63970 Federal Register / Vol. 62, No. 232 / Wednesday, December 3, 1997 / Notices

picture, and television cameras during

this meeting will be limited to selected

portions of the meeting as determined

by the ACNW Chairman. Information

regarding the time to be set aside for this

purpose may be obtained by contacting

the Chief, Nuclear Waste Branch, prior

to the meeting. In view of the possibility

that the schedule for ACNW meetings

may be adjusted by the Chairman as

necessary to facilitate the conduct of the

meeting, persons planning to attend

should notify Mr. Major as to their

particular needs.

Further information regarding topics

to be discussed, whether the meeting

has been canceled or rescheduled, the

Chairman’s ruling on requests for the

opportunity to present oral statements

and the time allotted therefor can be

obtained by contacting Mr. Richard K.

Major, Chief, Nuclear Waste Branch

(telephone 301/415–7366), between 8:00

A.M. and 5:00 P.M. EST.

ACNW meeting agenda, meeting

transcripts, and letter reports are

available for downloading or reviewing

on the internet at http://www.nrc.gov/

ACRSACNW.

The ACNW meeting dates for

Calendar Year 1998 are provided below:

ACNW

meeting

No.

1998 ACNW meeting date

................ No Meeting in January.

98 ........... February 24–26, 1998.

99 ........... March 24–26, 1998.

100 ......... April 21–23, 1998.

................ No Meeting in May

101 ......... June 10–12, 1998.

102 ......... July 21–23, 1998.

................ No Meeting in August.

103 ......... September 22–24, 1998 (Las

Vegas, NV).

104 ......... October 20–22, 1998.

................ No Meeting in November.

105 ......... December 15–17, 1998.

Dated: November 26, 1997.

Andrew L. Bates,

Advisory Committee Management Officer.

[FR Doc. 97–31730 Filed 12–2–97; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Biweekly Notice; Applications and

Amendments to Facility Operating

Licenses Involving No Significant

Hazards Considerations

I. Background

Pursuant to Public Law 97–415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 97–415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from November 7,

1997, through November 20, 1997. The

last biweekly notice was published on

November 19, 1997 (62 FR 61836).

Notice of Consideration of Issuance of

Amendments to Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a

proposed determination that the

following amendment requests involve

no significant hazards consideration.

Under the Commission’s regulations in

10 CFR 50.92, this means that operation

of the facility in accordance with the

proposed amendment would not (1)

involve a significant increase in the

probability or consequences of an

accident previously evaluated; or (2)

create the possibility of a new or

different kind of accident from any

accident previously evaluated; or (3)

involve a significant reduction in a

margin of safety. The basis for this

proposed determination for each

amendment request is shown below.

The Commission is seeking public

comments on this proposed

determination. Any comments received

within 30 days after the date of

publication of this notice will be

considered in making any final

determination.

Normally, the Commission will not

issue the amendment until the

expiration of the 30-day notice period.

However, should circumstances change

during the notice period such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received before

action is taken. Should the Commission

take this action, it will publish in the

Federal Register a notice of issuance

and provide for opportunity for a

hearing after issuance. The Commission

expects that the need to take this action

will occur very infrequently.

Written comments may be submitted

by mail to the Chief, Rules and

Directives Branch, Division of Freedom

of Information and Publications

Services, Office of Administration, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555–0001, and

should cite the publication date and

page number of this Federal Register

notice. Written comments may also be

delivered to Room 6D22, Two White

Flint North, 11545 Rockville Pike,

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

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

written comments received may be

examined at the NRC Public Document

Room, the Gelman Building, 2120 L

Street, NW., Washington, DC. The filing

of requests for a hearing and petitions

for leave to intervene is discussed

below.

By January 2, 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
Federal Register / Vol. 62, No. 232 / Wednesday, December 3, 1997 / Notices 63971

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

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

factors specified in 10 CFR

2.714(a)(1)(i)–(v) and 2.714(d).

For further details with respect to this

action, see the application for

amendment which is available for

public inspection at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, and at the local public

document room for the particular

facility involved.

Carolina Power & Light Company, et al.,

Docket Nos. 50–325 and 50–324,

Brunswick Steam Electric Plant, Units 1

and 2, Brunswick County, North

Carolina

Date of amendments request:

November 6, 1997.

Description of amendments request:

The proposed amendments change the

Technical Specifications (TS) for the

Brunswick Steam Electric Plant (BSEP)

Units 1 and 2 to allow three 18-month

diesel generator (DG) surveillance

requirements (SR) to be performed

during both plant operation

(Operational Conditions 1 and 2) and

shutdown (Operational Conditions 3, 4,

and 5) rather than, as currently required,

only during shutdown. The first SR is

an inspection of the DG involving a

partial disassembly. The second ensures

that non-critical DG protective functions

are bypassed on an Emergency Core

Cooling system actuation signal. The

third verifies that the DG operates for

greater than or equal to 60 minutes

while loaded to at least 3500 kw, which

bounds the maximum expected postaccident diesel generator loading. The

proposed amendments additionally

remove an expired footnote from the

BSEP Unit 2 DG TS.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

10 CFR 50.92 provides standards for

determining whether a significant

hazards consideration exists. A

proposed amendment to an operating

license for a facility involves no

significant hazards consideration if

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, (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. Carolina Power & Light Company

has reviewed these proposed license

amendment requests and has concluded

that their adoption would not involve a

significant hazards consideration. The

basis for this determination follows.

1. The proposed license amendments

do not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

The proposed license amendments

add a footnote to SR 4.8.1.1.2.d to allow

performance of SR 4.8.1.1.2.d.1, SR

4.8.1.1.2.d.4, and SR 4.8.1.1.2.d.5 in

OPERATIONAL CONDITION 1, 2, 3, 4,

or 5 rather than only during shutdown.

The footnote requires the unit to be in

OPERATIONAL CONDITION 3, 4, or 5

when performing SR 4.8.1.1.2.d.2, SR

4.8.1.1.2.d.3, SR 4.8.1.1.2.d.6, and SR

4.8.1.1.2.d.7 for its associated diesel

generators. No such limitation is placed

on SR 4.8.1.1.2.d.1, SR 4.8.1.1.2.d.4, or

SR 4.8.1.1.2.d.5.

There is no relaxation of any limiting

condition for operation (LCO) and no

decrease in surveillance requirements as

a result of the proposed amendments.

