Federal Register / Vol. 62, No. 78 / Wednesday, April 23, 1997 / Notices 19825

For the Nuclear Regulatory Commission.

John W. N. Hickey,

Chief, Low-Level Waste and Decommissioning

Projects Branch, Division of Waste

Management, Office of Nuclear Material

Safety and Safeguards.

[FR Doc. 97–10523 Filed 4–22–97; 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 March 29,

1997, through April 11, 1997. The last

biweekly notice was published on April

9, 1997 (62 FR 17223).

Notice of Consideration Of Issuance Of

Amendments To Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

and Opportunith For A Hearing

The Commission has made a

proposed determination that the

following amendment requests involve

no significant hazards consideration.

Under the Commission’s regulations in

10 CFR 50.92, this means that operation

of the facility in accordance with the

proposed amendment would not (1)

involve a significant increase in the

probability or consequences of an

accident previously evaluated; or (2)

create the possibility of a new or

different kind of accident from any

accident previously evaluated; or (3)

involve a significant reduction in a

margin of safety. The basis for this

proposed determination for each

amendment request is shown below.

The Commission is seeking public

comments on this proposed

determination. Any comments received

within 30 days after the date of

publication of this notice will be

considered in making any final

determination.

Normally, the Commission will not

issue the amendment until the

expiration of the 30-day notice period.

However, should circumstances change

during the notice period such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received before

action is taken. Should the Commission

take this action, it will publish in the

Federal Register a notice of issuance

and provide for opportunity for a

hearing after issuance. The Commission

expects that the need to take this action

will occur very infrequently.

Written comments may be submitted

by mail to the Chief, Rules Review and

Directives Branch, Division of Freedom

of Information and Publications

Services, Office of Administration, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555-0001, and

should cite the publication date and

page number of this Federal Register

notice. Written comments may also be

delivered to Room 6D22, Two White

Flint North, 11545 Rockville Pike,

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

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

written comments received may be

examined at the NRC Public Document

Room, the Gelman Building, 2120 L

Street, NW., Washington, DC. The filing

of requests for a hearing and petitions

for leave to intervene is discussed

below.

By May 23, 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
19826 Federal Register / Vol. 62, No. 78 / Wednesday, April 23, 1997 / Notices

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555-0001, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington DC, by

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

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

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

operator should be given Datagram

Identification Number N1023 and the

following message addressed to (Project

Director): petitioner’s name and

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555-0001, and to the attorney for

the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

factors specified in 10 CFR

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

For further details with respect to this

action, see the application for

amendment which is available for

public inspection at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, and at the local public

document room for the particular

facility involved.

Carolina Power & Light Company, et

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

Nuclear Power Plant, Unit 1, Wake and

Chatham Counties, North Carolina

Date of amendment request: March

17, 1997

Description of amendment request:

The proposed change would revise eight

specifications for 18-month tests to

delete a conditional statement that the

testing be done while the unit is shut

down and to clarify that Harris Nuclear

Plant (HNP) may take credit for tests on

some components which are performed

while the unit is at power.

Basis for proposed no significant

Hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed changes permit HNP to

evaluate the conditions required to safely

perform a test, but the changes do not

directly affect the functioning or operation of

any plant equipment. Since no equipment

operation is involved there is no increase in

the probability or consequence of any

previously identified accident.

2. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed changes to the conditional

statements on the surveillance frequencies do

not involve any physical alterations or

additions to plant equipment or alter the

manner in which any safety-related system

performs itsfunction or is operated.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any previously

evaluated.

3. The proposed amendment does not

involve a significant reduction in the margin

of safety.

The proposed changes to the conditional

statements on the surveillance frequency

allows HNP to evaluate the conditions

needed to safely perform the required testing.

There is no change in the frequency of testing

or in the testing which is required. There is

no change in the responsibility of HNP

toperform tests in a safe and responsible

manner, and any changes to procedures will

have to be individually evaluated to ensure

that they do not reduce the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh,

North Carolina 27605

Attorney for licensee: William D.

Johnson, Vice President and Senior

Counsel, Carolina Power & Light

Company, Post Office Box 1551,

Raleigh, North Carolina 27602

NRC Project Director: Mark Reinhart,

Acting

Commonwealth Edison Company,

Docket Nos. STN 50-454 and STN 50-

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

Ogle County, Illinois Docket Nos. STN

50-456 and STN 50-457, Braidwood

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

Illinois

Date of amendment request: January

30, 1997

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) 1.0,

‘‘Definitions;’’ TS 3/4.6.1, ‘‘Primary

Containment’’ and associated Bases; and

TS 5.4.2, ‘‘Reactor Coolant System

Volume’’ for Byron and Braidwood to

support steam generator replacement.

ComEd will be replacing the original

Westinghouse D4 steam generators at

Byron and Braidwood with Babcock and

Wilcox International steam generators.

The replacement steam generators

increase the Reactor Coolant System

volume which results in a higher

calculated peak containment pressure

(Pa) value.

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.

Each of the [replacement steam generators]

RSGs has a larger [reactor coolant system]

RCS side volume than the original steam

generators (OSGs). As a result of the RCS
Federal Register / Vol. 62, No. 78 / Wednesday, April 23, 1997 / Notices 19827

volume increase, the mass and energy release

during the blowdown phase of the large

break loss of coolant accident (LBLOCA) is

increased. Additionally, the heat transfer rate

of the RSGs is greater than the OSGs, and the

RSGs will operate at a slightly higher

pressure than that for the OSGs.

Consequently, the steam enthalphy exiting

the break during the reflood period, with the

RSG, will be greater than that for the OSG.

This results in an increase in the

containment building peak pressure, Pa.

The proposed revisions to the Technical

Specifications involve the specified value of

Unit 1 RCS volume and the defined value of

Unit 1 Pa. Several editorial changes are also

being made to improve clarity and

consistency of the TS.

RCS volume is not an initiator for any

event and an increase in volume does not

affect any operating margin or requirements.

Therefore, increasing the primary volume

does not increase the probability of any event

previously analyzed.

The revised value of Pa continues to be less

than the design basis pressure for the

containment building structure. The change

represents only a revision to the containment

test pressure for containment leakage testing.

Such testing is only performed with the

affected unit in the shutdown condition.

Therefore, the proposed change in Pa does

not involve a significant increase in the

probability of an accident previously

evaluated.

All accidents in the Updated Final Safety

Analysis Report (UFSAR) were evaluated to

determine the effect of an increase in primary

volume on accident consequences. The

events identified that may be impacted by an

increase in primary volume are the Waste

Gas System Leak or Failure and LBLOCA. For

the Waste Gas System Leak or Failure, the

activity of the decay tank is controlled to

Technical Specification limits which are

unaffected by RCS volume. Therefore, an

increase in RCS volume would not increase

the offsite dose.

The offsite dose calculation for the

LBLOCA is unaffected by the proposed

change. The license basis offsite dose

calculation is in accordance with NRC Reg

Guide 1.4 ‘‘Assumptions Used for Evaluating

The Potential Radiological Consequences of a

Loss of Coolant Accident for Pressurized

Water Reactors.’’ This Regulatory Guide

states, in part, ’’...a number of appropriately

conservation assumptions, based on

engineering judgment and on applicable

experimental results from safety research

programs conducted by the AEC.’’ These

conservatisms include (but are not limited to)

the following assumptions:

• Twenty five percent of the equilibrium

radioactive full power inventory is

immediately available for leakage from the

primary containment.

• 100% of the equilibrium full power

radioactive noble gas inventory is

immediately available for leakage from the

primary containment.

• The primary containment should be

assumed to leak at the (maximum) leak rate

specified in the technical specifications for

the first 24 hours and at 50% of this value

for the remaining 29 days of the accident

duration.

The design basis leakage corresponding to

a peak containment pressure of 50 psig

utilized in the design basis accident analysis

is 0.10% per day of the containment free air

mass. Therefore, the offsite dose calculation

was performed with a leakage of .1% per day

for day one and .05% per day for days two

through 30. Isotopic inventories are

unaffected by the increase in reactor coolant

volume. Thus, the offsite dose is unaffected

by the increase in the peak containment

pressure. Therefore, this proposed change to

Pa does not involve a significant increase in

the consequences of an accident previously

evaluated.

The editorial changes proposed are for

clarity and consistency within the Technical

Specifications and do not affect either the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed change in RCS volume is a

change in a plant parameter within the

‘‘Design Features’’ section of the Technical

Specifications. Increasing the RCS volume

does not create any new or different failure

modes. The existing RCS design

requirements continue to be met.

The revised value of Pa continues to be less

than the design basis pressure for the

containment building structure. The change

represents only a revision to the test pressure

for containment leakage testing. Such testing

is only performed with the affected unit in

the shutdown condition. Therefore, no new

or different failure modes are being

introduced by modification of the testing

parameters.

The editorial changes proposed are for

clarity and consistency within the Technical

Specifications and do not result in any

physical changes to the facility or how it is

operated. No new or different failure modes

are being introduced by these changes.

Therefore, these proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

Changing the RCS volume in the Technical

Specifications does not reduce the margin of

safety. RCS volume is a design feature. The

change in RCS volume does not involve a

change to any setpoint or design

requirements. An evaluation of all UFSAR

accidents was performed to determine the

effect of an increase in RCS volume. This

evaluation is summarized as follows:

An evaluation of the Chemical and Volume

Control System Malfunction was performed

to determine the effect of the increased RCS

volume due to the RSGs. The larger RCS

volume of the RSGs reduces the reactivity

insertion for a given dilution flow rate.

Therefore, the UFSAR analyses remain

bounding for Byron Unit 1 and Braidwood

Unit 1 with the RSGs and there is no

reduction in the margin of safety.

An evaluation of the Inadvertent Actuation

of the Emergency Core Cooling System

During Power Operation Event was

performed to determine the effect of the

increased RCS volume due to the RSGs. For

this event, the injection of borated water

causes a negative reactivity insertion, which

increases DNBR. For a given Refueling Water

Storage Tank (RWST) boron concentration,

the larger RCS volume will cause a reduction

in the negativity insertion rate as compared

to the current UFSAR analysis. However,

negative reactivity would still be inserted, no

fuel pins would experience DNB, and there

is no reduction in the margin of safety.

An evaluation of the Small Break LOCA

was performed to determine the effect of

increased RCS volume. The additional RCS

volume will cause a delay in the loop seal

clearing which in turn delays the core

uncovery as compared with the UFSAR

analysis. A delay in core uncovery reduces

the amount of core heatup which results in

a lower peak clad temperature (PCT) because

the core decay heat would be less than in the

UFSAR analysis. The benefit is considered

small, but there is still a benefit. Therefore,

the increased RCS volume does not result in

a reduction in the margin of safety.

An evaluation of the Large Break LOCA

was performed to determine the effect of

increased RCS volume. For a LBLOCA, the

increased RCS volume causes the blowdown

phase of the event to be longer. Increased

blowdown phase, alone, could potentially

result in a higher PCT. However, the RSGs

also have less resistance to flow due to

increased primary side steam generator flow

area, which results in a higher blowdown

flow compared to the OSGs. The increased

blowdown flow more than compensates for

the longer blowdown phase associated with

the increased RCS volume. The net effect is

a decrease in PCT for the RSG compared to

the OSG. Therefore, there is no reduction in

the margin of safety.

An evaluation of the Gas Waste System

Leak or Failure was performed to determine

the effect of the increased RCS volume.

Because the activity of the decay tank is

controlled within Technical Specification

limits, an increase in RCS volume would not

change the results of the event. Therefore,

there is no reduction in the margin of safety.

An evaluation was performed to determine

the effect of the increased RCS volume on the

peak containment pressure following a

LBLOCA. The increased RCS volume caused

the peak containment pressure to increase to

47.8 psig. This is still below the containment

design pressure of 50.0 psig. Therefore, there

is no reduction in the margin of safety.

This proposed change involves testing

requirements designed to demonstrate

adequate leakage rates are maintained. If

adequate leakage rates are maintained as

outlined in the Technical Specifications,

there will be no reduction in the margin of

safety. In the event of degradation of a

containment seal that results in unacceptable

leakage, plant shutdown will occur as

required by Technical Specifications and

administrative requirements in accordance

with approved plant procedures. Therefore,

this proposed change does not involve a

significant reduction in a margin of safety.

The editorial changes proposed are for

clarity and consistency within the Technical

Specifications and do not result in any

physical changes to the facility or how it is
19828 Federal Register / Vol. 62, No. 78 / Wednesday, April 23, 1997 / Notices

operated. Therefore, the changes have no

effect on the margin of safety.

Thus, this amendment request does not

result in any decrease 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

requested amendments involve no

significant hazards consideration.

Local Public Document Room

location: For Byron, the Byron Public

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

Box 434, Byron, Illinois 61010; for

Braidwood, the Wilmington Public

Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481.

