27792 Federal Register / Vol. 62, No. 98 / Wednesday, May 21, 1997 / Notices

operation with higher enriched fuel

have been evaluated by the NRC staff.

Based on its review, the NRC staff has

concluded that such changes would not

adversely affect plant safety. The

proposed changes have no adverse affect

on the probability of any accident. The

higher enrichment, with increased fuel

burnup, may slightly change the mix of

fission products that might be released

in the event of a serious accident, but

such small changes would not

significantly affect the consequences of

serious accidents. No changes are being

made in the types or amounts of any

radiological effluents that may be

released offsite. There is no significant

increase in the allowable individual or

cumulative occupational radiation

exposure. Accordingly, the Commission

concludes that the proposed action

would result in no significant

radiological environmental impact.

The environmental impacts of

transportation resulting from the use of

higher enrichment fuel and extended

irradiation were published and

discussed in the staff assessment

entitled ‘‘NRC Assessment of the

Environmental Effects of Transportation

Resulting from Extended Fuel

Enrichment and Irradiation,’’ dated July

7, 1988. This assessment was published

in connection with an Environmental

Assessment related to the Shearon

Harris Nuclear Plant, Unit 1, which was

published in the Federal Register (53

FR 30355) on August 11, 1988, as

corrected on August 24, 1988 (53 FR

32322). As indicated therein, the

environmental cost contribution of an

increase in the fuel enrichment of up to

5.0 weight percent Uranium-235 and

irradiation limits of up to 60,000

gigawatt-days-per-metric-ton (GWD/MT)

are either unchanged or may, in fact, be

reduced from those summarized in

Table S–4 as set forth in 10 CFR

51.52(c). These findings are applicable

to the proposed increase at BVPS–1

given that the proposal involves 5% and

burnup of less than 60,000 GWD/MT.

Accordingly, the Commission concludes

that there are no significant radiological

environmental impacts associated with

the proposed amendment.

With regard to potential

nonradiological impacts of reactor

operation with higher enrichment and

extended irradiation, the proposed

action involves features located entirely

within the restricted area as defined in

10 CFR part 20. It does not affect

nonradiological plant effluents and has

no other environmental impact.

Accordingly, the Commission concludes

that there are no significant

nonradiological environmental impacts

associated with the proposed action.

Alternative to the Proposed Action

Since the Commission has concluded

that there are no significant

environmental effects that would result

from the proposed action, any other

alternative would have equal or greater

environmental impacts and need not be

evaluated.

The principal alternative would be to

deny the requested amendment. This

would not reduce environmental impact

of plant operations and would result in

reduced operational flexibility.

Alternative Use of Resources

The action does not involve the use of

any resources not previously considered

in the Final Environmental Statement

for the Beaver Valley Power Station,

Unit No. 1 dated July 1973.

Agencies and Persons Consulted

In accordance with its stated policy,

on April 14, 1997, the staff consulted

with the Pennsylvania State official, Mr.

Michael P. Murphy of the Bureau of

Radiation Protection, Department of

Environmental Protection, regarding the

environmental impact of the proposed

action. The State official had no

comments.

Finding of No Significant Impact

The Commission has determined not

to prepare an environmental impact

statement for the proposed license

amendment.

Based upon the foregoing

environmental assessment, we conclude

that the proposed action will not have

a significant effect on the quality of the

human environment.

For further details with respect to this

proposed action, see the application for

amendment dated February 27, 1997,

that 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—5- room located

at the B. F. Jones Memorial Library, 663

Franklin Avenue, Aliquippa,

Pennsylvania 15001.

Dated at Rockville, Maryland, this 15th day

of May 1997.

For the Nuclear Regulatory Commission.

John F. Stolz,

Director, Project Directorate I–2, Division of

Reactor Projects—I/II, Office of Nuclear

Reactor Regulation.

[FR Doc. 97–13271 Filed 5–20–97; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Biweekly Notice

Applications And Amendments To

Facility Operating Licenses Involving

No Significant Hazards Considerations

I. Background

Pursuant to Public Law 97-415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 97-415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from April 28,

1997 through May 9, 1997. The last

biweekly notice was published on May

7, 1997 (62 FR 24984).

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
Federal Register / Vol. 62, No. 98 / Wednesday, May 21, 1997 / Notices 27793

expiration of the 30-day notice period.

However, should circumstances change

during the notice period such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received before

action is taken. Should the Commission

take this action, it will publish in the

Federal Register a notice of issuance

and provide for opportunity for a

hearing after issuance. The Commission

expects that the need to take this action

will occur very infrequently.

Written comments may be submitted

by mail to the Chief, Rules Review and

Directives Branch, Division of Freedom

of Information and Publications

Services, Office of Administration, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555-0001, and

should cite the publication date and

page number of this Federal Register

notice. Written comments may also be

delivered to Room 6D22, Two White

Flint North, 11545 Rockville Pike,

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

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

written comments received may be

examined at the NRC Public Document

Room, the Gelman Building, 2120 L

Street, NW., Washington, DC. The filing

of requests for a hearing and petitions

for leave to intervene is discussed

below.

By June 20, 1997, the licensee may

file a request for a hearing with respect

to issuance of the amendment to the

subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555-0001, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington DC, by

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

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

1-(800) 342-6700). The Western Union

operator should be given Datagram

Identification Number N1023 and the

following message addressed to (Project

Director): petitioner’s name and

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555-0001, and to the attorney for

the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

factors specified in 10 CFR

2.714(a)(1)(i)-(v) and 2.714(d).
27794 Federal Register / Vol. 62, No. 98 / Wednesday, May 21, 1997 / Notices

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.

Arizona Public Service Company, et al.,

Docket Nos. STN 50-528, STN 50-529,

and STN 50-530, Palo Verde Nuclear

Generating Station, Units Nos. 1, 2, and

3, Maricopa County, Arizona

Date of amendments request:

December 27, 1996

Description of amendments request:

The proposed amendments would

revise Technical Specification (TS)

3.6.1.3.b (peak containment internal

pressure for the design basis loss of

coolant accident (LOCA)) from 49.5 psig

to 52 psig and the associated Bases

Sections. The proposed amendments

reflect values based on a revised LOCA

analysis. The LOCA analysis was

revised to reflect the maximum primary

containment internal pressure specified

in other TS. This maximum primary

containment internal pressure was not

used in the original LOCA 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

consideration. The NRC staff’s analysis

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 amendment

increases the peak calculated containment

internal pressure for the design basis LOCA

from 49.5 psig to 52 psig. The maximum

pressure occurs following an accident. Since

the pressure is a consequence of an accident,

this change has no effect on the probability

of accident initiation, and therefore, the

probability of an accident previously

evaluated has not been significantly

increased.

The consequences of an accident

previously evaluated in the Updated Final

Safety Analysis Report (UFSAR) will not be

significantly increased. UFSAR Section

15.6.5.6, ‘‘Analyses of Effects and

Consequences - Large Break LOCA,’’ states

that ‘‘It is assumed that the containment

leaks at the maximum rates allowed by the

Technical Specifications, i.e., 0.1 vol. %/d

for the first 24 hours and half of that rate

thereafter.’’ The dose calculation assumes

that under accident conditions, the release of

radionuclides to the containment is

instantaneously homogenized within the

containment free air volume. This results in

a constant radioactivity per volume (curies/

cc) regardless of containment internal

pressure. Since radioactivity is assumed to be

homogenized in the containment free air

volume, the volume percent leaked per day

is equivalent to the fraction of radioactivity

which leaks from the containment per day.

Therefore, the increase in the peak calculated

containment internal pressure for the design

basis LOCA from 49.5 psig to 52 psig does

not effect dose consequences associated with

the design basis LOCA. The proposed change

to the peak calculated containment internal

pressure for the design basis LOCA does not

impact the radiological consequences of a

LOCA as analyzed in Chapters 6 and 15 of

the UFSAR.

The proposed amendments do not,

therefore 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 possibility of a new or different

kind of accident has not been created.

The increase in the peak calculated

containment internal pressure for the

design basis LOCA does not affect the

design or operation of existing plant

equipment, nor involve new plant

equipment. 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 does not involve

a significant reduction in a margin of safety.

The containment design pressure is 60

psig. The acceptance criteria in NRC

Standard Review Plan, Section 6.2.1.1.A,

‘‘PWR Dry Containments, including

Subatmospheric Containments,’’ requires in

Item 11.1 that ‘‘the containment design

pressure should provide at least a 10%

margin above the accepted peak calculated

containment pressure following a loss of

coolant accident.’’ For PVNGS to maintain

the required margin, this requires that the

peak calculated containment internal

pressure for the design basis LOCA would be

no higher than 54 psig. Since the revised

peak calculated containment internal

pressure for the design basis LOCA remains

below the 54 psig limit, the proposed change

does not involve a significant reduction in

the margin of safety.

Based on this review, it appears that

the three standards of 10 CFR50.92(c)

are satisfied. Therefore, the NRC staff

proposes to determine that the

amendments request involve no

significant hazards consideration. Local

Public Document Room location:

Phoenix Public Library, 1221 N. Central

Avenue, Phoenix, Arizona 85004

Attorney for licensee: Nancy C. Loftin,

Esq., Corporate Secretary and Counsel,

Arizona Public Service Company, P.O.

Box 53999, Mail Station 9068, Phoenix,

Arizona 85072-3999

NRC Project Director: William H.

Bateman

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 9,

1997

Description of amendment request:

The proposed change will extend the

existing Technical Specifications

surveillance intervals from 7 days to 14

days for the Channel Functional Tests

for the refueling equipment interlocks

and for the one-rod-out interlock. The

change will permit, under most normal

circumstances, a complete offloading,

shuffling, or onloading of fuel, without

the need to halt refueling activities

solely for the performance of these

surveillance tests.

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 extends the

Technical Specification Surveillance

Requirement (SR) Frequency for the Channel

Functional Tests (CFTs) for the refueling

equipment interlocks and the one-rod-out

interlock. The refueling equipment interlocks

and the one-rod-out interlock are explicitly

assumed in the analysis of the control rod

removal error during refueling. Criticality,

and therefore, subsequent prompt reactivity

excursions are prevented during the insertion

of fuel, provided all control rods are fully

inserted during the fuel insertion. The

refueling equipment interlocks accomplish

this by preventing loading fuel into the core

with any control rod withdrawn, or by

preventing withdrawal of a control rod from

the core during fuel loading. The one-rod-out

interlock and adequate shutdown margin

prevent criticality by preventing withdrawal

of more than one control rod. With one

control rod withdrawn, the core will remain

subcritical, thereby preventing any prompt

critical excursion. The proposed change does

not change the function of any of these

interlocks, only the frequency at which the

interlocks undergo channel functional

testing. A review of past test performances

has demonstrated that extending the

Frequency from 7 days to 14 days will not

result in any increase in test failures.

Therefore, the proposed change will not

change the ability of these interlocks to

perform when required. Based on this, there

can be no significant increase in the

radiological consequences of any previously

evaluated accident since all interlocks will

continue to perform as presently analyzed.

Therefore, the proposed change does not

involve a significant increase in the
Federal Register / Vol. 62, No. 98 / Wednesday, May 21, 1997 / Notices 27795

probability or consequences of an accident

previously evaluated.

2. The proposed change would not create

the possibility of a new or different kind of

accident from any previously evaluated.

The proposed change extends the SR

Frequency for performing CFTs for refueling

equipment and one-rod-out interlocks. This

change does not result in a modification to

the plant or to the manner in which the plant

is operated. The testing will still

demonstrated the operability of the

interlocks. Thus, the interlocks will still

function in the same manner. 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 will not involve a

significant reduction in the margin of safety.

