Federal Register / Vol. 63, No. 145 /Wednesday, July 29, 1998 / Notices 40551

appearance test in CTS 4.9.A.2.e.1.d)

(Change ITS 5.5–M.4)

Environmental Impacts of the Proposed

Action

The Commission has completed its

evaluation of the proposed conversion

of the CTS to the ITS for the CNS,

including the six beyond-scope issues

identified above. Changes which are

administrative in nature have been

found to have no effect on the technical

content of the TS.

The increased clarity and

understanding these changes bring to

the TS are expected to improve the

operators control of the CNS in normal

and accident conditions.

Relocation of requirements from the

CTS to other licensee-controlled

documents does not change the

requirements themselves. Future

changes to these requirements may then

be made by the licensee under 10 CFR

50.59 and other NRC-approved control

mechanisms which will ensure

continued maintenance of adequate

requirements. All such relocations have

been found consistent with the

guidelines of NUREG–1433 and the

Commission’s Final Policy Statement.

Changes involving more restrictive

requirements have been found to

enhance station safety.

Changes involving less restrictive

requirements have been reviewed

individually. When requirements have

been shown to provide little or no safety

benefit, or to place an unnecessary

burden on the licensee, their removal

from the TS is justified. In most cases,

relaxations previously granted to

individual plants on a plant-specific

basis were the result of a generic action,

or of agreements reached during

discussions with the OG and found to

be acceptable for the station. Generic

relaxations contained in NUREG–1433

have been reviewed by the NRC staff

and found to be acceptable.

In summary, the proposed revisions to

the TS have been found to provide

control of station operations such that

reasonable assurance will be provided

that the health and safety of the public

will be adequately protected.

The proposed amendment will not

increase the probability or consequences

of accidents, will not change the

quantity or types of any effluent that

may be released offsite, and will not

significantly increase occupational or

public doses. Also, these changes do not

affect the design of the station, do not

involve any modifications to the station,

and do not increase the licensed power

and allowable effluents for the station.

The changes will not create any new or

unreviewed environmental impacts that

were not considered in the Final

Environmental Statement (FES) related

to the operation of the CNS dated

February 1973. Therefore, there are no

significant radiological impacts

associated with the proposed

amendment.

With regard to potential nonradiological impacts, the proposed

amendment involves features located

entirely within the restricted area

defined in 10 CFR Part 20. They do not

affect non-radiological station effluents

and have no other environmental

impact. Therefore, there are no

significant non-radiological

environmental impacts associated with

the proposed amendment.

Accordingly, the Commission

concludes that there are no significant

environmental impacts associated with

the proposed amendment.

Alternatives to the Proposed Action

Since the Commission has concluded

there is no significant environmental

impact associated with the proposed

amendment, any alternatives with equal

or greater environmental impact need

not be evaluated. The principal

alternative to the proposed amendment

would be to deny the amendment.

Denial of the licensee’s application

would not reduce the environmental

impacts of the CNS operations, but it

would prevent the safety benefits to the

station from the conversion to the ITS.

The environmental impacts of the

proposed action and the alternative

action are similar.

Alternative Use of Resources

This action does not involve the use

of any resources not previously

considered in the FES for the CNS.

Agencies and Persons Consulted

In accordance with its stated policy,

on July 22, 1998, the staff consulted

with the Nebraska State official, Cheryl

Rogers of the State Department of

Health, regarding the environmental

impact of the proposed action. The State

official had no comments.

Finding of No Significant Impact

Based upon the environmental

assessment, the Commission concludes

that the proposed action will not have

a significant effect on the quality of the

human environment. Accordingly, the

Commission has determined not to

prepare an environmental impact

statement for the proposed action.

For further details with respect to the

proposed action, see the licensee’s

application dated March 27, 1997, as

supplemented by the letters dated

September 29 and December 22, 1997,

and February 9, March 13, March 26,

April 16, and May 6, 1998, which is

available for public inspection at the

Commission’s Public Document Room,

The Gelman Building, 2120 L Street,

NW., Washington, DC, and at the local

public document room located at the

Auburn Memorial Library, 1810

Courthouse Avenue, Auburn, Nebraska

68305.

Dated at Rockville, Maryland, this 23rd day

of July 1998.

For the Nuclear Regulatory Commission.

David L. Wigginton,

Acting Director, Project Directorate IV–1,

Division of Reactor Projects III/IV, Office of

Nuclear Reactor Regulation.

[FR Doc. 98–20235 Filed 7–28–98; 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 Pub. L. 97–415, the U.S.

Nuclear Regulatory Commission (the

Commission or NRC staff) is publishing

this regular biweekly notice. Pub. L. 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 July 3, 1998,

through July 17, 1998. The last biweekly

notice was published on July 15, 1998

(63 FR 38198).

Notice of Consideration of Issuance of

Amendments to Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a

proposed determination that the

following amendment requests involve

no significant hazards consideration.

Under the Commission’s regulations in

10 CFR 50.92, this means that operation

of the facility in accordance with the
40552 Federal Register / Vol. 63, No. 145 /Wednesday, July 29, 1998 / Notices

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 and

Directives Branch, Division of

Administration 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 August 14, 1998, the licensee may

file a request for a hearing with respect

to issuance of the amendment to the

subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555–0001, Attention:

Rulemakings and Adjudications Staff, or

may be delivered to the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington DC, by the above date. 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
Federal Register / Vol. 63, No. 145 /Wednesday, July 29, 1998 / Notices 40553

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.

Duke Energy Corporation (DEC), et al.,

Docket Nos. 50–413 and 50–414,

Catawba Nuclear Station, Units 1 and 2,

York County, South Carolina

Date of amendment request: May 27,

1997, as supplemented by letters dated

March 9, March 20, April 20, May 27,

and June 24, 1998

Description of amendment request:

The proposed amendments would

revise the current Technical

Specifications (TS) of each unit to

conform with NUREG–1431, Revision 1,

‘‘Standard Technical Specifications—

Westinghouse Plants.’’ The staff had

previously issued a Notice of

Consideration of Issuance of

Amendments published in the Federal

Register on July 14, 1997 (62 FR 37628)

covering all the proposed changes that

were indeed within the scope of

NUREG–1431. The staff subsequently

published two Notices of Consideration

of Issuance of Amendments and

Proposed No Significant Hazards

Determination (63 FR 25106, dated May

6, 1998; 63 FR 27760 dated May 20,

1998) to cover DEC’s March 9, March

20, April 20, and May 27, 1998,

supplements, which proposed changes

that are beyond the scope of NUREG–

1431. On June 24, 1998, DEC identified

additional beyond-scope changes. The

following descriptions and proposed no

significant hazard analyses cover only

those beyond-scope changes. Associated

with each change are administrative/

editorial changes such that the new or

revised requirements would fit into the

format of NUREG–1431.

1. Current TS 4.8.1.1.2.f specifies that the

fuel for the emergency diesel generators

(EDGs) be periodically sampled for

particulate contamination strictly in

accordance with the industry standard

ASTM-D2276–78. DEC proposed to relax this

requirement, adopting only the guidance of

the standard, but using a larger particulate

filter for sampling (change from 0.8-to 3-

micron). The revised requirement would

show up as TS 5.5.13.c of the Improved TS.

No changes to the design and functions of the

EDGs are proposed.

2. DEC proposed to revise current TS

Table 4.3–1, Functions 16 and 17. The

revised requirements, to show up as

Table 3.3.1–1, Functions 15 and 16.b, of

the Improved TS, would add an

actuation logic test surveillance for the

reactor trip system interlocks and the

safety injection input from the

engineered safety feature actuation

system. No changes to the design and

functions of these systems are involved.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), DEC

has provided its analyses of the issue of

no significant hazards consideration for

each of the above proposed changes.

The NRC staff has reviewed DEC’s

analyses against the standards of 10 CFR

50.92(c). The NRC staff’s analysis is

presented below.

1. Will the changes involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

For all the changes the answer is

‘‘no.’’ The proposed changes will not

affect the safety function of the subject

systems. There will be no direct effect

on the design or operation of any plant

structures, systems, or components. No

previously analyzed accidents were

initiated by the functions of these

systems, and the systems will continue

to perform their functions in mitigating

consequences of previously analyzed

accidents. Therefore, the proposed

changes will have no impact on the

consequences or probabilities of any

previously evaluated accidents.

2. Will the changes create the

possibility of a new or different kind of

accident from any accident previously

evaluated?

For all the changes the answer is

‘‘no.’’ The proposed changes would not

lead to any design or operating

procedure change. Hence, no new

equipment failure modes or accidents

from those previously evaluated will be

created.

3. Will the changes involve a

significant reduction in a margin of

safety?

For all the changes the answer is

‘‘no.’’ Margin of safety is associated with

confidence in the design and operation

of the plant. The proposed changes to

the TS do not involve any change to

plant design, operation, or analysis.

Thus, the margin of safety previously

analyzed and evaluated is maintained.

Based on this analysis, it appears that

the three standards of 10 CFR 50.92(c)

are satisfied for each of the proposed

changes. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: York County Library, 138 East

Black Street, Rock Hill, South Carolina.

Attorney for licensee: Mr. Paul R.

Newton, Legal Department (PB05E),

Duke Energy Corporation, 422 South

Church Street, Charlotte, North

Carolina.

NRC Project Director: Herbert N.

Berkow.

Duke Energy Corporation (DEC), et al.,

Docket Nos. 50–413 and 50–414,

Catawba Nuclear Station, Units 1 and 2,

York County, South Carolina

Date of amendment request:

September 15, 1997, as supplemented

by letters dated March 5, April 27, and

June 15, 1998.

Description of amendment request:

The staff had previously published a

Notice of Consideration of Amendments

and Proposed No Significant Hazards

Consideration Determination on the

licensee’s September 15, 1997,

application in the Federal Register on

October 8, 1997 (62 FR 52580). As a

result of the staff’s requests for

additional information, DEC expanded

its original amendment application by

letter dated June 15, 1998. Specifically,

the June 15, 1998, letter proposes

requirements regarding the Low

Temperature Overpressure Protection

System to be added to the units’

Technical Specifications. There is,

however, no change to plant design.

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, addressing the three

standards of 10 CFR 50.92(c):

First Standard

Implementation of this amendment would

not involve a significant increase in the

probability or consequences of an accident

previously evaluated. The Low Temperature

Overpressure Protection System is not an

accident initiating system; it is an accident

mitigating system. Therefore, the addition of

supplemental Technical Specification

required controls pertaining to this system

cannot impact accident initiating

probabilities. The Low Temperature

Overpressure Protection System will remain

fully capable of performing its design

accident mitigation function for the modes in

which it is required. Therefore, no accident

consequences will be impacted.

Second Standard

Implementation of this amendment would

not create the possibility of a new or different

kind of accident from any accident

previously evaluated. As noted previously,

the Low Temperature Overpressure

Protection System is not an accident

initiating system. The addition of the

supplemental Technical Specification
40554 Federal Register / Vol. 63, No. 145 /Wednesday, July 29, 1998 / Notices

controls pertaining to this system as specified

will not impact any plant systems that are

accident initiators. No other modifications

are being proposed to the plant which would

result in the creation of new accident

mechanisms.

