific test

acceptance criteria with reference to the

requirements of the Inservice Testing

(IST) Program. In addition, the 18-

month test frequency would be replaced

with the test frequency requirements

specified in the IST Program. The

current footnote (1) pertaining to the

performance of recirculation spray

pump 2RSS*P21A would be deleted.

Date of publication of individual

notice in Federal Register: April 18,

1995 (60 FR 19417)

Expiration date of individual notice:

May 18, 1995

Local Public Document Room

location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa,

Pennsylvania 15001.

Maine Yankee Atomic Power Company,

Docket No. 50-309, Maine Yankee

Atomic Power Station, Lincoln County,

Maine

Date of amendment request: April 14,

1995

Description of amendment request:

The proposed amendment would revise

the Technical Specifications to allow

the use of the Westinghouse Electric

Corporation sleeving process for

repairing steam generator tubes.

Date of publication of individual

notice in Federal Register: April 21,

1995 (60 FR 19969)

Expiration date of individual notice:

May 22, 1995

Local Public Document Room

location: Wiscasset Public Library, High

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

04578.

Notice Of Issuance Of Amendments To

Facility Operating Licenses

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application

complies with the standards and

requirements of the Atomic Energy Act

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

Commission’s rules and regulations.

The Commission has made appropriate

findings as required by the Act and the

Commission’s rules and regulations in

10 CFR Chapter I, which are set forth in

the license amendment.

Notice of Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for A Hearing in

connection with these actions was

published in the Federal Register as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the applications for

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

the Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, the Gelman Building, 2120 L

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

local public document rooms for the

particular facilities involved.

Carolina Power & Light Company, et

al., Docket Nos. 50-325 and 50-324,

Brunswick Steam Electric Plant, Units

1 and 2, Brunswick County, North

Carolina

Date of application for amendments:

July 22, 1994, as supplemented on

March 6, 1995

Brief description of amendments: The

amendments change the Technical

Specifications to implement a

performance based assessment program,

including corresponding organizational

and functional changes. Specifically, the

changes affect the independent review

function, the independent assessment of

plant activity and the Independent

Safety Engineering Group. These

functions will be performed by the

Nuclear Assessment Section (NAS). The

NAS’s fundamental role will be to: (1)

assist plant management in the early

identification of issues that may prevent

the plant from achieving quality, and (2)

ensure effective correction of

deficiencies.

Date of issuance: April 18, 1995

Effective date: April 18, 1995

Amendment Nos.: 177 and 208

Facility Operating License Nos. DPR71 and DPR-62. Amendments revise the

Technical Specifications.

Date of initial notice in Federal

Register: August 31, 1994 (59 FR 45017)

The March 6, 1995, submittal added

Radiation Protection to the list of

assessments in TS 6.5.5.2 and reworded

Section 6.5.4.4, but did not change the

no significant hazards consideration

determination as published in the

Federal Register. The Commission’s

related evaluation of the amendments is

contained in a Safety Evaluation dated

April 18, 1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: University of North Carolina at

Wilmington, William Madison Randall

Library, 601 S. College Road,

Wilmington, North Carolina 28403-

3297.

Carolina Power & Light Company,

Docket No. 50-261, H. B. Robinson

Steam Electric Plant, Unit No. 2,

Darlington County, South Carolina

Date of application for amendment:

June 18, 1992, as supplemented

December 8, 1992 and February 3, 1995

Brief description of amendment: The

amendment adds limiting conditions of

operation and surveillance requirements

for the pressurizer power-operated relief

valves and their associated block valves

whenever average temperature is above

350 degrees F or the reactor is critical.

Specifications are also added for lowtemperature overpressure protection
24926 Federal Register / Vol. 60, No. 90 / Wednesday, May 10, 1995 / Notices

whenever average temperature is less

than 350 degrees F and the reactor

coolant system is not vented to the

containment.

Date of issuance: April 14, 1995

Effective date: April 14, 1995

Amendment No.: 162

Facility Operating License No. DPR23. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: September 2, 1992 (57 FR

40208). Renoticed on March 1, 1995 (60

FR 11127) The December 8, 1992, letter

corrected a typographical error and did

not affect the no significant hazards

consideration. The licensee’s letter

dated February 3, 1995, proposed a

revision to the TS regarding block valve

testing in accordance with Generic

Letter 90-06 recommendations. The

proposed change was noticed on March

1, 1995 (60 FR 11127). The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated April 14, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Hartsville Memorial Library,

147 West College Avenue, Hartsville,

South Carolina 29550.

Carolina Power & Light Company,

Docket No. 50-261, H. B. Robinson

Steam Electric Plant, Unit No. 2,

Darlington County, South Carolina

Date of application for amendment:

November 4, 1994, as supplemented

April 6, 1995.

Brief description of amendment: The

amendment changes the testing

frequency of the turbine overspeed

protection valves from monthly to

quarterly to implement an enhancement

recommended by Generic Letter 93-05,

‘‘Line-Item Technical Specification

Improvements to Reduce Surveillance

Requirements for Testing During Power

Operation.’’ The April 6, 1995 submittal

provided clarifying information only,

and did not change the proposed no

significant hazards determination.

Date of issuance: April 27, 1995

Effective date: April 27, 1995

Amendment No.: 164

Facility Operating License No. DPR23. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: December 7, 1994 (59 FR

63115) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

April 27, 1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: Hartsville Memorial Library,

147 West College Avenue, Hartsville,

South Carolina 29550.

Carolina Power & Light Company, et

al., Docket No. 50-400, Shearon Harris

Nuclear Power Plant, Unit 1, Wake and

Chatham Counties, North Carolina

Date of application for amendment:

January 19, 1995, as supplemented

March 20, 1995

Brief description of amendment: The

amendment revises Technical

Specification 4.0.3 and its associated

Bases to provide for a delay period in

which to perform a surveillance that

was not performed within its specified

frequency.