The proposed change extends the SR

Frequency for performing CFTs on the

refueling equipment and one-rod-out

interlocks from 7 days to 14 days. Reviews

of past test results indicate that extending the

test interval to 14 days will not result in an

increase in the number of CFT failures for

these interlocks. This implies that extending

the SR Frequency to 14 days will not result

in an increase in the amount of time the

instrument channels will be inoperable when

required to be operable. Since the proposed

change does not result in any reduction in

the amount of time the instrument channels

will be operable, the proposed change does

not involve a significant reduction in the

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081.

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts, and Trowbridge,

2300 N Street, NW., Washington, DC

20037.

NRC Project Director: Gail H. Marcus

GPU Nuclear Corporation, et al.,

Docket No. 50-289, Three Mile Island

Nuclear Station, Unit No. 1, Dauphin

County, Pennsylvania

Date of amendment request: April 21,

1997

Description of amendment request:

The proposed amendment would revise

the Technical Specifications that would

(1) reduce the volume of borated water

in the core flood tank (CFT) from 1040

cubic feet to 940 cubic feet, (2) reduce

the surveillance acceptance criteria for

the emergency core cooling system

(ECCS) high pressure injection (HPI)

flowrate from 500 gallons per minute

(GPM) to 431 GPM, and (3) revise a

limiting condition for operation (LCO)

which currently allows either local or

remote manual operability of decay heat

valves to delete the local manual valve

operability option.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration (SHC), which is presented

below:

1. State the basis for the determination that

the proposed activity will not represent a

significant increase in the probability of

occurrence or consequences of an accident.

This TSCR [Technical Specification change

request] revises the LCO for RB [reactor

building] sump isolation valves, the LCO for

the core flood tank level, and the surveillance

requirement for HPI injection flow rate. The

Core Flood and HPI systems are not actuated

until an event occurs. The CFT level used in

the new accident analysis is that level

required to be maintained in the CFT

throughout operation (i.e., pre-accident). The

new CFT level does not prevent safe accident

mitigation.

Likewise, the reduced HPI flow cannot

cause an event to occur, and while such flow

results in less injection to the RCS [reactor

coolant system] when actuated, this is

acceptable as demonstrated in the LOCA

[loss-of-coolant accident] analyses. Changes

to the LCO for the RB sump isolation valves

support the safety analysis assumptions. The

action statements related to both the level

requirement and flow rates remain

unchanged by this request. The function,

operation and surveillance intervals for the

isolation valves (DH-V-6A/B), the CFT level

and HPI injection system are not changed by

this request. Therefore, this activity does not

increase the probability of occurrence of an

accident, previously evaluated in the SAR

[safety analysis report].

Reducing the CFT nominal volume and

reducing the HPI flow acceptance criteria in

the Technical Specifications will not increase

the radiological consequences of any LOCA

evaluated in the SAR. The results of analyses

using the reduced CFT inventory and

reduced HPI flow demonstrate that the

consequences are within the limits of 10 CFR

50.46. No fuel failure in addition to that

assumed in the evaluation of the dose

consequences would occur. Therefore, the

radiological consequences would not

increase.

The editorial changes described above have

no impact upon the probability of occurrence

or consequences of an accident.

2. State the basis for the determination that

the activity does not create the possibility of

an accident of a new or different type than

any previously analyzed in the SAR.

This TSCR revises the LCO for RB sump

isolation valves, the LCO for the core flood

tank level, and the surveillance requirement

for HPI injection flow rate. This change will

not adversely affect the capability of the

emergency core cooling systems in the event

of a LOCA. The function, operation and

surveillance intervals for both the borated

water level in the core flood tank, and ECCS

systems are not changed by this request and

no physical changes or modifications are

being made to Core Flood and HPI system

boundaries. Therefore, because there are no

configuration changes this activity does not

create the possibility of an accident or

malfunction of a different type than

previously analyzed in the SAR.

In addition, the editorial changes described

above do not create the possibility of an

accident of a new or different type than any

previously analyzed in the SAR.

3. State the basis for the determination that

the margin of safety is not significantly

reduced.

This TSCR revises the LCO for RB sump

isolation valves, the LCO for the core flood

tank level, and the surveillance requirement

for HPI injection flow rate. No system

configuration changes (hardware

modifications) will be made to implement

the change request, upon approval of the

license amendment. The action requirements

for these technical specifications have not

changed. Actions to be taken if operability

requirements are not met include plant

shutdown under certain conditions.

Furthermore, impact upon the margin to

safety is limited because the results of the

LOCA analyses demonstrate that the 10 CFR

50.46 acceptance criteria are met,

specifically: the PCT [peak clad temperature]

limit and the core-wide oxidation limit of 1

percent of the fuel cladding, as identified in

the Technical Specification bases. Hence the

margin of safety as defined in the bases of

any technical specification is not

significantly reduced or impacted by the

implementation of this change request, or the

editorial changes described above.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Law/Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Walnut

Street and Commonwealth Avenue, Box

1601, Harrisburg, PA 17105.

Attorney for licensee: Ernest L. Blake,

Jr., Esquire, Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Patrick D.

Milano, Acting

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: April 22,

1997

Description of amendment request:

The proposed amendments would

revise Technical Specifications 5.3.1,

Fuel Assemblies, and 6.9.1.6, Core

Operating Limits Report, to allow use of
27796 Federal Register / Vol. 62, No. 98 / Wednesday, May 21, 1997 / Notices

an alternate zirconium-based fuel

cladding, ZIRLO, and limited

substitution of fuel rods by ZIRLO filler

rods.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

A. The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The methodologies used in the accident

analyses remain unchanged. With the

exception of a reduction in the heat flux hot

channel factor (FQ), the operating limits will

not be changed. The proposed changes will

not result in any equipment exceeding its

design limits under normal or accident

conditions. The calculated doses presented

in the UFSAR will remain bounding. Other

than the changes to the fuel assemblies, there

are no physical changes to the plant

associated with this Technical Specification

change. A reload safety analysis will

continue to be performed for each cycle to

demonstrate compliance with fuel safety

design bases.

VANTAGE+ fuel assemblies with ZIRLO

clad fuel rods meet the same fuel assembly

and fuel rod design bases as VANTAGE 5H

fuel assemblies. Since the original design

criteria are met, the ZIRLO clad fuel rods will

not be an initiator for any new accident. The

clad material is similar in chemical

composition and has similar physical and

mechanical properties to Zircaloy. Thus,

cladding integrity is maintained and the

structural integrity of the fuel assembly is not

affected. ZIRLO cladding improves corrosion

performance and dimensional stability. No

concerns have been identified with respect to

the mixed core of Zircaloy and ZIRLO clad

assemblies. Also, no concerns have been

identified with respect to the use of an

individual assembly containing a

combination of Zircaloy and ZIRLO clad fuel

rods.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

B. The proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes do not result in any

equipment exceeding its design limits under

normal or accident conditions. All design

and performance criteria continue to be met

and no new failure mechanisms have been

identified. The ZIRLO cladding material

offers improved corrosion resistance and

structural integrity.

The proposed changes do not affect the

operation of any system or component in the

plant. The safety functions of the related

structures, systems, or components are not

changed, nor is the reliability of any

structure, system, or component reduced.

The changes do not affect the manner by

which the facility is operated and do not

change any facility design feature, structure,

or system. No new or different type of

equipment will be installed. Since there is no

other change to the facility or operating

procedures, and the safety functions and

reliability of structures, systems, or

components are not affected, the proposed

changes do not create the possibility of a new

accident or an accident different from those

previously evaluated.

C. The proposed changes do not involve a

significant reduction in a margin of safety.

Use of ZIRLO fuel cladding material will

not result in any equipment exceeding its

design or licensing bases limits under normal

or accident conditions. VANTAGE 5H reload

design and safety analysis limits are

unchanged. For each cycle reload core, the

fuel assemblies will be evaluated using NRCapproved reload design methods, including

consideration of the core physics analysis

peaking factors and core average linear heat

rate effects. ZIRLO fuel assemblies will be

assessed for use under conditions consistent

with normal core operating conditions

allowed in the Technical Specifications.

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 standards of

10 CFR 50.92(c) are satisfied. Therefore,

the NRC staff proposes to determine that

the request for amendments involves no

significant hazards consideration.

Local Public Document Room

location: Wharton County Junior

College, J. M. Hodges Learning Center,

911 Boling Highway, Wharton, TX

77488

Attorney for licensee: Jack R.

Newman, Esq., Morgan, Lewis &

Bockius, 1800 M Street, N.W.,

Washington, DC 20036-5869

NRC Project Director: William D.

Beckner

Indiana Michigan Power Company,

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

C. Cook Nuclear Plant, Unit Nos. 1 and

2, Berrien County, Michigan

Date of amendment request: March

26, 1997

Description of amendment request:

The proposed amendment would

modify the technical specifications

(TSs) which describe the control room

ventilation system autostart functions.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Per 10 CFR 50.92, the proposed changes do

not involve a significant hazards

consideration if the proposed changes do not:

1. involve a significant increase in the

probability or consequences of an accident

previously evaluated;

2. create the possibility of a new or

different kind of accident from any accident

previously evaluated; or

3. involve a significant reduction in a

margin of safety.

Criterion 1

These changes are administrative in nature,

intended to correct and clarify the TS

description of control room ventilation

system operation. Because no changes to

plant operations or physical changes to the

plant will occur due to these changes, they

do not involve a significant increase in the

probability or consequences of a previously

evaluated accident.

Criterion 2

Because no changes to plant operations or

the physical plant will occur due to these

changes, the changes will not create the

possibility of a new or different kind of

accident from any previously evaluated.

Criterion 3

These changes are administrative in nature,

intended to correct and clarify the present

TSs with regard to system operation

descriptions. Thus, the changes involve no

reduction in margins of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Maud Preston Palenske

Memorial Library, 500 Market Street, St.

Joseph, Michigan 49085.

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts, and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Gail H. Marcus

Indiana Michigan Power Company,

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

C. Cook Nuclear Plant, Unit Nos. 1 and

2, Berrien County, Michigan

Date of amendment request: March

26, 1997

Description of amendment request:

The proposed amendment would make

three administrative changes to the

technical specifications (TSs) dealing

with a grammatical error, an

inadvertently deleted frequency

requirement, and a footnote which is no

longer 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:

Per 10 CFR 50.92, the proposed changes do

not involve a significant hazards

consideration if the proposed changes do not:

1. involve a significant increase in the

probability or consequences of an accident

previously evaluated;
Federal Register / Vol. 62, No. 98 / Wednesday, May 21, 1997 / Notices 27797

2. create the possibility of a new or

different kind of accident from any accident

previously evaluated; or

3. involve a significant reduction in a

margin of safety.

Criterion 1

This amendment request does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated because the proposed changes to

the TSs do not affect the assumptions,

parameters, or results of any UFSAR accident

analysis. The firstproposed change, ‘‘A’’, is a

grammatical correction; the second proposed

change, ‘‘B’’, reformats the page, and returns

a frequency requirement that, while

inadvertently deleted from the TSs, was still

met via procedure; the third proposed change

deletes a footnote which is no longer

applicable. As described in Section II.C. of

licensee’s application request dated March

26, 1997, a load drop analysis is not required

for single-failure-proof load blocks.

Criterion 2

The proposed changes do not involve

physical changes to the plant or changes in

plant operating configuration. The changes

described above are essentially

administrative in nature, and thus do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

Criterion 3

The proposed changes are essentially

administrative in nature. Per NUREG-0612,

single-failure-proof cranes are exempt from

the requirements of a load drop analysis;

therefore, there is no 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: Maud Preston Palenske

Memorial Library, 500 Market Street, St.