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60603

NRC Project Director: Robert A. Capra

Consumers Power Company, Docket

No. 50-255, Palisades Plant, Van Buren

County, Michigan

Date of amendment request: March

27, 1997

Description of amendment request:

The proposed amendment would alter

the company name in the Facility

Operating License DPR-20 and

Technical Specifications for the

Palisades Plant. Specifically, the

proposed amendment would revise the

name from ‘‘Consumers Power

Company’’ to ‘‘Consumers Energy

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:

A. Do the proposed changes involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

Since the proposed changes do not alter

the technical content of any Facility

Operating License or Technical

Specifications requirements, they do not alter

any feature of plant equipment, settings,

operation, or configuration.

Therefore, they cannot involve a significant

increase in the probability of an accident

previously evaluated.

The proposed changes alter the company

name in the Facility Operating License and

Technical Specifications to reflect the change

from ‘‘Consumers Power Company’’ to

‘‘Consumers Energy Company’’. The

proposed change will not affect any

obligations. The company will continue to

own all of the same assets, will continue to

serve the same customers, and will continue

to honor all existing obligations and

commitments. The proposed changes will not

alter plant operation or configuration, or its

ability to respond to accidents.

Therefore, they will not involve a

significant increase in the consequences of

any accident previously evaluated.

B. Do the proposed changes create the

possibility of a new or different kind of

accident from any previously evaluated?

Since the proposed changes do not alter

the technical content of any Facility

Operating License or Technical

Specifications requirements, they do not alter

any feature of plant equipment, settings,

operation or configuration.

Therefore, they cannot create the

possibility of a new or different kind of

accident from any previously evaluated.

C. Do the proposed changes involve a

significant reduction in a margin of safety?

Since the proposed changes do not alter

the technical content of any Facility

Operating License or Technical

Specifications requirements, they do not alter

any feature of plant equipment, settings,

operation, or configuration.

Therefore, they cannot involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Van Wylen Library, Hope

College, Holland, Michigan 49423

Attorney for licensee: Judd L. Bacon,

Esquire, Consumers Power Company,

212 West Michigan Avenue, Jackson,

Michigan 49201

NRC Project Director: John N. Hannon

Duke Power Company, Docket Nos. 50-

269, 50-270 and 50-287, Oconee

Nuclear Station, Units 1, 2, and 3,

Oconee County, South Carolina

Date of amendment request: March

31, 1997 (TSC 96-10)

Description of amendment request:

The proposed amendments would

modify and clarify the High Pressure

Injection (HPI) System operability

requirements in Specification 3.3.1,

impose additional HPI system

operability requirements for operation

above 75 percent power, incorporate the

new Standard Technical Specifications

format for the HPI system, revise

Specification 3.3.2 to clarify that the

Reactor Building Emergency Sump

isolation valves are remote-manually

operated valves, and add new

specifications and a surveillance test to

address operability requirements of the

atmospheric dump valves. In addition,

corresponding Bases changes would be

incorporated.

Basis for proposed no significant

Hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) Involve a significant increase in the

probability or consequences of an accident

previously evaluated:

No. None of the proposed changes has any

impact upon the probability of any accident

which has been evaluated in the UFSAR

[Updated Final Safety Analysis Report]. The

only potential change in operating

configuration is allowing operation with the

HPI [High Pressure Injection] System pump

discharge header crossconnected. This operating mode does not

affect the probability of a LOCA [Loss-ofCoolant Accident] or of any other accident

evaluated in the UFSAR.

None of these changes have any impact

upon the ability of the HPI System to add

soluble poison to the Reactor Coolant

System, as addressed by Specification 3.2.

The remaining potential impact is upon the

ability to mitigate the consequences of a

small break LOCA, which is addressed

below. The small break LOCA is the limiting

design basis accident with respect to HPI

System operability requirements.

The proposed changes to Specification

3.3.1 provide appropriate actions to address

any degradation in the operability of the HPI

System. The operability requirements for the

HPI System are supported by a spectrum of

small break LOCA analyses based on the

approved Evaluation Model described in FTI

[Framatome Technologies, Incorporated]

topical report BAW-10192P. These small

break LOCA analyses demonstrate that the

acceptance criteria of 10CFR 50.46 are not

violated.

Two trains of HPI are required to mitigate

a small break LOCA above 75% FP [full

power]. Operability requirements in the

proposed Technical Specifications assure

that the HPI System can withstand the worst

single failure and still result in two HPI

pumps injecting through two trains. The full

power small break LOCA analyses supporting

this proposed license amendment have been

performed in accordance with the approved

Evaluation Model described in FTI topical

report BAW-10192P. The proposed Technical

Specifications limit operation above 75% FP

with a degraded HPI System to 72 hours

before a power reduction to less than 75% FP

(or a reactor shutdown) must be initiated.

The required actions depend on the HPI

System components that are inoperable. The

72 hour completion time is consistent with

the time requirements for HPI specified in

NUREG-1430.

When at or below 75% FP, one HPI train

provides sufficient flow to mitigate a small

break LOCA. The 75% power level is

justified by analyses using the Evaluation

Model described in FTI topical report BAW10192P, considering the worst case break

location and size described in LER [Licensee

Event Report] 269/90-15 and Attachment 3 to

this submittal. The proposed Technical

Specifications require two HPI trains to be

operable at or below 75% FP. These

requirements ensure that, following the worst

single failure, one train of HPI would remain
Federal Register / Vol. 62, No. 78 / Wednesday, April 23, 1997 / Notices 19829

available to mitigate a small break LOCA.

Operation with less than two HPI trains

operable is restricted to 72 hours before

shutdown requirements are imposed. This

completion time is consistent with the time

requirements specified for an HPI System in

NUREG-1430.

The additional HPI system restriction that

requires the HPI pump discharge header to be

cross-connected when all three HPI pumps

are operable does not increase the

consequences of a small break LOCA. If a

single failure prevents one HPI train from

actuating, this lineup results in at least two

HPI pumps initially injecting through the

automatically actuating train. This increases

the amount of cooling flow initially delivered

to the core as compared to the current system

configuration.

The impact of this alignment has been

evaluated, considering the potential single

active failures, including the failure of any

powered component to operate and any

single failure of electrical equipment.

It has been determined that, when each of

the three HPI pumps is either running or is

capable of automatic actuation upon an

Engineered Safeguards signal, crossconnection of the HPI pump discharge

header does not introduce susceptibility to

any single failure. Therefore, the potential

consequences of a small break LOCA are not

increased. If fewer than three HPI pumps are

either running or are capable of automatic

actuation, and the HPI pump discharge

header were cross-connected, a single failure

of one pump could cause a single pump to

be aligned to both HPI trains. In this

condition, the single pump could experience

runout conditions prior to corrective operator

action. However, proposed Specification

3.3.1 requires the discharge header to be

isolated between the two remaining operable

HPI pumps. The proposed BASES provide

guidelines to ensure that the requirements for

redundancy are properly implemented.

Therefore, the proposed specifications ensure

that the consequences of a small break LOCA

are not increased by allowing the HPI pump

discharge header to be cross-connected.

In addition, proposed Specification 3.4.7

requires new operability requirements for the

main steam atmospheric dump valves. These

operability requirements do not impact the

probability or consequences of any accident.

The proposed specification for the

atmospheric dump valves provides

additional assurance that these valves will be

operable in the event of a small break LOCA.

In summary, the proposed Technical

Specifications provide adequate controls to

assure that operability of the HPI System is

maintained in a manner consistent with the

requirements of the design basis accidents.

Therefore, it is concluded that this

amendment request will not significantly

increase the probability or consequences of

an accident previously evaluated.

(2) Create the possibility of a new or

different kind of accident from any kind of

accident previously evaluated:

No. Of the proposed substantive changes,

only cross-connection of the HPI pump

discharge header represents any change to

the way in which the facility is normally

operated. Operation with the discharge

header cross-connected is not a new

configuration, as it has always been used for

HPI pump testing both at power and during

shutdown conditions. Potential failure modes

have already been considered as described

earlier. No new initiating events or

potentially unanalyzed conditions have been

created. Therefore, this proposed amendment

will not create the possibility of any new or

different kind of accident.

(3) Involve a significant reduction in a

margin of safety.

No. The HPI restrictions associated with

the proposed Technical Specifications are

supported by analyses which demonstrate

that the acceptance criteria of 10 CFR 50.46

are not violated for any small break LOCA.

These analyses were performed in

accordance with the Evaluation Model

described in FTI topical report BAW-10192P.

Therefore, it is concluded that the proposed

amendment request will not result in a

significant decrease in the margin of safety.

Duke has concluded, based on the above,

that there are no significant hazards

considerations involved in this amendment

request.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Oconee County Library, 501

West South Broad Street, Walhalla,

South Carolina 29691

Attorney for licensee: J. Michael

McGarry, III, Winston and Strawn, 1200

17th Street, NW., Washington, DC 20036

NRC Project Director: Herbert N.

Berkow

Duquesne Light Company, et al., Docket

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

Power Station, Unit Nos. 1 and 2,

Shippingport, Pennsylvania

Date of amendment request: March

10, 1997

Description of amendment request:

The proposed amendment would

modify Technical Specification 3.4.5,

‘‘Steam Generators,’’ and associated

Bases to allow repair of steam generator

tubes by installation of sleeves with the

tungsten inert gas (TIG) welded sleeve

developed by ABB Combustion

Engineering. In addition, the proposed

amendment would delete the option for

using the kinetic sleeving methodology

previously approved for use at Beaver

Valley, but is not currently

recommended by Framatome

Technologies, Inc.

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 amendment allows the ABB

Combustion Engineering (ABB/CE) tungsten

inert gas (TIG) welded tubesheet sleeves and

tube support plate sleeves to be used as an

alternate steam generator tube repair method.

The sleeve configuration was designed and

analyzed in accordance with the criteria of

Regulatory Guide (RG) 1.121 and Section III

of the ASME [American Society of

Mechanical Engineers] Code. Fatigue and

stress analyses of the sleeved tube assemblies

produce acceptable results for both types of

sleeves as documented in ABB/CE Topical

Report CEN-629-P, Revision 02 and CEN-629-

P Addendum 1. Mechanical testing has

shown that the structural strength of the

sleeves under normal, faulted, and upset

conditions is within the acceptable limits

specified in RG 1.121. Leakage rate testing for

the tube sleeves has demonstrated that

primary to secondary leakage is not expected

during any plant condition. The

consequences of leakage through the sleeved

region of the tube is fully bounded by the

existing steam generator tube rupture (SGTR)

analysis included in the Updated Final

Safety Analysis Report (UFSAR).

The sleeves are designed to allow inservice

inspection of the pressure retaining portions

of the sleeve and parent tube. Inservice

inspection is performed on all sleeves

following installation to ensure that each

sleeve has been properly installed and is

structurally sound. Periodic inspections are

performed in subsequent refueling outages to

monitor sleeve degradation on a sample

basis. The eddy current technique used for

inspection will be capable of detecting both

axial and circumferential flaws. Specific

guidance for steam generator sleeve

inspection is provided in the current

technical specification surveillance

requirements. Tubes that contain defects in a

sleeve, which exceed the repair limit, will be

removed from service. This ensures that

sleeve and tube structural integrity is

maintained.

The proposed TS change to support the

installation of TIG welded sleeves does not

adversely impact any previously evaluated

design basis accident. The effect of sleeve

installation on the performance of the SG

[steam generator] was analyzed for heat

transfer, flow restriction, and steam

generation capacity. The sleeves reduce the

risk of primary to secondary leakage in the

SG. The installation of ABB/CE sleeves

results in a hydraulic flow restriction that is

dependent on the number and types of

sleeves installed. The reduction in primary

system flow rate is a small percentage of the

flow rate reduction seen from plugging one

tube and is a preferable alternative when

considering core margins based on minimum

reactor coolant system flow rates. The

sleeving installation will result in a

resistance to primary coolant flow through

the tube for other evaluated accidents. The

results of the analyses and testing, as well as

industry operating experience, demonstrate

that the sleeve assembly is an acceptable
19830 Federal Register / Vol. 62, No. 78 / Wednesday, April 23, 1997 / Notices

means of maintaining tube integrity. In

summary, installation of sleeves does not

substantially affect the primary system flow

rate or the heat transfer capability of the

steam generators.

Installation of the sleeves can be used to

repair degraded tubes by returning the

condition of the tubes to their original design

basis condition for tube integrity and leak

tightness during all plant conditions. The

tube bundle overall structural and leakage

integrity will be increased with the

installation of the sleeves reducing the risk

of primary to secondary leakage in the SG

while maintaining acceptable reactor coolant

system flow rates. Therefore, sleeving will

not increase the probability of occurrence of

an accident previously evaluated.

Removal of the kinetically welded sleeve

process as an approved SG tube repair

methodology will have no effect on plant

operations. There are currently no kinetically

welded sleeves installed in the steam

generators. Had there been, plant operations

would have still been bounded by the

existing SGTR analysis in the UFSAR.