As such, the proposed license

amendments will not affect the ability of

the diesel generators to perform their

intended safety function. Performance

of SR 4.8.1.1.2.d.1, SR 4.8.1.1.2.d.4, and
63972 Federal Register / Vol. 62, No. 232 / Wednesday, December 3, 1997 / Notices

SR 4.8.1.1.2.d.5, during power

operations, will not adversely affect

overall nuclear safety. Diesel generator

capacity is such that any three of the

four diesel generators can supply the

required loads for the safe shutdown of

one unit and a design basis accident on

the other unit without relying on offsite

power. The diesel generator is not tied

to the emergency bus (E bus) during

performance of SR 4.8.1.1.2.d.1 or SR

4.8.1.1.2.d.4. Therefore, performance of

SR 4.8.1.1.2.d.1 and SR 4.8.1.1.2.d.4,

during power operation, will not affect

the operability of any other safetyrelated systems nor will it create any

perturbations of the electrical

distribution system that could challenge

plant operation.

Performance of SR 4.8.1.1.2.d.5,

during power operation, will not

adversely affect overall nuclear safety.

SR 4.8.1.1.2.d.5 is performed in a

similar manner to SR 4.8.1.1.2.a.5,

which requires that, at least once per 31

days on a staggered test basis, a diesel

generator be synchronized to the E bus

and loaded to 1750 kw for 15 minutes.

The critical portions of these

surveillances are when the diesel

generators are being synchronized to the

E bus or disconnected from the E Bus.

As such, performance of SR 4.8.1.1.2.d.5

during power operation does not create

an additional opportunity of a

perturbation of the electrical

distribution system that could challenge

plant operation than currently exists as

a result of the performance of SR

4.8.1.1.2.a.5. The existing design of the

electrical distribution system ensures

that a grid problem will not result in

failure of a diesel generator when it is

synchronized to the E bus. The E buses

are normally supplied by offsite power,

via a 4160 V balance of plant (BOP) bus,

through a master/slave breaker

combination. When performing SR

4.8.1.1.2.d.5, the diesel generator is

started in manual mode and

synchronized to the E bus. With a diesel

generator synchronized to the E bus, the

diesel generator is protected from a

potential overload condition. Class 1E

protective relaying, at the E bus, is

aligned to the trip circuit of the slave

breaker to protect the diesel from an

overload condition should the normal

source of power be lost. These relays

sense E bus voltage, E bus frequency,

and directional power from the E bus to

the BOP bus. Actuation of any of these

relays, with the diesel in manual, will

trip the slave and master breakers to

separate the diesel generator from the

BOP bus. This separates the diesel

generator from the potential overload

condition. In addition, either a loss of

offsite power or loss of coolant accident

results in the diesel generator output

breaker opening, E bus loads stripping,

and the diesel generator reverting to

automatic mode. This allows the diesel

generator to tie back to the E bus and

carry the E bus loads.

The proposed license amendments

reflect the clarification, previously made

to Bases Section 3/4.8, ‘‘Electrical Power

Sources,’’ in SR 4.8.1.1.2.d itself.

Accordingly, SR 4.8.1.1.2.d.2, SR

4.8.1.1.2.d.3, SR 4.8.1.1.2.d.6, and SR

4.8.1.1.2.d.7 are performed for diesel

generator 1 or 2 with BSEP, Unit No. 1

in OPERATIONAL CONDITION 3, 4, or

5 and for diesel generator 3 or 4 with

BSEP, Unit No. 2 in OPERATIONAL

CONDITION 3, 4, or 5. Defining the

term ‘‘during shutdown’’ as

‘‘OPERATIONAL CONDITION 3, 4, or

5’’ is consistent with the current TS

requirements of SR 4.8.1.1.2.d. TS Table

1.2, ‘‘OPERATIONAL CONDITIONS,’’

defines five OPERATIONAL

CONDITIONS for the BSEP. There are

two OPERATIONAL CONDITIONS

applicable to power operation with the

unit critical (i.e., POWER OPERATION

and STARTUP) and three

OPERATIONAL CONDITIONS

applicable to a subcritical, shutdown

unit (i.e., HOT SHUTDOWN, COLD

SHUTDOWN, and REFUELING).

Therefore, ‘‘during shutdown’’ and ‘‘in

OPERATIONAL CONDITION 3, 4, or 5’’

have equivalent meaning.

Eliminating the expired BSEP, Unit

No. 2 footnote to SR 4.8.1.1.2.d.1 is an

administrative change and, therefore,

cannot increase the probability or

consequences of an accident previously

evaluated.

Based on the above, the proposed

license amendments do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed license amendments

will not create the possibility of a new

or different kind of accident from any

accident previously evaluated.

The proposed license amendments to

allow performance of SR 4.8.1.1.2.d.1,

SR 4.8.1.1.2.d.4, and SR 4.8.1.1.2.d.5 in

OPERATIONAL CONDITION 1, 2, 3, 4,

or 5, rather than only during shutdown,

do not affect the operation or response

of any plant equipment, including the

diesel generators, or introduce any new

failure mechanism. Plant systems and

equipment will continue to respond in

accordance with design and as

analyzed. There will not be a

malfunction of a new or different type

introduced by the proposed license

amendments.

The proposed license amendments

reflect the clarification, previously made

to Bases Section 3/4.8, in SR 4.8.1.1.2.d

itself. Accordingly, SR 4.8.1.1.2.d.2, SR

4.8.1.1.2.d.3, SR 4.8.1.1.2.d.6, and SR

4.8.1.1.2.d.7 are performed for diesel

generator 1 or 2 with BSEP, Unit No. 1

in OPERATIONAL CONDITION 3, 4, or

5 and for diesel generator 3 or 4 with

BSEP, Unit No. 2 in OPERATIONAL

CONDITION 3, 4, or 5. Defining the

term ‘‘during shutdown’’ as

‘‘OPERATIONAL CONDITION 3, 4, or

5’’ is consistent with the current TS

requirements of SR 4.8.1.1.2.d. TS Table

1.2, ‘‘OPERATIONAL CONDITIONS,’’

defines five OPERATIONAL

CONDITIONS for the BSEP. There are

two OPERATIONAL CONDITIONS

applicable to power operation with the

unit critical (i.e., POWER OPERATION

and STARTUP) and three

OPERATIONAL CONDITIONS

applicable to a subcritical, shutdown

unit (i.e., HOT SHUTDOWN, COLD

SHUTDOWN, and REFUELING).

Therefore, ‘‘during shutdown’’ and ‘‘in

OPERATIONAL CONDITION 3, 4, or 5’’

have equivalent meaning.

Eliminating the expired BSEP, Unit

No. 2 footnote to SR 4.8.1.1.2.d.1 is an

administrative change and, therefore,

cannot create the possibility of a new or

different kind of accident from any

accident previously evaluated.