Date of issuance: April 17, 1995

Effective date: April 17, 1995

Amendment No.: 56

Facility Operating License No. NPF63. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: February 15, 1995 (60 FR

8742) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

April 17, 1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh,

North Carolina 27605.

Carolina Power & Light Company, et

al., Docket No. 50-400, Shearon Harris

Nuclear Power Plant, Unit 1, Wake and

Chatham Counties, North Carolina

Date of application for amendment:

July 22, 1994, as supplemented March 6,

1995.

Brief description of amendment: The

amendment implements a performancebased assessment program, including

corresponding organizational and

functional changes. Specifically, the

changes affect the Independent Review

(IR) function, the independent

assessment of plant activity and the

Independent Safety Engineering Group.

These functions will be performed by

the proposed Nuclear Assessment

Section (NAS). The NAS will perform

internal evaluations and assessment

activities and serve as plant

management’s staff for the objective

oversight of plant performance relating

to nuclear safety, reliability, and quality.

The NAS’s fundamental role will be to:

(1) assist plant management in the early

identification of issues which may

prevent the plant from achieving quality

performance on a sustained basis; and

(2) ensure effective correction of

deficiencies.

Date of issuance: April 21, 1995

Effective date: April 21, 1995

Amendment No.: 57

Facility Operating License No. NPF63. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: August 31, 1994 (59 FR 45019)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated April 21, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh,

North Carolina 27605.

Commonwealth Edison Company,

Docket No. 50-374, LaSalle County

Station, Unit 2, LaSalle County, Illinois

Date of application for amendment:

March 31, 1995

Brief description of amendment: The

amendment revises the safety/relief

valve (SRV) safety function lift setting

allowable tolerance band from -3/+1%

to plus or minus 3% and includes a

requirement for the lift settings to be

within plus or minus 1% of the

technical specification limit following

testing.

Date of issuance: April 25, 1995

Effective date: Immediately, to be

implemented prior to restart from the

sixth refueling outage.

Amendment No.: 89

Facility Operating License No. NPF18: The amendment revised the

Technical Specifications. Public

comments requested as to proposed no

significant hazards consideration: Yes

(60 FR 17590 dated April 6, 1995). That

notice provided an opportunity to

submit comments on the Commission’s

proposed no significant hazards

consideration determination. No

comments have been received. This

notice also provided for an opportunity

to request a hearing by May 8, 1995, but

indicated that if the Commission makes

a final no significant hazards

consideration determination, any such

hearing would take place after issuance

of the amendment. The Commission’s

related evaluation of the amendment,

finding of exigent circumstances, and

final determination of no significant

hazards consideration is contained in a

Safety Evaluation dated April 25, 1995.

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60690

Local Public Document Room

location: Jacobs Memorial Library,

Illinois Valley Community College,

Oglesby, Illinois 61348.
Federal Register / Vol. 60, No. 90 / Wednesday, May 10, 1995 / Notices 24927

Consolidated Edison Company of New

York, Docket No. 50-247, Indian Point

Nuclear Generating Unit No. 2,

Westchester County, New York

Date of application for amendment:

June 1, 1994, as supplemented on

January 25, 1995, April 7, April 19, and

April 26, 1995.

Brief description of amendment: The

amendment revises Technical

Specification Section 3.10 to allow

extended Rod Position Indication (RPI)

deviation limits and on-line calibration

of the RPI channels for cycle 13 only.

Date of issuance: April 28, 1995

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 182

Facility Operating License No. DPR26: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 20, 1994 (59 FR 37069).

The January 25, April 7, April 19, and

April 26, 1995, submittals provided

clarifying information that did not affect

the initial no significant hazards

determination. The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

April 28, 1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10610.

Consumers Power Company, Docket

No. 50-255, Palisades Plant, Van Buren

County, Michigan

Date of application for amendment:

February 10, 1995, as supplemented

March 27 and 30, 1995

Brief description of amendment: This

amendment revises the Technical

Specifications to allow a one-time

deferral of several 18-month interval

surveillance tests until the upcoming

scheduled refueling outage to avoid the

necessity of imposing a plant shutdown

solely for the sake of their performance.

In the March 30, 1995, letter the license

also withdrew its request for deferral of

several surveillance tests.

Date of issuance: April 20, 1995

Effective date: April 20, 1995

Amendment No.: 164

Facility Operating License No. DPR20. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 1, 1995 (60 FR 11131)

The March 27 and 30, 1995, letters

provided clarifying information which

was within the scope of the initial

notice and did not affect the staff’s

original proposed no significant hazards

consideration determination.The

Commission’s related evaluation of the

amendment and of the withdrawalof

certain surveillance test deferrals is

contained in a Safety Evaluation dated

April 20, 1995.No significant hazards

consideration comments received: No.

Local Public Document Room

location: Van Wylen Library, Hope

College, Holland, Michigan 49423.

Duke Power Company, Docket Nos. 50-

369 and 50-370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg

County, North Carolina

Date of application for amendments:

February 23, 1995, as supplemented by

letter dated March 21, 1995.

Brief description of amendments: The

amendments revise Technical

Specification (TS) 3.8.2.1 and TS 3.8.3.1

to allow installation of replacement

equipment in response to an Electrical

Distribution Systems Functional

Inspection, conducted by the NRC in

July 1991. The existing breaker

arrangement could result in a trip of

both the battery and main breakers if a

fault occurs on one of the 125-V dc

panelboards. The licensee committed to

have these breakers replaced in 1995

with a better coordinated design to

eliminate the concern.

Date of issuance: April 14, 1995

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment Nos.: 155 and 137

Facility Operating License Nos. NPF9 and NPF-17: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: March 8, 1995 (60 FR 12791)

The March 21, 1995, letter provided

clarifying information that did not

change the scope of the February 23,

1995, 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 April 14, 1995.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Atkins Library, University of

North Carolina, Charlotte (UNCC

Station), North Carolina 28223.