Joseph, Michigan 49085.

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts, and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Gail H. Marcus

North Atlantic Energy Service

Corporation, Docket No. 50-443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request: February

12, 1997

Description of amendment request:

The proposed amendment would

change position titles in certain

Seabrook Station, Unit No. 1 (Seabrook)

Appendix A Technical Specifications

(TS) to reflect the present Seabrook

organization, would clarify the approval

authority for the Station Qualified

Reviewer Program, and would correct a

reference. Specifically, the proposed

amendment would:

1. Change TS 6.0, ‘‘Administrative

Controls’’ to reflect accurately the

current North Atlantic Management

organization, their assigned duties as

previously reported to the NRC, and

their proper titles,

2. Corrects an incorrect reference in

TS 6.4.3.9.b., and

3. Clarifies the term ‘‘Manager’’ in TS

6.4.2, ‘‘Station Qualified Reviewer

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. The NRC staff has

reviewed the licensee’s analysis against

the standards of 10 CFR 50.92(c). The

NRC staff’s review is presented below.

A. The changes do not involve a significant

increase in the probabilityor consequences of

an accident previously evaluated (10 CFR

50.92(c)(1)) because the proposed changes are

merely administrative or editorial in nature.

The proposed changes involve position title

changes to reflect current organization,

correct an incorrect reference, and provide

clarification with regard to the organizational

level for certain approvals. The changes do

not affect the manner by which the facility

is operated and do not change any facility

design feature or equipment. Since there is

no change to the facility or operating

procedures, there is no effect upon the

probability or consequences of any accident

previously analyzed.

B. The changes do not create the possibility

of a new or different kind of accident from

any accident previously evaluated (10 CFR

50.92(c)(2)) because they do not affect the

manner by which the facility is operated or

involve any changes to equipment or features

which affect the operational characteristics of

the facility. Therefore, no new accident

initiator is introduced that could cause a new

or different kind of accident from those

previously evaluated. The proposed changes

merely involve position title changes to

reflect current organization, correct an

incorrect reference, and provide clarification

with regard to the organizational level for

certain approvals.

C. The changes do not involve a significant

reduction in a margin of safety (10 CFR

50.92(c)(3)) because the proposed changes do

not affect the manner by which the facility

is operated or involve equipment or features

which affect the operational characteristics of

the facility.Based on this review, it appears

that the three standards of 10 CFR 50.92(c)

are satisfied. Therefore, the NRC staff

proposes to determine that the amendment

request involves no significant hazards

consideration.

Local Public Document Room

location: Exeter Public Library,

Founders Park, Exeter, NH 03833.

Attorney for licensee: Lillian M.

Cuoco, Esquire, Northeast Utilities

Service Company, Post Office Box 270,

Hartford CT 06141-0270.

NRC Project Director: Patrick D.

Milano

Northeast Nuclear Energy Company

(NNECO), et al., Docket No. 50-423,

Millstone Nuclear Power Station, Unit

No. 3, New London County, Connecticut

Date of amendment request: April 15,

1997

Description of amendment request:

The proposed amendment would make

changes to Technical Specification

Sections 4.3.3.6 and 4.6.4.1, which

require that the hydrogen monitors be

periodically tested. Specifically, the

changes to the surveillances would

increase the testing of the monitor’s

hydrogen sensor, correct inconsistencies

between surveillances, and make

changes to the Bases of the

surveillances.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

NNECO [Northeast Nuclear Energy

Company] has reviewed the proposed

changes in accordance with 10CFR 50.92 and

has concluded that the change does not

involve a significant hazards consideration

(SHC). The bases for this conclusion is that

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

satisfied. The proposed changes do not

involve [an] SHC because the changes would

not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change to Technical

Specification Surveillances 4.3.3.6 and

4.6.4.1 to perform a hydrogen sensor

calibration test once per 92 days on a

staggered test basis is consistent with the

design and operation of the hydrogen

monitor system. The hydrogen monitoring

system is independent of the reactor coolant

system boundary, has no effect on the

probability of occurrence of a loss of coolant

accident and performing surveillance testing

does not significantly increase the probability

of an accident previously evaluated.

The proposed change to Technical

Specification Surveillances 4.3.3.6 and

4.6.4.1 to perform a hydrogen sensor

calibration test will not require the opening

of a containment isolation valve and

conducting surveillance testing does not

significantly increase the consequence of an

accident previously evaluated.

The proposed change to Technical

Specification Surveillances 4.3.3.6 and

4.6.4.1 to change the channel check

frequency from once per 31 days to once per

12 hours on Table 4.3-7 Item 18, add an

analog channel operational test to

surveillance 4.3.3.6.2 and make editorial

changes to the surveillances and bases

sections are considered administrative

changes. Administrative changes do not

involve a significant increase in the
27798 Federal Register / Vol. 62, No. 98 / Wednesday, May 21, 1997 / Notices

probability or consequence of an accident

previously evaluated.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequence 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 to Technical

Specification Surveillances 4.3.3.6 and

4.6.4.1 to perform a hydrogen sensor

calibration test do not add any new

equipment to the plant and do not affect the

way any system important to safety is

operated either in normal or under accident

conditions.

The proposed changes to Technical

Specification Surveillances 4.3.3.6 and

4.6.4.1 to change the channel check

frequency from once per 31 days to once per

12 hours on Table 4.3-7 Item 18, add an

analog channel operational test to

surveillance 4.3.3.6.2 and make editorial

changes to the surveillances and bases

sections are considered administrative

changes.

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 to Technical

Specification Surveillances 4.3.3.6 and

4.6.4.1 to perform a hydrogen sensor

calibration test will provide assurance of

expected instrument performance under

accident conditions and performing

surveillance testing do not involve a

significant reduction in a margin of safety.

The proposed changes to Technical

Specification Surveillances 4.3.3.6 and

4.6.4.1 to change the channel check

frequency from once per 31 days to once per

12 hours on Table 4.3-7 Item 18, add an

analog channel operational test to

surveillance 4.3.3.6.2 and make editorial

changes to the surveillances and bases

sections are considered administrative

changes. Administrative changes do not

involve a significant reduction in a margin of

safety.

Therefore, the proposed changes do not

involve a significant reduction in a margin of

safety.

In conclusion, based on the information

provided, it is determined that the proposed

changes do not involve an SHC.

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 Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, Connecticut, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

P.O. Box 270, Hartford, CT 06141-

0270NRC Deputy Director: Phillip F.

McKee

Northeast Nuclear Energy Company

(NNECO), et al., Docket No. 50-423,

Millstone Nuclear Power Station, Unit

No. 3, New London County, Connecticut

Date of amendment request: April 17,

1997

Description of amendment request:

The proposed amendment would

modify Technical Specification 3.7.14

by clarifying the actions to be taken

when an area temperature exceeds its

temperature limit.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

NNECO [Northeast Nuclear Energy

Company] has reviewed the proposed change

in accordance with 10CFR 50.92 and has

concluded that the change does not involve

a significant hazards consideration (SHC).

The bases for this conclusion is that the three

criteria of 10CFR 50.92(c) are not satisfied.

The proposed change does not involve [an]

SHC because the change would not:

1. Involve a significant increase in the

probability or consequence of an accident

previously evaluated.

The proposed change to Technical

Specification 3.7.14 will establish allowable

tolerances to ensure that the applicable

systems, structures and components are

operated within their existing design bases.

Technical Specification 3.7.14 specifies the

actions to be taken when an area temperature

exceeds its temperature limit. The action

taken is dependent on the amount and

duration by which the area temperature

exceeds its limit. Actions are currently

specified for exceeding area temperature by

less than 20 °F and greater than 20°F for

periods less than 8 hours and for periods

greater than 8 hours. This change clarifies the

actions to be taken when the temperature

exceeds its limit by exactly 20 °F or exceed

its limit for exactly 8 hours. It is concluded

that this change is a clarification only in that

it causes the more conservative actions to be

taken at greater than or equal to 20 °F, or at

greater than or equal to 8 hours.

The proposed change, therefore, does not

involve a significant increase in the

probability or consequence of an accident

previously evaluated.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

Establishment of tolerances and

clarification of actions at a specific value

does not [ ] change the operation of any

system, structure or component during

normal or accident conditions.

Therefore, the proposed change 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 change is administrative in nature in

that it resolves a discontinuity in the range

of temperatures and in the duration period

above the applicable limit for which action

is required. Establishment of tolerances

ensures parameters are set and maintained

within allowable design constraints.

Clarification of applicability for the required

actions ensures that action is proscribed for

all possible conditions thereby not permitting

operation outside of allowable design.

Therefore, the proposed change does not

involve a significant reduction in a margin of

safety.

In conclusion, based on the information

provided, it is determined that the proposed

change does not involve an SHC.

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 Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, Connecticut, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

P.O. Box 270, Hartford, CT 06141-0270

NRC Deputy Director: Phillip F.

McKee

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of amendment request: March

31, 1997

Description of amendment request:

The proposed amendment would

change Technical Specification (TS)

Sections 3/4.6.5.3.2, ‘‘Filtration,

Recirculation, and Ventilation System

(FRVS),’’ to (1) provide an appropriate

Limiting Condition for Operation and

ACTION Statement that reflects the

design basis for the FRVS, and (2)

clarify the manner in which FRVS

testing is performed.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed changes do not involve a

significant increase in the probability or
Federal Register / Vol. 62, No. 98 / Wednesday, May 21, 1997 / Notices 27799

consequences of an accident previously

evaluated.

The proposed TS revisions involve: 1) no

hardware changes; 2) no significant changes

to the operation of any systems or

components in normal or accident operating

conditions; and 3) no changes to existing

structures, systems or components. Therefore

these changes will not increase the

probability of an accident previously

evaluated. Since the plant systems associated

with these proposed changes will still be

capable of: 1) meeting all applicable design

basis requirements; and 2) retaining the

capability to mitigate the consequences of

accidents described in the HC [Hope Creek]

UFSAR [Updated Final Safety Analysis

Report], the proposed changes were

determined to be justified. As a result, these

changes will not involve a significant

increase in the 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 changes contained in this

submittal will not adversely impact the

operation of any safety related component or

equipment. Since the proposed changes

involve: 1) no hardware changes; 2) no

significant changes to the operation of any

systems or components; and 3) no changes to

existing structures, systems or components,

there can be no impact on the potential

occurrence of any accident. Furthermore,

there is no change in plant testing proposed

in this change request which could initiate

an event. Therefore, these changes will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed changes for the TS related to

the Filtration Recirculation and Ventilation

System (FRVS) Recirculation Subsystem

provide consistency between the Hope Creek

TS and post-accident descriptions of the

FRVS Recirculation Subsystem operation

already contained in the UFSAR and

reflected in the Hope Creek SER [Safety

Evaluation Report] (NUREG-1048). PSE&G

[Public Service Electric & Gas] believes that

the proposed allowed outage times and

ACTION Statements for the FRVS

Recirculation Subsystem: 1) will ensure that

the required minimum number of FRVS

recirculation units will be available to

mitigate the consequences of accidents

described in the UFSAR; and 2) provide

appropriate direction and time requirements

for placing the unit in a safe shutdown

condition when the system is degraded.

Therefore, the changes contained in this

request do not result in a significant

reduction in a margin of safety.