Based on the above, the proposed

license amendments do not create the

possibility of a new or different kind of

accident from any previously evaluated.

3. The proposed license amendments

do not involve a significant reduction in

a margin of safety.

Bases Section 3/4.8, ‘‘Electrical Power

Systems,’’ states that the operability of

the alternating current (ac) and direct

current power sources and associated

distribution systems during operation

ensures that sufficient power will be

available to supply the safety-related

equipment required for the safe

shutdown of the facility and the

mitigation and control of accident

conditions within the facility. Diesel

generator capacity is such that any three

of the four diesel generators can supply

the required loads for the safe shutdown

of one unit and a design basis accident

on the other unit without relying on

offsite power. Performance of SR

4.8.1.1.2.d.1, SR 4.8.1.1.2.d.4, and SR

4.8.1.1.2.d.5 during power operation

will not affect the operability of any

other safety-related systems, nor will it

create any perturbations of the electrical

distribution system that could challenge

plant operation. Class 1E protective

relaying, at the E bus, protects the diesel

from an overload condition should the

normal source of power be lost while

performing SR 4.8.1.1.2.d.5. There is no

relaxation of any LCO as a result of the
Federal Register / Vol. 62, No. 232 / Wednesday, December 3, 1997 / Notices 63973

proposed license amendments. If an

additional ac power source becomes

inoperable during the performance of

SR 4.8.1.1.2.d.1, SR 4.8.1.1.2.d.4, and

SR 4.8.1.1.2.d.5, the units will be placed

in the appropriate OPERATIONAL

CONDITION in accordance with TS

3.8.1.1, ‘‘A.C. Sources Operating.’’

Therefore, the diesel generators’ ability

to perform their intended safety

function, as described in Section

8.3.1.1.6.1 of the BSEP Updated Final

Safety Analysis Report, is not adversely

affected by the proposed license

amendments.

The proposed license amendments are

consistent with the guidance of Generic

Letter 91–04, ‘‘Changes In Technical

Specification Surveillance Intervals To

Accommodate A 24-Month Fuel Cycle,’’

which concludes that TSs need not

restrict surveillances to only being

performed during shutdown provided

that performance of the surveillance

during power operations does not

adversely affect safety.

The proposed license amendments

reflect the clarification, previously made

to Bases Section 3/4.8, in SR 4.8.1.1.2.d

itself. Accordingly, SR 4.8.1.1.2.d.2, SR

4.8.1.1.2.d.3, SR 4.8.1.1.2.d.6, and SR

4.8.1.1.2.d.7 are performed for diesel

generator 1 or 2 with BSEP, Unit No. 1

in OPERATIONAL CONDITION 3, 4, or

5 and for diesel generator 3 or 4 with

BSEP, Unit No. 2 in OPERATIONAL

CONDITION 3, 4, or 5. Defining the

term ‘‘during shutdown’’ as

‘‘OPERATIONAL CONDITION 3, 4, or

5’’ is consistent with the current TS

requirements of SR 4.8.1.1.2.d. TS Table

1.2, ‘‘OPERATIONAL CONDITIONS,’’

defines five OPERATIONAL

CONDITIONS for the BSEP. There are

two OPERATIONAL CONDITIONS

applicable to power operation with the

unit critical (i.e., POWER OPERATION

and STARTUP) and three

OPERATIONAL CONDITIONS

applicable to a subcritical, shutdown

unit (i.e., HOT SHUTDOWN, COLD

SHUTDOWN, and REFUELING).

Therefore, ‘‘during shutdown’’ and ‘‘in

OPERATIONAL CONDITION 3, 4, or 5’’

have equivalent meaning.

Eliminating the expired BSEP, Unit

No. 2 footnote to SR 4.8.1.1.2.d.1 is an

administrative change and, therefore,

cannot involve a significant reduction in

a margin of safety.

Based on the above, the proposed

license amendments do not involve a

significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of North Carolina at

Wilmington, William Madison Randall

Library, 601 S. College Road,

Wilmington, North Carolina 28403–

3297.

Attorney for licensee: William D.

Johnson, Vice President and Senior

Counsel, Carolina Power & Light

Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: James E. Lyons.

Carolina Power & Light Company, et al.,

Docket Nos. 50–325 and 50–324,

Brunswick Steam Electric Plant, Units 1

and 2, Brunswick County, North

Carolina

Date of amendments request:

November 6, 1997.

Description of amendments request:

The proposed amendments to Technical

Specification (TS) Limiting Conditions

for Operation (LCO) 3.3.5.5,

Instrumentation for Control Room

Emergency Ventilation System (CREVS)

and 3.7.2, Control Room Emergency

Ventilation System, and associated

Bases for the Brunswick Steam Electric

Plant (BSEP) Units 1 and 2 would be

limited in duration (approximately 3

months) and would allow operation of

both BSEP units to continue while

upgrades to the control building

ventilation system, including new air

conditioning (AC) units, are being

installed. Part of the planned work

requires opening the ductwork at the

evaporative (i.e. cooling) coils.

Temporary barriers will be constructed

to preserve the leakage integrity of the

control room pressure boundary;

however, the temporary barriers will not

be seismically qualified. While the

permanent AC units are out of service,

temporary AC units will be utilized.

During the upgrade installation, the AC

for the control room will not be

protected from certain external events

(e.g., seismic events, environmental

hazards such as tornadoes and

hurricanes, radiological sabotage, and

missile hazards), as required by the

system design and licensing basis, and

will not fully meet single failure criteria.

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 amendments do not

involve a significant increase in the

probability or consequences of an

accident previously evaluated.

The proposed changes do not affect

any component of any of the barriers to

radiation release, any of the systems

which protect the core from

overheating, nor any system used to

shut down the reactor. The proposed

changes do not affect any of the

chlorination system piping or the tank

car, which would be the initiating

components of a chlorine release event.

The proposed changes affect the CREVS

and CREVS instrumentation, neither of

which are accident or event causing

systems. Therefore, the proposed

changes do not increase the probability

of an accident or toxic gas release

previously analyzed in the Updated

Final Safety Analysis Report (FSAR).

The proposed changes do not affect

the ability of the CREVS to mitigate the

consequences of a design basis accident

or event involving a release of

radioactive material. In addition, the

proposed changes do not significantly

affect the ability of the system to

mitigate the consequences of a toxic gas

release. The following measures will be

taken to minimize the consequences of

accidents and events:

Temporary isolation barriers will be

constructed to provide integrity of the

duct during design basis radiation

release events. These temporary barriers

will ensure that 10 CFR Part 50, General

Design Criterion 19 for Control Room

operator doses is met for all design basis

radiation release accidents.