Third Standard

Implementation of this amendment would

not involve a significant reduction in a

margin of safety. Margin of safety is related

to the confidence in the ability of the fission

product barriers to perform their design

functions during and following an accident

situation. These barriers include the fuel

cladding, the reactor coolant system, and the

containment system. The performance of the

fission product barriers will not be impacted

by implementation of this proposed

amendment supplement. The Low

Temperature Overpressure Protection System

will remain fully capable of performing its

design function for the modes in which it is

required. Therefore, no safety margin will be

significantly impacted.

The staff reviewed the licensee’s

analysis, and agrees that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: York County Library, 138 East

Black Street, Rock Hill, South Carolina.

Attorney for licensee: Mr. Paul R.

Newton, Legal Department (PB05E),

Duke Energy Corporation, 422 South

Church Street, Charlotte, North

Carolina.

NRC Project Director: Herbert N.

Berkow.

Duke Energy Corporation, Docket Nos.

50–369 and 50–370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg

County, North Carolina

Date of amendment request: May 8,

1998.

Description of amendment request:

The proposed amendments would

revise the Technical Specifications (TS)

for the Power Range Neutron Flux High

Trip setpoints in the event of inoperable

main steam safety valves. The licensee

has determined that the new values are

more conservative than the values in the

current TS. Also, the proposed changes

would delete the references to the 3-

loop operation. The proposed changes

are consistent with the proposed

Improved Standard TS submitted by the

licensee on May 27, 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. Will operation of the facility in

accordance with the proposed amendment

involve an increase in the probability or

consequences of an accident previously

evaluated?

The proposed amendment involves a

reduction in the maximum allowable power

range neutron flux high setpoints in case of

inoperable main steam safety valves. All

applicable UFSAR [Updated Final Safety

Analysis Report] Chapter 15 transient

acceptance criteria are met with the proposed

change. Therefore, operation of the facility in

accordance with the proposed amendment

will not involve an increase in the

probability or consequences of an accident

previously evaluated.

2. Will operation of the facility in

accordance with the proposed amendment

create the possibility of a new or different

kind of accident from any accident

previously evaluated:

No new equipment or operating practice is

involved with this proposed amendment. No

alteration to any existing hardware is

involved with this proposed amendment.

Power Range high neutron flux setpoint

calibration is continued to be performed by

the same approved procedure. Therefore,

operation of the facility in accordance with

the proposed amendment will not create the

possibility of any new or different kind of

accident from any accident previously

evaluated.

3. Will operation of the facility in

accordance with the proposed amendment

involve a reduction in a margin of safety?

The proposed change is in a moreconservative direction. All applicable UFSAR

Chapter 15 transient acceptance criteria are

met with the proposed amendment.

Therefore, operation of the facility in

accordance with the proposed amendment

will not involve a reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: J. Murrey Atkins Library,

University of North Carolina at

Charlotte, 9201 University City

Boulevard, Charlotte, North Carolina.

Attorney for licensee: Mr. Albert Carr,

Duke Energy Corporation, 422 South

Church Street, Charlotte, North

Carolina.

NRC Project Director: Herbert N.

Berkow.

Duke Energy Corporation (DEC), et al.,

Docket Nos. 50–369 and 50–370,

McGuire Nuclear Station, Units 1 and 2,

Mecklenburg County, North Carolina

Date of amendment request: May 27,

1997, as supplemented by letters dated

March 9, March 20, April 20, May 27,

June 3, June 24, and July 7, 1998.

Description of amendment request:

The proposed amendments would

revise the current Technical

Specifications (TS) of each unit to

conform with NUREG–1431, Revision 1,

‘‘Standard Technical Specifications—

Westinghouse Plants.’’ The staff had

previously issued a Notice of

Consideration of Issuance of

Amendments published in the Federal

Register on July 15, 1997 (62 FR 37940)

covering all the proposed changes that

were indeed within the scope of

NUREG–1431. The staff subsequently

published additional Notices of

Consideration of Issuance of

Amendments and Proposed No

Significant Hazards Determination on

May 6, 1998 (63 FR 25107 and 63 FR

25108 (two notices)) and on May 20,

1998 (63 FR 27761) to cover DEC’s

March 9, March 20, April 20, and May

27, 1998, supplements, which proposed

changes that are beyond the scope of

NUREG–1431.

On June 24, 1998, DEC identified

additional beyond-scope changes. The

following descriptions and proposed no

significant hazard analyses cover only

those beyond-scope changes. Associated

with each change are administrative/

editorial changes such that the new or

revised requirements would fit into the

format of NUREG–1431.

1. Current TS 4.8.1.1.2.f specifies that

the fuel for the emergency diesel

generators (EDGs) be periodically

sampled for particulate contamination

in accordance with ASTM–D2276–78.

DEC proposed to relax this requirement,

adopting instead the guidance of

ASTM–D2276, Method A. The revised

requirement would show up as TS

5.5.13.c of the Improved TS. No changes

to the design and functions of the EDGs

are proposed.

2. DEC proposed to change the

required action due to inoperable

channels of the containment pressure

control system as currently contained in

Table 3.3–3, Item 7. The revised

requirement would show up as Action

Item 16b in Table 3.3.2–1 of the

Improved TS. No changes to the design

and functions of the containment

pressure control system are involved.

3. DEC proposed to revise current TS

Table 4.3–1, Functions 16 and 17. The

revised requirements, to show up as

Table 3.3.1–1 Functions 15 and 16.b,

would add an actuation logic test

surveillance for the reactor trip system

interlocks and the safety injection input

from the engineered safety feature

actuation system. No changes to the

design and functions of these systems

are involved.

Basis for proposed no significant

hazards consideration determination:
Federal Register / Vol. 63, No. 145 /Wednesday, July 29, 1998 / Notices 40555

As required by 10 CFR 50.91(a), DEC

has provided its analyses of the issue of

no significant hazards consideration for

each of the above proposed changes.

The NRC staff has reviewed DEC’s

analyses against the standards of 10 CFR

50.92(c). The NRC staff’s analysis is

presented below.

1. Will the changes involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

For all the changes the answer is

‘‘no.’’ The proposed changes will not

affect the safety function of the subject

systems. There will be no direct effect

on the design or operation of any plant

structures, systems, or components. No

previously analyzed accidents were

initiated by the functions of these

systems, and the systems will continue

to perform their functions in mitigating

consequences of previously analyzed

accidents. Therefore, the proposed

changes will have no impact on the

consequences or probabilities of any

previously evaluated accidents.

2. Will the changes create the

possibility of a new or different kind of

accident from any accident previously

evaluated?

For all the changes the answer is

‘‘no.’’ The proposed changes would not

lead to any hardware or operating

procedure change. Hence, no new

equipment failure modes or accidents

from those previously evaluated will be

created.

3. Will the changes involve a

significant reduction in a margin of

safety?

For all the changes the answer is

‘‘no.’’ Margin of safety is associated with

confidence in the design and operation

of the plant. The proposed changes to

the TS do not involve any change to

plant design, operation, or analysis.

Thus, the margin of safety previously

analyzed and evaluated is maintained.

Based on this analysis, it appears that

the three standards of 10 CFR 50.92(c)

are satisfied for each of the proposed

changes. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: York County Library, 138 East

Black Street, Rock Hill, South Carolina.

Attorney for licensee: Mr. Paul R.

Newton, Legal Department (PB05E),

Duke Energy Corporation, 422 South

Church Street, Charlotte, North

Carolina.

NRC Project Director: Herbert N.

Berkow.

Duke Energy Corporation, Docket No.

50–287, Oconee Nuclear Station, Unit 3,

Oconee County, South Carolina

Date of amendment request: July 16,

1998.

Description of amendment request:

The proposed change would extend, on

a one-time basis, certain specified

Technical Specification surveillances

that are required to be performed at a

frequency of 18 months from the

maximum allowed frequency of 22

months, 15 days, to a maximum of 24

months. The following surveillances are

involved: (a) Standby Shutdown Facility

(SSF) Reactor Coolant System (RCS)

Pressure Instrument Calibration; (b) SSF

RCS Pressurizer Level Instrument

Calibration; (c) SSF RCS Makeup Pump

Flow Instrument Calibration; (d) Reactor

Protective System (RPS) RCS Flow

Instrument Calibration; (e) RPS RCS

Pressure Instrument Calibration; and (f)

Low Pressure Injection System Pump

Discharge Valves LP–17 and LP–18

Manual Cycle.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

This proposed change has been evaluated

against the standards in 10 CFR 50.92 and

has been determined to involve no significant

hazards, in 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?

No. A review of the previous two

instrument channel tests and calibrations,

and two manual valve cycle tests discussed

in this amendment request concluded that no

adverse effects should occur as a result of the

one-time extension.

There is a high level of confidence that the

instruments and valves should be available to

perform their intended function during the

requested extension period. Thus, the

probability and consequences of an accident

previously evaluated will not be significantly

increased.

(2) Create the possibility of a new or

different kind of accident from the accidents

previously evaluated?

No. Since the one-time extension should

not cause any adverse effects on Standby

Shutdown Facility, Reactor Protective

System or the Low Pressure Injection system,

a new or different kind of accident from the

accidents which were previously evaluated

will not occur. The Standby Shutdown

Facility, Reactor Protective System or the

Low Pressure Injection system should be

available to perform their intended function

during the requested extension period.

(3) Involve a significant reduction in a

margin of safety?

No. The margin of safety will not be

significantly reduced by this amendment

request because the Standby Shutdown

Facility, Reactor Protective System or the

Low Pressure Injection system should be

available to perform their intended function

during the requested extension period. In

addition, the review of the previous tests and

calibrations which are discussed in the

amendment request concluded that no

adverse effects should occur as a result of the

one-time extension.

Duke [Energy Corporation] has concluded,

based on the above information, that there

are no significant hazards 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.

Attorney for licensee: J. Michael

McGarry, III, Winston and Strawn, 1200

17th Street, NW., Washington, DC.

NRC Project Director: Herbert N.

Berkow.

Florida Power Corporation, et al.,

Docket No. 50–302, Crystal River

Nuclear Generating Plant, Unit No. 3,

Citrus County, Florida

Date of amendment request: April 28,

1998.

Description of amendment request:

The proposed amendment would

change the scope and frequency of

volumetric and surface inspections for

the reactor coolant pump motor

flywheels. The current prescribed

frequency and scope are contained in

U.S. NRC Regulatory Guide 1.14,

Regulatory Positions C.4.b.1 and C.4.b.2.

The proposed revision reflects the

frequency and scope of volumetric and

surface examinations, which has been

reviewed and approved by the NRC, as

stated in the Safety Evaluation for

Topical Report WCAP–14535A.

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 CR–3 [Crystal River Unit 3]

components addressed by this proposed

change are the Reactor Coolant Pumps

(RCPs), identified by plant tagging

procedures as RCP–1A, RCP–1B, RCP–1C,
40556 Federal Register / Vol. 63, No. 145 /Wednesday, July 29, 1998 / Notices

and RCP–1D. The RCPs are vertical, single

stage, single suction, shaft seal, centrifugal

pumps. The RCPs ensure that adequate

cooling water is circulated through the

reactor coolant system. Following loss of

power to the RCP motor, the flywheel, in

conjunction with the impeller and motor

rotating assembly, provide sufficient

rotational inertia to assure adequate coolant

flow during RCP coastdown, thus providing

adequate core cooling. The maximum loading

on the RCP motor flywheel results from

overspeed following a large loss of coolant

accident (LOCA). The estimated maximum

speed in the event of a LOCA was established

conservatively. The proposed change does

not affect that analysis. Reduced coastdown

times due to a single failed flywheel is

bounded by the locked rotor analysis,

therefore it will not place the plant in an

unanalyzed condition.