Therefore, these proposed changes do 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 implementation of the proposed

sleeving process will not introduce

significant or adverse changes to the plant

design basis. Stress and fatigue analyses of

the repair has shown the ASME Code Section

III and RG 1.121 allowable values are met.

Implementation of TIG welded sleeving

maintains overall tube bundle structural and

leakage integrity at a level consistent with

that of the originally supplied tubing. Leak

and mechanical testing of the sleeves support

the conclusions that the sleeve retains both

structural and leakage integrity during all

conditions. Repair of a tube with a sleeve

does not provide a mechanism that would

result in an accident outside of the area

affected by the sleeve.

Any hypothetical accident as a result of

potential tube or sleeve degradation in the

repaired portion of the tube is bounded by

the existing SGTR analysis. The SGTR

analysis accounts for the installation of

sleeves and the impact on current plugging

level analyses. The sleeve design does not

affect any other component or location of the

tube outside of the immediate area repaired.

The current primary to secondary leakage

limit ensures that SG tube integrity is

maintained in the event of an MSLB [main

steam line break] or LOCA [loss-of-coolant

accident]. The limit will provide for leakage

detection and a plant shutdown in the event

of the occurrence of an unexpected single

crack resulting in excessive tube leakage. The

leakage limit also provides for early detection

and a plant shutdown prior to a postulated

crack reaching critical crack lengths for

MSLB conditions.

Inservice inspections are performed

following sleeve installation to ensure proper

weld fusion has occurred to maintain

structural integrity. The post installation

inspection also serves as baseline data to be

used for comparison during future

inspections. Periodic eddy current

inspections monitor the pressure retaining

portions of the sleeve and parent tube for

degradation. Eddy current techniques will be

employed that are sensitive to axial and

circumferential degradation.

Increasing the sample size of tubes

repaired using either sleeving process during

each scheduled inservice inspection will

increase the monitoring of these tubes for any

further degradation. The improved

monitoring and evaluation of the tube and

the sleeves assures tube structural integrity is

maintained or the tube is removed from

service.

Corrosion testing of typical sleeve-tube

configurations was performed to evaluate

local stresses, sleeve life, and resistance to

primary and secondary side corrosion. The

tests were performed on stress relieved and

as-welded (non-stress relieved) sleeve-tube

joints. Using the corrosion test data in

conjunction with finite element analyses of

the local stress, the stress relieved joint life

was determined to be in excess of 40 years.

The ABB/CE TIG welded sleeve operating

experience in the industry has shown no

sleeve failures due to service induced

degradation in sleeves that were installed

with acceptable inspection results. This

experience includes the stress relieved and

as-welded sleeve configurations. All sleeves

will be stress relieved as specified in the

topical report.

Removal of the kinetically welded sleeve

process as an approved SG tube repair

methodology and not completing the

additional corrosion testing necessary to

establish the design life for the kinetically

welded sleeve in the presence of a crevice

will not create the possibility of a new or

different type of accident from any accident

previously evaluated.

Repair of an SG tube with a kinetically

welded sleeve would not have provided a

mechanism that resulted in an accident

outside of the area affected by the sleeve.

Any hypothetical accident as a result of

potential tube or sleeve degradation in the

repaired portion of the tube would have been

bounded by the existing SGTR analysis. The

SGTR analysis accounts for the installation of

sleeves and the impact on current plugging

level analyses. The sleeve design does not

affect any other component or location of the

tube outside of the immediate area repaired.

Furthermore, there are currently no

kinetically welded sleeves installed in either

plant.

Therefore, the proposed changes do not

create the possibility of a new or different

type of accident from any accident

previously evaluated.

3. Does the change involve a significant

reduction in a margin of safety?

The TIG welded sleeving repair of

degraded steam generator tubes has been

shown by analysis to restore the integrity of

the tube bundle to its original design basis

condition. The safety factors used in the

design of the sleeves for the repair of

degraded tubes are consistent with the safety

factors in the ASME Boiler and Pressure

Vessel Code Section III used in steam

generator design. The design of the ABB/CE

SG sleeves has been verified by testing to

preclude leakage during normal and

postulated accident conditions.

The portion of the installed sleeve

assembly which represents the reactor

coolant pressure boundary can be monitored

for the initiation and progression of sleeve/

tube wall degradation, thus satisfying the

requirement of RG 1.83. The portion of the

SG tube bridged by the sleeve joints is

effectively removed from the pressure

boundary, and the sleeve then forms the new

pressure boundary. The sleeve enhances the

safety of the plant by reestablishing the

protective boundaries of the steam generator.

Keeping the tube in service with the use of

a sleeve instead of plugging the tube and

removing it from service increases the heat

transfer efficiency of the steam generator.

During each scheduled inservice inspection,

each sleeve inspected and found to have

unacceptable degradation shall be removed

from service.

The effect on the design transients and the

accident analyses have been revised based on

the installation of sleeves equal to the tube

plugging level coincident with the minimum

reactor coolant flow rate. Evaluation of the

installation of sleeves was based on the

determination that LOCA evaluations for the

licensed minimum reactor coolant flow

bound the combined effect of tube plugging

and sleeving up to an equivalent of the actual

plugging limit. Sleeving results in a fractional

amount of the plugging limitation of one tube

and is a preferable alternative when

considering core margins based on minimum

reactor coolant system flow rates. The

sleeving installation will result in a

resistance to primary coolant flow through

the tube. The primary coolant flow through

the ruptured tube is reduced by the influence

of the installed sleeve; therefore, the

consequences to the public due to an SGTR

event have not increased.

As SG sleeve removes an indication of a

possible leak source from the reactor coolant

system (RCS) pressure boundary, eliminating

the potential of a primary-to-secondary leak.

The structural integrity of the tube is

maintained by the sleeve and sleeve-to-tube

joint.

Installation of either tube sheet or tube

support plate sleeves will increase the

protective boundaries of the steam generators

and will not reduce the margin of safety.

Removal of the kinetically welded sleeve

process as an approved SG tube repair

methodology will not result in a reduction in

the margin of safety. There are currently no

kinetically welded sleeves installed in either

plant. SG tube integrity will be maintained

by applying an alternate NRC approved

repair methodology or removing the SG tube

from service by plugging.

Therefore, the proposed changes do not

involve a significant reduction in the margin

of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.
Federal Register / Vol. 62, No. 78 / Wednesday, April 23, 1997 / Notices 19831

Local Public Document Room

location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, PA

15001

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: John F. Stolz

Duquesne Light Company, et al., Docket

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

Power Station, Unit Nos. 1 and 2,

Shippingport, Pennsylvania

Date of amendment request: March

10, 1997

Description of amendment request:

The proposed amendment would revise

Technical Specifications 3.4.5, ‘‘Steam

Generators,’’ and associated Bases to

allow repair of steam generator tubes by

installation of sleeves with the

Electrosleeving process developed by

Framatome Technologies, Inc. (FTI).

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 Electrosleeve configuration has been

designed and analyzed in accordance with

the requirements of the ASME [American

Society of Mechanical Engineers] Code. The

applied stresses and fatigue usage for the

Electrosleeve are bounded by the limits

established in the ASME Code. Minimum

material property values are used for the

structural and plugging limit analysis.

Mechanical testing has shown that the

structural strength of nickel Electrosleeves

under normal, upset, and faulted conditions

provides margin to the acceptance limits.

These acceptance limits bound the most

limiting (3 times normal operating pressure

differential) burst margin recommended by

Regulatory Guide 1.121. Leakage testing has

shown that the Electrosleeve is essentially

leaktight during all plant conditions.

The Electrosleeve nominal wall thickness

depth-based plugging limit is determined

using the guidance of Regulatory Guide 1.121

and the pressure stress equation of Section III

of the ASME Code. The limiting requirement

of Regulatory Guide 1.121 for the

Electrosleeve, which applies to part through

wall degradation, is the minimum acceptable

wall thickness to maintain a safety factor of

three against tube failure under normal

operating conditions. A bounding set of

design and transient loading input conditions

was used for the minimum wall thickness

evaluation in the generic evaluation.

Evaluation of the minimum acceptable wall

thickness for normal, upset and postulated

accident condition loading per the ASME

Code indicates these conditions are bounded

by the design minimum wall thickness.

Bounding tube wall degradation growth

rate per cycle and nondestructive

examination uncertainty has been assumed

for determining the Electrosleeve technical

specification plugging limit. Electrosleeve

wall degradation extent determined by

nondestructive examination, which would

require plugging Electrosleeved tubes, is

developed using the guidance of Regulatory

Guide 1.121 and is defined in FTI Topical

Report BAW-10219P, Revision 1, to be 20%

throughwall of the nominal sleeve wall

thickness.

The effect of Electrosleeving and plugging

will remain below the plugging limit

assumed in the UFSAR [Updated Final Safety

Analysis Report]. The proposed change will

not increase the consequences of these

accidents.

The results of the analyses and testing

demonstrate that the Electrosleeve is an

acceptable means of maintaining tube

integrity. Further, per Regulatory Guide 1.83

recommendations, the Electrosleeved tube

can be monitored through periodic

inspections with present NDE

[nondestructive examination] techniques.

These measures demonstrate that installation

of Electrosleeves spanning degraded areas of

the tube will restore the tube to a condition

consistent with its original design basis.

Since the main steamline break postaccident primary-to-secondary leakage is not

increased by the presence of Electrosleeves,

the consequences of an accident previously

evaluated in the UFSAR are not increased.

Conformance of the Electrosleeve design with

the applicable sections of the ASME Code

and results of the leakage and mechanical

tests support the conclusion that installation

of Electrosleeves does not increase 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?

Electrosleeving will not adversely affect

any plant component. Stress and fatigue

analysis of the repair has shown that the

ASME Code and Regulatory Guide 1.121

criteria are not exceeded. Implementation of

Electrosleeving maintains overall tube

bundle structural and leakage integrity at a

level consistent with that of the original

tubing during all plant conditions. Leak and

mechanical testing of Electrosleeves support

the conclusions of the calculations that each

Electrosleeve retains both structural and

leakage integrity during all conditions.

Electrosleeving of tubes does not provide a

mechanism resulting in an accident outside

of the area affected by the Electrosleeves.

Any accident resulting from potential tube or

Electrosleeve degradation in the repaired

portion of the tube is bounded by the existing

tube rupture accident analysis.

Implementation of Electrosleeving will

reduce the potential for primary-to-secondary

leakage while not significantly impacting

available primary coolant flow area in the

event of a LOCA. By effectively isolating

degraded areas of the tube through repair, the

potential for steamline break leakage is

reduced. These degraded intersections now

are returned to a condition consistent with

the Design Basis. While the installation of an

Electrosleeve reduces primary coolant flow,

the reduction is far below that caused by

plugging. Greater primary coolant flow area

is maintained through Electrosleeving versus

plugging. Therefore, the possibility of a new

or different kind of accident from any

accident previously evaluated is not created.

3. Does the change involve a significant

reduction in a margin of safety?

The Electrosleeve repair of degraded steam

generator tubes has been shown by analysis

to restore the integrity of the tube bundle

consistent with its original design basis

condition. The tube/Electrosleeve operational

and faulted condition stresses are bounded

by the ASME Code requirements and the

Electrosleeved tubes are leaktight. The safety

factors used in the design of Electrosleeves

for the repair of degraded tubes are consistent

with the safety factors in the ASME Code

used in steam generator design. The portions

of the installed Electrosleeve assembly which

represent the reactor coolant pressure

boundary can be monitored for the initiation

and progression of Electrosleeve/tube wall

degradation, thus satisfying the requirements

of Regulatory Guide 1.83. The portion of the

tube bridged by the Electrosleeve is

effectively removed from the pressure

boundary, and the Electrosleeve then forms

the new pressure boundary. The areas of the

Electrosleeved tube assembly which require

inspection are defined in Framatome

Technologies Inc. Topical Report BAW10219P, Revision 1.

In addition, since the installed

Electrosleeve represents a portion of the

pressure boundary, a baseline inspection of

these areas is required prior to operation with

Electrosleeves installed. The effect of

sleeving on the design transients and

accident analyses has been reviewed based

on the installation of Electrosleeves up to the

level of steam generator tube plugging

coincident with the minimum reactor coolant

flow rate and UFSAR and has been found

acceptable.