During the time that the temporary

barrier is used, the chlorine tank car

will be removed from the exclusion

area. Analyses have shown that with the

chlorine tank car outside of the

exclusion area, there is no threat to

Control Room habitability. Removal of

the chlorine tank car from the exclusion

area is the current Technical

Specification requirement for

inoperability of the Control Room

chlorine isolation mode.

The temporary condensing units for

the Control Room Air Conditioning

system will be installed to high quality

standards, and a spare condensing unit

will be provided such that two units can

be maintained functional. These units

will each be powered from a separate

division of Class 1E power. The

operation of the units will be monitored

to ensure that they are in good operating

order.

If two or more of the condensing units

should fail, instructions have been

provided to the operators for increased

monitoring of temperatures, and

mitigating actions are available to the

operators if temperatures rise above a

predetermined limit.

Therefore, the consequences of an

accident or an event involving a release
63974 Federal Register / Vol. 62, No. 232 / Wednesday, December 3, 1997 / Notices

of radiation, toxic gas, or smoke will not

be significantly increased. In addition,

the change will not significantly affect

the consequences of a seismic event or

other severe natural phenomena, as

previously analyzed in the Updated

FSAR.

2. The proposed amendments would

not create the possibility of a new or

different kind of accident from any

accident previously evaluated.

The proposed changes involve

adjustments to the LCO requirements for

CREVS relative to protection from

severe natural phenomena. The

proposed changes do not introduce any

new modes of plant operation. The

proposed changes do not involve any

new modes of system operation, except

that temporary condensing units will be

used in place of the permanent

condensing units. The temporary

condensing units will interface with the

permanent Control Building Heating

Ventilation and Air Conditioning

system in a similar manner to the

permanent system. The piping

connections to the permanent system

will be the same, and the controls

interface will be the same. No new

cross-ties will be created and no new

piping will be run though the

habitability boundary. Therefore, this

change will not create the possibility of

a new or different kind of accident from

any accident previously analyzed.

3. The proposed license amendments

do not involve a significant reduction in

a margin of safety.

The proposed changes do not

represent a significant change in the

assumptions and inputs to the analyses

for Control Room operator doses. No

increase in the doses to the Control

Room operators is expected after a

seismic event or tornado, since the

integrity of existing barriers to release of

radioactive material are not affected.

Therefore, this change does not result in

a significant reduction in the margin of

safety for a radiological event.

The proposed change does not

represent a change to the leakage criteria

for the Control Room, or the Control

Room ventilation ductwork, following

either a toxic gas or external smoke

event. The bounding analysis remains

valid, unless the failure is caused by a

tornado or seismic event. Due to the low

probability of such an event occurring

during the short time frame involved in

this modification, the occurrence of

such an event is not of significant

concern.

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: University of North Carolina at

Wilmington, William Madison Randall

Library, 601 S. College Road,

Wilmington, North Carolina 28403–

3297.

Attorney for licensee: William D.

Johnson, Vice President and Senior

Counsel, Carolina Power & Light

Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: James E. Lyons.

Consumers Energy Company, Docket

No. 50–155, Big Rock Point Nuclear

Plant, Charlevoix County, Michigan

Date of amendment request:

September 19, 1997 (Accession No.

9709240373).

Description of amendment request:

The amendment request propose

changes to the Facility Operating

License and technical specifications

(TS) to reflect the permanent cessation

of power operations and permanent

transfer of nuclear reactor fuel to the

spent fuel pool (SFP). In particular,

Consumers Energy requests to change:

safety limits; limiting safety system

settings; limiting control system

settings; limiting conditions for

operation; surveillance requirements;

design features; and administrative

controls. On November 12, 1997,

Consumers Energy provided

supplemental information regarding

their no significant hazards

determination, as requested by NRC

request for additional information letter

dated October 12, 1997. By letters dated

June 26 and September 23, 1997, the

licensee certified permanent cessation

of power operations and permanent

removal of all fuel from the reactor,

respectively.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The proposed change provides the

applicable requirements to assure safe

storage of spent nuclear fuel during

decommissioning following the

permanent cessation of power

operations at the Big Rock Point Nuclear

Plant (BRP) on August 30, 1997 [see

Consumers Energy letter to NRC dated

June 26, 1997] and permanent removal

of all fuel from the reactor vessel on

September 20, 1997 [see Consumer

Energy letter to NRC dated September

23, 1997]. Decommissioning activities

conducted using these controls do not

present undue risk to the public, and do

not impact common defense and

security. As such, these changes will

not:

1. Involve a significant increase in the

probability or consequences of an

accident previously evaluated.

No accidents previously evaluated in

the Updated Final Hazards Summary

Report (UFHSR) will have their

probability of occurrence increased

because the proposed controls

effectively preclude the occurrence of

criticality, fuel temperature exceeding

limits, or fuel handling accidents. The

probability of plant accidents associated

with power operations have been

significantly reduced. Accidents

associated with spent fuel handling,

including cask and single bundle drop

and spent fuel cooling capability loss

events, are still pertinent and were

reviewed using new data on pool

inventory and revised 10 CFR 20

radiological limit determinations. The

probability of occurrence of accidents

associated with storing 441 spent fuel

assemblies in the SFP (current license

limit) have not been affected by the

changes in the proposed TSs.

The consequences of a fuel handling

and cask drop accidents were evaluated

based on the removal of all fuel from the

reactor and loading spent nuclear fuel in

the SFP. The removal of all fuel from

the reactor vessel to storage in the SFP

and the subsequent decay of the fuel in

the pool result in no increase in the

probability of these accidents and

continuously reduced consequences

from these accidents.

Analyses using the techniques in

Branch Technical Position APCSB 9–2

provide the heat rate from a freshlyremoved full core off-load in the SFP

whose racks are filled with a total of 441

fuel assemblies as the most limiting

cooling condition. Existing cooling

equipment under the current TSs

provide sufficient cooling to preclude

spent fuel pool temperatures reaching

150 degrees-Fahrenheit with a complete

loss of spent fuel cooling for 72 hours.

This precludes entry into an unanalyzed

condition for the SFP and provides 3

days to recover cooling flow of

‘‘approximately 30’’ gallons per minute.

Since this specification change is

intended for implementation following

93 days after shutdown (approximately

November 30, 1997), this analysis

justifies the allowance of 24 hours to reestablish cooling flow provided in

specification 3.1.2.

2. Create the possibility of a new or

different kind of accident from any

accident previously evaluated.