Reducing the frequency of inspection, as

proposed, will not significantly increase the

probability of an accident previously

evaluated. CR–3 is not specifically analyzed

for a flywheel failure accident. The design,

fabrication, and testing of the flywheels in

accordance with the guidance found in

Regulatory Guide 1.14 minimizes the

potential for flywheel failure. Nevertheless,

the topical report indicates that the flywheels

could be operated for forty years without

inspection, and there would be no significant

increase in the probability of failure of the

flywheel. However, inspections are proposed

to continue at a frequency of once every ten

years as a conservative measure. Therefore,

these changes do not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

(2) Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The purpose of the RCP motor flywheel

inspection is to identify flaws that could lead

to failure of the flywheel. The design,

fabrication, and testing of the flywheels in

accordance with the guidance found in

Regulatory Guide 1.14 minimizes the

potential for flywheel failure. No new failure

mode is introduced due to the change in

flywheel inspection frequency since the

proposed changes do not involve the

addition or modification of equipment, nor

alter the design or operation of affected plant

systems, structures or components.

Therefore, these changes do not create a

possibility of a new or different kind of

accident from any previously evaluated.

(3) Involve a significant reduction in a

margin of safety.

As shown in the topical report, RCP motor

flywheels have been inspected for twenty

years without any service induced flaws

being identified. Additionally, the analyses

demonstrated that the flywheels are

manufactured from excellent quality steel,

have a high fracture toughness, and have a

very high flaw tolerance. The topical report

indicates that the flywheels could be

operated for forty years without inspection,

and there would be no significant increase in

the probability of failure of the flywheels.

However, inspections are proposed to

continue at a frequency of once every ten

years as a conservative measure. The nondestructive examination acceptance criteria

is not changing as a result of the proposed

LAR. Thus, the margin of safety is not

reduced significantly by the proposed change

in inspection frequency.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida

34428.

Attorney for licensee: R. Alexander

Glenn, General Counsel, Florida Power

Corporation, MAC—A5A, P. O. Box

14042, St. Petersburg, Florida 33733–

4042.

NRC Project Director: Frederick J.

Hebdon.

Florida Power and Light Company, et

al., Docket No. 50–389, St. Lucie Plant,

Unit No. 2, St. Lucie County, Florida

Date of amendment request: May 27,

1998.

Description of amendment request:

The proposed amendment would

modify the Technical Specifications

(TS) to remove the requirement for

safety injection tanks (SITs) to be

operable in reactor operational Mode 4.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) Operation of the facility in accordance

with the proposed amendment would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed amendment does not involve

changes to previously evaluated accident

initiators. The proposed TS changes related

to removal of the requirement for safety

injection tanks to be operable in MODE 4 do

not impact the results of existing accident

analyses, and have no adverse impact on any

plant system performance.

The function of each SIT is to provide early

reactor core reflood in the event of a LBLOCA

[large break loss-of-coolant accident]. Safety

injection tanks are not required for mitigating

the consequences of large RCS pipe ruptures

in MODE 4, and the proposed change to TS

3.5.1 will delete the requirement for SIT

operability when in this mode. Due to the

reduced initial stored energy and decay heat

generation rate consistent with operation in

the shutdown modes, the required operable

HPSI [high-pressure safety injection] pump is

sufficient to perform the function of reactor

vessel reflood and coolant inventory makeup. Therefore, operation of the facility in

accordance with the proposed amendment

will not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

(2) Operation of the facility in accordance

with the proposed amendment would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed amendment will not change

the physical plant or the modes of operation

defined in the facility license. The changes

do not involve the addition of new

equipment or the modification of existing

equipment, nor do they alter the design of St.

Lucie plant systems described in the Updated

Final Safety Analysis Report (UFSAR). There

are no adverse effects on any system

performance due to the proposed TS changes,

and the plant configuration will continue to

remain consistent with assumptions used in

the existing accident analyses. Therefore,

operation of the facility in accordance with

the proposed amendment would not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

(3) Operation of the facility in accordance

with the proposed amendment would not

involve a significant reduction in a margin of

safety.

The proposed TS changes have been

evaluated with respect to the applicable

safety analyses. FPL [Florida Power and Light

Co.] determined from this new evaluation

that safety injection tanks are not required to

prevent core uncovery during a loss of

coolant accident initiated in MODE 4. Due to

the reduced core heat removal requirements

in this lower mode and in the absence of

substantial core uncovery, fuel cladding

temperatures and clad oxidation will remain

at low levels, long term cooling will be

maintained, and 10 CFR 50.46 acceptance

criteria will be satisfied. Therefore, operation

of the facility in accordance with the

proposed amendment would not involve a

significant reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Indian River Community

College Library, 3209 Virginia Avenue,

Fort Pierce, Florida 34981–5596.

Attorney for licensee: M.S. Ross,

Attorney, Florida Power & Light, P.O.

Box 14000, Juno Beach, Florida 33408–

0420.

NRC Project Director: Frederick J.

Hebdon.

GPU Nuclear, Inc. et al., Docket No. 50–

219, Oyster Creek Nuclear Generating

Station, Ocean County, New Jersey

Date of amendment request: June 29,

1998.

Description of amendment request:

This Technical Specification change
Federal Register / Vol. 63, No. 145 /Wednesday, July 29, 1998 / Notices 40557

request replaces in its entirety, a

previously submitted request dated

February 22, 1996, and published in the

Federal Register on March 27, 1996 (61

FR 13525). This request greatly reduces

the scope of the previous request. It

retains the provision to delete the

requirement that the biennial inspection

of the Emergency Diesel Generators

(EDGs) be performed during shutdown,

permits skipping diesel starting battery

capacity test for recently installed

batteries, and increases the minimum

loading during diesel testing from 20%

to 80%. In addition, there are wording

changes to enhance clarity, and a

typographical error is corrected.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. State the basis for the determination that

the proposed activity will or will not increase

the probability of occurrence or the

consequences of an accident.

The proposed activity deletes the

requirement to inspect EDGs during shut

down from the Technical Specifications and

permits skipping diesel starting battery

capacity tests of recently installed batteries.

The minimum loading during the testing of

the diesels has been increased from 20% to

80%. In addition, wording changes were

made to enhance clarity and a minor

typographical error was corrected. During

reactor operations other power sources are

available to compensate for one diesel being

out of service. The inspections and testing

will continue to be done with the same

intervals and the 80% loading is a more

stringent requirement. Therefore, these

changes do not affect the design or

performance of the EDGs or their ability to

perform their design function.

2. State the basis for the determination that

the activity does or does not create a

possibility of an accident or malfunction of

a different type than any previously

identified in the [safety analysis report] SAR.

The EDGs are not the source of any

accident described in the SAR. These

changes do not modify the design or

performance of the EDGs and do not affect

plant functions or actions. Current

specifications permit one diesel generator to

be inoperable for up to 7 days and this

change will not impact that time frame.

Therefore, the proposed change does not

create the possibility of an accident or

malfunction of a different type than those

previously identified.

3. State the basis for the determination that

the margin of safety is not reduced.

The proposed changes are designed to

improve EDG reliability and availability

during shutdown periods by providing

flexibility in the scheduling and performance

of maintenance. The surveillance intervals

are unchanged and operability requirements

are not modified. The proposed activity does

not alter the basis of any technical

specification that is related to the

establishment or maintenance of a nuclear

safety margin. Therefore, the margin of safety

is not significantly reduced by this action.

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: Ocean County Library,

Reference Department, 101 Washington

Street, Toms River, NJ 08753.

Attorney for licensee: Ernest L. Blake,

Jr., Esquire. Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Cecil O.

Thomas.

Niagara Mohawk Power Corporation,

Docket No. 50–220, Nine Mile Point

Nuclear Station, Unit 1 (NMP1), Oswego

County, New York

Date of amendment request: June 19,

1998.

Description of amendment request:

The proposed amendment would

update Technical Specification (TS)

3.2.2, ‘‘Minimum Reactor Vessel

Temperature for Pressurization,’’ and

the associated TS Bases pages. TS 3.2.2

contains tables and figures that limit the

minimum reactor vessel temperature for

a given pressure. The limits are based

upon the number of Effective Full

Power Years (EFPY) of core operation.

The current tables and figures are valid

for up to 18 EFPYs of core operation.

The proposed amendment will

substitute new tables and figures that

are valid for 20, 24 and 28 EFPYs. The

word ‘‘leakage’’ would be added to

clarify that this TS applies to both

leakage and hydrostatic tests.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The operation of Nine Mile Point Unit

1, in accordance with the proposed

amendment, will not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

The changes to the P–T [pressure and

temperature] curves are being proposed to

preclude brittle fracture of RPV [reactor

pressure vessel] materials for up to 28 EFPYs.

In addition to the leakage/hydrostatic test

curve for 28 EFPYs, leakage/hydrostatic test

curves have been prepared for exposures up

to 20 EFPYs and up to 24 EFPYs to shorten

outage time for startups conducted prior to

these exposures. Safety margins specified in

10 CFR Part 50, Appendix G and Appendix

G to Section III of the ASME [American

Society of Mechanical Engineers] Code will

continue to be met for each of these curves.

Also, the proposed changes do not affect the

probability of any accident precursors.

Therefore, operation in accordance with the

proposed change will not involve a

significant increase in the probability of an

accident previously evaluated.

The RPV, as part of the reactor coolant

system, provides a barrier to the release of

reactor coolant and subsequent radiological

consequences. Operation in accordance with

the proposed amendment will preclude

brittle fracture of the RPV consistent with

current requirements, and consequently, not

affect the consequences of any accidents.

Therefore, operation of NMP1 [Nine Mile

Point Unit 1] in accordance with the

proposed amendment will not involve a

significant increase in the consequences of an

accident previously evaluated.

2. The operation of Nine Mile Point Unit

1, in accordance with the proposed

amendment, will not create the possibility of

a new or different kind of accident from any

accident previously evaluated.

The proposed change does not involve any

physical alterations to plant configurations or

introduce any new accident precursors

which could initiate a new or different kind

of accident. The proposed change does not

affect the intended function of the RPV nor

does it affect the operation of the RPV in a

way which would create a new or different

kind of accident. The changes to the P–T

curves are being proposed to preclude brittle

fracture of RPV materials for up to 28 EFPYs.

Safety margins specified in 10 CFR Part 50,

Appendix G and Appendix G to Section III

of the ASME Code will continue to be met.

Therefore, operation of NMPI in accordance

with the proposed change will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The operation of Nine Mile Point Unit

1, in accordance with the proposed

amendment, will not involve a significant

reduction in a margin of safety.