Duquesne Light Company, et al., Docket

Nos. 50-334 and 50-412, Beaver Valley

Power Station, Unit Nos. 1 and 2,

Shippingport, Pennsylvania

Date of application for amendments:

September 2, 1992

Brief description of amendments:

These amendments revise the Appendix

A Technical Specifications relating to

the required surveillance frequency for

comparing the incore and excore axial

imbalance. The revision requires

comparison of the incore to excore axial

imbalance at least once every 31

Effective Full Power Days above 15

percent of rated thermal power rather

than once every 31 days above 15

percent of rated thermal power as was

previously required.

Date of issuance: April 26, 1995

Effective date: April 26, 1995

Amendment Nos.: 186 and 67

Facility Operating License Nos. DPR66 and NPF-73: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: October 14, 1992 (57 FR

47128) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

April 26, 1995.No significant hazards

consideration comments received: No.

Local Public Document Room

location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa,

Pennsylvania 15001.

Entergy Operations, Inc., Docket Nos.

50-313 and 50-368, Arkansas Nuclear

One, Unit Nos. 1 and 2, Pope County,

Arkansas

Date of amendment request: June 20,

1994

Brief description of amendments: The

amendments relocated the requirements

of the quality assurance program and

the security and emergency plans from

the administrative controls section of

the technical specifications to the

respective licensee-controlled

documents.

Date of issuance: April 25, 1995

Effective date: 90 days from date of

issuance

Amendment Nos.: 179 and 160

Facility Operating License Nos. DPR51 and NPF-6. Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: August 17, 1994 (59 FR 42340)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated April 25, 1995.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas

72801.

Entergy Operations, Inc., Docket No.

50-382, Waterford Steam Electric

Station, Unit 3, St. Charles Parish,

Louisiana

Date of amendment request: August 5,

1993

Brief description of amendment: The

amendment removed the requirements

associated with loose-part detection
24928 Federal Register / Vol. 60, No. 90 / Wednesday, May 10, 1995 / Notices

system from the Technical

Specifications for Waterford Steam

Electric Station, Unit 3. These

requirements will be incorporated into

the Waterford 3 Updated Final Safety

Analysis Report and maintained under

the provisions of 10 CFR 50.59.

Date of issuance: April 20, 1995

Effective date: April 20, 1995

Amendment No.: 104

Facility Operating License No. NPF38. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: September 15, 1993 (58 FR

48382) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

April 20, 1995.No significant hazards

consideration comments received: No.

Local Public Document Room

location: University of New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, Louisiana 70122.

Entergy Operations, Inc., Docket No.

50-382, Waterford Steam Electric

Station, Unit 3, St. Charles Parish,

Louisiana

Date of amendment request: April 4,

1995, as supplemented by letter dated

April 5, 1995

Brief description of amendment: The

amendment changed the Appendix A

Technical Specifications (TSs) by

revising the TSs for moderator

temperature coefficient. The

amendment approves a one time

deviation by excluding the two-thirds

end-of-cycle moderator temperature

coefficient test requirement for Cycle 7.

Date of issuance: April 27, 1995

Effective date: April 27, 1995

Amendment No.: 105

Facility Operating License No. NPF38. Amendment revised the Technical

Specifications.Public comments

requested as to proposed no significant

hazards consideration: Yes (60 FR

18431, dated April 11, 1995). The notice

provided an opportunity to submit

comments on the Commission’s

proposed no significant hazards

consideration determination. No

comments have been received. The

notice also provided for an opportunity

to request a hearing by May 11, 1995,

but stated that any such hearing would

take place after issuance of the

amendment. The Commission’s related

evaluation of the amendments, finding

of exigent circumstances, and final

determination of no significant hazards

consideration is contained in a Safety

Evaluation dated April 27, 1995.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: University of New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, Louisiana 70122.

Entergy Operations, Inc., System

Energy Resources, Inc., South

Mississippi Electric Power Association,

and Mississippi Power & Light

Company, Docket No. 50-416, Grand

Gulf Nuclear Station, Unit 1, Claiborne

County, Mississippi

Date of application for amendment:

October 12, 1994

Brief description of amendment: The

amendment removed License Condition

2.C.(26) related to Turbine Disk

Integrity.

Date of issuance: April 17, 1995

Effective date: April 17, 1995

Amendment No: 121

Facility Operating License No. NPF29. Amendment revises the license.

Date of initial notice in Federal

Register: November 9, 1994 (59 FR

55868) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

April 17, 1995. No significant hazards

consideration comments received: No

Local Public Document Room

location: Judge George W. Armstrong

Library, 220 S. Commerce Street,

Natchez, MS 39120.

Florida Power and Light Company, et

al., Docket Nos. 50-335 and 50-389, St.

Lucie Plant, Unit Nos. 1 and 2, St. Lucie

County, Florida

Date of application for amendments:

May 23, 1994

Brief description of amendments:

These amendments will relocate the

seismic monitoring instrumentation

Limiting Conditions of Operation,

Surveillance Requirements and the

associated tables contained in Technical

Specifications 3.3.3.3, 4.3.3.3.1 and

4.3.3.3.2 to the Updated Final Analysis

Report.

Date of issuance: April 25, 1995

Effective date: April 25, 1995

Amendment Nos.: 135 and 74

Facility Operating License Nos. DPR67 and NPF-16: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: July 6, 1994 (59 FR 34664) The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated April 25, 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:

January 20, 1995

Brief description of amendments: The

amendments revise the administrative

requirements of Technical Specification

(TS) 6.4.1.2 related to the areas of

technical expertise that must be

represented on the Plant Review Board

(PRB). The licensee proposed this

change in order to maintain an

appropriate level of PRB expertise after

the implementation of a planned

reorganization that includes combining

certain departments that are listed

separately in the current TS 6.4.2.1

requirements.