The permanent cessation of power

operation and removal of fuel from the
Federal Register / Vol. 62, No. 232 / Wednesday, December 3, 1997 / Notices 63975

reactor eliminates the possibility of the

following categories of accidents and

transients to create a hazard to the

health and safety of the public: increase

in heat removal by the secondary

system; increase in reactor coolant

inventory; decrease in heat removal by

the secondary system; decrease in

reactor coolant inventory; reactivity and

power distribution anomalies;

anticipated transient without scram;

and, single loop operation. These

revised TSs, in combination with

requirements in the UFHSR, provide

assurance that fuel handling and spent

fuel cask drop accident, which represent

the remaining specific pertinent

accidents analyzed in the ‘‘radioactive

release from a subsystem of component’’

category, will not occur. Because the

revised TSs related to fuel handling,

spent nuclear fuel storage, and handling

of the spent fuel cask satisfy current

license and UFHSR requirements, no

new accidents are created.

3. Involve a significant reduction in a

margin of safety.

The safety margins for analyzed

accidents are maintained because the

containment structures and redundant

control established by the plant remain

in place until the decay of spent fuel has

reduced the source term to levels that

analysis confirms do not require the

containment features. ninety three days

after permanent cessation of operations,

the spent nuclear fuel at BRP will have

decayed to the point where the added

margin from this decay more than

compensates for the removal of the

containment as a safety feature, and

allows relaxed controls for the cooling

of the SFP.

The Big Rock Point Plant Safety

Committee has reviewed this Facility

Operating License and TS change

request and has determined this change

does not involve an unreviewed safety

question and, therefore, involves no

significant hazards consideration. The

proposed change has been reviewed by

the BRP Nuclear Performance

Assessment Department.

The NRC staff has reviewed the

licensee’s analysis, as provided by

licensee letters dated September 19 and

November 12, 1997, and, based on this

review, it appears that the three

standards or 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: North

Central Michigan College, 1515 Howard

Street, Petosky, MI 49770.

Attorney for licensee: Judd L. Bacon,

Esquire, Consumers Energy Company,

212 West Michigan Avenue, Jackson,

Michigan 49201.

NRC Project Director: Seymour H.

Weiss.

Duke Energy Corporation, Docket Nos.

50–269, 50–270, and 50–287, Oconee

Nuclear Station, Units 1, 2, and 3,

Oconee County, South Carolina

Date of amendment request: March

11, 1993; supplemented August 26,

November 29, December 6, 1993,

October 3, 1995, February 27, and

September 3, 1997 (TSC 93–03).

Description of amendment request:

The proposed changes would replace

the present Electrical Power Systems

section of the Technical Specifications,

Sections 3.7 and 4.6, by consolidating

and rearranging the present

specifications, incorporating new

specifications, and formating the section

similar to the Babcock and Wilcox

Standard Technical Specifications. The

proposed changes would address such

concerns as Keowee hydro station

operability, Lee gas turbine operability,

overhead and underground emergency

power path operability, Keowee and

Keowee main step-up transformer

outage requirements, surveillance

requirements of various components

and systems, Oconee distribution

system requirements, protective

instrumentation system requirements,

operability of 125 VDC Vital Instrument

and Control power and limiting

condition for operation, inverter

requirements, Oconee shutdown

requirements related to various

components, Keowee unit extended

outage, dc power operability

requirements, battery cell parameter

requirements, and various editorial and

related Bases changes.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below.

Duke Power Company (Duke)

[currently Duke Energy Corporation] has

made the determination that this

amendment request involves a No

Significant Hazards Consideration by

applying the standards established by

NRC regulations in 10 CFR 50.92. This

ensures 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:

Each accident analysis addressed

within the Oconee Final Safety Analysis

Report (FSAR) has been examined with

respect to the changes proposed within

this amendment request. Changes

included in this amendment request are

provided to assure availability of

electrical power systems for mitigation

of Design Basis Accidents (DBAs). As

described within the technical

justification, the following types of

changes are included:

(1) Editorial and administrative

changes associated with reformatting

the Technical Specification

requirements;

(2) Additional restrictions not

presently included in the Technical

Specifications such as the addition of

requirements for electrical power

systems during cold shutdown and

refueling, for the 230 kV switchyard

degraded grid protection system and to

delete the special inoperability period

for the Keowee CX transformer;

(3) Technical changes to current

requirements to provide clarity and

operational flexibility. These changes

maintain the ability of the electrical

power systems to mitigate the

consequences of DBAs without a

significant reduction in availability.

These changes include the definition of

emergency power paths to include the

associated DC sources and auxiliary

transformers, the combination of special

inoperability periods for ‘‘planned’’ and

‘‘unplanned’’ reasons, and the ability to

use the Keowee special inoperability

period more than once in a three year

period; and

(4) Relocation of requirements which

are unnecessary for the mitigation of

DBAs to licensee controlled documents.

Relocated requirements include

surveillance requirements for the

External Grid Trouble Protection

system.

Based on the above and the technical

justification * * *, there is no

significant increase in the probability of

DBA as a result of this change, nor is

there a significant increase in the

consequences of a DBA as a result of

this change since the proposed

amendment assures availability of

electrical power systems.

(2) Create the possibility of a new or

different kind of accident from any kind

of accident previously evaluated:

The proposed changes make no

physical changes to the plant

configuration and do not adversely

affect the performance of any

equipment. Operation of ONS [Oconee

Nuclear Station] in accordance with

these Technical Specifications will not

create any failure modes not bounded

by previously evaluated accidents.

Consequently, this change will not

create the possibility of a new or

different kind of accident from any kind

of accident previously evaluated.
63976 Federal Register / Vol. 62, No. 232 / Wednesday, December 3, 1997 / Notices

(3) Involve a significant reduction in

a margin of safety:

Margins of safety associated with

these Technical Specifications have

been evaluated. These changes include

editorial and administrative changes

associated with reformatting the

Technical Specification requirements,

additional restrictions not presently

included in the Technical

Specifications, technical changes to

current requirements which maintain

the ability of the electrical power

systems to mitigate the consequences of

DBAs, and relocation of requirements

which are unnecessary for the

mitigation of DBAs to licensee

controlled documents. The design basis

of auxiliary electrical systems is to

supply the required ES [emergency

system] loads of one Unit and safe

shutdown loads of the other two units.

The proposed amendment does not

affect any safety limits, setpoints, or

design parameters and assures the

continued availability of electrical

power systems; thus preserving the

existing margin of safety. Therefore,

there will be no significant reduction in

any margin of safety.

Duke has concluded based on the

above, and the technical justification

* * * that there are no significant

hazards considerations involved in this

amendment request.