The revisions to Surveillance Requirement

4.6.5.3.2.b provide an accurate and clearly

defined basis for performing this surveillance

test. The proposed changes implement

PSE&G—s existing interpretation of the TS

requirements and therefore do not alter the

manner in which this surveillance test is

currently being performed. PSE&G has

concluded that this surveillance test method

appropriately tests the FRVS Recirculation

Subsystem. Since the FRVS recirculation

units will continue to be tested with the

heaters: 1) operable; and 2) set at the demand

necessary to ‘‘reduce the buildup of

moisture,’’ PSE&G believes that the proposed

changes to clarify the TS are justified.

Therefore, the changes contained in this

request do not result in a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey

08070

Attorney for licensee: M. J.

Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW.,

Washington, DC 20005-3502

NRC Project Director: John F. Stolz

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of amendment request: March

31, 1997

Description of amendment request:

The proposed amendment would

provide changes to Technical

Specification (TS) 2.1.2, ‘‘THERMAL

POWER, High Pressure and High Flow,’’

ACTION a.1.c for TS 3.4.1.1,

‘‘Recirculation Loops,’’ and the Bases

for TS 2.1, ‘‘Safety Limits.’’ These

changes are being made to implement

an appropriately conservative Safety

Limit Minimum Critical Power Ratio

(SLMCPR) for all Hope Creek core and

fuel designs.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The derivation of the revised SLMCPRs for

Hope Creek for incorporation into the

Technical Specifications, and its use to

determine cycle—specific thermal limits,

have been performed using NRC approved

methods. Additionally, interim

implementing procedures which incorporate

cycle—specific parameters have been used

which result in a more restrictive value for

SLMCPR. These calculations do not change

the method of operating the plant and have

no effect on the probability of an accident

initiating event or transient.

There are no significant increases in the

consequences of an accident previously

evaluated. The basis of the MCPR [Minimum

Critical Power Ratio] Safety Limit is to ensure

that no mechanistic fuel damage is calculated

to occur if the limit is not violated. The new

SLMCPRs preserve the existing margin to

transition boiling and the probability of fuel

damage is not increased. Therefore, the

proposed change does not involve an

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 changes contained in this

submittal result from an analysis of the Cycle

7 core reload using the same fuel types as

previous cycles. These changes do not

involve any new method for operating the

facility and do not involve any facility

modifications. No new initiating events or

transients result from these changes.

Therefore, the proposed Technical

Specification 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 a margin of safety.

The margin of safety as defined in the

Technical Specification bases will remain the

same. The new SLMCPRs are calculated

using NRC approved methods which are in

accordance with the current fuel design and

licensing criteria. Additionally, interim

implementing procedures, which incorporate

cycle—specific parameters, have been used.

The MCPR Safety Limit remains high enough

to ensure that greater than 99.9% of all fuel

rods in the core will avoid transition boiling

if the limit is not violated, thereby preserving

the fuel cladding integrity. Therefore, the

proposed Technical Specification changes do

not involve a reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070

Attorney for licensee: M. J.

Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW.,

Washington, DC 20005-3502

NRC Project Director: John F. Stolz

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: April 11,

1997

Description of amendment request:

The proposed amendments would

change Technical Specification 3.6.2.3,

‘‘Containment Cooling System’’ and the
27800 Federal Register / Vol. 62, No. 98 / Wednesday, May 21, 1997 / Notices

associated bases. The changes would

increase the cooling water flow rate for

the 31-day and 18-month surveillances

and specify that during the 31-day

surveillance the fans are started and

operated in low speed. The changes are

being proposed to ensure that the

cooling water flow rate and the fan

speed being verified are representative

of the Containment Fan Cooling Unit

post-accident mode of operation.

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 ensure that the fan

speed and cooling water flow rate being

verified is representative of the fan speed and

cooling water flow rate required for the postaccident mode of operation. The proposed

changes affect an accident mitigation system

and are being made to assure that the system

is being tested in its accident mitigation

mode. There are no new accident initiators

created by the proposed changes. Therefore,

the proposed changes do not involve a

significant increase in the probability of an

accident previously evaluated.

The proposed changes provide assurance

that the CFCUs will be capable of

maintaining peak containment pressure and

temperature within design limits by verifying

the proper post-accident cooling water flow

to the CFCUs. No physical changes to the

plant result from the proposed changes to the

surveillance requirements. Therefore, the

proposed changes do not involve a

significant increase in the 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 changes for demonstrating

operability of the CFCUs in the low speed

mode, with the required post-accident

cooling water flow rate, are consistent with

the existing safety function of the CFCUs

following a Design Basis Accident (DBA).

The proposed changes to the surveillance

requirements do not involve any physical

changes to plant components, systems or

structures, or the operation of the CFCUs in

the post-accident mode. Therefore, the

proposed changes do 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 changes to the surveillance

requirements provide assurance that the

CFCUs will perform their intended design

function of maintaining peak containment

pressure and temperature consistent with the

current design basis following a DBA by

verifying the proper post-accident cooling

water flow to the CFCUs. Since the high

speed and low speed control circuits are

independent and there are separate breakers

used to energize the CFCU motors in high

and low speed, the CFCUs would be capable

of starting in the low speed mode following

a DBA although the high speed breaker and

control circuit may not be available.

Verification of the post-accident flow rate

during the 31 day surveillance also ensures

that the required supporting system, Service

Water, is available for normal operation. To

ensure that the containment air temperature

is maintained below the initial temperature

condition assumed in the accident analysis

during normal operation, Technical

Specification 3/4.6.1.5 requires verification

of the average containment temperature once

every 24 hours in Modes 1 through 4.

The proposed changes to the CFCU

surveillance requirements do not affect the

ability of the CFCUs to perform their normal

and post-accident functions. These proposed

changes ensure the verification of the proper

post-accident service water flow rate to the

CFCUs. Therefore, the proposed changes do

not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Salem Free Public library, 112

West Broadway, Salem, NJ 08079

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW, Washington,

DC 20005-3502

NRC Project Director: John F. Stolz

South Carolina Electric & Gas Company

(SCE&G), South Carolina Public Service

Authority, Docket No. 50-395, Virgil C.

Summer Nuclear Station, Unit No. 1,

Fairfield County, South Carolina

Date of amendment request: March

26, 1997

Description of amendment request:

The proposed amendment would revise

the Virgil C. Summer Nuclear Station

Technical Specifications to change the

definition of ‘‘Core Alteration.’’ The

proposed definition will not consider

movement of components other than

fuel, sources, or reactivity control

components. These proposed changes

are technically consistent with the

requirements of NUREG-1431, Revision

1, ‘‘Westinghouse Standard Technical

Specifications,’’ issued on April 7, 1995.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

The proposed changes revise the

definition of Core Alteration to be the

movement of fuel, sources, or reactivity

control components; and to delete ‘‘or

manipulation’’ and ‘‘conservative’’ from

the text. These changes do not affect the

probability of an accident previously

evaluated. The movement of

components other than fuel, sources,

and reactivity control components,

within the reactor vessel is enveloped

by the analyzed event. Deleting the

words ‘‘or manipulation’’ and

‘‘conservative’’ from the definition of

Core Alteration are administrative

changes and also do not impact

initiators of analyzed events. The only

component assumed to be an initiator of

an analyzed event is dropping an

irradiated fuel assembly, however, fuel

is still part of the definition.

Furthermore, a fuel handling accident is

minimized by administrative controls

and physical limitations imposed on

fuel handling operations. The

movement of components other than

fuel, sources, and reactivity control

components within the reactor vessel

will be controlled under plant

administrative controls. This change has

no effect on the boron dilution event

because when boron concentration is

below limits, Core Alterations are

restricted to maintain the maximum

Shutdown Margin. Movement of other

components will have a negligible

impact on core reactivity.

The changes to the definition of Core

Alteration do not increase the consequences

of an accident previously evaluated. The

accident analysis assumes an irradiated fuel

assembly is dropped with the consequences

well within the 10 CFR 100 limits. The

dropping of other components was not

addressed in the plant safety analyses,

however, the analysis of the dropped fuel

assembly encompasses other components.

The consequences of a boron dilution event

are not addressed because Core Alterations

are not allowed when the boron

concentration is below limits. These changes

do not affect the mitigation capabilities of

any component or system nor do they affect

the assumptions relative to the mitigation of

accidents or transients. Therefore, the change

does not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

The proposed changes revise the

definition of Core Alteration to be the

movement of fuel, sources, or reactivity

control components; and to delete ‘‘or

manipulation’’ and ‘‘conservative’’ from

the text. The change does not involve a
Federal Register / Vol. 62, No. 98 / Wednesday, May 21, 1997 / Notices 27801

significant change in the design or

operation of the plant. The changes do

not involve a physical alteration of the

plant (no new or different type of

equipment will be installed), or new or

unusual operator actions. The changes

will not impose any new or different

requirements or eliminate any existing

requirements. The definition of Core

Alteration is being clarified and made

consistent with NUREG-1431, Rev. 1.

Therefore, the change does not create

the possibility of a new or different kind

of accident from any accident

previously evaluated.

3. Does this change involve a significant

reduction in margin of safety?

The proposed changes revise the definition

of Core Alteration to be the movement of

fuel, sources, or reactivity control

components; and to delete ‘‘or manipulation’’

and ‘‘conservative’’ from the text. The safety

analysis assumes an irradiated fuel assembly

is dropped. Controls for handling

components other than fuel, sources, or

reactivity control components within the

reactor vessel are in plant administrative

controls. The effect of a boron dilution event

on Shutdown Margin is limited due to the

requirement to suspend Core Alterations. The

movement of other components have a

negligible impact on core reactivity. No

change is being proposed, in the applicability

of the definition, to the movement of

components which factor in the design basis

analyses (fuel handling accident). Deleting

the terms ‘‘or manipulation’’ and

‘‘conservative’’ from the definition of Core

Alteration results in a clarification to the

definition that does not technically alter the

meaning. Therefore, the change does not

involve a significant reduction in a margin of

safety

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Fairfield County Library, 300

Washington Street, Winnsboro, SC

29180

Attorney for licensee: Randolph R.

Mahan, South Carolina Electric & Gas

Company, Post Office Box 764,

Columbia, South Carolina 29218

NRC Project Director: F. Mark

Reinhart, Acting

South Carolina Electric & Gas Company

(SCE&G), South Carolina Public Service

Authority, Docket No. 50-395, Virgil C.

Summer Nuclear Station, Unit No. 1,

Fairfield County, South Carolina

Date of amendment request: March

26, 1997

Description of amendment request:

The proposed amendment would revise

the Virgil C. Summer Nuclear Station

Technical Specifications (TS),

Surveillance Requirement (SR) 4.5.2.a,

to add (1) the charging/high head safety

injection (HHSI) pump cross connect

valves, and (2) the charging pump miniflow header isolation valve, to the SR

valve list. The proposed change is an

administrative change to meet the

recommendations of NRC Branch

Technical Position (BTP) EICSB 18,

which establishes the acceptability of

disconnecting power to electrical

components of fluid systems as one

means of designing against a single

failure that might cause an undesirable

component action. TS SR 4.5.2.a

includes a list of the required positions

of manually-controlled, electricallyoperated valves, and identify those

valves to which the requirements for

removal of electrical power is applied in

order to satisfy the single failure

criterion.

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. Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

The proposed change adds the charging/

HHSI pump cross connect valves and the

charging pump mini-flow header isolation

valve to the ECCS [Emergency Core Cooling

System] Subsystems - Tavg (greater than or

equal to) 350°F Technical Specification

Surveillance Requirement. This Surveillance

Requirement will require the valves to be

verified open with power to the valve

operators removed once per 12 hours. ... The

charging/HHSI pump cross connect valves

and the charging mini-flow header isolation

valve are not initiators of any analyzed event.