Date of issuance: April 27, 1995

Effective date: As of the date of

issuance to be implemented within 30

days

Amendment Nos.: 84 and 62

Facility Operating License Nos. NPF68 and NPF-81: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: February 6, 1995 (60 FR 7077)

The April 4, 1995, letter provided

additional and clarifying information

that did not change the scope of the

January 20, 1995, application or the

initial proposed no significant hazards

consideration determination.The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated April 27, 1995. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Burke County Library, 412

Fourth Street, Waynesboro, Georgia

30830.

Houston Lighting & Power Company,

City Public Service Board of San

Antonio, Central Power and Light

Company, City of Austin, Texas, Docket

Nos. 50-498 and 50-499, South Texas

Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request:

November 8, 1994, as supplemented by

letter dated March 14, 1995.

Brief description of amendments: The

amendments require that only one of the

two battery chargers associated with

each Class 1E 125-VDC Channel I and

Channel IV is operable.

Date of issuance: April 17, 1995

Effective date: April 17, 1995, to be

implemented within 31 days.

Amendment Nos.: Unit 1 -

Amendment No. 73; Unit 2 -

Amendment No. 62
Federal Register / Vol. 60, No. 90 / Wednesday, May 10, 1995 / Notices 24929

Facility Operating License Nos. NPF76 and NPF-80. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: December 7, 1994 (59 FR

63123) The March 14, 1995, supplement

withdrew that portion of the proposed

amendments where the required

wording was already incorporated into

the Technical Specifications by

amendments issued on February 14,

1995, in response to another

amendment request. The March 14,

1995, letter also provided clarifying

information and did not change the

original no significant hazards

consideration determination. The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated April 17, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Wharton County Junior

College, J. M. Hodges Learning Center,

911 Boling Highway, Wharton, TX

77488.

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

Duane Arnold Energy Center, Linn

County, Iowa

Date of application for amendment:

November 10, 1994, as supplemented

March 1, 1995

Brief description of amendment: The

amendment revises the Duane Arnold

Energy Center Technical Specification

Section 3.2.A to refer to the Offsite Dose

Assessment Manual for the setpoint of

the Offgas Stack Radiation Monitor and

makes the ‘‘Applicable Operating

Mode’’ and the ‘‘Action’’ statements for

these instruments consistent with the

required function. The Action statement

for the other instruments which initiate

Secondary Containment isolation is also

revised to be consistent with the current

practice and with the function of those

instruments. The Basis is also revised to

add further description of the function

and requirements.

Date of issuance: April 25, 1995

Effective date: April 25, 1995

Amendment No.: 209

Facility Operating License No. DPR49. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: December 21, 1994 (59 FR

65815) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

April 25, 1995.The March 1, 1995,

submittal provided supplemental

information that did not change the

initial proposed no significant hazards

consideration determination.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Cedar Rapids Public Library,

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

Iowa 52401.

Indiana Michigan Power Company,

Docket Nos. 50-315 and 50-316, Donald

C. Cook Nuclear Plant, Unit Nos. 1 and

2, Berrien County, Michigan

Date of application for amendments:

April 6, 1994

Brief description of amendments: The

amendments delete part of License

Condition 2.C.(4) to Operating License

No. DPR-58 and part of License

Condition 2.C.(3)(o) to Operating

License No. DPR-74 on fire protection.

The related fire protection safety

evaluation also changes three of the

modifications listed in Table 1 of the

Safety Evaluation Report of July 31,

1979, that supported amendments nos.

31 and 12 to Operating Licenses No.

DPR-58 and No. DPR-74, respectively.

Date of issuance: April 19, 1995

Effective date: April 19, 1995

Amendment Nos.: 194 and 180

Facility Operating License Nos. DPR58 and DPR-74. Amendments revised

the Facility Operating Licenses.

Date of initial notice in Federal

Register: September 28, 1994 (59 FR

49429) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

April 19, 1995.No significant hazards

consideration comments received: No.

Local Public Document Room

location: Maud Preston Palenske

Memorial Library, 500 Market Street, St.

Joseph, Michigan 49085.

Niagara Mohawk Power Corporation,

Docket No. 50-410, Nine Mile Point

Nuclear Station, Unit 2, Oswego

County, New York

Date of application for amendment:

March 9, 1995

Brief description of amendment: The

amendment revises Technical

Specification Section 4.6.1.2.a, Primary

Containment/Containment Leakage.

This change allows the second Type A

containment leak rate test to be

performed at refueling outage 5 instead

of refueling outage 4, consistent with an

exemption to 10 CFR Part 50, Appendix

J which has been granted.

Date of issuance: April 24, 1995

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 65

Facility Operating License No. NPF69: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: March 23, 1995 (60 FR 15310)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated April 24, 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.

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:

December 2, 1994

Brief description of amendment: The

amendment changes the Millstone 3

Technical Specification Table 4.3-1 by

adding a note for certain Functional

Units which would allow an entry into

Mode 2 or Mode 1 before performing

calibration for the power range

detectors.

Date of issuance: April 26, 1995

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 109

Facility Operating License No. NPF49. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 1, 1995 (60 FR 6304)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated April 26, 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 LondonTurnpike, Norwich, CT

06360.

Northern States Power Company,

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

Island Nuclear Generating Plant, Unit

Nos. 1 and 2, Goodhue County,

Minnesota

Date of application for amendments:

February 23 and March 3, 1995

Brief description of amendments: The

amendments revise the Prairie Island

Technical Specifications section 4.4.A.5

to add the phrase ‘‘and all approved

exemptions.’’ after the reference to 10

CFR Part 50, Appendix J. This revision

will allow implementation of approved

exemptions from the testing schedule

requirements of 10 CFR Part 50,

Appendix J, Section III.D.1.(a).