The NRC 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.

Duquesne Light Company, et al., Docket

No. 50–334, Beaver Valley Power

Station, Unit No. 1, Shippingport,

Pennsylvania

Date of amendment request:

November 4, 1997.

Description of amendment request:

The proposed amendment would

change Item 6.a.2, ‘‘4.16 Emergency Bus

(Start Diesel),’’ of Table 3.3–4 of

Technical Specification 3.3.2.1. The

proposed change would reduce the trip

setpoint for starting the emergency

diesel generators on emergency bus

undervoltage from a trip setpoint of

greater than or equal to 83 percent with

a 12-cycle delay time to greater than or

equal to 75 percent of nominal bus

voltage with a time delay of less than

0.9 second including auxiliary relay

times. The proposed change would also

reduce the allowable value from greater

than or equal to 81 percent of nominal

bus voltage to greater than or equal to

74 percent of nominal bus voltage with

a time delay of less than 0.9 second

including auxiliary relay times.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the change involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

The proposed change replaces the

current Engineered Safety Feature

setpoint, allowable value and delay time

for the diesel generator start on loss of

power function. An analysis has been

performed to develop the new values to

minimize the diesel generator starts

when a Reactor Coolant Pump (RCP) is

being started or a fast bus transfer

occurs. The heat generated by an

increase in motor current, in response to

reduced voltage, will be less than the

heat generated during motor starting.

The analysis results show that bus

voltages may dip below the allowable

setpoint value and then recover to the

pick-up setpoint within the proposed

delay time without stalling motors.

The proposed change does not affect

the design and reliability of any plant

equipment; therefore, the probability of

occurrence of a previously evaluated

accident is not increased. The operation

of the plant will not be changed as a

result of this proposed amendment,

except that fewer diesel generator starts

will be initiated.

This function anticipates the loss of

voltage to protect equipment connected

to the 4.16 Kv emergency bus. The

UFSAR [Updated Final Safety Analysis

Report] accident analyses do not take

credit for this function; therefore, the

consequences of an accident previously

evaluated is not increased.

2. Does the change create the

possibility of a new or different kind of

accident from any accident previously

evaluated?

The proposed change to the trip

setpoint, allowable value and delay time

will continue to ensure that the safetyrelated equipment connected to the

emergency bus is adequately protected

from a low voltage condition. These

setting changes will minimize the diesel

generator starts due to voltage drops

when an RCP is started or a fast bus

transfer occurs.

The new setpoint and time delay

allow normal voltage drops to occur

during expected plant operations

without causing any thermal damage to

safety-related equipment. The

performance of the safety system will

remain unchanged and will not alter

any plant equipment, performance

requirements or safety analysis.

Therefore, the proposed change does not

create the possibility of a new or

different kind of accident from any

accident previously evaluated.

3. Does the change involve a

significant reduction in a margin of

safety?

The proposed change does not

involve a significant reduction in a

margin of safety since an analysis has

been performed to verify that safetyrelated equipment connected to the

emergency bus is adequately protected

from a low voltage condition with the

proposed settings. The proposed

changes do not affect the UFSAR design

bases, accident assumptions, or

technical specification bases. In

addition, the proposed changes do not

affect release limits, monitoring

equipment or plant operating practices.

Therefore, the proposed change will not

involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, PA

15001.

Attorney for Licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: John F. Stolz.

Entergy Operations, Inc., et al., Docket

No. 50–416, Grand Gulf Nuclear Station,

Unit 1, Claiborne County, Mississippi

Date of amendment request: October

28, 1997.

Description of amendment request:

The amendment would (1) revise the

frequency of conducting five

Surveillance Requirements (SRs) and (2)

add a 10 CFR Part 50, Appendix J

Testing Program for Primary

Containment Systems in the Technical

Specifications (TSs) for Grand Gulf

Nuclear Station, Unit 1 (GGNS). The

five SRs are the following: SR 3.6.1.1.1

for primary containment, SR 3.6.1.2.1
Federal Register / Vol. 62, No. 232 / Wednesday, December 3, 1997 / Notices 63977

for primary containment air locks, and

SRs 3.6.1.3.5, 3.6.1.3.8, and 3.6.1.3.9 for

primary containment isolation valves.

The proposed revisions for each of the

five SRs are to delete the references to

SR 3.0.2 not being applicable and

change the surveillance frequency from

being ‘‘in accordance with 10 CFR 50,

Appendix J, as modified by approved

exemptions’’ to ‘‘in accordance with 10

CFR Part 50, Appendix J, testing

program.’’ The testing program would

be added to Section 5.0, Administrative

Controls, of the TSs. Changes to the

Bases of the TSs were also provided in

the submittal.

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:

I. The proposed change does not

significantly increase the probability or

consequences of an accident previously

evaluated.

[On April 26, 1995, the licensee was

granted an exemption to Appendix J of

10 CFR Part 50 that allowed

performance-based containment leak

rate testing. This exemption will expire

on the startup from Refueling Outage 9,

currently scheduled for the spring of

1998. The licensee’s proposed changes

to the TSs are to adopt Option B,

Performance-Based Requirements, that

is now in Appendix J, but was not in

Appendix J in 1995 when the exemption

was granted. The technical findings that

support the rulemaking for Option B are

in NUREG–1493, ‘‘Performance-Based

Containment Leak Rate Test Program,’’

dated September 1995. The licensee

stated in its submittal that its current

containment leak rate testing program

meets the requirements of Option B.]

Two initiating events were identified

which could be affected by the proposed

changes [in the submittal of October 28,

1997]. An interfacing system LOCA

[(loss-of-coolant accident)] could be

caused by significant leakage of both

normally closed isolation valves in

systems with high pressure/low

pressure interfaces. Interfacing systems

LOCAs were considered for the LPCI,

LPCS, HPCS, and RCIC systems [(i.e.,

low pressure coolant injection, low

pressure core spray, high pressure core

spray, and reactor core isolation

cooling)]. Because the frequency for

testing of these valves will not be

changed under this proposal, there is no

increase in the probability or

consequences of an accident [previously

evaluated].

The second event evaluated was a

LOCA outside containment. In this case

the probability for failure of the MSIVs

[(main steam isolation valves)] and the

feedwater isolation valves were

calculated and combined with the

frequency of a pipe break outside

containment and the conditional

probability of a core melt given a LOCA.

The increase in core damage is

extremely small and therefore does not

significantly increase the probability of

any previously evaluated accident.

Further, because the testing frequency

for MSIVs and feedwater isolation

valves are not being changed, the LOCA

outside containment events can be

discounted.