It is concluded that the proposed license

amendment request does not result in a

significant reduction in the margin of safety

as defined in the UFSAR or technical

specifications.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, PA

15001

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: John F. Stolz

Northeast Nuclear Energy Company, et

al., Docket No. 50-336, Millstone

Nuclear Power Station, Unit No. 2, New

London, Connecticut

Date of amendment request: March

27, 1997
19832 Federal Register / Vol. 62, No. 78 / Wednesday, April 23, 1997 / Notices

Description of amendment request:

The proposed changes to the Technical

Specifications (TSs) would modify the

limiting condition for operation (LCO)

and surveillance requirements (SR) for

the ultimate heat sink. The ultimate heat

sink for Millstone Unit No. 2 is the Long

Island Sound that transfers heat from

safety-related systems during normal

and accident conditions. Specifically,

TS LCO 3.7.11 would be changed to

indicate that the ultimate heat sink is

operable at a water temperature of less

than or equal to 75 °F instead of an

average value. TS SRs 4.7.11.a and .b

would also delete the use of average

when verifying the water temperature

and delete the reference to a specific

monitoring location, the Unit No. 2

intake structure. These proposed

changes do not change the ultimate heat

sink temperature limit, which remains

at a maximum of 75 °F.

The TS Bases 3/4.7.11 would also be

modified to reflect the above changes

and to identify the various locations that

the ultimate heat sink temperature can

be measured.

Basis for proposed no significant

Hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The proposed changes do not involve an

SHC [significant hazards consideration]

because the changes would not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed changes remove the

reference to a monitoring location where the

temperature of the ultimate heat sink is

measured and eliminate the use of an average

ultimate heat sink temperature. The

instruments used provide information to the

operators which will permit them to ensure

that the plant is operated within the design

basis of the plant. The subject instruments

will provide an accurate representation of the

ultimate heat sink temperature. This role is

passive; thus, these instruments cannot

initiate or mitigate any accident.

The locations used to monitor the ultimate

heat sink temperature will be maintained in

the Bases. This is a licensee controlled

document which is maintained under the

requirements of 10CFR50.59. The details

being removed from the Technical

Specifications are not assumed to be an

initiator of any analyzed event. Since any

changes to the relocated details will be

evaluated per 10CFR50.59, any possible

increase in the probability or consequences

of an accident previously evaluated will be

addressed.

The proposed changes do not revise the

ultimate heat sink temperature limit of 75 °F.

The current analysis is based on the ultimate

heat sink temperature limit of 75 °F.

Therefore, there is no effect on the

consequences of any accident previously

evaluated.

Thus, the license amendment request does

not impact the probability of an accident

previously evaluated nor does it involve a

significant increase in the consequences of an

accident previously evaluated.

2. Created the possibility of a new or

different kind of accident from any

previously evaluated.

The proposed changes remove the

reference to a monitoring location where the

temperature of the ultimate heat sink is

measured and eliminate the use of an average

ultimate heat sink temperature. The

instruments used provide information to the

operators which will permit them to ensure

that the plant is operated within the design

basis of the plant. The subject instruments

will provide an accurate representation of the

ultimate heat sink temperature. This role is

passive, thus, these instruments cannot

initiate or mitigate any accident.

The proposed changes will not alter the

plant configuration (no new or different type

of equipment will be installed) or require any

new or unusual operator actions. They do not

alter the way any structure, system, or

component functions and do not alter the

manner in which the plant is operated. The

proposed changes do not introduce any new

failure modes. They will not alter

assumptions made in the safety analysis and

licensing basis.

The locations used to monitor the ultimate

heat sink temperature will be maintained in

the Bases. This is a licensee controlled

document which is maintained under the

requirements of 10CFR50.59. Thus, adequate

control of information will be ensured.

Therefore, the changes will 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 remove the

reference to a monitoring location where the

temperature of the ultimate heat sink is

measured and eliminate the use of an average

ultimate heat sink temperature. They do not

change the ultimate heat sink temperature

limit of 75 °F, which is assumed by the

current analysis. Therefore, there is no effect

on the consequences of any accident

previously evaluated and no significant

impact on offsite doses associated with

previously evaluated accidents. Thus, there

is no significant reduction in the margin of

safety for the design basis accident analysis.

The license amendment request does not

result in a reduction of the margin of safety

as defined in the Bases for Technical

Specification 3.7.11. The instruments used

provide information to the operators which

will permit them to ensure that the plant is

operated within the design basis of the plant.

The subject instruments will provide an

accurate representation of the ultimate heat

sink temperature. The proposed changes do

not alter the way any structure, system, or

component functions and do not alter the

manner in which the plant is operated. They

do not have any impact on the protective

boundaries (e.g., fuel matrix and cladding,

reactor coolant system pressure boundary,

and primary and secondary containment), or

on the safety limits for these boundaries.

The locations used to monitor the ultimate

heat sink temperature will be maintained in

the Bases. The Bases are a licensee controlled

document which is maintained under the

requirements of 10CFR50.59. Since any

future changes to this license controlled

document will be evaluated per the

requirements of 10CFR50.59, any possible

reduction (significant or insignificant) in a

margin of safety will be addressed.

Thus, the license amendment request does

not involve a significant reduction in the

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360, and the Waterford

Library, ATTN: Vince Juliano, 49 Rope

Ferry Road, Waterford, CT 06385

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: March

31, 1997

Description of amendment request:

The proposed amendment would

modify Technical Specification

Surveillance Requirement 4.7.1.2.1.b

which requires the testing of the

auxiliary feedwater motor-driven and

turbine-driven pumps on recirculation

flow at least once per 92 days. The

proposed amendment would also makes

changes to the appropriate Bases

section.

Basis for proposed no significant

Hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

NNECO has reviewed the proposed

changes in accordance with 10CFR 50.92 and

has concluded that the changes do 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:
Federal Register / Vol. 62, No. 78 / Wednesday, April 23, 1997 / Notices 19833

1. Involve a significant increase in the

probability or consequence of an accident

previously evaluated.

The proposed changes to Technical

Specification Surveillance 4.7.1.2.1.b to

increase the required test parameter for the

motor driven pumps from 1460 psid to 1468

psid and replacing the current parameters for

the motor driven and turbine driven pumps

from differential pressure measured in psid

[pounds per square inch differential] to total

head measured in feet are consistent with

equipment design criteria and does not

significantly increase the probability of an

accident previously evaluated.

The proposed changes to increase the

required test parameter for the motor driven

pumps from 1460 psid to 1468 psid and

replacing the current parameters for the

motor driven and turbine driven pumps from

differential pressure measured in psid to total

head measured in feet provides the necessary

assurance that the pumps will function as

required in accident analyses and does not

significantly increase the consequence of an

accident previously evaluated.

The moving of the reference to

Specification 4.0.5 in order to clarify that it

applies to the testing of the motor driven and

turbine driven pumps and the modifications

to the bases section are administrative and do

not involve a significant increase in the

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 Surveillance 4.7.1.2.1.b to

increase the required test parameter for the

motor driven pumps from 1460 psid to 1468

psid and replacing the current parameters for

the motor driven and turbine driven pumps

from differential pressure measured in psid

to total head measured in feet does not

change the operation of the auxiliary

feedwater system or any of its components

during normal or accident evaluations.

The moving of the reference to

Specification 4.0.5 in order to clarify that it

applies to the testing of the motor driven and

turbine driven pumps and the modifications

to the bases section are administrative and do

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

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 change to Technical

Specification Surveillance 4.7.1.2.1.b to

increase the referenced total head of the

motor

driven auxiliary feedwater pumps during

surveillance testing provides an acceptable

margin between the required surveillance

and design pump performance to provide

assurance that the pumps will operate

consistent with system evaluations and does

not involve a significant reduction in a

margin of safety.

The change in the referenced units from

differential pressure measured in psid to total

head measured in feet for the motor driven

auxiliary and turbine driven auxiliary

feedwater pumps during surveillance testing

is to account for the effect of water density

on pump performance during each test and

does not involve a significant reduction in a

margin of safety.

The moving of the reference to

Specification 4.0.5 in order to clarify that it

applies to the testing of the motor driven and

turbine driven pumps and the modifications

to the bases section are administrative and 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: March

31, 1997

Description of amendment request:

The proposed amendment would

separate the required testing of motoroperated valve thermal overload

protection into two new 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 has reviewed the proposed change

in accordance with 10CFR50.92 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 10CFR50.92(c) are not satisfied.

The proposed change does not involve a SHC

because the change would not:

1. Involve a significant increase in the

probability or consequence of an accident

previously evaluated.

The proposed changes to the surveillance

testing of the motor-operated valve thermal

overload protection are consistent with

equipment design criteria and performing

surveillance testing does not significantly

increase the probability of an accident

previously evaluated. The proposed changes

to the surveillance testing provides the

necessary assurance that the motor operated

valve thermal overload protection will

function as required and does not involve a

significant increase in the 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 the surveillance

testing of the motor-operated valve thermal

overload protection does not change the

operation of any system or system

component during normal or accident

evaluations.

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 proposed changes to the surveillance

testing of the motor-operated valve thermal

overload protection are administrative in that

the changes to the surveillance only clarify

that following maintenance on the motor

starter, a channel calibration is required only

on that valve. The surveillance continues to

require periodic representative sample

testing.

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-

0270NRC Deputy Director: Phillip F.

McKee
19834 Federal Register / Vol. 62, No. 78 / Wednesday, April 23, 1997 / Notices

Public Service Electric & Gas Company,

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

Nuclear Generating Station, Unit Nos. 1

and 2, Salem County, New Jersey

Date of amendment request: March 4,

1997

Description of amendment request:

The amendments would modify the

Emergency Core Cooling System (ECCS)

surveillance test acceptance criteria in

Technical Specification 3/4.5.2 for the

Centrifugal Charging (CH) and the

Safety Injection (SI) pumps. The

changes to the specified flow values

would account for system alignments

that effect the suction pressure to the

pumps. In the recirculation mode,

increased flow occurs when the CH and

SI pumps take suction from the

discharge of the Residual Heat Removal

pumps.

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 evaluations performed by

Westinghouse determined that, with the

proposed changes, the subject pumps remain

operable and the safety analyses criteria

remain valid.

Previous conclusions under LCR [License

Change Request] 91-03 evaluating the

consequences of the LOCA [loss-of-coolantaccident] considered in the Salem Units 1 &

2 licensing basis remain unchanged. With

respect to the LOCA, the Peak Cladding

Temperature (PCT) continues to conform to

the 10CFR50.46 guidelines of less than

2200*F. Evaluation of LOCA mass and

energy releases previously found acceptable

remain valid. Decreasing the acceptance

window to accommodate the potential of an

increase to pump runout flow, assures that

the current limits on pump runout flows

continue to be met. This change ensures

pump integrity is maintained during the

accident. The reduction of the flow by

throttling valves to compensate for the

potential suction boost remains within the

current analyses and therefore more

conservative values are being proposed.

Additionally, the proposed change balances

the pump flows more appropriately by

differentiating between the hot and cold leg

alignments. Flow to the reactor core is

unaffected by the very slight reduction in the

upper flow limits. Since the design

limitations continue to be met and the

integrity of the reactor coolant system

pressure boundary is not challenged, offsite

dose assumptions and calculations remain

valid. Further, the ECCS is post-accident

mitigation system and probability of an

accident is not increased by this proposed

change. Lastly, the correction of double use

of the word ‘‘the’’ in Salem Unit 1 Technical

Specification section 4.5.2.h.1.a is of editorial

nature.

Based on the above information, the

proposed changes do not increase the risk 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. No new single failures are

initiated. The proposed changes will

therefore not create the possibility of a new

or different kind of accident from any

accident previously evaluated. The proposed

change addresses suction boost by changing

the Technical Specification surveillance

acceptance criteria. The typographical

correction is of editorial nature.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The evaluation of LOCA accident analysis

previously performed by Westinghouse

continues to be met and verifies that, with

the proposed changes to the TS, plant

operations will be maintained within the

bounds of safe, analyzed conditions as

defined in the UFSAR [Updated Final Safety

Analysis Report] and that conclusions

presented in the UFSAR remain valid. The

peak cladding temperatures (PTC) remains

unchanged as no effective differences in the

operating parameters have occurred. The

typographical correction is of editorial

nature. The proposed changes will therefore

not reduce 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: 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

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: March

7, 1997

Description of amendments request:

The proposed amendments would allow

operability testing for the containment

isolation valves listed in Table 3.6-1 of

the Technical Specifications during a

defueled status. 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. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.]

The proposed changes do not significantly

increase the probability or consequences of

an accident previously evaluated in the

FSAR [Final Safety Analysis Report]. The

proposed changes have no impact on the

probability of an accident. The containment

isolation valves will continue to require

operability testing. Allowing the testing to be

performed when the unit is in a defueled

status will have no impact on any accidents

previously evaluated. The net effect of these

changes is not significant and, as a result,

does not involve a significant increase in the

consequences of an accident previously

evaluated.

[2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.]

The proposed changes to the Technical

Specifications do not increase the possibility

of a new or different kind of accident than

any accident already evaluated in the FSAR.

No new limiting single failure or accident

scenario has been created or identified due

to the proposed changes. Safety-related

systems will continue to perform as

designed. The proposed changes do not

create the possibility of a new or different

kind of accident from any previously

evaluated.

[3. Involve a significant reduction in a

margin of safety.]

The proposed changes do not involve a

significant reduction in the margin of safety.