The existing NMP1 P–T curves were

developed using safety margins for brittle

fracture found in 10 CFR PART 50 Appendix

G and Appendix G to Section III of the ASME

Code. The proposed NMP1 P–T operation

curves, which are valid for up to 28 EFPYs

of operation, were also developed using the

safety margins for brittle fracture found in 10

CFR PART 50, Appendix G and Appendix G

to Section III of the ASME Code.

Accordingly, operation of NMPI in

accordance with the revised P–T operating

limits will continue to preclude brittle

fracture of the RPV materials during plant

heatup, cooldown, and leakage/hydrostatic

test conditions with the same margin of

safety that currently exists. Therefore,

operation of NMP1 in accordance with the

proposed amendment will not involve a

significant reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three
40558 Federal Register / Vol. 63, No. 145 /Wednesday, July 29, 1998 / Notices

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston & Strawn,

1400 L Street, NW., Washington, DC

20005–3502.

NRC Project Director: S. Singh Bajwa.

Pacific Gas and Electric Company,

Docket Nos. 50–275 and 50–323, Diablo

Canyon Nuclear Power Plant, Unit Nos.

1 and 2, San Luis Obispo County,

California

Date of amendment request: May 22,

1997, as supplemented by letters dated

June 12, 1997, August 28, 1997 and

January 29, 1998.

Description of amendment request:

The proposed amendments would

revise the combined Technical

Specifications (TS) for the Diablo

Canyon Power Plant Unit Nos. 1 and 2

TS 3/4.7.3.1, ‘‘Plant Systems—Vital

Component Cooling Water System,’’ to

add new action statements and

surveillance requirements for the

component cooling water (CCW) surge

tank pressurization system. CCW surge

tank pressurization system requirements

currently exist in an equipment control

guideline, but are proposed for

inclusion in TS because the CCW surge

tank pressurization system is required to

support licensing basis assumptions for

a design basis loss-of-coolant accident.

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 component cooling water (CCW) surge

tank pressurization system is designed to

mitigate the consequences of an accident,

and cannot initiate an accident.

The proposed changes to the Technical

Specifications (TS) incorporate requirements

for the CCW surge tank pressurization system

to assure that the consequences of an

accident are not increased. The CCW surge

tank pressurization system was installed to

restore the component cooling water system

to its original design and licensing basis. The

design of the CCW surge tank pressurization

system ensures that a minimum pressure of

17 psig is maintained in the surge tank at the

initiation of a design basis loss of coolant

accident. This minimum pressure is

sufficient to ensure that boiling will not

occur in the containment fan cooler units

(CFCUs), assuming the worst case accident

conditions with a concurrent loss of offsite

power (LOOP).

Therefore, the addition of these new

requirements does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The CCW surge tank pressurization system

is designed to mitigate the consequences of

an accident, and cannot initiate an accident.

The proposed TS changes incorporate

requirements for the CCW surge tank

pressurization system. Installation of the

CCW surge tank pressurization system

provides assurance that boiling in the CFCUs

will not occur, assuming the worst case

accident, with a concurrent LOOP.

Therefore, addition of these requirements

does not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed changes to the TS

incorporate requirements for the CCW surge

tank pressurization system to assure that the

consequences of an accident are not

increased. The design of the CCW surge tank

pressurization system ensures that a

minimum pressure of 17 psig is maintained

in the surge tank at the initiation of a design

basis accident. The minimum pressure is

sufficient to ensure that boiling will not

occur in the CFCUs, assuming the worst case

accident conditions with a concurrent LOOP.

Therefore, the proposed changes do not

involve a reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment requests

involve no significant hazards

consideration.

Local Public Document Room

Location: California Polytechnic State

University, Robert E. Kennedy Library,

Government Documents and Maps

Department, San Luis Obispo, California

93407.

Attorney for Licensee: Christopher J.

Warner, Esq., Pacific Gas & Electric

Company, P.O. Box 7442, San

Francisco, California 94120.

NRC Project Director: William H.

Bateman.

PECO Energy Company, Public Service

Electric and Gas Company, Delmarva

Power and Light Company, and Atlantic

City Electric Company, Docket Nos. 50–

277 and 50–278, Peach Bottom Atomic

Power Station, Unit Nos. 2 and 3, York

County, Pennsylvania

Date of application for amendments:

March 20, 1998, as revised by letter

dated June 26, 1998.

Description of amendment request:

The proposed amendment would revise

the Technical Specifications (TS) to

permit incorporation of an End-of-Cycle

Recirculation Pump Trip.(EOC–RPT)

System at Peach Bottom Atomic Power

Station (PBAPS), Units 2 and 3.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed TS changes do not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The addition of the EOC–RPT System will

not involve a significant increase in the

probability or consequences of an accident

previously evaluated. The EOC–RPT System

has been designed to appropriate standards

and specifications to ensure that the ability

of the plant to mitigate the effects of

accidents is maintained. Each division is

electrically, mechanically, and physically

independent to meet the single failure

criterion.

The EOC–RPT System will improve the

reactor core thermal response following a

turbine trip transient caused by either a

turbine control valve fast closure or a turbine

stop valve closure. The EOC–RPT will be

relied upon to reduce the fuel thermal

mechanical transient excursion such that fuel

thermal limits are not violated. Under

conditions when the system is inoperable,

more conservative thermal limits will be

enforced.

The new system will utilize existing RPS

[Reactor Protection System] logic to initiate

the Reactor Recirculation System (RRS)

pump trips on a turbine generator trip and a

generator load rejection event. The inputs to

RPS used by EOC–RPT will be from turbine

stop valve (TSV) limit switches and turbine

control valve (TCV) oil pressure switches.

There will be no direct interface between the

EOC–RPT System and the main turbine

control system. Thus the new system can not

initiate a turbine trip or generator load

rejection event. This change does not result

in significant increase in the probability of

events described in the UFSAR [Updated

Final Safety Analysis Report]. Additionally,

the probability of inadvertent single or dual

recirculation pump trips due to the addition

of the EOC–RPT components will not be

significantly increased by this modification.

No new challenges to the reactor coolant

pressure boundary will result from the

incorporation of the EOC–RPT System which
Federal Register / Vol. 63, No. 145 /Wednesday, July 29, 1998 / Notices 40559

could result in a significant increase in the

consequences of an accident.

2. The proposed TS changes do not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The ECO-RPT System has been designed to

appropriate standards and specifications to

ensure that no new sequence of events or

failure modes will occur such that a transient

event will escalate into a new or different

type of accident.

The PBAPS UFSAR evaluates several

recirculation pump trip events, including the

limiting case of a pump seizure. A spurious

dual EOC–RPT pump trip is similar to other

RRS pump trip events evaluated in the

UFSAR and does not represent a different

type of accident.

Additionally, this modification will not

create any new failure mode or sequences of

events that could lead to a different type of

accident than previously evaluated. The new

EOC–RPT System will not involve any new

challenges to a fission product barrier. The

EOC–RPT System does not make any changes

to the design function of the RRS. Therefore,

the new equipment installed by this

modification cannot create the possibility of

a different type than previously evaluated in

the SAR [Safety Analysis Report].

The EOC–RPT System is classified as

important-to-safety. Failure or malfunction of

the new equipment will not prevent or affect

the ability of safety-related or important-tosafety systems to respond to the design basis

accidents described in the FSAR [Final

Safety Analysis Report].

There will be no software used in the EOC–

RPT System. The system logic consists of two

electrically and physically separated trip

systems; one will be used to trip one EOC–

RPT System breaker, and the other will be

used to trip the second EOC–RPT System

breaker for each pump.

The design of this modification assures

that the new system is not susceptible to

electromagnetic (EM) emissions and will not

cause inadvertent operation of existing plant

equipment due to EM emissions.

Based on the previous discussion, the

possibility of a new or different kind of

accident from any accident previously

evaluated will not be created.

3. The proposed TS changes do not involve

a significant reduction in a margin of safety.

There are no significant reductions in any

margin of safety previously approved by the

USNRC as a result of this change to the TS.

The EOC–RPT System will ensure that fuel

thermal limits are not exceeded during the

limiting transient. In the event that the EOC–

RPT System is determined to be inoperable,

specific operating limits are provided in the

COLR. In all cases, thermal limits are not

exceeded and the margin of safety is not

significantly reduced.

The plant LOCA response will not change

for present core configurations (i.e., 9 x 9

fuel) with the EOC–RPT System installed.

For GE 8 x 8 fuel, which could be used at

a future time, there could be a small increase

in Peak Cladding Temperature (PCT). This

increase would still be well below the 2200°

F acceptance limit defined in 10 CFR 50.46.

There will be no significant reduction in

the margin of safety as previously approved

by the USNRC, since the calculated increase

in peak cladding temperature for a core

containing limit 8x8 fuel design (BP/P8 x 8R)

is a small increase above the previously

analyzed peak cladding temperature.

Additionally, this modification does not

impact the safety function of the RRS piping,

thus reactor coolant pressure boundary safety

limits are not affected.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Education

Building, Walnut Street and

Commonwealth Avenue, Box 1601,

Harrisburg, PA 17105.

Attorney for Licensee: J.W. Durham,

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

Counsel, PECO Energy Company, 2301

Market Street, Philadelphia, PA 19101.

NRC Project Director: Robert A. Capra.

PECO Energy Company, Public Service

Electric and Gas Company, Delmarva

Power and Light Company, and Atlantic

City Electric Company, Dockets Nos. 50–

277 and 50–278, PeachBottom Atomic

Power Station, Units Nos. 2 and 3, York

County, Pennsylvania

Date of application for amendments:

May 1, 1998

Description of amendment request:

The proposed amendment would revise

the Technical Specifications (TS) to

delete requirements for the functional

testing of the safety relief valves (SRVs)

during each unit startup at Peach

Bottom Atomic Power Station (PBAPS),

Units 2 and 3.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed TS changes do not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed Technical Specification

changes to the requirement for functional

testing of the SRVs during each unit startup

will not significantly increase the probability

or consequences of an accident previously

evaluated. Elimination of the functional test

will not prevent the SRVs from performing

their intended safety function. The proposed

change to delete the SRV functional test at

power should delete a potential initiator of

SRV leakage. The remaining testing and

inspections will continue to adequately

demonstrate the operability of the SRVs for

both the safety and depressurization modes.

As a result of deleting the requirement for

functional testing of the SRVs during each

unit startup and replacing these requirements

with the proposed tests contained TS SR

[surveillance requirement] 3.4.3.2 and

3.5.1.12, the only change in the frequency of

testing of the SRV components is that the

main valve disc of the SRVs will be tested

every two cycles (approximately four years)

as compared to the current one cycle

(approximately two years) frequency. As

described above, the lift test of the main

valve disc is currently performed at an offsite

facility. A review of offsite testing data for

the years 1987 through 1998 was performed

for the PBAPS, Units 2 and 3 SRVs. Since the

design of the SRVs is to ensure operation of

the overpressurization protection and the

ADS [Automatic Depressurization System]

function is to reduce reactor pressure during

a small break LOCA [loss-of-coolant

accident], the review consisted of looking for

any failures of the main valve disc to stroke

open during setpoint actuation. This review

consisted of reviewing ‘‘as-found’’ test data

since any failures following a rebuild would

be found during the final certification testing.

Based on a review of as-found data, it was

concluded that there were no reported cases

of the main disc failing to open during

setpoint pressure testing. Therefore, deleting

the requirement for functional testing of the

SRVs during each unit startup is not

expected to negatively impact these test

results.