Date of issuance: April 18, 1995

Effective date: April 18, 1995, with

full implementation within 30 days.

Amendment Nos.: 117 and 110

Facility Operating License Nos. DPR42 and DPR-60. Amendments revised

the Technical Specifications.
24930 Federal Register / Vol. 60, No. 90 / Wednesday, May 10, 1995 / Notices

Date of initial notice in Federal

Register: March 15, 1995 (60 FR 14025).

The March 3, 1995, letter provided

clarifying information within the scope

of the original submittal and did not

change the staff’s initial proposed no

significant hazards consideration

determination. The Commission’s

related evaluation of the amendments is

contained in a Safety Evaluation dated

April 18, 1995.No significant hazards

consideration comments received: No.

Local Public Document Room

location: Minneapolis Public Library,

Technology and Science Department,

300 Nicollet Mall, Minneapolis,

Minnesota 55401.

Pacific Gas and Electric Company,

Docket Nos. 50-275 and 50-323, Diablo

Canyon Nuclear Power Plant, Unit Nos.

1 and 2, San Luis Obispo County,

California

Date of application for amendments:

August 17, 1994 (Reference LAR 94-06)

Brief description of amendments: The

proposed amendments increase the

allowed outage time of the refueling

water storage tank (RWST) for

adjustment of boron concentration from

one to eight hours as contained in

Technical Specifications Section 3.5.5.

Date of issuance: April 14, 1995

Effective date: April 14, 1995, to be

implemented within 30 days of issuance

Amendment Nos.: Unit 1 -

Amendment No. 101; Unit 2 -

Amendment No. 100

Facility Operating License Nos. DPR80 and DPR-82: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: October 12, 1994 (59 FR

51621) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

April 14, 1995.No significant hazards

consideration comments received: No.

Local Public Document Room

location: California Polytechnic State

University, Robert E. Kennedy Library,

Government Documents and Maps

Department, San Luis Obispo, California

93407.

PECO Energy Company, Public Service

Electric and Gas Company Delmarva

Power and Light Company, and

Atlantic City Electric Company, Docket

Nos. 50-277 and 50-278, Peach Bottom

Atomic Power Station, Unit Nos. 2 and

3, York County, Pennsylvania

Date of application for amendments:

October 25, 1994 as supplemented

February 13, 1995.

Brief description of amendments: The

amendment clarifies the technical

specification surveillance requirements

and bases for high pressure coolant

injection system testing at low reactor

pressure.

Date of issuance: April 18, 1995

Effective date: April 18,

1995Amendments Nos.: 200 and 202

Facility Operating License Nos. DPR44 and DPR-56: The amendments

revised the Technical Specifications.

Public comments requested as to

proposed no significant hazards

consideration: Yes (59 FR 55498 dated

November 7, 1994). That notice

provided an opportunity to submit

comments on the Commission’s

proposed no significant hazards

consideration determination, and also

provided an opportunity to request a

hearing by December 7, 1994. No

comments or requests for hearings have

been received. The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

April 18, 1995.

Local Public Document Room

location: Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Education

Building, Walnut Street and

Commonwealth Avenue, Box 1601,

Harrisburg, Pennsylvania 17105.

Philadelphia Electric Company, Public

Service Electric and Gas Company,

Delmarva Power and Light Company,

and Atlantic City Electric Company,

Docket No. 50-278, Peach Bottom

Atomic Power Station, Unit No. 3, York

County, Pennsylvania

Date of application for amendment:

January 13, 1995 as supplemented by

letters dated March 14, 1995 and April

12, 1995.

Brief description of amendment: The

requested changes would modify Tables

3.7.1 and 3.7.4 of the Technical

Specifications (TS) to reflect a change in

the number of primary containment

penetrations and isolation valves

associated with the traversing in-core

probe (TIP) system. In order to prevent

confusion with the staff’s review of

PECO’s September 29, 1994 application

to implement improved TS at Peach

Bottom, the staff is issuing the license

amendment regarding the TIP system for

Unit 3 only.

Date of issuance: April 24, 1995

Effective date: April 24, 1995

Amendment No.: 203

Facility Operating License No. DPR56: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 1, 1995 (60 FR 11139)

The March 14, 1995 and April 12, 1995,

letters provided clarifying information

that did not change the initial proposed

no significant hazards consideration

determination. The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

April 24, 1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Education

Building, Walnut Street and

Commonwealth Avenue, Box 1601,

Harrisburg, Pennsylvania 17105.

Public Service Electric & Gas Company,

Docket Nos. 50-272 and 50-311, Salem

Nuclear Generating Station, Unit Nos. 1

and 2, Salem County, New Jersey

Date of application for amendments:

September 20, 1994

Brief description of amendments: The

amendments modify the Technical

Specifications for auxiliary feedwater to

reduce the secondary side steam

pressure required for testing the turbine

driven auxiliary feedwater pump and to

allow 24 hours to perform the test after

reaching the minimum test pressure.

Date of issuance: April 17, 1995

Effective date: As of the date of

issuance to be implemented within 60

days.

Amendment Nos.: 165 and 146

Facility Operating License Nos. DPR70 and DPR-75. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: November 9, 1994 (59 FR

55889) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

April 17, 1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: Salem Free Public Library, 112

West Broadway, Salem, New Jersey

08079.

Public Service Electric & Gas Company,

Docket No. 50-311, Salem Nuclear

Generating Station, Unit No. 2, Salem

County, New Jersey

Date of application for amendment:

February 3, 1994, as supplemented

September 19, 1994, and November 23,

1994

Brief description of amendment: The

amendment changes the Technical

Specifications to reflect a reduction in

Reactor Coolant System flow.