... The charging pump/HHSI pump cross

connect valves are being modified to meet

the recommendations of the BTP (including

this Technical Specification change). The

charging pump mini-flow header isolation

valve meets the requirements of the BTP

except it is not located in the Technical

Specifications. ... Requiring the valves to be

verified open with power removed from the

valve operator once per 12 hours does not

affect the assumptions relative to the

mitigation of accidents or transients. This

requirement ensures that the valves are in a

position with power removed so that a failure

will not occur that will affect the mitigation

of an accident. These valves are required to

be open during a LOCA [loss-of-coolant

accident]. This change will ensure that the

valves are open with power removed.

Therefore, the change does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. Does this change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

...This change does not involve a

significant change in the design or operation

of the plant. This change is a result of BTP

EICSB 18. The charging/HHSI pump cross

connect valves are being modified to have

power lockout capability, redundant

indication on the main control board, and be

included in the Technical Specifications.

This will ensure that a single failure (hot

short in the controls of either valve) will not

cause spurious actuation of the valves during

the injection or recirculation phase of the

ECCS. The charging pump mini-flow header

isolation valve meets the requirements of the

BTP except it is not located in the Technical

Specifications. The charging/HHSI pump

cross connect valves and charging pump

mini-flow header isolation valve are required

to remain open during a LOCA. This

modification will ensure that the valves will

remain open during an accident which

requires ECCS operation. The proposed

change will not introduce any new accident

initiators. Therefore, the change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Does the change involve a significant

reduction in margin of safety?

...The ECCS is required to operate upon

receipt of a safety injection signal. The

charging/HHSI pump cross connect valves

and the charging pump mini-flow header

isolation valve are required to remain open

during ECCS operation. However, a single

failure may cause a spurious actuation

(closure) of the valves which could hinder

HHSI flow. The modification to the charging/

HHSI cross connect valves (the addition of a

power lockout feature and redundant

position indication) and the added TS

Surveillance Requirement will eliminate this

failure scenario and ensure the valves remain

in their safety function position (open). The

charging pump mini-flow header isolation

valves already contain a power lockout

feature and redundant position indication.

These valves are being added to the

Technical Specifications to meet the

requirements of BTP EICSB 18. Therefore, the

change does not involve a significant

reduction in a margin of safety[.]

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Fairfield County Library, 300

Washington Street, Winnsboro, SC

29180

Attorney for licensee: Randolph R.

Mahan, South Carolina Electric & Gas

Company, Post Office Box 764,

Columbia, South Carolina 29218

NRC Project Director: Mark Reinhart,

Acting
27802 Federal Register / Vol. 62, No. 98 / Wednesday, May 21, 1997 / Notices

Tennessee Valley Authority, Docket

Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of application for amendments:

March 13, 1997 (TS 97-01)

Brief description of amendments: The

amendments change the Technical

Specifications by raising the allowable

U-235 enrichment, as specified in

Section 5.6.1.2, of fuel stored in the new

fuel pit storage racks from 4.5 to 5.0

weight percent.

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:

Operation of Sequoyah Nuclear Plant

(SQN) in accordance with the proposed

amendment will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change to the allowed

enrichment of new fuel stored in the new

fuel storage racks does not change the

criticality potential with the proposed fuel

arrangement requirements for the storage

racks. The potential keff values are

maintained the same as the current TS

[Technical Specification] requirements. In

addition, the storage racks are not modified,

other than the locations that cannot be filled

with fuel assemblies, and the processes for

loading and unloading fuel in these racks and

the controls for these racks remain the same.

Since the keff limits and operating processes

are unchanged by the proposed revision,

there is no increase in the probability of an

accident previously evaluated. Likewise,

there is no impact to the consequences of an

accident or increase in offsite dose limits as

a result of the proposed TS change because

the criticality requirements are unchanged

and plant equipment will be utilized and

operated without change considering the fuel

storage location limits imposed by this

request.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

As stated above, the plant equipment and

operating processes will not be altered by the

proposed TS change with the exception of

allowed fuel storage locations in the new fuel

storage racks. The limitations on acceptable

fuel storage locations in the racks ensure that

the keff limits are maintained at the same

limits as currently required. TVA has not

postulated a criticality event at SQN for the

spent or new fuel storage locations because

the design of the associated storage racks,

potential moderation, and TS allowable fuel

enrichments do not support the potential for

this condition. Considering the physical

barriers that will be installed and verified to

be in place prior to initial loading of fuel in

the new fuel storage racks, the new fuel

storage rack physical limitations will

continue to ensure that criticality events are

not credible for the proposed change.

Therefore, this change does not create the

potential for a new accident from any

previously analyzed.

3. Involve a significant reduction in a

margin of safety.

The proposed TS change maintains the

existing requirements for criticality by

utilizing limited storage locations in the new

fuel pit storage racks. There is no change to

operating practices associated with the use

and control of these racks except for the

storage limitations. For these reasons, there

will be no reduction in the margin of safety

as a result of implementing the proposed TS

change.

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

Wisconsin Electric Power Company,

Docket Nos. 50-266 and 50-301, Point

Beach Nuclear Power Plant, Unit Nos.

1 and 2, Town of Two Creeks,

Manitowoc County, Wisconsin

Date of amendment request: April 4,

1997 (TSCR 197)

Description of amendment request:

The proposed amendments revise TS

15.6, ‘‘Administrative Controls,’’ and

15.7, ‘‘Radiological Effluent Technical

Specifications,’’ to change the corporate

officer responsible for nuclear

operations from ‘‘Vice PresidentNuclear Power,’’ to ‘‘Chief Nuclear

Officer,’’ and to require that the position

be an officer of the company.

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 Point Beach Nuclear

Plant in accordance with the proposed

amendments will not result in a significant

increase in the probability or consequences

of an accident previously evaluated.

The proposed changes are administrative

only. There are no physical changes to the

facility or its operation. All Limiting

Conditions of Operation, Limiting Safety

System Settings, and Safety Limits specified

in the Technical Specification remain

unchanged. Additionally, there are no

changes in the Quality Assurance Program,

Emergency Plan, Security Plan, and Operator

Training and Requalification Program.

Therefore, an increase in the probability or

consequences of an accident previously

evaluated cannot occur.

2. Operation of the Point Beach Nuclear

Plant in accordance with the proposed

amendments will not create the possibility of

a new or different kind of accident from any

accident previously evaluated.

The proposed changes are administrative

only. No changes to the facility structures,

systems and components or their operation

will result. The design and design basis of

the facility remain unchanged. The plant

safety analyses remain current and accurate.

No new or different failure mechanisms are

introduced. Therefore, the possibility of a

new or different kind of accident from any

accident previously evaluated is not

introduced.

3. Operation of the Point Beach Nuclear

Plant in accordance with the proposed

amendments does not involve a significant

reduction in a margin of safety.

The proposed amendments are

administrative only. All safety margins

established through the design and facility

license including the Technical

Specifications remain unchanged. In

addition, the proposed amendments ensure

continued emphasis and assignment of

responsibility for overall nuclear safety.

Therefore, all margins of safety are

maintained.

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: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin

54241

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts, and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037

NRC Project Director: John N. Hannon

Wisconsin Electric Power Company,

Docket Nos. 50-266 and 50-301, Point

Beach Nuclear Power Plant, Unit Nos.

1 and 2, Town of Two Creeks,

Manitowoc County, Wisconsin

Date of amendment request: April 14,

1997 (TSCR 198)

Description of amendment request:

The proposed amendments revise TS

15.3.1, ‘‘Reactor Coolant System,’’ to

require both reactor coolant pumps to be

operable when the reactor is critical and

to require that the reactor be placed in

hot shutdown within 6 hours if one or

both reactor coolant pumps cease

operating. This revision eliminates the

current provision which allows single

pump operation up to 3.5 percent

power.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the
Federal Register / Vol. 62, No. 98 / Wednesday, May 21, 1997 / Notices 27803

licensee has provided its analysis of the

issue of no significant hazards

consideration which is presented below:

1. Operation of the Point Beach Nuclear

Plant in accordance with the proposed

amendments will not result in a significant

increase in the probability or consequences

of an accident previously evaluated.

The amendments proposed eliminate an

inconsistency in the Technical Specifications

in a conservative manner. The proposed

changes ensure that required protection

functions remain operable in all required

modes of operation. Since the protection

functions remain operable in accordance

with existing Technical Specification

requirements and serve to mitigate analyzed

events no increase in the consequences of a

previously analyzed accident results. The

protective functions are not accident

initiators and are maintained and tested in

accordance with existing Technical

Specification requirements, therefore the

probability of a previously analyzed accident

cannot increase. Therefore, operation of the

Point Beach Nuclear Plant in accordance

with the proposed changes does not result in

an increase in probability or consequences of

a previously analyzed accident.

2. Operation of the Point Beach Nuclear

Plant in accordance with the proposed

amendments will not create the possibility of

a new or different kind of accident from any

accident previously evaluated.

The proposed amendments restore

consistency within the Technical

Specifications thus ensuring the protections

functions remain operable as required and

the units are operated within the bounds of

the existing safety analyses. Therefore,

operation of the Point Beach Nuclear Plant in

accordance with the proposed amendments

does not result in a new or different kind of

accident from any accident previously

evaluated.

3. Operation of the Point Beach Nuclear

Plant in accordance with the proposed

amendments does not involve a significant

reduction in a margin of safety.

Margins of safety are defined by the

bounds of the design and in the safety

analyses performed for the Point Beach

Nuclear Plant. The proposed amendments

eliminate an inconsistency within the

Technical Specifications and ensure the

plant will respond as analyzed in the Safety

Analyses. There is no physical change in the

facility or operation. Therefore, operation of

the Point Beach Nuclear Plant in accordance

with the proposed amendments does not

involve a reduction in 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: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin

54241

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts, and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037

NRC Project Director: John N. Hannon

Wolf Creek Nuclear Operating

Corporation, Docket No. 50-482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request: March

21, 1997, as supplemented by letter

dated April 15, 1997.

Description of amendment request:

This amendment request proposes to

revise the technical specifications

associated with the inspection of the

reactor coolant flywheel to provide an

exception to the recommendations of

Regulatory Guide 1.14, Revision 1,

‘‘Reactor Coolant Pump Flywheel

Integrity.’’ The proposed exception

would allow either an ultrasonic

volumetric examination or surface

examination to be performed at

approximately 10-year intervals. In

addition, a correction of the issuance

date of a referenced regulatory guide is

included.

This amendment would also allow

delaying the complete flywheel

examination for the ‘‘D’’ reactor coolant

pump until the Fall 1997 outage.

This supersedes the staff’s proposed

no significant hazards consideration

determination evaluation for the

requested changes that was published

on January 2, 1997 (62 FR 133).

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 safety function of the RCP [reactor

coolant pump] flywheels is to provide a

coastdown period during which the RCPs

would continue to provide reactor coolant

flow to the reactor after loss of power to the

RCPs. The maximum loading on the RCP

flywheel results from overspeed following a

LOCA [loss-of-coolant accident]. The

maximum obtainable speed in the event of a

LOCA was predicted to be less than 1500

rpm. Therefore, a peak LOCA speed of 1500

rpm is used in the evaluation of RCP

flywheel integrity in WCAP-14535. This

integrity evaluation shows a very high flaw

tolerance for the flywheels. The proposed

change does not affect that evaluation.

Reduced coastdown times due to a single

failed flywheel is bounded by the locked

rotor analysis, therefore, it would not place

the plant in an unanalyzed condition.