Failure of, or leakage through[,] a

containment barrier can[,] however,

increase the consequences of those

accidents previously evaluated. Because

the leakage probability for two valves in

series to fail is very small and because

all lines isolated by a single

containment isolation valve always have

a water seal and cannot act as a release

pathway unless the integrity of the

connected system is compromised, there

is no significant increase in the

consequences of any previously

evaluated accident.

Containment bypass can also increase

the consequences of [previously]

evaluated accidents. Accident

sequences involving containment have

been shown to be relatively insignificant

by the GGNS IPE [(Individual Plant

Examination]). The potential for

[containment] bypass was analyzed. The

analysis showed that the probabilities

for bypass were dominated by failure to

close scenarios. Many programs are in

place at GGNS to monitor containment

component performance[,] and to ensure

that proper maintenance and repairs are

made during the service life of the

containment. Other routine

surveillances are performed periodically

to ensure that the valves will close on

demand. In fact, all valves that are

required to close for containment

isolation and that are not maintained

closed at all times during power

operations are stroke tested quarterly

or[,] at a minimum, during each

refueling outage in accordance with

ASME [(American Society of

Mechanical Engineers) Boiler and

Pressure Vessel Code,] Section XI,

Subsection IWV.

[Based on the above, the proposed

changes do not significantly increase the

probability or consequences of an

accident previously evaluated.]

II. The proposed change does not

create the possibility of a new or

different kind of accident from any

accident previously evaluated.

The request involves the reduction in

the local leak rate and the integrated

leak rate testing frequencies [in

accordance with Option B of Appendix

J to 10 CFR Part 50]. Extending the test

frequencies has no influence on, nor

does it contribute in any way to, the

possibility of a new or different kind of

accident or malfunction from those

previously analyzed. The method of

performing the test is not changed. No

new accident modes are created by

extending the testing intervals. No

safety-related equipment or safety

functions are altered as a result of this

change.

[Therefore, the proposed changes do not

create the possibility of a new or

different kind of accident from any

accident previously evaluated.]

III. The proposed change does not

involve a significant reduction in a

margin of safety.

The only margin of safety that has the

potential of being impacted by the

proposed changes involves the offsite

dose consequences of postulated

accidents which are directly related to

containment leakage rate. The

containment isolation system is

designed to limit leakage to La which is

defined by the GGNS TSs to be 0.437

percent by weight of the containment air

[volume] per 24 hours at [the

containment pressure of] 11.5 psig (Pa).

The limitation on containment leakage

rate is designed to ensure that total

leakage volume will not exceed the

value assumed in the accident analyses

at the peak accident pressure (11.5 psig,

Pa).

To provide additional conservatism,

the measured overall integrated leakage

rate is further limited to less than or

equal to 0.75 La during performance of

the periodic integrated leakage rate test

and to less than or equal to 0.60 La for

type B and C leakage rate tests [of

Appendix J]. This is done to account for

the possible degradation of the

containment leakage barriers between

[the Appendix J] tests. This acceptance

criteria ensures that an acceptable

margin of safety is being maintained and

will not be altered by the proposed

changes. The preservation of this margin

will continue to provide for potential

degradation of the leakage barriers

between tests.

No change in the method of testing is

being proposed. The tests will continue

to be done at full pressure (Pa) or greater

[pressure]. The test pressure for primary

containment isolation valves will

continue to be applied in the same

direction as would be required for the

valve to perform its safety function

(unless a different direction can be

shown to be equivalent or conservative).

Primary containment penetrations
63978 Federal Register / Vol. 62, No. 232 / Wednesday, December 3, 1997 / Notices

which require Type B leakage rate tests

will be performed in the same manner

as before. The Type A test [of Appendix

J] will continue to be performed at full

pressure (Pa). Other programs are in

place to ensure that proper maintenance

and repairs are performed during the

service life of the primary

containment[,] and systems and

components penetrating the primary

containment.

No change in the owners allowable

leakage rate is being proposed. These

conservative leakage rates ensure that[,]

if every penetration were at its

maximum allowable leakage rate, the

total containment leakage would still be

below 0.60 La. The effect of multiple

penetration barriers is not considered

which provides further conservatism.

The assessment of risk analysis for the

proposed changes concluded that the

overall risk impact of the changes are

neutral and essentially negligible. Any

containment isolation barrier allowed to

be tested at less frequent intervals

[through performance-based testing of

proposed Option B of Appendix J] will

have demonstrated enhanced

performance which minimizes the

potential for increased leakage. The

assessment further shows that there is

reasonable assurance that an acceptable

level of performance for the

containment isolation function can be

maintained. The overall risk impact for

the proposed changes are small enough

to be almost indeterminate. No change

to the leakage rate specified in the TSs

is being proposed.

[The proposed changes to the TSs are in

accordance with Option B of Appendix

J of 10 CFR Part 50.]

[Therefore, the proposed changes do not

involve a significant reduction in a

margin of safety.]

Based on the above evaluation,

operation in accordance with the

proposed amendment involves no

significant hazards consideration.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 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: Judge George W. Armstrong

Library, 220 S. Commerce Street,

Natchez, MS 39120.

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, NW., 12th Floor,

Washington, DC 20005–3502.

NRC Project Director: David A.

Wigginton, Acting.

Maine Yankee Atomic Power Company,

Docket No. 50–309, Maine Yankee,

Atomic Power Station, Lincoln County,

Maine

Date of amendment request:

September 30, 1997.

Description of amendment request:

The proposed amendment would

eliminate certain license conditions of

the Maine Yankee operating license that

are no longer appropriate in the

permanently defueled condition of the

plant. These conditions include

restrictions on the Fire Protection

Program and implementation of leakage

reduction, airborne iodine monitoring,

secondary water chemistry, and cooling

water discharge monitoring programs.

By letter dated August 7, 1997, the

licensee certified permanent cessation

of power operations and permanent

removal of fuel from the reactor vessel.

Most of the provisions of the Maine

Yankee operating license were

established to ensure protection of the

public health and safety during power

operations. Maine Yankee has proposed

to eliminate those license requirements

that are not relevant to the permanently

defueled plant condition to allow the

Maine Yankee staff to focus on those

provisions which are still appropriate

during decommissioning.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The proposed change does not:

1. Involve a significant increase in the

probability or consequence of an

accident previously evaluated.