Date of issuance: April 17, 1995

Effective date: April 17, 1995

Amendment No.: 147

Facility Operating License No. DPR75: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: March 15, 1995 (60 FR 14028)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated April 17, 1995.No
Federal Register / Vol. 60, No. 90 / Wednesday, May 10, 1995 / Notices 24931

significant hazards consideration

comments received: No

Local Public Document Room

location: Salem Free Public Library, 112

West Broadway, Salem, New Jersey

08079.

Rochester Gas and Electric

Corporation, Docket No. 50-244, R. E.

Ginna Nuclear Power Plant, Wayne

County, New York

Date of application for amendment:

March 13, 1995

Brief description of amendment: This

amendment revises Technical

Specification 4.4.2.4.a to replace

specific leakage rate testing frequencies

for containment isolation valves that

require Type C testing for the 1995

refueling outage to be completed prior

to exiting Cold Shutdown tentatively

scheduled for April 27, 1995.

Date of issuance: April 26, 1995

Effective date: April 26, 1995

Amendment No.: 59

Facility Operating License No. DPR18: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 22, 1995 (60 FR 15167)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated April 26, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Rochester Public Library, 115

South Avenue, Rochester, New York

14610.

TU Electric Company, Docket Nos. 50-

445 and 50-446, Comanche Peak Steam

Electric Station, Unit Nos. 1 and 2,

Somervell County, Texas

Date of amendment request: February

23, 1994 (LAR 94-005, TXX-94034)

Brief description of amendments:

These amendments changed Technical

Specification (TS) 3/4.5.1, ‘‘Emergency

Core Cooling Systems, Accumulators,

Cold Leg Injection,’’ to: 1) allow a one

hour allowed outage time following

discovery of a closed cold leg injection

accumulator discharge isolation valve in

Modes 1, 2, or 3; 2) eliminate the

redundant requirement to reverify

accumulator boron concentration

following fill from the refueling water

storage tank RWST; 3) remove the

accumulator water level and pressure

channel analog channel operational test

and channel calibration from the TSs;

and 4) change the accumulator limits to

analysis values rather than indicated

values. Also these amendments

modified TS 3/4.5.2, ‘‘ECCS Subsystems

- Tavg ≤ 350°F’’ to reduce the visual

inspection frequency following

containment entries.

Date of issuance: April 27, 1995

Effective date: April 27, 1995, to be

implemented within 30 days.

Amendment Nos.: Unit 1 -

Amendment No. 40; Unit 2 -

Amendment No. 26

Facility Operating License Nos. NPF87 and NPF-89. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: August 3, 1994 (59 FR 39597)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated April 27, 1995.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: University of Texas at

Arlington Library, Government

Publications/Maps, 702 College, P.O.

Box 19497, Arlington, TX 76019.

TU Electric Company, Docket Nos. 50-

445 and 50-446, Comanche Peak Steam

Electric Station, Unit Nos. 1 and 2,

Somervell County, Texas

Date of amendment request: August 9,

1994, (LAR 94-013, TXX-94211)

Brief description of amendments:

These amendments eliminated ‘‘High

Negative Neutron Flux Rate’’ reactor trip

function based on analyses which

demonstrate that the protection

provided by the reactor trip function is

not required. The affected Technical

Specifications were: 2.2.1, ‘‘Reactor Trip

System Instrumentation Setpoints,’’ and

3/4.3.1, ‘‘Reactor Trip System

Instrumentation.’’ Also affected was

Bases Section 2.2.1.

Date of issuance: April 17, 1995

Effective date: April 17, 1995, to be

implemented within 30 days.

Amendment Nos.: Unit 1 -

Amendment No. 39; Unit 2 -

Amendment No. 25

Facility Operating License Nos. NPF87 and NPF-89. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: September 28, 1994 (59 FR

49438) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

April 17, 1995.No significant hazards

consideration comments received: No.

Local Public Document Room

location: University of Texas at

Arlington Library, Government

Publications/Maps, 702 College, P.O.

Box 19497, Arlington, TX 76019.

Union Electric Company, Docket No.

50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application for amendment:

September 9, 1994, as supplemented on

December 22, 1994

Brief description of amendment: The

amendment revises the Technical

Specification (TS) 3/4.8.2.1, 3/4.8.2.2, 3/

4.8.3.1, and 3/4.8.3.2. The changes

address the 125-volt DC buses and adds

provisions for swing battery chargers,

and removes provisions for the 4160-

volt and 480-volt AC emergency buses.

Date of issuance: April 18, 1995

Effective date: April 18, 1995

Amendment No.: 99

Facility Operating License No. NPF30. Amendment revises the Technical

Specification Bases and FSAR.

Date of initial notice in Federal

Register: January 4, 1995 (60 FR 506)

The December 22, 1994, letter provided

supplemental information that did not

change the initial proposed no

significant hazards consideration

determination. The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

April 18, 1995. No significant hazards

consideration comments received: No.

Local Public Document Room

location: Callaway County Public

Library, 710 Court Street, Fulton,

Missouri 65251.

Virginia Electric and Power Company,

Docket Nos. 50-280 and 50-281, Surry

Power Station, Unit Nos. 1 and 2, Surry

County, Virginia.

Date of application for amendments:

February 14, 1995

Brief description of amendments:

These amendments modify the

Technical Specifications (TS) to revise

Section 4.4.D of the TS to permit

approved exemptions to the

containment integrated leak rate test

frequency requirements.

Date of issuance: April 18, 1995

Effective date: April 18, 1995

Amendment Nos.: 196 and 196

Facility Operating License Nos. DPR32 and DPR-37: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: March 15, 1995 (60 FR 14029)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated April 18, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Swem Library, College of

William and Mary, Williamsburg,

Virginia 23185.

Virginia Electric and Power Company,

Docket Nos. 50-280 and 50-281, Surry

Power Station, Unit Nos. 1 and 2, Surry

County, Virginia.