Therefore, these 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 accident previously

evaluated.

The proposed amendment does not create

the possibility of a new or different kind of

accident from any previously evaluated since

the proposed amendments will not change

the physical plant or the modes of plant

operation defined in the facility operating

license. No new failure mode is introduced

due to the proposed change, since the

proposed change does not involve the

addition or modification of equipment, nor

do they alter the design or operation of

affected plant systems, structures, or

components.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The operating limits and functional

capabilities of the affected systems,

structures, and components are basically

unchanged by the proposed amendment. The

results of the flywheel inspections performed

have identified no indications affecting

flywheel integrity. As identified in WCAP14535, detailed stress analysis as well as risk

analysis have been completed with the

results indicating that there would be no

change in the probability of failure for RCP

flywheels if all inspections were eliminated.

Therefore these 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

locations: Emporia State University,

William Allen White Library, 1200

Commercial Street, Emporia, Kansas

66801 and Washburn University School

of Law Library, Topeka, Kansas 66621

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts and Trowbridge,

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

20037

NRC Project Director: William H.

Bateman

Previously Published Notices Of

Consideration Of Issuance Of

Amendments To Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

And Opportunity For A Hearing

The following notices were previously

published as separate individual

notices. The notice content was the

same as above. They were published as

individual notices either because time

did not allow the Commission to wait

for this biweekly notice or because the

action involved exigent circumstances.

They are repeated here because the

biweekly notice lists all amendments

issued or proposed to be issued
27804 Federal Register / Vol. 62, No. 98 / Wednesday, May 21, 1997 / Notices

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.

Commonwealth Edison Company,

Docket Nos. 50-237 and 50-249,

Dresden Nuclear Power Station, Units 2

and 3, located in Grundy County,

Illinois

Date of amendment request: January

24, 1997.

Description of amendment request:

The application proposed to change the

Technical Specifications to reflect the

installation of new reactor water level

instrumentation for the Emergency Core

Cooling System actuation.

Date of publication of individual

notice in Federal Register: April 18,

1997 (62 FR 19143).Expiration date of

individual notice: May 19, 1997

Local Public Document Room

location: The Morris Area Public Library

District, 604 Liberty Street, Morris,

Illinois 60450

Commonwealth Edison Company,

Docket Nos. 50-237 and 50-249,

Dresden Nuclear Power Station, Units 2

and 3, located in Grundy County,

Illinois

Date of amendment request: March 5,

1997.

Description of amendment request:

The application proposed to remove the

Main Steam Line Radiation Monitor

High scram and the Main Steam Line

Tunnel Radiation High input to the

Main Steam Line Isolation function

requirement from the Technical

Specifications (TS). The proposed

changes are a result of a Boiling Water

Reactor Owners Group initiative to

minimize inadvertent scrams and Main

Steam Isolation Valve closure due to

erroneous radiation monitor actuation.

Date of publication of individual

notice in Federal Register: April 18,

1997 (62 FR 19141).Expiration date of

individual notice: May 19, 1997

Local Public Document Room

location: The Morris Area Public Library

District, 604 Liberty Street, Morris,

Illinois 60450

Commonwealth Edison Company,

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

Cities Nuclear Power Station, Units 1

and 2, located in Rock Island County,

Illinois

Date of amendment request: April 21,

1997

Description of amendment request:

The amendments would reflect a change

in the Quad Cities, Unit 2, Minimum

Critical Power Ratio (MCPR) Safety

Limit and add the Siemens Power

Corporation (SPC) methodology for

application of the Advanced Nuclear

Fuel for Boiling Water Reactors (ANFB)

Critical Power Correlation to coresident

General Electric fuel for Quad Cities,

Unit 2, Cycle 15, to Technical

Specification Section 6.9.A.6.b.

Date of publication of individual

notice in Federal Register: April 30,

1997 (62 FR 23499)

Expiration date of individual notice:

May 30, 1997

Local Public Document Room

location: Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois 61021

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 31, 1997

Brief description of amendment: The

proposed amendment would revise the

Ginna Station Improved Technical

Specifications to reflect a planned

modification to the spent fuel pool

storage racks.Date of publication of

individual notice in Federal Register:

April 30, 1997 (62 FR 23502)

Expiration date of individual notice:

May 30, 1997

Local Public Document Room

location: Rochester Public Library, 115

South Avenue, Rochester, New York

14610

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:

March 5, 1997, as supplemented May 9,

1997. The May 9, 1997, letter provided

clarifying information that did not

change the initial proposed no

significant hazards consideration

determination.

Brief description of amendments: The

amendments incorporate a new

Technical Specification for

instrumentation associated with

automatic isolation of a pathway for

release of non-condensible gases from

the main condenser.

Date of issuance: May 9, 1997

Effective date: May 9, 1997

Amendment Nos.: 185 and 216

Facility Operating License Nos. DPR71 and DPR-62: Amendments change

the Technical Specifications.

Date of initial notice in Federal

Register: April 9, 1997 (62 FR 17224)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated May 9, 1997.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, et

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

Nuclear Power Plant, Unit 1, Wake and

Chatham Counties, North Carolina

Date of application for amendment:

March 14, 1997

Brief description of amendment: The

amendment extends the allowed outage

time for its refueling water storage tank
Federal Register / Vol. 62, No. 98 / Wednesday, May 21, 1997 / Notices 27805

while performing surveillance testing of

its reactor coolant system pressure

isolation valves (Surveillance 4.4.6.2.2).

Date of issuance: May 6, 1997

Effective date: May 6, 1997

Amendment No. 71

Facility Operating License No. NPF63. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: March 26, 1997 (62 FR 14459)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated May 6, 1997.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:

April 18, 1997, as supplemented April

29, 1997.

Brief description of amendment: The

amendment approves the modification

to the protection circuitry for emergency

diesel generators. The associated Safety

Evaluation delineates the staff’s review

and findings that the modification and

related Final Safety Analysis Report

(FSAR) changes are acceptable.

Date of issuance: May 8, 1997

Effective date: May 8, 1997

Amendment No. 72

Facility Operating License No. NPF63. The amendment approves

modification to the protection circuitry

for emergency diesel generators and

related FSAR changes.

Date of initial notice and proposed no

significant hazards consideration in

Federal Register: (62 FR 19818 dated

April 23, 1997). 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 23, 1997,

but indicated that if the Commission

makes a final no significant hazards

consideration determination any such

hearing would take place after issuance

of the 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

Attorney for licensee: William D.

Johnson, Vice President and Senior

Counsel, Carolina Power & Light

Company, Post Office Box 1551,

Raleigh, North Carolina 27602

Local Public Document Room

location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh,

North Carolina 27605.

Commonwealth Edison Company,

Docket Nos. STN 50-454 and STN 50-

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

Ogle County, IllinoisDocket Nos. STN

50-456 and STN 50-457, Braidwood

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

Illinois

Date of application for amendments:

November 4, 1996, as supplemented on

December 4, 1996, and March 20, 1997.

Brief description of amendments: The

amendments revise the technical

specifications (TS) to permit the

removal of containment tendon

sheathing filler grease in up to 35

tendons for Byron, Unit 1, and

Braidwood, Unit 1, in advance of the

steam generator replacement outages.

The grease will be removed

approximately 6 months prior to the

respective steam generator replacement

outages. In addition, in Amendment No.

80 issued on April 16, 1997, the title in

Braidwood’s TS 6.9.1.7 was

unintentionally left uncorrected. The

corrected page is included in this

amendment.

Date of issuance: May 6, 1997

Effective date: Immediately, to be

implemented within 30 days.

Amendment Nos.: 89, 89 and 81, 81

Facility Operating License Nos. NPF37, NPF-66, NPF-72 and NPF-77: The

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 15, 1997 (62 FR 2186).

The March 20, 1997, submittal provided

additional clarifying information that

did not change the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated May 6, 1997No

significant hazards consideration

comments received: No

Local Public Document Room

location: For Byron, the Byron Public

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

Box 434, Byron, Illinois 61010; for

Braidwood, the Wilmington Public

Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481.

Commonwealth Edison Company,

Docket Nos. 50-237 and 50-249,

Dresden Nuclear Power Station, Units 2

and 3, Grundy County, Illinois

Date of application for amendments:

February 17, 1997, as supplemented

February 27, March 12, March 26, April

2, and April 10, 1997

Brief description of amendments: The

amendments evaluate the Unreviewed

Safety Question (USQ) associated with

the use of containment pressure to

compensate for the deficiency in Net

Positive Suction Head (NPSH) for the

Emergency Core Cooling System (ECCS)

pumps following a Design Basis

Accident (DBA). In the resolution of the

USQ, the licensee changed the Updated

Final Safety Analysis Report (UFSAR)

in the following areas:

1. containment analysis,

2. decay heat model,

3. increase in the suppression pool

temperature and the effect on other

associated systems following a DBA,

and

4. ECCS heat exchanger duty and

containment cooling service water

(CCSW) system flow.In addition, the

proposed amendments would change

the Technical Specification (TS)

allowable water temperature limits for

the suppression chamber and the

ultimate heat sink from less than or

equal to 75 degrees Fahrenheit to less

than or equal to 95 degrees Fahrenheit.

The original licensing basis water

temperature for both the suppression

chamber and ultimate heat sink was 95

degrees Fahrenheit. Both values were

changed in the TS in Amendment Nos.

152 and 147 for Dresden, Units 2 and 3,

respectively, issued on January 28,

1997. The amendments to lower the

ultimate heat sink and suppression pool

temperature limits in the TS was in

response to the resolution of a USQ

associated with the operation of

Dresden, Units 2 and 3, following the

discovery of a calculational error

concerning the head loss across the

ECCS suction strainers. The proposed

amendments will return both units to

normal operating conditions allowing

for continued power operations when

the ultimate heat sink temperature goes

above 75 degrees Fahrenheit during

warm weather.

Date of issuance: April 30, 1997

Effective date: Immediately, to be

implemented within 30 days.

Amendment Nos.: 157; 152.

Facility Operating License Nos. DPR19 and DPR-25: The amendments

revised the licenses, TS and USFAR.

Date of initial notice in Federal

Register: February 27, 1997 (62 FR

8998). The February 27, March 12,

March 26, April 2 and April 10, 1997,

submittals provided additional

clarifying information that did not

change the initial proposed no

significant hazards consideration

determination.The Commission’s

related evaluation of the amendments is

contained in a Safety Evaluation dated
27806 Federal Register / Vol. 62, No. 98 / Wednesday, May 21, 1997 / Notices

April 30, 1997No significant hazards

consideration comments received: No

Local Public Document Room

location: Morris Area Public Library

District, 604 Liberty Street, Morris,

Illinois 60450.

Commonwealth Edison Company,

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

Cities Nuclear Power Station, Units 1

and 2, Rock Island County, Illinois

Date of application for amendments:

February 17, 1997

Brief description of amendments: The

amendments would change the

Technical Specifications by increasing

the load test values of the emergency

diesel generators in Surveillance

Requirement 4.9.A.8.h from between

2625 kW and 2750 kW to 2730 kW and

2860 kW.

Date of issuance: May 1, 1997

Effective date: Immediately, to be

implemented within 60 days.

Amendment Nos.: 176 and 172

Facility Operating License Nos. DPR29 and DPR-30: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: March 26, 1997 (62 FR 14460).

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated May 1, 1997.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois

61021.

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:

August 22, 1996, as supplemented

March 28, 1997.

Brief description of amendment: The

amendment revises Technical

Specification Sections 3.3 and 4.5 to

allow the deletion of the requirement to

utilize sodium hydroxide (NaOH) as an

additive in the post-accident

containment spray system.