Therefore, eliminating the functional test is

not expected to negatively impact these test

results or involve a significant increase in the

probability of an accident previously

evaluated.

As discussed in the PBAPS, Units 2 and 3

Updated Final Safety Analyses Report

(UFSAR), analyzed events resulting in a

nuclear system pressure increase, such as

MSIV [main steam line isolation valve]

closure, generator load rejection, turbine trip,

failure of the turbine bypass valves to open,

and loss of main condenser vacuum, take

credit for the SRVs opening to mitigate the

consequences of these events. The proposed

changes will not increase the consequences

of these events, since a series of remaining

tests will ensure all SRV components will

function. The SRVs will therefore be capable

of performing their design functions.

SRV second stage valve leakage can be

increased as a result of corrosion/debris

introduced on the seating area surface.

Second stage leakage, if allowed to

continually increase, will eventually start to

depressurize the volume above the SRV main

valve piston to the extent that sufficient

differential pressure will lift the main valve

disc. Reactor vessel coolant inventory

decrease due to an inadvertent opening of a

Safety Relief Valve is an abnormal operating

transient event. This event can be a precursor

to fuel failure due to gradual loss of coolant,

and the mitigation is similar to the small

break LOCA. Under the proposed change, it

is expected that the probability of SRV

leakage will decrease, thus the probability of

occurrences of an inadvertent SRV actuation

is reduced, therefore reducing the probability

or
40560 Federal Register / Vol. 63, No. 145 /Wednesday, July 29, 1998 / Notices

consequences of an accident previously

evaluated.

2. The proposed TS changes do not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The SRVs will not be operated or tested in

a manner contrary to their design. As a result,

no new mode of operation is introduced.

Therefore, the revised testing will not create

a new failure mode of the SRVs which could

create the possibility of a new or different

kind of accident from any previously

evaluated. Since other tests, taken together,

confirm the entire SRV assembly functions

adequately, this proposed change is justified.

The proposed change to delete the SRV

functional test at power will not impact the

ability of the SRV to open and provide their

intended safety function.

3. The proposed TS changes do not involve

a significant reduction in a margin of safety.

By removing the Technical Specification

requirements to perform the in-situ

functional testing during startup, the

probability of inadvertently opening of a SRV

should be reduced through the elimination of

a potential initiator of SRV second stage disc

leakage and subsequent erosion. This

Technical Specification change will aid in

decreasing SRV leakage and improve SRV

reliability at power operations. Eliminating

the SRV in-situ functional test during startup

will increase the margin of safety during

operations, transients, or accidents.

Remaining surveillance testing and

inspections assure each component necessary

for successful opening of the SRV function

properly as designed.

Removal of the functional test will not

negatively impact the Technical

Specifications lift setpoints of the SRVs

necessary for the function of the safety mode.

The functional test does not completely test

the safety mode of the SRV which is based

on the Technical Specifications lift setpoints.

Offsite testing at operating steam pressure

ensures the operability of the SRV pilot,

second stage, and main valve function. The

valves are refurbished and post maintenance

testing is performed at a steam pressure of

1040 psig. Upon successful test completion,

the valve receives written certification from

the lab and is returned to PBAPS for

reinstallation. To receive certification, the

valve must have zero main seat leakage and

meet the acceptance criteria for setpoint

pressure. These tests satisfy the requirements

of the PBAPS IST [Inservice Testing] Program

and TS. The tests contained in the proposed

TS SR 3.4.3.2 and 3.5.1.12 will verify the

operation of the solenoid and second stage

disc movement of all 11 SRVs in the

depressurization mode.

The remaining segments of the SRV tests

verify the ability of the SRV logic. In

summary, this change will not involve a

significant reduction in the margin of safety,

because of the reduction in SRV degradation,

and the remaining tests confirm the valves

will function properly when required.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, this appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Education

Building, Walnut Street and

Commonwealth Avenue, Box 1601,

Harrisburg, PA 17105.

Attorney for Licensee: J.W. Durham,

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

Counsel, PECO Energy Company, 2301

Market Street, Philadelphia, PA 19101.

NRC Project Director: Robert A. Capra.

Power Authority of The State of New

York, Docket No. 50–286, Indian Point

Nuclear Generating Unit No. 3,

Westchester County, New York

Date of amendment request: June 16,

1998.

Description of amendment request:

The proposed change would relocate the

Safety Review Committee (SRC) review,

audit, and related record keeping

requirements from the Technical

Specifications (TSs) to Chapter 17 of the

Final Safety Analysis Report (FSAR).

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 proposed license amendment

involve a significant increase in the

probability or consequences of an accident

previously analyzed?

Response: This amendment application

does not involve a significant increase in the

probability or consequences of an accident

previously analyzed. The relocation of the

SRC review, audit, and related record

keeping requirements from the TS to the

FSAR does not alter the performance or

frequency of these activities. Future changes

to the QA [Quality Assurance] program,

located in Chapter 17 of the FSAR, which

constitute a reduction in commitments, are

governed by 10 CFR 50.54(a). Therefore,

sufficient controls for these requirements

exist and these changes do not involve a

significant increase in the probability or

consequences of an accident previously

analyzed.

(2) Does the proposed license amendment

create the possibility of a new or different

kind of accident from any accident

previously evaluated?

Response: This amendment application

does not create the possibility of a new or

different kind of accident from any accident

previously evaluated. The proposed changes

involve the relocation of SRC requirements

from the TS to the FSAR. Relocation of these

requirements does not affect plant equipment

or the way the plant operates. The reviews,

audits, and record keeping will continue to

be performed in the identical manner as they

are currently being performed. Therefore, the

proposed revisions cannot create a new or

different kind of accident.

(3) Does the proposed amendment involve

a significant reduction in a margin of safety?

Response: This amendment application

does not involve a significant reduction in a

margin of safety. The requested Technical

Specification revisions relocate SRC review,

audit and related record keeping

requirements from the TS to the FSAR. These

requirements are not being altered by this

relocation. The reviews, audits, and record

keeping will continue to be performed in the

identical manner as they are currently being

performed. Any changes to these

requirements which constitute a reduction in

commitments will be processed in

accordance with 10 CFR 50.54(a). Therefore,

the proposed changes do not involve a

significant reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10601.

Attorney for licensee: Mr. David E.

Blabey, 10 Columbus Circle, New York,

New York 10019.

NRC Project Director: S. Singh Bajwa,

Director.

Power Authority of the State of New

York, Docket No. 50–333, James A.

FitzPatrick Nuclear Power Plant,

Oswego County, New York.

Date of amendment request: July 10,

1998.

Description of amendment request:

The proposed changes would relocate

portions of reactor coolant chemistry

requirements from the technical

specifications (TSs) to licenseecontrolled procedures. Changes to the

relocated requirements will then be

controlled by the provisions of 10 CFR

50.59.

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 TS amendment will not

significantly increase the probability or

consequences of any previously evaluated

accidents.

The proposed changes simplify the TS,

meet regulatory requirements for relocated

TS, and implement the recommendations of

the Commission’s Final Policy Statement on

TS improvements. Future changes to these

requirements will be controlled by 10 CFR
Federal Register / Vol. 63, No. 145 /Wednesday, July 29, 1998 / Notices 40561

50.59. The proposed changes are

administrative in nature and do not involve

any modification to any plant equipment or

affect plant operation. Therefore, the

proposed changes do not involve a

significant increase in the probability or

consequences of any previously evaluated

accident.

2. The proposed TS amendment will not

create the possibility of a new or different

kind of accident.

The proposed changes are administrative

in nature, do not involve any physical

alterations to any plant equipment, and cause

no change in the method by which any safety

related system performs its function.

Therefore, this proposed TS amendment will

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed TS amendment will not

involve a significant reduction in a margin of

safety.

The proposed changes are administrative

in nature, will not alter the basic regulatory

requirements, and do not affect any safety

analyses. Therefore, no margin of safety is

reduced as a result of these changes.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Attorney for licensee: Mr. David E.

Blabey, 1633 Broadway, New York, New

York 10019.

NRC Project Director: S. Singh Bajwa,

Director.

Tennessee Valley Authority, Docket No.

50–390 Watts Bar Nuclear Plant, Unit 1,

Rhea County, Tennessee

Date of amendment request: June 5,

1997, and supplemented April 21, 1998.

Description of amendment request:

Part 1—DG Online Testing:

The proposed amendment involves

the testing of the standby diesel

generators (DGs) and revises the Watts

Bar Unit 1 (WBN) Technical

Specifications (TSs) to allow additional

testing of the DGs on-line during

MODES 1 and 2. The proposed changes

affect Surveillance Requirement (SR)

3.8.1.14. The testing performed for this

surveillance fulfills the requirements of

Regulatory Guide 1.9, ‘‘Selection,

Design, Qualification, and Testing of

Emergency Diesel Generator Units Used

as Class 1E Onsite Electric Power

Systems at Nuclear Power Plants.’’ This

testing is performed once every 18

months to ensure that the DGs can start

and run continuously for an interval of

not less than 24 hours. Specifically, the

proposed amendment revises SR

3.8.1.14 and its associated Bases to

delete the note which prohibits the

performance of the on-line 24 hour test

during MODES 1 or 2.

Part 2—DG Battery Testing:

As currently written, the TSs permit

testing of the DG batteries and chargers

only during MODES 5 and 6 when

operability of all four DGs is not

required. The proposed amendment

would revise the Watts Bar Unit 1 TSs

to allow testing of the DG batteries and

battery chargers during MODES 1, 2, 3,

and 4 as well. Implementation of these

changes will require entry into Action

B.4 of TS 3.8.1 for the affected diesel.

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:

Part 1—DG Online Testing

1. Operation of the facility in accordance

with the proposed amendment would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed amendment to allow the 24-

hour DG endurance run to be conducted

during any mode of operation does not

significantly increase the probability or

consequences of an accident previously

evaluated in Chapter 15 of the FSAR [Final

Safety Analysis Report] since the capability

to safely shutdown the plant following a

LOOP [loss of offsite power], LOCA [loss-ofcoolant accident] or LOCA/LOOP coincident

with a single failure is maintained

throughout the surveillance test. The 24-hour

endurance test does not disable any of the

automatic actuations and interlocks of the DG

control functions, nor prevent the satisfactory

completion of the LOOP or LOCA/LOOP

loading sequence if a LOOP or LOCA signal

is received at any time during the test.

Required Class-1E onsite power operability

during normal operation, shutdown cooling,

loss of offsite power, and accident conditions

will be the same.

In addition, the performance of proposed

Surveillance Requirement 3.8.1.14 during

MODES 1 or 2 will not significantly increase

the consequences of perturbations to any of

the electrical distribution systems that could

result in a challenge to steady state operation

or to plant safety systems. Performance of

proposed Surveillance Requirement 3.8.1.14

during MODES 1 or 2, or failure of the

surveillance, will not cause, or result in, an

anticipated operational occurrence with

attendant challenges to plant safety systems

that has not been previously analyzed for the

existing monthly surveillances.