Date of application for amendments:

September 6, 1994, as supplemented

March 7, 1995

Brief description of amendments:

These amendments modify the

Technical Specifications to revise the
24932 Federal Register / Vol. 60, No. 90 / Wednesday, May 10, 1995 / Notices

review responsibilities of the Station

Nuclear Safety and Operating

Committee and the Management Safety

Review Committee.

Date of issuance: April 21, 1995

Effective date: April 21, 1995

Amendment Nos.: 197 and 197

Facility Operating License Nos. DPR32 and DPR-37: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: October 12, 1994 (59 FR

51631) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

April 21, 1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: Swem Library, College of

William and Mary, Williamsburg,

Virginia 23185.

Washington Public Power Supply

System, Docket No. 50-397, Nuclear

Project No. 2, Benton County,

Washington

Date of application for amendment:

April 1, 1993

Brief description of amendment: This

amendment revises TS 3.8.1, ‘‘A.C.

Sources’’ by increasing the minimum

required level of diesel generator fuel

storage capacity. This change is based

on testing and revised calculations that

demonstrated that the existing levels of

DG fuel storage were inadequate to meet

the post-loss of coolant accident fuel

consumption requirements for seven

days of operation.

Date of issuance: April 25, 1995

Effective date: April 25, 1995, to be

implemented within 30 days of issuance

Amendment No.: 136

Facility Operating License No. NPF21: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: May 12, 1993 (58 FR 28065)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated April 25, 1995.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Richland Public Library, 955

Northgate Street, Richland, Washington

99352.

Wisconsin Public Service Corporation,

Docket No. 50-305, Kewaunee

NuclearPower Plant, Kewaunee County,

Wisconsin

Date of application for amendment:

August 24, 1994 as supplemented on

January 23, 1995.

Brief description of amendment: The

amendment revises Kewaunee Nuclear

Power Plant (KNPP) Technical

Specification (TS) 3.1.b.1 and Figure TS

3.1-4 regarding Low Temperature

Overpressure (LTOP) protection for the

reactor coolant pressure boundary. The

change extends the LTOP requirements

through the end of operating cycle 21 or

18.40 effective full power years. The

Basis Section has also been modified to

reflect these changes.

Date of issuance: April 26, 1995

Effective date: April 26, 1995

Amendment No.: 120

Facility Operating License No. DPR43. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 12, 1994 (59 FR

51632). The January 23, 1995, submittal,

provided additional reference material

which did not change the initial no

significant hazards consideration

determination. The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

April 26, 1995.No significant hazards

consideration comments received:

None.

Local Public Document Room

location: University of Wisconsin

Library Learning Center, 2420 Nicolet

Drive, Green Bay, Wisconsin 54301.

Wisconsin Public Service Corporation,

Docket No. 50-305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of application for amendment:

November 8, 1994, as supplemented on

January 9, February 14, March 8, and

April 3, 1995.

Brief description of amendment: The

amendment revises Kewaunee Nuclear

Power Plant (KNPP) Technical

Specification (TS) 3.1.d, ‘‘Leakage of

Reactor Coolant,’’ TS 4.2.b, ‘‘Steam

Generator Tubes,’’ and TS 3.4.a, ‘‘Steam

Generators,’’ to allow application of a

voltage-based repair limit for the steam

generator (SG) tube support plate (TSP)

intersections experiencing outside

diameter stress corrosion cracking

(ODSCC). The amendment also reduces

the allowed primary-to-secondary

operational leakage from any one SG

from 500 gallons per day (gpd) to 150

gpd. These changes to the tube repair

criteria are applicable for the 1995 to

1996 operating cycle (Cycle 21) only.

Date of issuance: April 17, 1995

Effective date: April 17, 1995

Amendment No.: 118

Facility Operating License No. DPR43. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: December 7, 1994 (59 FR

63127). The January 9, February 14, and

March 8, and April 3, 1995, submittals

provided clarifying information which

did not change the initial no significant

hazards consideration determination.

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated April 17, 1995. No

significant hazards consideration

comments received: None.

Local Public Document Room

location: University of Wisconsin

Library Learning Center, 2420 Nicolet

Drive, Green Bay, Wisconsin 54301.

Wisconsin Public Service Corporation,

Docket No. 50-305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of application for amendment:

September 7, 1994

Brief description of amendment: The

amendment revises Kewaunee Nuclear

Power Plant (KNPP) Technical

Specifications (TS) by adding two new

sections, TS Section 3.0 and TS Section

4.0, with associated bases. TS Section

3.0 establishes the general requirements

applicable to each of the Limiting

Conditions for Operation (LCOs) within

Section 3 of the KNPP TS. TS Section

4.0 establishes the general requirements

applicable to Surveillance

Requirements. The new requirements of

TS 4.0.b also affect TS Sections 4.5, 4.6,

4.7, and Tables TS 4.1-2 and 4.1-3.

Date of issuance: April 18, 1995

Effective date: April 18, 1995

Amendment No.: 119

Facility Operating License No. DPR43. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 12, 1994 (59 FR

51632)The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

April 18, 1995.No significant hazards

consideration comments received: No.

Local Public Document Room

location: University of Wisconsin

Library Learning Center, 2420 Nicolet

Drive, Green Bay, Wisconsin 54301.

Notice Of Issuance Of Amendments To

Facility Operating Licenses And Final

Determination Of No Significant

Hazards Consideration And

Opportunity For A Hearing (Exigent

Public Announcement Or Emergency

Circumstances)

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application for the

amendment complies with the

standards and requirements of the

Atomic Energy Act of 1954, as amended

(the Act), and the Commission’s rules

and regulations. The Commission has

made appropriate findings as required
Federal Register / Vol. 60, No. 90 / Wednesday, May 10, 1995 / Notices 24933

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 June

9, 1995, the licensee may file a request

for a hearing with respect to issuance of

the amendment to the subject facility

operating license and any person whose

interest may be affected by this

proceeding and who wishes to

participate as a party in the proceeding

must file a written request for a hearing

and a petition for leave to intervene.