Although, as a result of these proposed

changes, the containment isolation valves

could be tested for operability while the unit

is in a defueled state, there is no impact in

the accident analyses. These proposed

changes are technically consistent with the

requirements of NUREG-1431, Revision 1

which has already received the requisite

review and approval of the NRC staff. Thus

the proposed changes do not involve a

significant reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Houston-Love Memorial

Library, 212 W. Burdeshaw Street, Post

Office Box 1369, Dothan, Alabama

36302

Attorney for licensee: M. Stanford

Blanton, Esq., Balch and Bingham, Post

Office Box 306, 1710 Sixth Avenue

North, Birmingham, Alabama 35201

NRC Project Director: Herbert N.

Berkow
Federal Register / Vol. 62, No. 78 / Wednesday, April 23, 1997 / Notices 19835

Tennessee Valley Authority, Docket

Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of amendment request: August

22, 1996, as supplemented on March 28,

1997 (TS 96-02)

Description of amendment request:

The proposed changes would revise

Section 3.6.5 of the Sequoyah Technical

Specifications (TS) and associated Bases

to lower the minimum TS ice basket

weight of 1,155 pounds to 1,071

pounds. This would reduce the overall

weight of ice required in the ice

condenser from 2,245,320 pounds to

2,082,024 pounds. The TVA license

amendment request also proposed to

extend the chemical analysis

surveillance interval for the ice

condenser ice bed from 12 months to 18

months based on the provisions of

Generic Letter 93-05.

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.

TVA proposes to modify the SQN Unit 1

and Unit 2 TSs [Technical Specifications] to

revise Surveillance Requirement (SR)

4.6.5.1.d to lower SQN’s minimum TS basket

weight from 1,155 pounds (lbs) to 1,071 lbs,

thus lowering the overall ice condenser

weight from 2,245,320 lbs to 2,082,024 lbs.

The ice condenser system is provided to

absorb thermal energy release following a

loss-of-coolant accident (LOCA) or high

energy line break (HELB) and to limit the

peak pressure inside containment. The

current containment analysis for SQN is

based on a minimum of 993 lbs of ice per

basket evenly distributed throughout the ice

condenser at the end of an 18-month

refueling cycle. The revised containment

analysis shows that for the predicted

sublimation rate of 15 percent for 18 months,

an average basket weight of 922 lbs at the end

of the 18-month period would ensure

containment design pressure is not exceeded.

Based on TVA’s evaluation and the revised

containment analysis, TVA considers the

reduction of ice weight to be acceptable for

satisfying the safety function of the ice

condenser for an 18-month ice weighing

interval. Therefore, the proposed change does

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

TVA is also proposing to extend the

surveillance interval as it pertains to the ice

bed chemical analysis. Based on test results,

both at SQN and the industry, the average

boron concentration and pH changes are

minimal; therefore, this change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Elimination of the temperature at which

the pH of the ice bed is determined is an

administrative change. Future testing will be

accomplished in accordance with American

Society for Testing and Materials Standards

recommendations. Therefore, this change

cannot increase the probability of an accident

and the consequences of an accident will not

increase.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

TVA’s request to lower the TS limit for ice

weight at the start of the surveillance interval

will not result in a new or different kind of

accident from that previously analyzed in

SQN’s Final Safety Analysis

Report. SQN’s ice condenser serves to limit

the peak pressure inside containment

following a LOCA. TVA has evaluated the

revised containment pressure analysis for

SQN (Enclosure 4, Westinghouse WCAP12455, Revision 1) and determined that

sufficient ice would be present at all times

to keep the peak containment pressure below

SQN’s containment design pressure of 12

pounds per square inch gage (psig).

Therefore, this change would not result in a

new or different kind of accident from any

previously analyzed.

The proposed reduced testing frequency of

the chemical composition of the ice bed does

not change the manner in which the plant is

operated. Additionally, the ice condenser is

a passive system that reacts to an accident,

but does not support plant operation on a

daily basis. The reduced testing frequency of

the ice bed chemical composition does not

generate any new accident precursors;

therefore, the possibility of a new or different

kind of accident from any previously

analyzed is not created.

Elimination of the temperature at which

the pH of the ice bed is determined is an

administrative change. This change cannot

create the possibility of a new or different

kind of accident.

3. Involve a significant reduction in a

margin of safety.

The ice condenser system is provided to

absorb thermal energy release following a

LOCA and to limit the peak pressure inside

containment. The current ice condenser

analysis for SQN is based on a minimum of

993 lbs of ice per basket. The revised

containment analysis changes the minimum

ice weight assumed in the analysis to 922 lbs

per basket.

The revised containment analysis shows

that using an average basket weight of 1,071

lbs and a sublimation allowance of 15

percent, all bays would have an average

basket weight of 922 lbs at the end of the 18-

month surveillance interval. The revised

analysis utilizes new mass and energy

releases (refer to Westinghouse WCAP-10325-

P-A), which substantially delays ice-bed

meltout and limits the initial containment

peak pressure to approximately 7.15 psig

during the blowdown phase. The ice-bed

meltout delay allows the second containment

pressure peak, which is driven mainly by the

decay heat, to be limited to approximately

11.45 psig, which is below the containment

design pressure of 12 psig.

Based on TVA’s evaluation and the revised

containment analysis, TVA considers the

reduction of the average basket weight to be

acceptable for satisfying the safety function

of the ice condenser for the current 18-month

interval. Therefore, the proposed change does

not involve a significant reduction in the

margin of safety.

The proposal to extend the surveillance

from 12 to 18 months does not change the

boron concentration or pH requirements.

Experience at Duke Power Company, as

stated in NUREG-1366, indicates that these

parameters do not change appreciably when

verified every 9 months. SQN has a similar

experience with a 12-month interval. Since

the boron concentration and the post-LOCA

pH requirements remain essentially the same,

there is no reduction in the margin of safety.

Elimination of the temperature at which

the pH of the ice bed is determined is an

administrative change. Future testing will be

accomplished in accordance with ASTM

recommendations. The difference between

the pH values determined at the current TS

specified temperature and the temperature

currently recommended by the ASTM

standards is insignificant. Therefore, there is

no reduction in the margin of safety.

The NRC has reviewed the licensee’s

analysis and, based on thisreview, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H,

Knoxville, Tennessee 37902

NRC Project Director: Frederick J.

Hebdon

Virginia Electric and Power Company,

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

Power Station, Unit Nos. 1 and 2, Surry

County, Virginia

Date of amendment request:

September 12, 1996

Description of amendment request:

The proposed change to the Technical

Specifications is administrative in

nature in that it would add the NRC

standard fire protection license

condition to each unit’s Operating

License and relocate the fire protection

requirements from the Technical

Specifications to the Updated Final

Safety Analysis Report (UFSAR).

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:
19836 Federal Register / Vol. 62, No. 78 / Wednesday, April 23, 1997 / Notices

Specifically, operation of Surry Power

Station with the proposed amendment will

not:

1. Involve a significant increase in either

the probability of occurrence or

consequences of any accident or equipment

malfunction scenario that is important to

safety and which has been previously

evaluated in the UFSAR. The requirements of

the Fire Protection Program have not been

changed by theproposed amendment.

Relocation of the Fire Protection Program

requirements into the UFSAR and station

procedures does not decrease any portion of

the program. The same fire protection

requirements exist as before the change.

2. Create the possibility of a new or

different type of accident than those

previously evaluated in the safety analysis

report. The requirements of the Fire

Protection Program have not been changed by

the proposed amendment. This is an

administrative change to relocate the Fire

Protection Program requirements from the

Technical Specifications to the UFSAR and

station procedures. Consequently, the

possibility of a new or different kind of

accident from any accident previously

evaluated has not been created.

3. Involve a significant reduction in a

margin of safety. Implementation of the Fire

Protection Program requirements is assured

by the UFSAR and station procedures. Since

the rogram is being retained intact, there is

no reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 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: Swem Library, College of

William and Mary, Williamsburg,

Virginia 23185.

Attorney for licensee: Michael W.

Maupin, Esq., Hunton and Williams,

Riverfront Plaza, East Tower, 951 E.

Byrd Street, Richmond, Virginia 23219

NRC Project Director: Mark Reinhart,

Acting

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: January

16, 1997

Description of amendment request:

The proposed amendments (Point Beach

Nuclear Plant (PBNP) Technical

Specifications (TS) Change Request

(TSCR) 191) would revise the minimum

boron concentration required in the

refueling water storage tank(s)(RWST),

boric acid storage tanks (BAST), and

safety injection (SI) accumulators during

normal operation; the minimum boron

concentration of primary coolant during

refueling conditions; and the minimum

boron concentration in the reactor when

positive reactivity could be added and/

or boron dilution could occur and

containment integrity is not intact.

These changes are necessary to

accommodate the planned extension of

the operating cycle from 12 months to

18 months. The licensee proposes to

change TS 15.3.2, ‘‘Chemical and

Volume Control System,’’ TS 15.3.3,

‘‘Safety Injection and Residual Heat

Removal Systems,’’ TS 15.3.6,

‘‘Containment System,’’ TS 15.3.8,

‘‘Refueling,’’ and associated Bases.

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 this facility under the

proposed Technical Specifications will not

create a significant increase in the probability

or consequences of an accident previously

evaluated.

The probabilities of accidents previously

evaluated are based on the probability of

initiating events for these accidents.

Initiating events for accidents previously

evaluated are described in the PBNP FSAR

[final safety analysis report].

In effect, the proposed changes will result

in: (1) higher boron concentrations of

primary coolant during refueling, and (2)

higher boron inventories in the RWSTs,

BASTs, and SI accumulators. These changes

do not require hardware changes or changes

to the operation of accident-mitigating

equipment. These changes relate to the

performance capability of particular accident

mitigation systems; equipment that is not

postulated to cause accidents. Therefore,

these proposed changes do not cause an

increase in the probabilities of any accidents

previously evaluated.

The consequences of accidents previously

evaluated in the PBNP FSAR are determined

by the results of analyses that are based on

initial conditions of the plant, the type of

accident, transient response of the plant, and

the operation and failure of equipment and

systems.

In effect, the proposed changes will result

in: (1) higher boron concentrations of

primary coolant during refueling, and (2)

higher boron inventories in the RWSTs,

BASTs, and SI accumulators. These

increased boron concentrations do not

increase the probability that engineered

safety features equipment will fail, nor do

these changes affect the capability of this

equipment to operate as required for the

accidents previously evaluated in the PBNP

FSAR. These changes do not require

hardware changes or changes to the operation

of accident-mitigating equipment.

The consequential effects of a lower

containment spray pH will not affect the

capability of the containment spray to

remove elemental iodine during design basis

LOCA [loss-of-coolant accident] accidents.

Also, the consequential reduction in

containment sump water pH will not affect

the fluid’s capability to retain elemental

iodine, nor will it adversely increase the

potential corrosion rates for materials inside

containment if the sump water is sprayed

into containment during the recirculation

phase of a LOCA.

Another consequence of injecting a higher

concentration boric acid solution into the

core during a LOCA may be an abbreviated

onset to boron precipitation in the postLOCA core. An incremental change in the

boron injection concentration would not

have significant effect on the postulated

onset, but each core reload safety evaluation

will continue to verify that the existing

emergency operating procedures

accommodate the potential for boron

precipitation.

Therefore, this proposed license

amendment does not affect the consequences

of any accident previously evaluated in the

PBNP FSAR, because the factors that are used

to determine the consequences of accidents

are not changed.

2. Operation of this facility under the

proposed Technical Specifications change

will not create the possibility of a new or

different kind of accident from any

previously evaluated.

New or different kinds of accidents can

only be created by new or different accident

initiators or sequences. New and different

types of accidents (different from those that

were originally analyzed for Point Beach)

have been evaluated and incorporated into

the licensing basis for PBNP. Examples of

different accidents that have been

incorporated into the PBNP licensing basis

include anticipated transients without scram

and station blackout.

The changes proposed by this TSCR do not

create any new or different accident initiators

or sequences because these changes to

minimum boron concentrations will not

cause failures of equipment or accident

sequences different than the accidents

previously analyzed. No new equipment

interfaces are created, and no new materials

or fluids are introduced. The incremental

increase in boron concentrations will not

create a failure mechanism not previously

known and evaluated. Therefore, these

proposed TS changes do not create the

possibility of an accident of a different type

than any previously evaluated in the PBNP

FSAR.

3. Operation of this facility under the

proposed Technical Specifications change

will not create a significant reduction in a

margin of safety.

The margins of safety for Point Beach are

based on the design and operation of the

reactor and containment and the safety

systems that provide their protection. Plant

safety margins are established through

Limiting Conditions for Operation, Limiting

Safety System Settings and Safety Limits

specified in the Technical Specifications.