Therefore, TVA concludes that the above

change does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

2. Operation of the facility in accordance

with the proposed amendment would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The requested changes do not result in a

new or different kind of accident from that

previously analyzed in WBN’s Final Safety

Analysis Report. The changes propose to

eliminate restrictions of the plant operating

modes in which standby DG system testing

may be performed but does not change the

type of testing performed and are not due to

modification of the system design. NRC’s

assessment of the testing of the DGs in the

configuration proposed is documented in

Section 8.3.1.12 of Supplements 13 and 14 of

the Safety Evaluation Report and in letters

dated June 20, 1991, and March 28, 1994.

3. Operation of the facility in accordance

with the proposed amendment would not

involve a significant reduction in the margin

of safety.

As previously stated, performance of

proposed Surveillance Requirement 3.8.1.14

during Modes 1 or 2 will not cause, or result

in, an anticipated operational occurrence

with attendant challenges to plant safety

systems that has not been previously

analyzed for the existing monthly

surveillances. Therefore, implementation of

the proposed amendment will not reduce the

margin of safety for this system.

Part 2—DG Battery Testing

1. Operation of the facility in accordance

with the proposed amendment would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed changes to the TSs apply

only to the DG battery system and do not in

any way affect the vital battery system or

safety system loads supplied by the vital

battery system. The changes do not result in

a condition where the design or function of

the DGs or DG battery systems would be

modified. The DG battery subsystems supply

only the control and field flashing power to

support a single DG and do not supply any

other unrelated system loads or functions.

Therefore, manipulation of the DG battery

system is not a credible means of perturbing

the vital power distribution system and

challenging safety systems. In addition, the

surveillances for the DG batteries are

required to be performed only once every 18

months.

A DG declared inoperable due to the

testing must be returned to operable status

within 72 hours in accordance with Action

B.4 of TS 3.8.1. To ensure this could be

achieved, the results of previous

performances of the SRs were reviewed.

From this review, it was established that in

accordance with LCO [limiting condition for

operation] 3.8.6, Table 3.8.6–1, Note c, the

batteries can be restored within 72 hours to

a condition where the charging current is less

than 1 ampere. Achieving this charging

current for the DG batteries is acceptable for

meeting specific gravity limits following a

battery recharge for a maximum of 31 days.

In addition, the DG sets are occasionally

removed from the standby condition to

perform preventative and/or corrective

maintenance. The intent is to perform this
40562 Federal Register / Vol. 63, No. 145 /Wednesday, July 29, 1998 / Notices

testing in conjunction with other required

maintenance activities such that adverse

effects on diesel unavailability are

minimized. Compliance with the 10 CFR

50.65 Maintenance Rule program

requirements for diesel unavailability

ensures that any diesel inoperability incurred

by this change is minimized.

2. Operation of the facility in accordance

with the proposed amendment would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The requested changes do not result in a

new or different kind of accident from that

previously analyzed in WBN’s Final Safety

Analysis Report. The changes propose to

eliminate restrictions of the plant operating

modes in which DG battery system testing

may be performed but does not change the

type of testing performed and are not due to

modification of the system design. The

requested changes will result in a DG being

declared inoperable in accordance with

Action B.4 of TS 3.8.1 for the duration of the

testing, but does not impact the existing time

limitations for the LCO. This change does not

alter system performance and does not

introduce any new accident initiators or

scenarios.

3. Operation of the facility in accordance

with the proposed amendment would not

involve a significant reduction in the margin

of safety.

The proposed amendment concerns only

the conduct of testing but does not in any

way affect the performance parameters of the

safety system or in any way affect the ability

of the system to perform its safety function

of providing control and field flashing power

for the DGs. Consequently, operation of the

facility in accordance with the requested

changes would 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: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga,

TN 37402.

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET l0H,

Knoxville, Tennessee 37902.

NRC Project Director: Frederick J.

Hebdon.

Tennessee Valley Authority, Docket No.

50–390 Watts Bar Nuclear Plant, Unit 1,

Rhea County, Tennessee

Date of amendment request: June 26,

1998

Description of amendment request:

The proposed amendment would revise

the Watts Bar Nuclear Plant (WBN)

Technical Specifications (TS) and

associated Bases to delete the power

range neutron flux high negative rate

reactor trip function based on the

analysis provided in Westinghouse

Electric Corporation topical report

WCAP–11394–P–A, ‘‘Methodology for

the Analysis of the Dropped Rod

Event.’’

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

A. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The negative flux rate trip deletion does

not increase the probability or consequences

of core damage accidents resulting from

dropped RCCA [rod cluster control assembly]

events previously analyzed. The safety

functions of other safety related systems and

components, which are related to accident

mitigation, have not been altered. All other

primary protection (reactor trip and ESF)

functions are not impacted by the

elimination of the negative flux rate trip

function. The consequences of accidents

previously evaluated in the FSAR [final

safety analysis report] are unaffected by this

proposed change because no change to any

equipment response or accident mitigation

scenario has resulted. There are no additional

challenges to fission product barrier integrity.

No new radiological analyses are required.

Therefore the proposed change will have no

effect on the probability or consequences of

accidents previously evaluated.

B. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The negative flux rate trip deletion does

not create the possibility of a new or different

kind of accident than any accident already

evaluated in the FSAR. No new accident

scenarios, failure mechanisms, or limiting

single failures are introduced as a result of

this proposed change. The proposed

modification does not challenge the

performance or integrity of any safety-related

systems.

It has been demonstrated that the function

of the negative flux rate trip can be

eliminated by the approved methodology

described in WCAP 11394–P–A. A Watts bar

specific analysis has confirmed that for the

dropped RCCA and dropped RCCA bank

event, no direct reactor trip or automatic

power reduction is required to meet the DNB

[departure from nucleate boiling] licensing

basis for this Condition II event. The negative

flux rate trip function is not credited as a

backup for any other Chapter 15 event. Thus,

this change does not create the possibility of

a new or different kind of accident from any

previously evaluated.

C. The proposed amendment does not

involve a significant reduction in a margin of

safety.

The margin of safety associated with the

acceptance criteria for any postulated WBN

accident is unchanged. It has been

demonstrated that the function of the

negative flux rate trip can be eliminated by

the approved methodology described in

WCAP 11394–P–A. Watts Bar specific

analysis has confirmed that the dropped

RCCA and dropped RCCA bank acceptance

criteria (DNB) continue to be met.

Conformance to the regulatory criteria for

plant operation with the negative flux rate

trip deletion is demonstrated, and regulatory

limits (DNB) are not exceeded. The

modification will have no effect on the

availability, operability, or performance of

the safety-related systems and components.

Therefore, the proposed license amendment

does not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga,

TN 37402.

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET l0H,

Knoxville, Tennessee 37902.

NRC Project Director: Frederick J.

Hebdon.

The Cleveland Electric Illuminating

Company, Centerior Service Company,

Duquesne Light Company, Ohio Edison

Company, Pennsylvania Power

Company, Toledo Edison Company,

Docket No. 50–440, Perry Nuclear Power

Plant, Unit 1, Lake County, Ohio

Date of amendment request: June 30,

1998.

Description of amendment request:

The proposed license amendment

revises Technical Specification 3.1.7,

‘‘Standby Liquid Control System.’’ The

purpose of the proposed change is to

increase the boron concentration in the

Standby Liquid Control System for the

Perry Nuclear Power Plant Cycle 8 fuel

design, and to provide margin for future

cycles as required.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration which is presented below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed change will not significantly

increase the probability or consequences of

an accident previously evaluated. The change

will only vary the ratio of borax to boric acid

that resides within the Standby Liquid
Federal Register / Vol. 63, No. 145 /Wednesday, July 29, 1998 / Notices 40563

Control System (SLCS) as the neutron

absorber.

Changing the definition of the solution

from a mixture of Sodium Pentaborate having

a molar ratio of 0.200, to a mixture of borax

and boric acid having a nominal molar ratio

of 0.229, does not degrade the stability of the

solution, change the mixing accuracy

requirements, or reduce the temperature

margins that might add to the risk of solution

crystallization. For each cycle, the reload

safety analysis confirms that the SLCS boron

concentration will satisfy the Technical

Specification requirements for the Perry

Nuclear Power Plant (PNPP).

The 5°F margin of safety for solution

solubility will continue to be maintained and

supported by the Containment Building

ambient temperatures and additionally

supplemented by auto initiated heating on

the SLCS tank and piping. The chosen borax

and boric acid molar ratio will continue to

maintain a limiting chemical addition mass,

to the plus 5°F solubility limit, greater than

or equal to the current 0.200 mixture. Any

inaccuracies associated with tank

temperature, tank volume, chemical analysis,

and initial and subsequent chemical

additions to the tank will also remain the

same.

The primary reactivity control system for

postulated accident conditions is the control

rod system. The SLCS is a redundant

reactivity control system to the control rod

system and is used in special plant capability

demonstration events cited in Appendix A of

the Updated Safety Analysis Report (USAR),

Chapter 15, which are extremely low

probability non-design basis postulated

incidents. There are no postulated accidents

evaluated in USAR Chapter 15 that take

credit for two or more reactivity control

systems preventing or mitigating each

accident. There is no increase to the

radiological consequences of postulated

incidents with the proposed change.

With the implementation of this proposed

change, the SLCS will continue to operate

and perform to all of its current requirements

for providing shutdown margin under

operating and ATWS conditions per 10 CFR

50, Appendix A, General Design Criterion

(GDC) 26 and 10 CFR 50.62. The proposed

change will not alter the operation of any

plant equipment assumed to function in

response to an analyzed event or otherwise

increase its failure probability. Therefore, the

proposed change does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed change does not impact the

operation of the SLCS or the function of any

of the active components. No new system

interactions are created by this change and

any parameters or conditions that could

contribute to the initiation of accidents

different than those already evaluated in the

USAR are not impacted. The change will

only vary the ratio of borax to boric acid that

resides within the SLCS as the neutron

absorber. The proposed values for solution

molar ratio and boron concentration ensure

that solution temperature margins are

maintained greater than or equal to the

current required margin to prevent solution

crystallization. As a result, no new failure

modes are being introduced.

Changing the definition of the solution

from Sodium Pentaborate having a nominal

molar ratio of 0.200, to a mixture of borax

and boric acid having a nominal molar ratio

of 0.229 does not degrade the stability of the

solution, change the mixing accuracy

requirements, or reduce the temperature

margins that might add to the risk of solution

crystallization.

Sufficient margin will be maintained to

allow for expected deviations in the molar

ratio and boron weight as the result of

variations in product composition, test

measurement inaccuracies, and for chemical

addition inaccuracies. The boron

concentration required within the SLC

system to meet the required shutdown

margin, will continue to be determined for

each fuel cycle as part of the reload safety

analysis per Technical Specifications. The

borax and boric acid concentration will

remain controlled via the Technical

Specification Surveillance Requirements and

the associated administrative procedures,

USAR text, and existing licensing

commitments.

The SLC system will meet its design basis

requirements for the weight of boron injected

and for maintaining the required temperature

margin for system operation. As the result of

the proposed change to increase the

minimum boron concentration, a new

minimum required SLC pump flow rate was

determined for compliance with the NRC

ATWS Rule 10 CFR 50.62.