Date of issuance: April 23, 1997

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 191

Facility Operating License No. DPR26: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 29, 1997 (62 FR 4345)

The March 28, 1997, supplemental letter

provided clarifying information that did

not change the initial proposed no

significant hazards consideration

determination or expand the scope of

the amendment request as originally

noticed. The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

April 23, 1997No significant hazards

consideration comments received: No

Local Public Document Room

location: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10610.

Duke Power Company, et al., Docket

Nos. 50-413 and 50-414, Catawba

Nuclear Station, Units 1 and 2, York

County, South Carolina

Date of application for amendments:

March 7, 1997, as supplemented by

letters dated April 2, 10, 16, 22, and 28,

1997

Brief description of amendments: The

amendment revise Section 3/4.7.1.6 of

the Technical Specifications to require

four instead of three steam generator

pressure operated relief valves operable.

Date of issuance: April 29, 1997

Effective date: As of the date of

issuance to be implemented within 30

days. Implementation of the

amendments include the incorporation

in the Updated Final Safety Analysis

Report (UFSAR) of the changes to the

description of the facility as set forth in

the licensee’s application dated March

7, 1997, as supplemented by letters

dated April 2, 10, 16, 22, and 28, 1997,

as evaluated in the staff’s Safety

Evaluation dated April 29, 1997.

Amendment Nos.: 159 and 151

Facility Operating License Nos. NPF35 and NPF-52: Amendments revised

the Technical Specifications and

License Conditions.

Date of initial notice in Federal

Register: March 13, 1997 (62 FR 11931)

The April 2, 10, 16, 22, and 28, 1997,

letters provided additional and

clarifying information that did not

change the scope of the March 7, 1997,

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 29,

1997.No significant hazards

consideration comments received: No

Local Public Document Room

location: York County Library, 138 East

Black Street, Rock Hill, South Carolina

29730

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:

September 30, 1994, as supplemented

by letters dated September 18, 1995, and

March 15, April 29, May 16, September

23, and October 28, 1996, and January

16, April 22, and May 2, 1997

Brief description of amendments: The

amendments revise the Technical

Specifications related to the

replacement of the Westinghouse Model

‘‘D’’ type preheat steam generators with

feedring steam generators designed by

Babcock and Wilcox International.

Date of issuance: May 5, 1997

Effective date: As of the date of

issuance to be implemented within 30

days for Unit 1; and effective upon

replacement of the steam generators for

Unit 2.

Amendment Nos.: 175 and 157

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

Technical Specifications.

Date of initial notice in Federal

Register: November 8, 1995 (60 FR

56366) The March 15, April 29, May 16,

September 23, and October 28, 1996,

and January 16, April 22, and May 2,

1997, letters provided clarifying

information that did not change the

scope of the September 30, 1994,

application and the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated May 5, 1997.No

significant hazards consideration

comments received: No

Local Public Document Room

location: J. Murrey Atkins Library,

University of North Carolina at

Charlotte, 9201 University City

Boulevard, North Carolina 28223-0001

Entergy Gulf States, Inc., Cajun Electric

Power Cooperative, and Entergy

Operations, Inc., Docket No. 50-458,

River Bend Station, Unit 1, West

Feliciana Parish, Louisiana

Date of amendment request:

November 15, 1996

Brief description of amendment: The

amemdment revises the technical

specifications to allow the performance

of the 24-hour emergency diesel

generator maintenance run while the

unit is in either Mode 1 or Mode 2.

Date of issuance: May 5, 1997

Effective date: May 5, 1997

Amendment No.: 94

Facility Operating License No. NPF47: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: January 2, 1997 (62 FR 127)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated May 5, 1997.No

significant hazards consideration

comments received. No.

Local Public Document Room

location: Government Documents
Federal Register / Vol. 62, No. 98 / Wednesday, May 21, 1997 / Notices 27807

Department, Louisiana State University,

Baton Rouge, LA 70803

Entergy Operations, Inc., Docket No.

50-313, Arkansas Nuclear One, Unit

No. 1, Pope County, Arkansas

Date of amendment request: April 11,

1997

Brief description of amendment: The

amendment would permit steam

generator tubes with intergranular

corrosion indications that may exceed

through-wall limits to remain in service

until the next refueling outage.

Date of issuance: May 7, 1997

Effective date: May 7, 1997

Amendment No.: 189

Facility Operating License No. DPR51: Amendment revised the Technical

Specifications.Public comments

requested as to proposed no significant

hazards consideration (NSHC): Yes (62

FR 19628 dated April 22, 1997). The

notice provided an opportunity to

submit comments on the Commission’s

proposed NSHC determination. No

comments have been received. The

notice also provided for an opportunity

to request a hearing by May 22, 1997,

but indicated that if the Commission

makes a final NSHC 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 NSHC are contained in a Safety

Evaluation dated May 7, 1997.

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

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

20005-3502

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Entergy Operations, Inc., Docket No.

50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of application for amendment:

December 19, 1996

Brief description of amendment: The

proposed changes revise Technical

Specification Table 4.3-1 to change the

power calibration requirements for the

linear power level, the Core Protection

Calculator (CPC) delta T power and the

CPC nuclear power signals between 15

and 80 percent power to allow more

conservative settings.

Date of issuance May 5, 1997

Effective date: May 5, 1997, to be

implemented within 30 days.

Amendment No.: 183

Facility Operating License No. NPF-6.

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 29, 1997 (62 FR 4348)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated May 5, 1997.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

GPU Nuclear Corporation, et al.,

Docket No. 50-219, Oyster Creek

Nuclear Generating Station, Ocean

County, New Jersey

Date of application for amendment:

November 12, 1996, as supplemented

November 27, 1996 (TSCR 224)

Brief description of amendment: The

amendment updates the technical

specifications to reflect the

implementation of the revised 10 CFR

Part 20, ‘‘Standards for Protection

Against Radiation.’’

Date of issuance : May 8, 1997

Date of issuance : May 8, 1997

Effective date: May 8, 1997, with full

implementation within 30 days.

Amendment No.: 191

Facility Operating License No. DPR16. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: December 18, 1996 (61 FR

66708). The Commission’s related

evaluation of this amendment is

contained in a Safety Evaluation dated

May 8, 1997.No significant hazards

consideration comments received: No.

Local Public Document Room

location: Ocean County Library,

Reference Department, 101 Washington

Street, Toms River, NJ 08753.

Maine Yankee Atomic Power Company,

Docket No. 50-309, Maine Yankee

Atomic Power Station, Lincoln County,

Maine

Date of application for amendment:

February 7, 1997

Brief description of amendment: The

amendment revises TS 3.12, ‘‘Station

Service Power,’’ to require both 115 kV

power circuits to be operable when the

reactor is critical and to limit or restrict

the time during which Maine Yankee

may continue to operate if one or both

of the 115 kV power circuits become

inoperable.

Date of issuance May 2, 1997

Effective date: May 2, 1997, to be

implemented within 30 days.

Amendment No.: 157

Facility Operating License No. DPR36: Amendment revised the Technical

Specifications and/or License.

Date of initial notice in Federal

Register: February 26, 1997 (FR 8799)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated May 2, 1997No

significant hazards consideration

comments received: No

Local Public Document Room

location: Wiscasset Public Library, High

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

04578.

Southern Nuclear Operating Company,

Inc., Docket Nos. 50-348 and 50-364,

Joseph M. Farley Nuclear Plant, Units

1 and 2, Houston County, Alabama

Date of amendments request February

24, 1997, as supplemented by letters

dated March 13, April 11, 23, and 29,

1997

Brief description of amendments: The

amendments change the Technical

Specification surveillance requirements

for the Control Room Emergency

Filtration System, the Penetration Room

Filtration System, and the Containment

Purge Exhaust Filter System.

Date of issuance May 1, 1997

Effective date: As of the date of

issuance to be implemented within 30

days

Amendment Nos.: 127 and 121

Facility Operating License Nos. NPF2 and NPF-8: Amendments revise the

Technical Specifications.

Date of initial notice in Federal

Register: March 6, 1997 (62 FR 10294)

The March 13, April 11, 23, and 29,

1997, letters provided clarifying

information that did not change the

scope of the February 24, 1997,

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

May 1, 1997.No significant hazards

consideration comments received: No

Local Public Document Room

location: Houston-Love Memorial

Library, 212 W. Burdeshaw Street, Post

Office Box 1369, Dothan, Alabama

36302

Tennessee Valley Authority, Docket

Nos. 50-260, and 50-296, Browns Ferry

Nuclear Plant, Units 2 and 3, Limestone

County, Alabama

Date of application for amendments:

June 21, 1996, supplemented February

7, 1997 (TS 377)

Brief description of amendments: The

amendments provide a new minimum

critical power ratio safety limit to

replace a nonconservative value.

Technical Specification Bases are also

updated to clarify usage of the residual

heat removal system supplemental

spent fuel pool cooling mode.

Date of issuance : May 7, 1997

Effective date: As of the date of

issuance to be implemented within 30

days from the date of issuance.

Amendment Nos.: 247 and 207
27808 Federal Register / Vol. 62, No. 98 / Wednesday, May 21, 1997 / Notices

Facility Operating License Nos. DPR52 and DPR-68: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register:

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 7, 1997.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Athens Public library, 405 E.

South Street, Athens, Alabama 35611

Virginia Electric and Power Company,

et al., Docket Nos. 50-338 and 50-339,

North Anna Power Station, Units No. 1

and No. 2, Louisa County, Virginia

Date of application for amendments:

September 4, 1996, as supplemented

February 3, 1997. The February 3, 1997

submittal provided clarifying

information only, and did not change

the proposed no significant hazards

consideration determination.

Brief description of amendments: The

amendments revise the license and

technical specifications (TS) to permit

the insertion of four demonstration fuel

assemblies into the reactor core of either

North Anna 1 or North Anna 2, as

described in the licensee’s submittal.

The four lead test assemblies, fabricated

by Framatome Cogema Fuels, will

incorporate several advanced design

features, including: a debris filter

bottom nozzle, mid-span mixing grids, a

floating top end grid, a quick disconnect

top nozzle, and use of advanced

zirconium alloys for fuel assembly

structural tubing and for fuel rod

cladding.

Date of issuance May 9, 1997

Effective date: May 9, 1997

Amendment Nos.: 204 and 185

Facility Operating License Nos. NPF4 and NPF-7. These amendments

revised the License and Technical

Specifications.

Date of initial notice in Federal

Register: December 4, 1996 (61 FR

64396) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

May 9, 1997.No significant hazards

consideration comments received: No.

Local Public Document Room

location: The Alderman Library, Special

Collections Department, University of

Virginia, Charlottesville, Virginia 22903-

2498.

Wisconsin Public Service Corporation,

Docket No. 50-305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of application for amendment:

July 18, 1996, as supplemented on

January 29, 1997.

Brief description of amendment: The

amendment revises Kewaunee Nuclear

Power Plant Technical Specification 3.8,

‘‘Refueling,’’ and its associated Basis, by

allowing the containment personnel air

lock doors to remain open during

refueling operations.

Date of issuance May 7, 1997

Effective date: May 7, 1997

Amendment No.: 132

Facility Operating License No. DPR43: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 14, 1996 (61 FR

42285). The January 29, 1997, submittal

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

May 7, 1997.No significant hazards

consideration comments received: No.

Local Public Document Room

location: University of Wisconsin,

Cofrin Library, 2420 Nicolet Drive,

Green Bay, Wisconsin 54311-7001.