The proposed Technical Specification

changes to refueling water storage tank

(RWST), SI accumulator, and BAST boron

inventory requirements have all been

evaluated to preserve the shutdown

capability described in the associated bases

(boration from just critical, hot zero or full

power, peak xenon with control rods at the
Federal Register / Vol. 62, No. 78 / Wednesday, April 23, 1997 / Notices 19837

insertion limit, to xenon-free cold shutdown

with the highest worth control rod assembly

fully withdrawn). Similarly, the proposed TS

change to the minimum boron concentration

of the primary coolant system for refueling

operations have been evaluated to preserve

the subcriticality margin described in the

associated TS bases (i.e., 5% [delta] k/k in

the cold condition with all rods inserted).

Because there are no changes to any of

these margins, the proposed license

amendment does not involve a reduction in

any 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: 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: January

21, 1997

Description of amendment request:

The proposed amendments (Point Beach

Nuclear Plant (PBNP) Technical

Specifications (TS) Change Request 195)

would revise TS Section 15.6.11,

‘‘Radiation Protection Program,’’ to

update all references to 10 CFR Part 20,

‘‘Standards for Protection Against

Radiation,’’ to restore consistency

between 10 CFR Part 20 regulations and

the PBNP TS.

Basis for proposed no significant

Hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration which is presented below:

1. Operation of the Point Beach Nuclear

Plant in accordance with the proposed

amendments will not create a significant

increase in the probability or consequences

of an accident previously evaluated.

The proposed amendments are

administrative in nature, providing

consistency between the Point Beach licenses

and Commission regulations. The

amendments do not affect the operation or

maintenance of any PBNP structure[,] system

or component. In addition, the regulations

and proposed changes provide more

conservative determinations of high radiation

areas, thereby potentially resulting in lower

personnel radiation exposures during normal

operation and post accident. The

consequences of an accident related to

personnel radiation exposures may be

reduced.

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 are

administrative only and do not affect the

operation or maintenance of any structure[,]

system or component at Point Beach Nuclear

Plant. No new systems or components are

introduced. Therefore, no new accident

initiators or sequences result from any

previously evaluated.

3. Operation of the Point Beach Nuclear

Plant in accordance with the proposed

amendments will not create a significant

reduction in a margin of safety.

The proposed amendments are

administrative and reflect regulatory

requirements that are more conservative than

those presently reflected in the PBNP

Technical Specifications. These more

conservative requirements result in more

conservative designation of high radiation

areas thereby providing additional margins of

safety related to the control of radiation

exposures to personnel. No structure[,]

system or component at PBNP at PBNP is

changed[,] thereby maintaining the margins

of safety for the operation of the Point Beach

Nuclear Plant.

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: January

24, 1997

Description of amendment request:

The proposed amendments (Point Beach

Nuclear Plant (PBNP) Technical

Specifications (TS) Change Request

(TSCR) 193) would revise TS 15.5.4,

‘‘Fuel Storage,’’ to increase fuel

assembly enrichment limits to 5.0 w/o

U-235 while maintaining Keff in the

storage pools (spent fuel pool and new

fuel storage racks) less than 0.95.

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 this facility under the

proposed Technical Specifications will not

create a significant increase in the probability

or consequences of an accident previously

evaluated.

The proposed changes do not involve a

change to structures, systems, or components

which would affect the probability or

consequences of an accident previously

evaluated in the PBNP Final Safety Analysis

Report (FSAR). The only relevant concern

with respect to increasing enrichment limits

in the spent fuel pool and new fuel storage

racks is one of criticality. The proposed

changes use the same criticality limit used in

the current Technical Specifications.

Therefore, margin to safe operation of Units

1 and 2 is maintained. The probability and

consequences of an accident previously

evaluated are dependent on this criticality

limit. Because the limit will not change, the

probability and consequences of those

accidents previously evaluated will not

change.

2. Operation of this facility under the

proposed Technical Specifications change

will not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed changes do not involve a

change to plant design. The proposed

increase in spent fuel pool and new fuel

storage racks fuel assembly enrichment limits

maintains the margin to safe operation of

Units 1 and 2 because the criticality limit for

the spent fuel pool and new fuel storage

racks will not change. These changes do not

affect any of the parameters or conditions

that contribute to the initiation of any

accidents. Because the criticality limit

remains the same, these changes have no

effect on plant operation, design, or initiation

of any accidents. Therefore, the proposed

changes will not create the possibility of a

new or different kind of accident from any

accident previously evaluated.

3. Operation of this facility under the

proposed Technical Specifications change

will not create a significant reduction in a

margin of safety.

The proposed changes maintain the margin

to safe operation of Units 1 and 2. The

margin of safety is based on the criticality

limit of the spent fuel pool and the new fuel

storage racks. Because this limit will not

change, the margin of safety will not be

affected. Therefore, the proposed changes

will not create 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: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin

54241
19838 Federal Register / Vol. 62, No. 78 / Wednesday, April 23, 1997 / Notices

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: February

12, 1997, as supplemented on March 11,

1997

Description of amendment request:

The proposed amendments (Point Beach

Nuclear Plant (PBNP) Technical

Specifications (TS) Change Request 196)

would relocate turbine overspeed

protection specifications, limiting

conditions for operation, surveillance

requirements, and associated bases from

TS Section 15.3.4, ‘‘Steam and Power

Conversion System,’’ and Section

15.4.1, ‘‘Operational Safety Review,’’ to

the Final Safety Analysis Report (FSAR)

in accordance with Generic Letter 95-10,

‘‘Relocation of Selected Technical

Specifications Requirements Related to

Instrumentation.’’

Basis for proposed no significant

Hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration which is presented below:

1. Operation of 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 amendments

administratively relocate turbine overspeed

protection Specifications to the Point Beach

Final Safety Analysis Report (FSAR). The

Specifications will be transferred verbatim,

except for the turbine load limit with the

crossover steam dump system inoperable,

which has already been evaluated under 10

CFR 50.59 and will be conservatively

reduced. In addition, the regulatory

requirements of 10 CFR 50.55a, ‘‘Codes and

Standards, ’’ will still apply to the relocated

Specifications. Therefore, operation of Point

Beach Nuclear Plant in accordance with the

proposed amendments cannot create an

increase in the probability or consequences

of an accident previously evaluated.

2. Operation of 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

administratively relocate Specifications to

the FSAR and in one case result in a more

conservative operating limit. Therefore,

operation of Point Beach Nuclear Plant in

accordance with the proposed amendments

cannot create a new or different kind of

accident from any accident previously

evaluated.

3. Operation of Point Beach Nuclear Plant

in accordance with the proposed

amendments will not create a significant

reduction in a margin of safety.

The proposed changes are administrative

in nature. There is no physical change to the

facility, its systems, or its operation, except

for the conservative reduction of the turbine

load limit with the crossover steam dump

system inoperable which has already been

justified via 10 CFR 50.59. Therefore,

operation of PBNP in accordance with the

proposed amendments cannot 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: 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: February

17, 1997; supersedes March 24, 1995, as

supplemented by letter dated August 16,

1995, amendment request.

Description of amendment request:

This amendment request proposes to

revise Technical Specification 1.7,

‘‘Containment Integrity,’’ Technical

Specification 3/4.6.1, ‘‘Containment

Integrity,’’ and Technical Specification

3/4.6.3, ‘‘Containment Isolation Valves.’’

These proposed changes would relocate

Technical Specification Table 3.6-1,

‘‘Containment Isolation Valves,’’ to the

Wolf Creek Generating Station (WCGS)

procedures. This proposed change is in

accordance with the guidance provided

in Generic Letter 91-08, ‘‘Removal of

Component Lists from Technical

Specifications,’’ dated May 6, 1991. In

addition, this request proposes that the

August 16, 1996, supplemental

submittal that provided an additional

footnote allowing an increased outage

time for certain component cooling

water system valves be withdrawn. The

determination that the additional

footnote is not required supersedes the

staff’s proposed no significant hazards

consideration determination evaluation

for the requested changes that was

published on September 27, 1995 (60 FR

49949).

Basis for proposed no significant

Hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes simplify the

technical specifications, meet the regulatory

requirements for control of containment

isolation, and are consistent with the

guidelines of GL 91-08. The procedural

details of Technical Specification Table 3.6-

1 have not been changed, but only relocated

to a different controlling document. The

proposed changes are administrative in

nature, should result in improved

administrative practices, and do not affect

plant operations.

The probability of occurrence of a

previously evaluated accident is not

increased because this change does not

introduce any new potential accident

initiating conditions. The consequences of an

accident previously evaluated is not

increased because the ability of containment

to restrict the release of any fission product

radioactivity to the environment will not be

degraded by this change.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes are administrative

in nature, do not result in physical

alterations or changes to the operation of the

plant, and cause no change in the method by

which any safety-related system performs its

function. Therefore, this proposed change

will not create the possibility of a new or

different kind of accident from any

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The administrative change to relocate

Technical Specification Table 3.6-1 to

appropriate plant procedures does not alter

the basic regulatory requirements for

containment isolation and will not adversely

affect containment isolation capability for

Coordinator credible accident scenarios.

Adequate control of the content of the table

is assured by existing plant procedures.

The proposed relocation of Technical

Specification Table 3.6-1 does not alter

current technical specification requirements

for containment isolation valve operability.

The LCO and Surveillance Requirements

would be retained in the revised technical

specifications. Therefore, the proposed

change will not affect the meaning,

application, and function of the current

technical specification requirements for the

valves in Table 3.6-1.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.
Federal Register / Vol. 62, No. 78 / Wednesday, April 23, 1997 / Notices 19839

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

Wolf Creek Nuclear Operating

Corporation, Docket No. 50-482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request: March

18, 1997

Description of amendment request:

This license amendment request revises

Technical Specification Surveillance

Requirement 4.5.2.c to clarify when a

containment entry visual inspection is

required. This proposed change to

reduce the visual inspection

requirement to at least once daily is in

accordance with the guidance provided

in Generic Letter 93-05, ‘‘Line-Item

Technical Specifications Improvements

to Reduce Surveillance Requirements

for Testing During Power 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.

Implementing the proposed change could

potentially increase the chances of loose

debris (trash, rags, clothing, etc.) being left in

containment for some period of time greater

than would be allowed under current

surveillance requirements. However, the

proposed change also clarifies that the visual

inspection must be performed at least once

daily. Therefore, the period of time that

debris could be left uncontrolled inside

containment would still be less than 24

hours. Based on work controls placed on

material entry/exit into containment and

personnel training on housekeeping controls

inside containment, and the results of past

surveillances, it is unlikely that a significant

amount of debris would be left uncontrolled

inside containment for this period of time.

Also, based on containment sump design,

relatively small amounts of debris would not

be sufficient to cause a significant amount of

blockage of the sump screens.

The probability of occurrence of a

previously evaluated accident is not

increased because the reduced frequency of

the visual inspection does not cause a

significant impact on the possibility of

containment sump screen blockage.

Therefore containment sump operability is

not affected by the proposed change. In

addition, the proposed change will not result

in any changes to the design or operation of

any plant systems or components.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed change decreases the

frequency of performing a visual inspection

for loose debris in containment, but does not

result in a change to the design or operation

of any plant system or component. The

purpose of the inspection is to ensure that

there is no loose debris, left in containment

following a containment entry, that could

potentially block the containment sump

screens during LOCA conditions. Delaying

this inspection until the last containment

entry each day will not result in a significant

amount of debris being left in containment,

based on housekeeping practices controlling

the entry/removal of trash and material into/

from containment; training of employees to

increase awareness of material control in

radiologically-controlled areas; and retaining

the requirement to perform a visual

inspection at least once per day when

containment entries are made (during periods

when containment integrity is established),

thereby ensuring that trash and debris can be

identified and removed on a daily basis (on

days containment entries are made).

Based on the above, this proposed change

will not create the possibility of a new or

different kind of accident from any

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The purpose of performing a visual

inspection of areas affected by a containment

entry is to ensure any debris or trash

generated by the activity during the

containment entry is identified and removed

from containment. This ensures that no trash

or debris is left in containment that could be

transported to and block the containment

sump screens during LOCA conditions.

Based on current material control and

housekeeping practices imposed on

containment entry/exit, and past inspection

results, reducing the surveillance

requirement to a once per day basis, on days

containment entries are made, would not

result in a significant amount of trash or

debris being left in containment following

completion of the entry, and any such

material would be identified and removed

prior to the end of the day. The amount of

trash or debris that could be left in

containment for a 24 hour period would be

significantly less than the amount that would

be required to cause sump screen blockage

sufficient to affect sump performance.

Therefore, the proposed change will not

result in a significant reduction in the margin

of safety of any plant system or equipment.

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

Wolf Creek Nuclear Operating

Corporation, Docket No. 50-482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request: March

18, 1997

Description of amendment request:

This license amendment request revises

Technical Specification Section 5.3.1,

Fuel Assemblies, to allow the use of an

alternate zirconium based fuel cladding

material, ZIRLO. Wolf Creek Nuclear

Operating Corporation (WCNOC) is

planning to insert Westinghouse fuel

assemblies containing ZIRLO fuel rod

cladding during the ninth refueling

outage, which is currently scheduled to

begin in late September 1997.