The proposed change meets current

regulations, maintains the fundamental safety

principles of plant design, and the associated

margins of safety. With the implementation

of the proposed change, the SLCS will

continue to operate and perform to all of its

current requirements for providing shutdown

margin under operating and ATWS

conditions per 10 CFR 50, Appendix A, GDC

26 and 10 CFR 50.62. As a result, no new

failure modes are being introduced. There are

no changes in the methods governing normal

plant operations, nor are the methods used to

respond to plant transients altered.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin to safety.

The required margin of safety for the SLCS

solution ensures an adequate margin of

solubility such that no precipitation will

occur in the SLC storage tank. The current

margin is provided by maintaining a

minimum solution temperature that is no less

than the saturation temperature

corresponding to the concentration of the

solution in the storage tank plus 5°F.

This 5°F provides the adequate margin for

inaccuracies associated with tank

temperature, tank volume, chemical analysis,

and initial and subsequent chemical

additions to the tank. The proposed change

does not impact the inaccuracies associated

with tank temperature, tank volume,

chemical analysis, and initial and subsequent

chemical additions to the tank. The new

analytical design values for the molar ratio

and boron concentration will continue to

maintain the solution temperature margins in

excess of the current minimum specified to

prevent solution crystallization.

Ambient temperatures within the building

that houses the SLC storage tank, the

Containment Building, will maintain the

solution temperature. Additionally, the

solution temperature is maintained by the

presence of auto initiated tank heaters and

pipe heat tracing. The 5°F margin will be

maintained with the new SLCS mole ratio

and higher boron concentration with the

existing instrument setpoints and

administrative controls.

The proposed change maintains the same

reactor shutdown margin for the next fuel

cycle and does not reduce the margin of

safety for any system parameter as defined in

the bases for the Technical Specifications.

The proposed change will not physically

alter the SLCS’s physical configuration or

components or introduce new system

interactions that could produce any

parameters or conditions that could

contribute to a reduction of safety for any

other system or scenario. The change will

only vary the ratio of borax to boric acid that

resides within the SLCS as the neutron

absorber.

Therefore, with the implementation of this

proposed change, the SLCS will continue to

operate and perform to all of its current

requirements for providing shutdown margin

under operating and ATWS conditions per 10

CFR 50, Appendix A, GDC 26 and 10 CFR

50.62 and does not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Perry Public Library, 3753

Main Street, Perry, OH 44081.

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts & Trowbridge,

2300 N Street, NW., Washington, DC

20037.

NRC Project Director: Ronald R.

Bellamy (Acting).

Vermont Yankee Nuclear Power

Corporation, Docket No. 50–271,

Vermont Yankee Nuclear Power Station,

Vernon, Vermont

Date of amendment request: June 30,

1998.

Description of amendment request:

The licensee proposes to delete the

calibration requirements for emergency

core cooling actuation

instrumentation—core spray (CS)

subsystem and low pressure coolant

injection (LPCI) system auxiliary power
40564 Federal Register / Vol. 63, No. 145 /Wednesday, July 29, 1998 / Notices

monitor since the relays operate from a

switched input and functional testing is

sufficient to demonstrate the relay

pickup/dropout capability.

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 will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated:

The proposed change does not involve a

change to the plant design or operation. The

Auxiliary Power Monitor logic relays

installed are tested to fully demonstrate

operability without performance of a

calibration on the pickup voltage value. The

design intent of the relays is to start LPCI and

CS pumps as soon as possible without

causing loss of the normal or emergency

power supplies and within the time frames

specified in the LOCA analysis of record. The

proposed change does not affect any of the

parameters or conditions that contribute to

initiation of any accidents previously

evaluated. Thus, the proposed change cannot

increase the probability of an accident

previously evaluated.

The proposed change does not involve a

change in the operation of the relays

controlling [Residual Heal Removal] RHR

and CS Pump start with normal power

available nor the initial RHR pump start on

a LOCA with normal power not available or

the time delay start of the remaining RHR or

CS pumps with normal power not available.

Failure of the relays to pickup would still

result in the start sequence for normal power

not available. The logic for both start

sequences is verified independent of an

instrument calibration and is consistent with

the LOCA analysis and the EDG load

analysis, therefore, the proposed change does

not significantly increase the consequences

of any accident previously evaluated.

2. The proposed amendment will not

create the possibility of a new or different

kind of accident from any previously

evaluated:

This proposed change will not involve any

physical changes to plant systems, structures

or components (SSC), or the manner in

which these SSCs are operated or

maintained. The calibration requirement has

previously been considered to be met by

performance of the Simulated Automatic

Actuation Test. Deletion of the calibration

requirement will not affect the RHR or CS

Pumps starting on a LOCA signal, with or

without an [Loss of Normal Power] LNP. The

operability of the Auxiliary Power Monitor

relays will still be tested under the

Functional test and Trip System Logic and

Simulated Automatic Actuation tests at the

frequencies specified. Therefore, this change

will not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

3. The proposed amendment will not

involve a significant reduction in the margin

of safety:

This proposed change to delete the

calibration requirement for the CS and LPCI

Auxiliary Power Monitor functions will not

change operation of the RHR or CS Pump

start sequences on a LOCA signal, with or

without normal power available. The

instantaneous logic sequence relays and time

delay relays will function to initiate RHR and

CS Pump start as designed. RHR and CS

Pump start times will remain within the

LOCA Safety Evaluation of record.

Operability of the relays and associated

circuitry are still demonstrated by the

Functional test and associated Trip System

Logic and Simulated Automatic Actuation

tests. Therefore, this change does not involve

a significant reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

Attorney for licensee: Mr. David R.

Lewis, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037–1128.

NRC Project Director: Cecil O.

Thomas.

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 No. 50–261, H.B.Robinson Steam

Electric Plant, Unit 2, Darlington

County, South Carolina

Date of amendment request: June 26,

1998.

Brief Description of amendment: The

proposed amendment would revise

Technical Specification (TS) 3.7.8,

‘‘Ultimate Heat Sink (UHS),’’ to permit

an 8-hour delay in UHS temperature

restoration period prior to entering the

plant shutdown required actions. Also,

for the duration of the restoration,

service water system (SWS) temperature

will be monitored hourly, and should

the temperature exceed 99 degrees F,

the plant will enter TS 3.7.8 required

action A.1, and be in MODE 3 within 6

hours.

Date of publication of individual

notice in the Federal Register: July 8,

1998 (63 FR 36967).

Expiration date of individual notice:

July 22, 1998, for comments; August 7,

1998, for hearings.

Local Public Document Room

location: Hartsville Memorial Library,

147 West College Avenue, Hartsville,

South Carolina 29550.

Duke Energy Corporation, Docket Nos.

50–369 and 50–370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg

County, North Carolina

Date of amendment request: March 3,

1998, as supplemented by letters dated

April 24 and May 7, 1998.

Description of amendment request:

The proposed amendments would

revise Figure 5.1–1 of the Technical

Specifications (TS) to show the new

location of the meteorological tower.

The meteorological tower will be

relocated to a new location to facilitate

use of the current location as a

construction site. The proposed TS

change does not change the related TS

Section 5.1.1.

Date of publication of individual

notice in Federal Register: June 29, 1998

(63 FR 35293).

Expiration date of individual notice:

July 29, 1998.

Local Public Document Room

location: J. Murrey Atkins Library,

University of North Carolina at

Charlotte, 9201 University City

Boulevard, Charlotte, North Carolina.

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
Federal Register / Vol. 63, No. 145 /Wednesday, July 29, 1998 / Notices 40565

Hazards Consideration Determination,

and Opportunity for A Hearing in

connection with these actions was

published in the Federal Register as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the applications for

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

the Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, the Gelman Building, 2120 L

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

local public document rooms for the

particular facilities involved.

Detroit Edison Company, Docket No.

50–341, Fermi 2, Monroe County,

Michigan

Date of application for amendment:

January 28, 1998 (NRC–98–0002).

Brief description of amendment: The

amendment revises technical

specification surveillance requirements

4.8.2.1.a.2, 4.8.2.1.b, and 4.8.2.1.c.4 to

accommodate new limits associated

with the design of the replacement

Division II 130/260-volt dc battery.

Date of issuance: July 9, 1998.

Effective date: July 9, 1998, with full

implementation prior to restart from the

sixth refueling outage.

Amendment No.: 121.

Facility Operating License No. NPF–

43: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: February 25, 1998 (63 FR

9597).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated July 9, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Monroe County Library

System, Ellis Reference and Information

Center, 3700 South Custer Road,

Monroe, Michigan 48161.

Detroit Edison Company, Docket No.

50–341, Fermi 2, Monroe County,

Michigan

Date of application for amendment:

December 10, 1997 (NRC–97–0105), as

supplemented January 28 and April 9,

1998.

Brief description of amendment: The

amendment revises Technical

Specification (TS) 2.2.1, ‘‘Reactor

Protection System Instrumentation

Setpoints,’’ TS 3.3.1, ‘‘Reactor

Protection System Instrumentation,’’ TS

3.3.6, ‘‘Control Rod Block

Instrumentation,’’ TS 3.4.1.1,

‘‘Recirculation Loops,’’ and the

associated Bases to accommodate an

upgrade of the power range neutron

monitoring system. The amendment

also revises the first page of Table 3.3.6–

2 to correct a typographical error in the

title.

NRC has also granted the request of

Detroit Edison Company to withdraw a

portion of its December 10, 1997,

application. The proposed change

would have revised TS Surveillance

Requirement 4.3.1.3 and its associated

Bases to indicate response time testing

is performed only on applicable

channels. However, following

discussions with the NRC staff, the

licensee withdrew the proposed change

in a letter dated April 9, 1998 (NRC–98–

0037). For further details with respect to

this action, see the application for

amendment dated December 10, 1997,

as supplemented above, and the

licensee’s letter dated April 9, 1998,

which withdrew this portion of the

application for license amendment, and

the staff’s Safety Evaluation enclosed

with the amendment. The above

documents 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

listed below.

Date of issuance: July 13, 1998.

Effective date: July 13, 1998, with full

implementation prior to restart from the

sixth refueling outage. Implementation

of this amendment shall include

preparation of Design Calculation DC–

5721, Volume I, and performance of a

human factors review for the

installation of the plant modification as

described in the licensee’s application

dated December 10, 1997, as

supplemented January 28 and April 9,

1998, and as evaluated in the staff’s

safety evaluation attached to this

amendment.

Amendment No.: 122.

Facility Operating License No. NPF–

43: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: January 14, 1998 (63 FR 2279).

The January 28 and April 9, 1998, letters

provided clarifying information and

updated TS pages that were within the

scope of the original Federal Register

notice and did not change the staff’s

initial proposed no significant hazards

considerations determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated July 13, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Monroe County Library

System, Ellis Reference and Information

Center, 3700 South Custer Road,

Monroe, Michigan 48161.

Duke Energy Corporation, 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:

April 20, 1998.

Brief description of amendments: The

amendments revise Tables 3.3–3 and

4.3–2 of the Technical Specifications of

each unit, correcting the operation mode

applicability of the control room area

ventilation actuation logic and relays

from ‘‘All’’ to ‘‘1, 2, 3, 4.’’

Date of issuance: July 9, 1998.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment Nos.: 167—Unit 1; 159—

Unit 2.