Notice Of Issuance Of Amendments To

Facility Operating LicensesAnd Final

Determination Of No Significant

Hazards ConsiderationAnd

Opportunity For A Hearing (Exigent

Public Announcement Or Emergency

Circumstances)

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application for the

amendment complies with the

standards and requirements of the

Atomic Energy Act of 1954, as amended

(the Act), and the Commission’s rules

and regulations. The Commission has

made appropriate findings as required

by the Act and the Commission’s rules

and regulations in 10 CFR Chapter I,

which are set forth in the license

amendment.

Because of exigent or emergency

circumstances associated with the date

the amendment was needed, there was

not time for the Commission to publish,

for public comment before issuance, its

usual 30-day Notice of Consideration of

Issuance of Amendment, Proposed No

Significant Hazards Consideration

Determination, and Opportunity for a

Hearing.

For exigent circumstances, the

Commission has either issued a Federal

Register notice providing opportunity

for public comment or has used local

media to provide notice to the public in

the area surrounding a licensee’s facility

of the licensee’s application and of the

Commission’s proposed determination

of no significant hazards consideration.

The Commission has provided a

reasonable opportunity for the public to

comment, using its best efforts to make

available to the public means of

communication for the public to

respond quickly, and in the case of

telephone comments, the comments

have been recorded or transcribed as

appropriate and the licensee has been

informed of the public comments.

In circumstances where failure to act

in a timely way would have resulted, for

example, in derating or shutdown of a

nuclear power plant or in prevention of

either resumption of operation or of

increase in power output up to the

plant’s licensed power level, the

Commission may not have had an

opportunity to provide for public

comment on its no significant hazards

consideration determination. In such

case, the license amendment has been

issued without opportunity for

comment. If there has been some time

for public comment but less than 30

days, the Commission may provide an

opportunity for public comment. If

comments have been requested, it is so

stated. In either event, the State has

been consulted by telephone whenever

possible.

Under its regulations, the Commission

may issue and make an amendment

immediately effective, notwithstanding

the pendency before it of a request for

a hearing from any person, in advance

of the holding and completion of any

required hearing, where it has

determined that no significant hazards

consideration is involved.

The Commission has applied the

standards of 10 CFR 50.92 and has made

a final determination that the

amendment involves no significant

hazards consideration. The basis for this

determination is contained in the

documents related to this action.

Accordingly, the amendments have

been issued and made effective as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the application for

amendment, (2) the amendment to
Federal Register / Vol. 62, No. 98 / Wednesday, May 21, 1997 / Notices 27809

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

20, 1997, the licensee may file a request

for a hearing with respect to issuance of

the amendment to the subject facility

operating license and any person whose

interest may be affected by this

proceeding and who wishes to

participate as a party in the proceeding

must file a written request for a hearing

and a petition for leave to intervene.

Requests for a hearing and a petition for

leave to intervene shall be filed in

accordance with the Commission’s

‘‘Rules of Practice for Domestic

Licensing Proceedings’’ in 10 CFR Part

2. Interested persons should consult a

current copy of 10 CFR 2.714 which is

available at the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington, DC and

at the local public document room for

the particular facility involved. If a

request for a hearing or petition for

leave to intervene is filed by the above

date, the Commission or an Atomic

Safety and Licensing Board, designated

by the Commission or by the Chairman

of the Atomic Safety and Licensing

Board Panel, will rule on the request

and/or petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses. Since the Commission has

made a final determination that the

amendment involves no significant

hazards consideration, if a hearing is

requested, it will not stay the

effectiveness of the amendment. Any

hearing held would take place while the

amendment is in effect.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555-001, 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 20555-001, and to the attorney for

the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of the

factors specified in 10 CFR

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

Commonwealth Edison Company,

Docket No. 50-265, Quad Cities Nuclear

Power Station, Unit 2, Rock Island

County, Illinois

Date of application for amendment:

April 29, 1997.

Brief description of amendment: The

proposed amendment modifies Section

5.3.A, ‘‘Design Features’’ of the

Technical Specifications (TS) to reflect

the ATRIUM-9B fuel design and would

include various Siemens Power

Corporation (SPC) topical reports in TS

Section 6.9.A.6, ‘‘Core Operating Limits

Report,’’ to reflect mechanical design

criteria for this fuel and topical reports

required for operation. This change

would allow this fuel to be loaded into

the core only under Operational Modes

3 (Hot Shutdown), 4 (Cold Shutdown),

and 5 (Refueling) and does not permit

startup or power operation using the

ATRIUM-9B fuel.

Date of issuance May 2, 1997

Effective date: May 2, 1997

Amendment No.: 173

Facility Operating License No. DPR30: The amendment revised the

Technical Specifications.Public

comments requested as to proposed no

significant hazards consideration:

No.The Commission’s related evaluation

of the amendment, finding of emergency

circumstances and final determination

of no significant hazards consideration

are contained in a Safety Evaluation

dated May 2, 1997.

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60603
27810 Federal Register / Vol. 62, No. 98 / Wednesday, May 21, 1997 / Notices

1 Investment Company Act Release Nos. 14220

(Oct. 31, 1984) (notice) and 14259 (Nov. 30, 1984)

(order).

Local Public Document Room

location: Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois

61021.

NRC Project Director: Robert A. Capra

Nebraska Public Power District, Docket

No. 50-298, Cooper Nuclear Station,

Nemaha County, Nebraska

Date of amendment request: May 2,

1997, as superseded May 5, 1997.

Brief description of amendment: The

proposed amendment relocates and

revises the requirements for the control

of the setpoint for the Standby Liquid

Control system relief valves. The

requirements would be relocated from

Section 4.4.A.2.a and Bases Section

3.4.A of the Cooper Technical

Specifications to the Updated Safety

Analysis Report and the Inservice

Testing Augmented Testing Program.

Date of issuance May 9, 1997

Effective date: May 9, 1997

Amendment No.: 176

Facility Operating License No. DPR46: The amendment revised the

Technical Specifications.Public

comments requested as to proposed no

significant hazards consideration:

No.The Commission’s related evaluation

of the amendment, finding of emergency

circumstances and final determination

of no significant hazards consideration

are contained in a Safety Evaluation

dated May 9, 1997.

Local Public Document Room

location: Auburn Memorial Library,

1810 Courthouse Avenue, Auburn, NE

68305.

Attorney for licensee: Mr. John R.

McPhail, Nebraska Public Power

District, Post Office Box 499, Columbus,

NE 68602-0499

NRC Project Director: William D.

Beckner

Dated at Rockville, Maryland, this 14th day

of May, 1997.

For the Nuclear Regulatory Commission

Elinor G. Adensam,

Deputy Director, Division of Reactor Projects

III/IV, Office of Reactor Regulation

[Doc. 97-13190 Filed 5-20-97; 8:45 am]

BILLING CODE 7590-01-F

RAILROAD RETIREMENT BOARD

Sunshine Act Meeting

The meeting of the Railroad

Retirement Board which was to be held

at 9:00 a.m. on May 21, 1997, at the

Board’s meeting room on the 8th floor

of its headquarters building, 844 North

Rush Street, Chicago, Illinois, 60611,

has been changed to 3:00 p.m. on May

21, 1997. The agenda for this meeting

was published at 62 FR 26342 on May

13, 1997.

The entire meeting will be open to the

public. The person to contact for more

information is Beatrice Ezerski,

Secretary to the Board, Phone No. 312–

751–4920.

Dated: May 16, 1997.

Beatrice Ezerski,

Secretary to the Board.

[FR Doc. 97–13423 Filed 5–20–97; 10:09 am]

BILLING CODE 7905–01–M

SECURITIES AND EXCHANGE

COMMISSION

[Investment Company Act Release No.

22663; 812–9440]

AIM Equity Funds, Inc., et. al.; Notice

of Application

May 15, 1997.

AGENCY: Securities and Exchange

Commission (‘‘SEC’’).

ACTION: Notice of application for

exemption under the Investment

Company Act of 1940 (the ‘‘Act’’).

APPLICANTS: AIM Equity Funds, Inc.,

AIM Funds Group, AIM International

Funds, Inc., AIM Investment Securities

Funds, AIM Summit Fund, Inc., AIM

Tax-Exempt Funds, Inc., AIM Variable

Insurance Funds, Inc., Short-Term

Investments Co., Short-Term

Investments Trust, and Tax-Free

Investments Co. (the ‘‘Funds’’), AIM

Advisors, Inc., and AIM Capital

Management, Inc. (the ‘‘Advisers,’’ and

collectively with the Funds, the

‘‘Applicants’’).

RELEVANT ACT SECTIONS: Order requested

under sections 6(c) and 17(b) of the Act

for an exemption from sections 17(a)

and 17(e) of the Act.

SUMMARY OF APPLICATION: Applicants

request an order amending a prior order

(the ‘‘Prior Order’’) under sections 6(c)

and 17(b) of the Act granting an

exemption from sections 17(a)(1),

17(a)(2) and 17(e) of the Act.1 The

requested order would let each Fund

engage in purchase and sale transactions

limited to U.S. government securities,

certain other high quality debt securities

and reverse repurchase agreements with

banks whose affiliated relationship with

the Funds arises solely out of their five

percent or greater share interest in a

Fund, except that no Fund will engage

in such transactions with a bank that

controls or advises that Fund. Any order

also would let each Fund compensate

an affiliated bank for acting as agent in

executing certain securities transactions.

FILING DATES: The application was filed

on January 19, 1995, and amended on

July 18, 1995, January 16, 1996, and

April 21, 1997. Counsel for applicants

has agreed to file another amendment

during the notice period, the substance

of which is incorporated herein.

HEARING OR NOTIFICATION OF HEARING: An

order granting the application will be

issued unless the SEC orders a hearing.

Interested persons may request a

hearing by writing to the SEC’s

Secretary and serving Applicants with a

copy of the request, personally or by

mail. Hearing requests should be

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

June 9, 1997, and should be

accompanied by proof of service on the

Applicants, in the form of an affidavit

or, for lawyers, a certificate of service.

Hearing requests should state the nature

of the writer’s interest, the reasons for

the request, and the issues contested.

Persons who wish to be notified of a

hearing may request such notification

by writing to the SEC’s Secretary.

ADDRESSES: Secretary, SEC, 450 Fifth

Street, NW., Washington, DC 20549.

Applicants, Eleven Greenway Plaza,

Suite 1919, Houston, Texas 77046.

FOR FURTHER INFORMATION CONTACT: H.R.

Hallock, Jr., Special Counsel, at (202)

942–0564 (Division of Investment

Management, Office of Investment

Company Regulation).

SUPPLEMENTARY INFORMATION: The

following is a summary of the

application. The complete application

may be obtained for a fee from the SEC’s

Public Reference Branch.

Applicants’ Representations

1. All of the Funds are registered

under the Act as open-end management

investment companies. AIM Advisors,

Inc., a wholly-owned subsidiary of AIM

Management Group Inc., a privatelyowned corporation, serves as

investment adviser for each Fund. AIM

Capital Management, Inc., a whollyowned subsidiary of AIM Advisors, Inc.,

serves as sub-adviser to three series

(‘‘Portfolios’’) of one of the Funds, AIM

Equity Inc. Both Advisers are registered

investment advisers under the

Investment Advisers Act of 1940.

2. The Prior Order granted the Funds

or certain of their predecessors a

conditional exemption, pursuant to

sections 6(c) and 17(b) of the Act, from

the provisions of section 17(a)(1),

section 17(a)(2) and section 17(e)

thereof. The Prior Order applies to

transactions by the Funds with a bank,

bank holding company or affiliate

thereof which may be deemed to be an