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 methodologies used in the accident

analysis remain unchanged. The proposed

changes do not change or alter the design

assumptions for the systems or components

used to mitigate the consequences of an

accident. Use of ZIRLO fuel cladding does

not adversely affect fuel performance or

impact nuclear design methodology.

Therefore accident analyses are not

impacted.

The operating limits will not be changed

and the analysis methods to demonstrate

operation within the limits will remain in

accordance with NRC approved

methodologies. Other than the changes to the

fuel assemblies, there are no physical

changes to the plant associated with this

technical specification change. A safety

analysis will continue to be performed for

each cycle to demonstrate compliance with

all fuel safety design basis.

VANTAGE 5H with IFMs fuel assemblies

with ZIRLO clad fuel rods meet the same fuel

assembly and fuel rod design bases as other

VANTAGE 5H with IFMs fuel assemblies. In

addition, the 10 CFR 50.46 criteria are

applied to the ZIRLO clad rods. The use of

these fuel assemblies will not result in a

change to the reload design and safety

analysis limits. The clad material is similar
19840 Federal Register / Vol. 62, No. 78 / Wednesday, April 23, 1997 / Notices

in chemical composition and has similar

physical and mechanical properties as

Zircaloy-4. Thus, the 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 use of an

assembly containing a combination of

Zircaloy-4 and ZIRLO clad fuel rods. Since

the dose predictions in the safety analyses

are not sensitive to fuel rod cladding

material, the radiological consequences of

accidents previously evaluated in the safety

analysis remain valid.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident or

malfunction of equipment important to safety

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

VANTAGE 5H with IFMs fuel assemblies

with ZIRLO clad fuel rods satisfy the same

design bases as those used for other

VANTAGE 5H with IFMs fuel assemblies. All

design and performance criteria continue to

be met and no new failure mechanisms have

been identified. Since the original design

criteria are met, the ZIRLO clad fuel rods will

not be an initiator for any new accident or

malfunction of equipment important to

safety. The ZIRLO cladding material offers

improved corrosion resistance and structural

integrity.

The proposed changes do not affect the

design or operation of any system or

component in the plant. The safety functions

of the related structures, systems or

components are not changed in any manner,

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 change to the

facility or operating procedures, and the

safety functions and reliability of structures,

systems and components are not affected, the

proposed changes do not create the

possibility of a new or different kind of

accident or malfunction of equipment

important to safety from any accident or

malfunction of equipment important to safety

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

Use of ZIRLO cladding material does not

change the VANTAGE 5H with IFMs reload

design and safety limits. The use of these fuel

assemblies will take into consideration the

normal core operating conditions allowed in

the Technical Specifications. For each cycle

reload core, the fuel assemblies will be

evaluated using NRC approved reload design

methods, including consideration of the core

physics analysis peaking factors and core

average linear heat rate effects.

The use of Zircaloy-4, ZIRLO or stainless

steel filler rods in fuel assemblies will not

involve a significant reduction in the margin

of safety because analyses using NRC

approved methodologies will be performed

for each configuration to demonstrate

continued operation within the limits that

assure acceptable plant response to accidents

and transients. These analyses will be

performed using NRC approved methods that

have been approved for application to the

fuel configuration.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

locations: Emporia State University,

William Allen White Library, 1200

Commercial Street, Emporia, Kansas

66801 and Washburn University School

of Law Library, Topeka, Kansas 66621

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts and Trowbridge,

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

20037

NRC Project Director: William H.

Bateman

Previously Published Notices Of

Consideration Of Issuance Of

Amendments To Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

And Opportunity For A Hearing

The following notices were previously

published as separate individual

notices. The notice content was the

same as above. They were published as

individual notices either because time

did not allow the Commission to wait

for this biweekly notice or because the

action involved exigent circumstances.

They are repeated here because the

biweekly notice lists all amendments

issued or proposed to be issued

involving no significant hazards

consideration.

For details, see the individual notice

in the Federal Register on the day and

page cited. This notice does not extend

the notice period of the original notice.

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 amendments request: March

27, 1997

Description of amendments request:

The proposed amendments would

revise the Technical Specifications for

the Brunswick Steam Electric Plant

Units 1 and 2 to eliminate certain

instrumentation response time testing

requirements in accordance with NRCapproved BWR Owners Group Topical

Report NEDO-32291-A, ‘‘System

Analysis for the Elimination of Selected

Response Time Testing

Requirements.’’Date of publication of

individual notice in Federal Register:

April 1, 1997 (62 FR 15542)

Expiration date of individual notice:

May 1, 1997

Local Public Document Room

location: University of North Carolina at

Wilmington, William Madison Randall

Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297

Southern Nuclear Operating Company,

Inc., Docket No. 50-348, Joseph M.

Farley Nuclear Plant, Unit No. 1,

Houston County, Alabama

Date of amendment request: March

25, 1997

Description of amendment request:

The proposed amendment would

modify Technical Specification 3/4.4.9,

‘‘Specific Activity,’’ and associated

Bases to reduce the limit associated

with dose equivalent iodine-131. The

steady-state dose equivalent iodine-131

limit would be reduced by 40 percent

from .5 [micro]Curie/gram to .3

[micro]Curie/gram.

Date of publication of individual

notice in Federal Register: April 4, 1997

(62 FR 16201)

Expiration date of individual notice:

May 5, 1997

Local Public Document Room

location: Appling County Public

Library, 301 City Hall Drive, Baxley,

Georgia 31513

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
Federal Register / Vol. 62, No. 78 / Wednesday, April 23, 1997 / Notices 19841

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.

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 19, 1997, as supplemented

April 3, 1997.

Brief description of amendments: The

amendments would delete the 24/48

Volt direct current (Vdc), batteries,

battery chargers and distribution

systems from the Technical

Specifications (TSs) for Unit 3, by

adding a footnote to indicate that these

TSs are only applicable to Unit 2. All

safety-related loads associated with the

24/48 Vdc batteries for Unit 3 will be

relocated to other safety-related battery

systems which are in the TSs.

Date of issuance: April 10, 1997

Effective date: Immediately, to be

implemented within 30 days.

Amendment Nos.: 156 and 151

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

revised the Technical Specifications.

Date of initial notice in Federal

Register: March 5, 1997 (62 FR 10088).

The April 3, 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

April 10, 1997. No significant hazards

consideration comments received: No

Local Public Document Room

location: Morris Area Public Library

District, 604 Liberty Street, Morris,

Illinois 60450

Consolidated Edison Company of New

York, Docket No. 50-247, Indian

PointNuclear Generating Unit No. 2,

Westchester County, New York

Date of application for amendment:

August 14, 1996, as supplemented

September 13, 1996

Brief description of amendment: The

amendment revises Technical

Specification Sections 3.3 and 6.9.1.9;

and the basis of Section 3.3, 3.6 and

3.10. The changes incorporate the best

estimate approach into the licensing

basis for the Indian Point Unit No. 2

loss-of-coolant accident analysis.

Date of issuance: March 31, 1997

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 188

Facility Operating License No. DPR26: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 29, 1997 (62 FR 4344)

The September 13, 1996, supplemental

letter did not change the initial

proposed no significant hazards

consideration.The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

March 31, 1997.No significant hazards

consideration comments received: No

Local Public Document Room

location: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10610

Consolidated Edison Company of New

York, Docket No. 50-247, Indian

PointNuclear Generating Unit No. 2,

Westchester County, New York

Date of application for amendment:

February 14, 1997, as supplemented

March 12, 1997.

Brief description of amendment: The

amendment revises Technical

Specification Section 4.13-2 to allow a

one-time extension of the interval for

steam generator tube inspection.

Specifically, the date for

commencement of the steam generator

tube inspection is extended from April

14, 1997 to May 2, 1997.

Date of issuance: April 9, 1997

Effective date: As of the date of

issuance to be implemented by April 14,

1997.

Amendment No.: 189

Facility Operating License No. DPR26: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 4, 1997 (62 FR 9816)

The March 12, 1997, supplemental letter

provided clarifying information that did

not change the initial proposed no

significant hazards consideration. The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated April 9, 1997.No

significant hazards consideration

comments received: No

Local Public Document Room

location: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10610

Consumers Power Company, Docket

No. 50-155, Big Rock Point Plant,

Charlevoix County, Michigan

Date of application for amendment:

November 7, 1996

Brief description of amendment: The

amendment revised Technical

Specification 4.2.9, Service and

Instrument Air System, to add an

additional air compressor.

Date of issuance: April 2, 1997

Effective date: Effective the date of

issuance.

Amendment No.: 118

Facility Operating License No. DPR-6:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: December 18, 1996 (61 FR

66706) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

April 2, 1997.No significant hazards

consideration comments received: No.

Local Public Document Room

location: North Central Michigan

College, 1515 Howard Street, Petoskey,

Michigan 49770

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:

January 3, 1997, as supplemented by

letter dated March 20, 1997

Brief description of amendments: The

amendments revise Technical

Specification Tables 3.3-2, 3.3-4, 3.3-5,

4.3-2 and Bases Sections 3/4.3.1 and 3/

4.3.2 to eliminate the safety injection

signal on low steam line pressure.

Date of issuance: April 3, 1997

Effective date: For Unit 1, as of the

date of issuance to be implemented

before startup from the next refueling

outage; For Unit 2, as of the date of

issuance to be implemented before

startup from the current refueling outage

Amendment Nos.: 158 and 150

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

the Technical Specifications.

Date of initial notice in Federal

Register: January 29, 1997 (62 FR 4345)

The March 20, 1997, letter provided

clarifying information that did not

change the scope of the original January

3, 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 3,

1997.No significant hazards

consideration comments received: No
19842 Federal Register / Vol. 62, No. 78 / Wednesday, April 23, 1997 / Notices

Local Public Document Room

location: York County Library, 138 East

Black Street, Rock Hill, South Carolina

29730

Duke Power Company, Docket Nos. 50-

269, 50-270 and 50-287, Oconee

Nuclear Station, Units 1, 2, and 3,

Oconee County, South Carolina

Date of application for amendments:

February 5, 1997

Brief description of amendments: The

amendments reflect replacement of the

existing source and intermediate range

nuclear instrumentation with a new

source range and wide range nuclear

instrumentation system that provides

more channels and continuous coverage

from the Source Range to above the

Power Range.

Date of issuance: March 31, 1997

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment Nos.: 223, 223, 220

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

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 26, 1997 (62 FR

8796) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

March 31, 1997.No significant hazards

consideration comments received:

Local Public Document Room

location: Oconee County Library, 501

West South Broad Street, Walhalla,

South Carolina 29691

Entergy Operations, Inc., Docket No.

50-382, Waterford Steam Electric

Station, Unit 3, St. Charles Parish,

Louisiana

Date of amendment request: October

16, 1996

Brief description of amendment: The

amendment changes the Appendix A

Technical Specifications by revising

Table 4.3-1 to expand the applicability

for Core Protection Calculator (CPC)

operability and to allow the use of a

cycle independent shape annealing

matrix in the CPCs.

Date of issuance: April 11, 1997

Effective date: April 11, 1997, to be

implemented within 60 days

Amendment No.: 125

Facility Operating License No. NPF38: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 12, 1997 (62 FR

6575) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

April 11, 1997No significant hazards

consideration comments received: No.

Local Public Document Room

location: University of New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, LA 70122

Entergy Operations, Inc., Docket No.

50-382, Waterford Steam Electric

Station, Unit 3, St. Charles Parish,

Louisiana

Date of amendment request:

December 2, 1996 as supplemented by

letter dated February 4 and March 14,

1997

Brief description of amendment: The

amendment changes the Technical

Specifications to reflect the approval for

the licensee to use of the new

Containment Leakage Rate Testing

Program as required by 10 CFR Part 50

Appendix J, Option B for Waterford

Steam Electric Station, Unit 3.

Date of issuance: April 10, 1997

Effective date: April 10, 1997

Amendment No.: 124

Facility Operating License No. NPF38: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 15, 1997 (62 FR 2189)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated April 10, 1997.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: University of New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, LA 70122

Florida Power and Light Company, et

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

Unit No. 1, St. Lucie County, Florida

Date of application for amendment:

December 9, 1996

Brief description of amendment: This

amendment modifies technical

specifications for selected cycle-specific

reactor physics parameters to refer to

the St. Lucie Unit 1 Core Operating

Limits Report for limiting values.