Facility Operating License Nos. NPF–

35 and NPF–52: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: May 20, 1998 (63 FR 27761).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated July 9, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: York County Library, 138 East

Black Street, Rock Hill, South Carolina.

Entergy Operations, Inc., Docket Nos.

50–313 and 50–368, Arkansas Nuclear

One, Units 1 and 2, Pope County,

Arkansas

Date of amendment request: October

2, 1996.

Brief description of amendments: The

amendments revise the ANO–1&2 TSs

by relocating selected TS requirements

related to instrumentation from the TS

to the Updated Final Safety Analysis

Report. The NRC provided guidance to

all holders of operating licenses or

construction permits for nuclear power

reactors on the proposed TS changes in

Generic Letter 95–10, ‘‘Relocation of
40566 Federal Register / Vol. 63, No. 145 /Wednesday, July 29, 1998 / Notices

Selected Technical Specifications

Requirements Related to

Instrumentation,’’ dated December 15,

1995.

Date of issuance: July 13, 1998.

Effective date: July 13, 1998, to be

implemented within 30 days.

Amendment Nos.: 192 and 191.

Facility Operating License Nos. DPR–

51 and NPF–6: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: January 15, 1997 (62 FR 2188).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated July 13, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801.

Entergy Operations, Inc., Docket No. 50–

382, Waterford Steam Electric Station,

Unit 3, St. Charles Parish, Louisiana

Date of amendment request: March

27, 1997, as supplemented by letters

dated April 3, July 21, October 23,

November 13, and December 12, 1997,

January 21, January 29, March 23, May

1, May 19, and May 21, May 28, and

June 12, 1998.

Brief description of amendment: The

amendment changes Appendix A

Technical Specification by increasing

the Spent Fuel Pool storage capacity

from 1088 to 2398 fuel assemblies and

by increasing the maximum fuel

enrichment from 4.9 w/o (weight

percent) to 5.0 w/o U–235.

Date of issuance: July 10, 1998.

Effective date: July 10, 1998.

Amendment No.: 144.

Facility Operating License No. NPF–

38: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: December 2, 1997 (62 FR

63732).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated July 10, 1998.

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,

Docket Nos. 50–250 and 50–251, Turkey

Point Plant Units 3 and 4, Dade County,

Florida

Date of application for amendments:

March 12, 1998.

Brief description of amendments: The

amendments revised Turkey Point Units

3 and 4 Facility Operating Licenses and

Technical Specifications to remove

certain license conditions and oudated

references, and to incorporate an

organizational change.

Date of issuance: July 9, 1998.

Effective date: July 9, 1998.

Amendment Nos.: 198 and 192.

Facility Operating Licenses Nos. DPR–

31 and DPR–41: Amendments revised

Turkey Point Units 3 and 4 Facility

Operating Licenses and Technical

Specifications.

Date of initial notice in Federal

Register: April 8, 1998 (67 FR 17225).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated July 9, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Florida International

University, University Park, Miami,

Florida 33199.

Niagara Mohawk Power Corporation,

Docket Nos. 50–220 and 50–410, Nine

Mile Point Nuclear Station Unit Nos. 1

and 2, Oswego County, New York

Date of applications for amendments:

May 15, 1998 (two letters, one for each

unit).

Brief description of amendment: The

amendments change administrative

sections of the Technical Specifications

to reflect a restructuring of licensee’s

Nuclear Division upper management

organization.

Date of issuance: July 7, 1998.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment Nos.: 162 and 83.

Facility Operating License Nos. DPR–

63 and NPF–69: Amendments revise the

Technical Specifications.

Date of initial notice in Federal

Register: June 2, 1998 (63 FR 30026).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated July 7, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston & Stawn,

1400 L Street, NW, Washington, DC

20005–3502.

NRC Project Director: S. Singh Bajwa.

Public Service Electric & Gas Company,

Docket Nos. 50–272 and 50–311, Salem

Nuclear Generating Station, Unit Nos. 1

and 2, Salem County, New Jersey

Date of application for amendments:

September 28, 1995, and April 23, 1998.

Brief description of amendments: The

amendments revise Technical

Specification 3/4.8.1.2, ‘‘Electrical

Power Sources—Shutdown,’’ by adding

a note to surveillance requirement

4.8.1.2 that identifies those

surveillances which are required to be

performed during Modes 5 and 6 (cold

shutdown and refueling, respectively).

Date of issuance: July 14, 1998.

Effective date: July 14, 1998.

Amendment Nos.: 212 and 192.

Facility Operating License Nos. DPR–

70 and DPR–75. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: November 8, 1995 (60 FR

56369).

The April 23, 1998, letter 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 July 14,1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Salem Free Public Library, 112

West Broadway, Salem, NJ 08079.

Rochester Gas and Electric Corporation,

Docket No. 50–244, R. E. Ginna Nuclear

Power Plant, Wayne County, New York

Date of application for amendment:

September 29, 1998, as supplemented

February 6, 1998, April 17, 1998, and

June 4, 1998.

Brief description of amendment: This

amendment revises the allowable value

and trip setpoint for the main steam

isolation high steam flow input into

limiting condition for operation.

Table 3.3.2–1, function 4.d.

Date of issuance: July 14, 1998.

Effective date: July 14, 1998.

Amendment No.: 71.

Facility Operating License No. DPR–

18: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 22, 1997 (62 FR

54876).

The February 6, 1998, April 17, 1998,

and June 4, 1998, letters provided

clarifying information that did not

change the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated July 14, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Rochester Public Library, 115

South Avenue, Rochester, New York

14610.
Federal Register / Vol. 63, No. 145 /Wednesday, July 29, 1998 / Notices 40567

Tennessee Valley Authority, Docket

Nos. 50–259, 50–260 and 50–296,

Browns Ferry Nuclear Plant, Units 1, 2

and 3, Limestone County, Alabama

Date of application for amendments:

June 6, 1996, as supplemented

September 26, 1997, January 23, 1998,

and May 19, 1998 (TS–372).

Brief description of amendments:

Changes to the technical specifications

administrative controls related to

quality assurance, and other

administrative and editorial changes.

Date of issuance: July 9, 1998.

Effective date: July 9, 1998.

Amendment Nos.: 233, 252, and 211.

Facility Operating License Nos. DPR–

33, DPR–52 and DPR–68: Amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: September 25, 1996 (61 FR

50346).

The supplemental letters dated

September 26, 1997, January 23, and

May 19, 1998 did not change the

original no significant hazards

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated July 9, 1998.

No significant hazards consideration

comments received: None.

Local Public Document Room

location: Athens Public Library, South

Street, Athens, Alabama 35611.

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50–346, Davis-Besse Nuclear Power

Station, Unit 1, Ottawa County, Ohio

Date of application for amendment:

April 24, 1998.

Brief description of amendment: This

amendment changed Technical

Specification (TS) Section 3/4.3.1.1,

‘‘Reactor Protection System

Instrumentation,’’ TS Section 3/4.3.2.1,

‘‘Safety Features Actuation System

Instrumentation,’’ TS Section 3/4.3.2.2,

‘‘Steam and Feedwater Rupture Control

System Instrumentation,’’ and the

associated TS bases. The TS tables of

response time limits were relocated to

the Davis-Besse Technical Requirements

Manual. Other changes in these TS

sections were also made consistent with

the relocation.

Date of issuance: July 7, 1998.

Effective date: July 7, 1998.

Amendment No.: 225.

Facility Operating License No. NPF–3:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

July 7, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of Toledo, William

Carlson Library, Government

Documents Collection, 2801 West

Bancroft Avenue, Toledo, OH 43606.

Wisconsin Electric Power Company,

Docket Nos. 50–266 and 50–301, Point

Beach Nuclear Plant, Units 1 and 2,

Town of Two Creeks, Manitowoc

County, Wisconsin

Date of application for amendments:

April 24, 1996, as supplemented

December 15, 1997, and June 22, 1998.

Brief description of amendments:

These amendments revise Technical

Specifications (TS) Section 15.7,

‘‘Radiological Effluent Technical

Specifications (RETS).’’ Portions of the

RETS are moved to licensee-controlled

documents consistent with Nuclear

Regulatory Commission guidance on TS

improvements. Other sections of the TSs

have also been revised consistent with

the removal of portions of the RETS.

Date of issuance: July 13, 1998.

Effective date: July 13, 1998, with full

implementation within 45 days.

Implementation shall include relocation

of certain Technical Specification

requirements to licensee-controlled

documents, as described in the

licensee’s application dated April 24,

1996, as supplemented by letter dated

December 15, 1997, and June 22, 1998,

and evaluated in the staff’s safety

evaluation attached to the amendments.

Amendment Nos.: 184 and 188.

Facility Operating License Nos. DPR–

24 and DPR–27: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: June 5, 1996 (61 FR 28620)

The December 15, 1997, and June 22,

1998, submittals provided additional

clarifying information and updated TS

pages. This information was within the

scope of the original Federal Register

notice and did not change the staff’s

initial no significant hazards

considerations determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated July 13, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: The Lester Public Library,

1001 Adams Street, Two Rivers,

Wisconsin 54241.

Dated at Rockville, Maryland, this 22nd

day of July 1998.

For the Nuclear Regulatory Commission.

Elinor G. Adensam,

Acting Director, Division of Reactor Projects—

III/IV, Office of Nuclear Reactor Regulation.

[FR Doc. 98–20111 Filed 7–28–98; 8:45 am]

BILLING CODE 7590–01–P

SECURITIES AND EXCHANGE

COMMISSION

Submission for OMB Review;

Comment Request

Upon Written Request, Copies Available

From: Securities and Exchange

Commission, Office of Filings and

Information Services, Washington, DC

20549.

Extension: Rule 17f–1(g)—SEC File No. 270–

30—OMB Control No. 3235–0290

Notice is hereby given that pursuant

to the Paperwork Reduction Act of 1995

(44 U.S.C. 3501 et seq.), the Securities

and Exchange Commission

(‘‘Commission’’) has submitted to the

Office of Management and Budget a

request for extension of the previously

approved collection of information

discussed below.

• Rule 17f–1(g) Requirements for

reporting and inquiry with respect to

missing, lost, counterfeit or stolen

securities.

Rule 17f–1(g), under the Securities

Exchange Act of 1934 (‘‘Act’’), requires

that all reporting institutions (i.e., every

national securities exchange, member

thereof, registered securities association,

broker, dealer, municipal securities

dealer, registered transfer agent,

registered clearing agency, participant

therein, member of the Federal Reserve

System and bank insured by the FDIC)

maintain and preserve a number of

documents related to their participation

in the Lost and Stolen Securities

Program (‘‘Program’’) under Rule 17f–1.

The following documents must be kept

in an easily accessible place for three

years, according to paragraph (g): (a)

copies or all reports of theft or loss

(Form X–17F–1A) filed with the

Commission’s designee: (b) all

agreements between reporting

institutions regarding registration in the

Program or other aspects of Rule 17f–1;

and (c) all confirmations or other

information received from the

Commission or its designee as a result

of inquiry.

Reporting institutions utilize these

records and reports (a) to report missing,

lost, stolen or counterfeit securities to

the data base, (b) to confirm inquiry of

the data base, and (c) to demonstrate

compliance with Rule 17f–1. The

Commission and the reporting

institutions’ examining authorities