Requests for a hearing and a petition for

leave to intervene shall be filed in

accordance with the Commission’s

‘‘Rules of Practice for Domestic

Licensing Proceedings’’ in 10 CFR Part

2. Interested persons should consult a

current copy of 10 CFR 2.714 which is

available at the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington, DC and

at the local public document room for

the particular facility involved. If a

request for a hearing or petition for

leave to intervene is filed by the above

date, the Commission or an Atomic

Safety and Licensing Board, designated

by the Commission or by the Chairman

of the Atomic Safety and Licensing

Board Panel, will rule on the request

and/or petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine
24934 Federal Register / Vol. 60, No. 90 / Wednesday, May 10, 1995 / Notices

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).

Carolina Power & Light Company,

Docket No. 50-261, H. B. Robinson

SteamElectric Plant, Unit No. 2,

Darlington County, South Carolina

Date of application for amendment:

April 13, 1995, as supplemented April

18, 1995.

Brief description of amendment:

Amendment revises TS Section 4.4.3.f,

g, and h to allow the post accident heat

removal system surveillance test

interval to be changed from a 12-month

interval to a refueling outage interval.

Date of issuance: April 19, 1995

Effective date: April 19, 1995

Amendment No.: 163

Facility Operating License No. DPR23. Amendment revises the Technical

Specifications.The Commission’s final

determination of significant hazards

consideration and related evaluation of

the amendment is contained in a Safety

Evaluation dated April 19, 1995.

Local Public Document Room

location: Hartsville Memorial Library,

147 West College Avenue, Hartsville,

South Carolina 29550.

Dated at Rockville, Maryland, this 3rd day

of May, 1995.

For The Nuclear Regulatory Commission

Elinor G. Adensam,

Acting Director, Division of Reactor Projects

- III/IV, Office of Nuclear Reactor Regulation

[Doc. 95–11367 Filed 5–9–95; 8:45 am]

BILLING CODE 7590–01–F

[Docket No. 50–160–Ren; ASLBP No. 95–

704–01–Ren]

Georgia Institute of Technology,

Atlanta, GA, Georgia Tech Research

Reactor, (Renewal of Facility License

No. R–97); Notice of Hearing

May 4, 1995.

On September 26, 1994, the Nuclear

Regulatory Commission published in

the Federal Register a notice of

opportunity for hearing with respect to

the proposed renewal of the facility

operating license for the Georgia Tech

Research Reactor, located on the

campus of the Georgia Institute of

Technology in Atlanta, Georgia (59 FR

49088. One request for a hearing and

petition for leave to intervene, filed by

Georgians Against Nuclear Energy

(GANE), was received. On November 18,

1994, an Atomic Safety and Licensing

Board was established to rule upon this

request and to preside over the

proceeding in the event that a hearing

were ordered.

After holding a prehearing conference

in Atlanta, Georgia, the Atomic Safety

and Licensing Board issued a Prehearing

Conference Order (LBP–95–6) on April

26, 1995, granting GANE’s request for a

hearing and petition for leave to

intervene.

Please take notice that a hearing will

be conducted it this proceeding. The

Atomic Safety and Licensing Board

designated to preside over the

proceeding consists of Dr. Jerry R. Kline,

Dr. Peter S. Lam, and Charles

Bechhoefer, who will serve as Chairman

of the Board.

During the course of the proceeding,

the Board may hold one or more

prehearing conferences pursuant to 10

CFR 2.752 and, if necessary, an

evidentiary hearing. The public is

invited to attend all these sessions,

except to the extent that information

protected by 10 CFR 2.790 (relevant to

one of the contentions accepted by the

Board) may be discussed.

Supplementing the opportunity

afforded at the first prehearing

conference, during some or all of these

sessions, and in accordance with 10

CFR 2.715(a), any person not a party to

the proceeding will be permitted to

make a limited appearance statement,

either in writing or (depending on time

availability) orally, setting forth his or

her position on the issues. These

statements do not constitute testimony

or evidence in these proceedings but

may assist the Board and/or parties in

the definition of issues being

considered. To the extent that oral

statements are permitted, the number of

persons making such statements and the

time allotted for each may be limited

depending upon the time available at

various sessions. Written statements

may be submitted at any time. Written

statements, and requests to make oral

limited appearance statements, should

be submitted to the Secretary, U.S.

Nuclear Regulatory Commission,

Washington, D.C. 20555, Attn:

Docketing and Service Branch. A copy

of such statement or request should be

served on the Chairman of this Atomic

Safety and Licensing Board, T3 F23,

U.S. Nuclear Regulatory Commission,

Washington, D.C. 20555.

Documents relating to this proceeding

are available for public inspection at the

Commission’s Public Document Room,

2120 L St. N.W., Washington, D.C.

20555.

Rockville, MD, May 4, 1995.

For the Atomic Safety and Licensing

Board.

Charles Bechhoefer,

Chairman, Administrative Judge.

[FR Doc. 95–11532 Filed 5–9–95; 8:45 am]

BILLING CODE 7590–01–M

[Docket Nos. 70–7001; 70–7002]

United States Enrichment Corporation:

Paducah Gaseous Diffusion Plant;

Portsmouth Gaseous Diffusion Plant;

Notice of Cancellation of Comment

Period and Cancellation of Public

Meetings Due to Inadequate

Application for Certification

The U.S. Nuclear Regulatory

Commission (NRC) received by letter

dated April 18, 1995, an application

from the United States Enrichment

Corporation (USEC) for the initial

certification of the gaseous diffusion

plants (GDPs) located near Paducah,

Kentucky and Piketon, Ohio. Notice of

receipt of this application along with

notice of comment period and public

meetings was published in The Federal

Register on April 28, 1995 (60 FR

21011). However, NRC’s preliminary