Date of issuance: April 1, 1997

Date of issuance: April 1, 1997

Effective date: April 1, 1997

Amendment No.: 150

Facility Operating License No. NPF16: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 15, 1997 (62 FR 2189)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated April 1, 1997. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort

Pierce, Florida 34954-9003

Northeast Nuclear Energy Company, et

al., Docket No. 50-423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of applications for amendment:

June 20, 1995, as supplemented August

30, 1995, and January 17, 1996

Brief description of amendment: The

amendment relocates the applicable

requirements of Technical Specification

(TS) 3.6.3 for the main steam line

isolation valves (MSIVs) to TS 3.7.1.5,

‘‘Main Steam Line Isolation Valves.’’ In

addition, the Applicability section of TS

3.7.1.5 is revised to indicate that

Specification 3.7.1.5 is applicable in

Mode 1 and in Modes 2, 3, and 4, except

where all MSIVs are closed and

deactivated (i.e., in Modes 2, 3, and 4,

TS 3.7.1.5 is applicable only if the

MSIVs are open). Also, the Action

Statement for the Limiting Condition for

Operation 3.7.1.5 has been revised using

the guidance of the Improved Standard

Technical Specifications for

Westinghouse plants (NUREG-1431).

The amendment also deletes a license

requirement to submit responses to and

to implement requirements of Generic

Letter 83-28, because the requirement

has been completed. Generic Letter 83-

28 pertains to the Salem anticipated

transient without scram event. In

addition, the amendment incorporates

TS Bases submitted by Northeast

Nuclear Energy Company by letters

dated June 20, 1995, February 5, 1996,

and March 21 and 26, 1997. Since all

four Bases changes affect Section B 3/

4.7 of the TS, the NRC staff is using

them in a group to avoid errors in

revising the TS.

Date of issuance: April 10, 1997

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment No.: 136

Facility Operating License No. NPF49: Amendment revised the Technical

Specifications and License Condition.

Date of initial notice in Federal

Register: August 2, 1995 (61 FR 39445)

and February 28, 1996 (61 FR 7555)The

August 30, 1995, letter provided

clarifying information that did not

change the scope of the June 20, 1995,

application and the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 10,

1997.No significant hazards

consideration comments received: No.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,
Federal Register / Vol. 62, No. 78 / Wednesday, April 23, 1997 / Notices 19843

Norwich, Connecticut and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut 06385

Northeast Nuclear Energy Company, et

al., Docket Nos. 50-245, 50-336, and 50-

423, Millstone Nuclear Power Station,

Unit Nos. 1, 2, and 3, New London,

Connecticut

Date of application for amendments:

February 3, 1997

Brief description of amendments: The

amendments revise Section 6,

‘‘Administrative Controls,’’ of the

Millstone Unit Nos. 1, 2, and 3

Technical Specifications to reflect

organizational changes that have been

implemented in the Nuclear Division.

Date of issuance: April 10, 1997

Effective date: As of the date of

issuance to be implemented within 60

days.

Amendment Nos.: 99, 206, and 135

Facility Operating License Nos. DPR21, DPR-65, and NPF-49: Amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: February 26, 1997 (62 FR

8800) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

April 10, 1997.No significant hazards

consideration comments received: No

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, Connecticut 06360, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut 06385

Pennsylvania Power and Light

Company, Docket Nos. 50-387 and 50-

388 Susquehanna Steam Electric

Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of application for amendments:

November 18, 1996

Brief description of amendments:

These amendments change the

Technical Specifications for

Susquehanna Steam Electric Station

(SSES), Units 1 and 2 by increasing the

maximum isolation times for reactor

core isolation cooling inboard warm-up

line isolation valves from 3 seconds to

12 seconds, high pressure core injection

inboard warm-up line siolation valves

from 3 seconds to 6 seconds and reactor

recirculation process sample line

isolation valves from 2 seconds to 9

seconds.

Date of issuance: April 7, 1997

Effective date: Both units, as of date

of issuance, to be implemented within

30 days.

Amendment Nos.: 164 and 135

Facility Operating License Nos. NPF14 and NPF-22: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: January 15, 1997 (61 FR 2191)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated April 7, 1997. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Osterhout Free Library,

Reference Department, 71 South

Franklin Street, Wilkes-Barre, PA 18701

Pennsylvania Power and Light

Company, Docket No. 50-388,

Susquehanna Steam Electric Station,

Unit 2, Luzerne County, Pennsylvania

Date of application for amendment:

March 17, 1997

Brief description of amendment: The

amendment modifies the Design

Features Section 5.3.1 of the Technical

Specifications to reflect the Atrium-10

design and would include a Siemens

Power Corporation topical report in

Section 6.9.3.2 to reflect mechanical

design criteria for this fuel. This change

would allow this fuel to be loaded into

the core only under Operational

Condition 5 (refueling) and does not

permit startup or power operation using

the Atrium-10 fuel.

Date of issuance: April 9, 1997

Effective date: As of date of issuance

to be implemented within 30 days.

Amendment No.: 136

Facility Operating License No. NPF22: This amendment revised the

Technical Specifications. Public

comments requested as to proposed no

significant hazards consideration: Yes

(62 FR 14167) March 25, 1997. That

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 April 24, 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 are contained in

a Safety Evaluation dated April 9, 1997.

Attorney for licensee: Jay Silbert,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street NW.,

Washington DC 20037.

Local Public Document Room

location: Osterhout Free Library,

Reference Department, 71 South

Franklin Street, Wilkes-Barre, PA 18701

Southern Nuclear Operating Company,

Inc., Georgia Power Company,

Oglethorpe Power Corporation,

Municipal Electric Authority of

Georgia, City of Dalton, Georgia, Docket

Nos. 50-321 and 50-366, Edwin I. Hatch

Nuclear Plant, Units 1 and 2, Appling

County, Georgia

Date of application for amendments:

September 19, 1996, as supplemented

December 17, 1996, January 23 and 31,

March 21 and April 4, 1997

Brief description of amendments: The

amendments revise the surveillance

requirements addressing the reactor

vessel pressure and temperature limits.

Date of issuance: April 4, 1997

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment Nos.: 206 and 147

Facility Operating

Local Public- Document -Room

locations: ments revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 2, 1997 (62 FR 128)

The December 17, 1996, January 23 and

31, March 21, 1997, and April 4, 1997,

letters provided 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 April 4,

1997.No significant hazards

consideration comments received: No

Local Public Document Room

location: Appling County Public

Library, 301 City Hall Drive, Baxley,

Georgia 31513

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:

October 18, 1996 as supplemented

March 12, March 17, April 4, and April

9, 1997 (TS 96-05)

Brief description of amendments: The

amendments change the Technical

Specifications (TS) by revising TS 3/

4.4.5 and 3.4.6.2 and associated Bases to

permanently incorporate requirements

for steam generator tube inspections and

repair in the Sequoyah Nuclear Plant,

Units 1 and 2 TS.

Date of issuance: April 9, 1997

Effective date: As of the date of

issuance to be implemented no later

than 45 days of its issuance.

Amendment Nos.: 222 and 213

Facility Operating License Nos. DPR77 and DPR-79: Amendments revise the

technical specifications and license

conditions.
19844 Federal Register / Vol. 62, No. 78 / Wednesday, April 23, 1997 / Notices

Date of initial notice in Federal

Register: February 11, 1997 (62 FR

6276) The March 12, March 17, April 4,

and April 9, 1997, letters provided

clarifying information that did not

change the scope of the October 18,

1996, application and the initial

proposed no significant hazards

consideration determination.The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated April 9, 1997.No

significant hazards consideration

comments received: None

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402

Notice Of Issuance Of Amendments To

Facility Operating Licenses And Final

Determination Of No Significant

Hazards Consideration And

Opportunity For A Hearing (Exigent

Public Announcement Or Emergency

Circumstances)

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application for the

amendment complies with the

standards and requirements of the

Atomic Energy Act of 1954, as amended

(the Act), and the Commission’s rules

and regulations. The Commission has

made appropriate findings as required

by the Act and the Commission’s rules

and regulations in 10 CFR Chapter I,

which are set forth in the license

amendment.

Because of exigent or emergency

circumstances associated with the date

the amendment was needed, there was

not time for the Commission to publish,

for public comment before issuance, its

usual 30-day Notice of Consideration of

Issuance of Amendment, Proposed No

Significant Hazards Consideration

Determination, and Opportunity for a

Hearing.

For exigent circumstances, the

Commission has either issued a Federal

Register notice providing opportunity

for public comment or has used local

media to provide notice to the public in

the area surrounding a licensee’s facility

of the licensee’s application and of the

Commission’s proposed determination

of no significant hazards consideration.

The Commission has provided a

reasonable opportunity for the public to

comment, using its best efforts to make

available to the public means of

communication for the public to

respond quickly, and in the case of

telephone comments, the comments

have been recorded or transcribed as

appropriate and the licensee has been

informed of the public comments.

In circumstances where failure to act

in a timely way would have resulted, for

example, in derating or shutdown of a

nuclear power plant or in prevention of

either resumption of operation or of

increase in power output up to the

plant’s licensed power level, the

Commission may not have had an

opportunity to provide for public

comment on its no significant hazards

consideration determination. In such

case, the license amendment has been

issued without opportunity for

comment. If there has been some time

for public comment but less than 30

days, the Commission may provide an

opportunity for public comment. If

comments have been requested, it is so

stated. In either event, the State has

been consulted by telephone whenever

possible.

Under its regulations, the Commission

may issue and make an amendment

immediately effective, notwithstanding

the pendency before it of a request for

a hearing from any person, in advance

of the holding and completion of any

required hearing, where it has

determined that no significant hazards

consideration is involved.

The Commission has applied the

standards of 10 CFR 50.92 and has made

a final determination that the

amendment involves no significant

hazards consideration. The basis for this

determination is contained in the

documents related to this action.

Accordingly, the amendments have

been issued and made effective as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the application for

amendment, (2) the amendment to

Facility Operating License, and (3) the

Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment, as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, the Gelman Building, 2120 L

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

local public document room for the

particular facility involved.

The Commission is also offering an

opportunity for a hearing with respect to

the issuance of the amendment. By May

23, 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. 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
Federal Register / Vol. 62, No. 78 / Wednesday, April 23, 1997 / Notices 19845

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).

Union Electric Company, Docket No.

50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application for amendment:

April 1, 1997

Brief description of amendment: The

amendment revises Technical

Specification Table 3.3-3 to correct

administrative errors associated with

the start logic of the turbine driven

auxiliary feedwater pump.

Date of issuance: April 2, 1997

Effective date: April 2, 1997

Amendment No.: 119

Facility Operating License No. NPF30: 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 April 2, 1997.

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 200379

Local Public Document Room

location: Callaway County Public

Library, 710 Court Street, Fulton,

Missouri 65251.

NRC Project Director: William H.

Bateman

Dated at Rockville, Maryland, this 16th day

of April, 1997.

For the Nuclear Regulatory Commission

Jack W. Roe,

Director ,Division of Reactor Projects III/IV,

Office of Nuclear Reactor Regulation

[Doc. 97–10334 Filed 4–22–97; 8:45 am]

BILLING CODE 7590–01–F

SECURITIES AND EXCHANGE

COMMISSION

Issuer Delisting; Notice of Application

to Withdraw From Listing and

Registration; (Donnelly Corporation,

Class A Common Stock, $0.10 Par

Value) File No. 1–9716

April 17, 1997.

Donnelly Corporation (‘‘Company’’)

has filed an application with the

Securities and Exchange Commission

(‘‘Commission’’), pursuant to Section

12(d) of the Securities Exchange Act of

1934 (‘‘Act’’) and Rule 12d2–2(d)

promulgated thereunder, to withdraw

the above specified security (‘‘Security’’)

from listing and registration on the

American Stock Exchange, Inc.

(‘‘Amex’’ or ‘‘Exchange’’).

The reasons cited in the application

for withdrawing the Security from

listing and registration include the

following:

The Company has complied with Rule

18 of the Amex by filing with such

Exchange a certified copy of preambles

and resolutions adopted by the

Company’s Board of Directors

authorizing the withdrawal of its

common stock from listing on the Amex

and by setting forth in detail to such

Exchange the reasons for such proposed

withdrawal, and the facts in support

thereof. The Company became listed for

trading on the New York Stock

Exchange, Inc. (‘‘NYSE’’) pursuant to a

Registration Statement on Form 8–A

effective March 6, 1997.

In making the decision to withdraw

its common stock from listing on the

Amex, the Company considered: (a) that

the Company believes that the NYSE

will offer the Company’s shareholders

more liquidity over time than is

presently available on the Amex; (b) that

the Company believes that listing on the

NYSE will offer greater visibility for the

Company and its stock potential for

greater institutional ownership; (c) that

as the Company becomes an

increasingly international company, it

believes there will be advantages to

having its stock listed on the NYSE, and

(d) many of the companies which it

regards as peers or leaders in its

industry are listed on the NYSE.

Any interested person may, on or

before May 8, 1997, submit by letter to

the Secretary of the Securities and

Exchange Commission, 450 Fifth Street,

NW., Washington, DC 20549, facts

bearing upon whether the application

has been made in accordance with the

rules of the exchanges and what terms,

if any, should be imposed by the

Commission for the protection of

investors. The Commission, based on