The purpose of the proposed change

is to eliminate requirements which are

not appropriate in the permanently

defueled plant condition. Since the

plant has permanently ceased operation

and will be maintained in a defueled

condition, many provisions of the

license related to operation of the plant

are no longer appropriate. Elimination

of these unnecessary requirements

allows the plant staff to focus on those

requirements which continue to be

appropriate to the existing plant

condition. The proposed change does

not affect those Chapter 14 accidents

which are appropriate to the current

plant conditions: fuel handling

accident, spent fuel cask drop, and

radioactive liquid waste system leaks

and failures, and therefore, does 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 this proposed change

is to eliminate requirements which are

not appropriate in the permanently

defueled plant condition. Since the

plant has permanently ceased operation

and will be maintained in a defueled

condition, many provisions of the

license related to operation of the plant

are no longer appropriate. Elimination

of these unnecessary requirements

allows the plant staff to focus on those

requirements which continue to be

appropriate to the existing plant

conditions. This proposed change does

not affect storage of spent fuel and,

therefore, does not create the possibility

of a new or different accident from any

accident previously evaluated.

3. Involve a significant reduction in a

margin of safety.

The purpose of the proposed change

is to eliminate requirements which are

not appropriate in the permanently

defueled plant condition. Since the

plant has permanently ceased operation

and will be maintained in a defueled

condition, many provisions of the

license related to operation of the plant

are no longer appropriate. Elimination

of these unnecessary requirements

allows the plant staff to focus on those

requirements which continue to be

appropriate to the existing plant

conditions. This proposed change does

not affect storage of spent fuel and,

therefore, does 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: Wiscasset Public Library, High

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

04578.

Attorney for licensee: Mary Ann

Lynch, Esquire, Maine Yankee Atomic

Power Company, P.O. Box 408,

Wiscasset, ME 04578.

NRC Project Director: Seymour H.

Weiss.

Maine Yankee Atomic Power Company,

Docket No. 50–309, Maine Yankee

Atomic Power Station, Lincoln County,

Maine

Date of amendment request: October

20, 1997.

Description of amendment request:

The proposed amendment would

replace in their entirety the existing

Technical Specifications incorporated
Federal Register / Vol. 62, No. 232 / Wednesday, December 3, 1997 / Notices 63979

in Facility Operating License No. DPR–

36 as Appendix A. Maine Yankee

developed the revised Technical

Specifications, titled Permanently

Defueled Technical Specifications, to

reflect the permanently shutdown and

defueled status of the plant. Changes are

proposed to the definitions, limiting

conditions for operation, surveillance,

and administrative control sections.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration. The NRC staff has

reviewed the licensee’s analysis against

the standards of 10 CFR 50.92(c). A

summary of the licensee’s review is

presented below:

The proposed change does not,

1. Involve a significant increase in the

probability or consequence of an

accident previously evaluated.

This proposed change is consistent

with the improved Standard Technical

Specifications. The relocation of

requirements from the technical

specifications to the licensee controlled

documents is consistent with the

criteria set forth in 10 CFR 50.36 for the

content of technical specifications. The

removal of definitions, generic LCO

actions and generic surveillance

requirements has no impact on facility

structures or equipment or the methods

of operation of such structures or

equipment. The deletion of design

features and safety limits not applicable

to the permanently shutdown and

defueled status of the Maine Yankee

reactor has no impact on the remaining

applicable design basis accidents. The

removal of LCO and Surveillance

specifications which are related only to

the operation of the nuclear reactor or

only to the prevention, diagnosis or

mitigation of transients or accidents

primarily involving the reactor, do not

affect the remaining applicable

accidents previously evaluated. The

critical safety functions involving core

reactivity control, reactor heat removal,

reactor coolant system inventory control

and containment integrity are no longer

necessary at the Maine Yankee facility.

The postulated accidents involving

damage to the reactor coolant system,

main steam lines, main feed lines, steam

generators or the reactor core and the

subsequent release of radioactive

material are no longer applicable at the

Maine Yankee facility. Spent fuel pool

cooling and makeup related equipment

and support equipment including

electrical power systems are not

required to be continuously available

since there is time available to effect

repairs or establish alternate sources of

makeup flow in the event of a loss of

cooling and makeup flow to the spent

fuel pool. The effect of radioactive

decay since the shutdown of the reactor

has reduced the consequences of the

fuel handling accident to levels below

those previously analyzed. The relevant

parameters associated with spent fuel

pool (level and boron concentration)

that make up the initial conditions

assumed in applicable analysis are

included in the technical specifications.

The deletion and modification of

provisions of administrative controls do

not directly affect the design of

structures or equipment necessary for

the safe storage of irradiated fuel or the

methods used for handling and storage

of such fuel in the spent fuel pool. The

changes to the administrative controls

are, in fact, administrative in nature and

do not affect any accident applicable to

the safe storage of irradiated fuel or the

permanently shutdown and defueled

condition of the reactor. Therefore, the

proposed changes to the Maine Yankee

Technical Specifications do not involve

any increase in the probability or

consequences of any accident

previously evaluated.

2. Create the possibility of a new or

different kind of accident from any

accident previously evaluated.

The proposed changes have no impact

on facility structures or equipment

affecting the safe storage of irradiated

fuel or the methods of operation of such

structures or equipment or handling and

storage of such fuel. These changes are

consistent with the improved Standard

Technical Specifications and add to the

clarity and ease of use of the proposed

PDTS. The removal of technical

specifications which are related only to

the operation of the nuclear reactor or

only to the prevention, diagnosis or

mitigation of transients or accidents

primarily involving the reactor, can not

result in different or more adverse

failure modes or accidents than

previously evaluated because the reactor

is permanently shutdown and defueled.

The proposed deletion of provisions of

the Maine Yankee Technical

Specifications do not affect systems

credited in the existing accident

analyses for the remaining applicable

postulated accidents at the Maine

Yankee facility. The proposed technical

specifications continue to require

proper control and monitoring of safety

significant parameters and activities.

The proposed restrictions on boron

concentration and level in the spent fuel

pool are fulfilled by normal operating

conditions and preserve initial

conditions assumed in the analyses of

postulated DBA’s. Therefore, the

proposed changes to the MYTS does not

create the possibility of a new or

different accident from any accident

previously evaluated.

3. Involve a significant reduction in a

margin of safety.

The deletion of provisions in the

technical specifications which are not

related to the storage of irradiated fuel

or which are inconsistent with the scope

of the improved Standard Technical

Specifications will not affect the

analyses of the design basis accidents

remaining applicable to the Maine

Yankee facility. The postulated design

basis accidents involving the reactor are

no longer possible due to the

permanently defueled status of the

Maine Yankee reactor. The requirements

for systems, structures and components

which have been deleted from the

Maine Yankee Technical Specifications

are not credited in the existing accident

analysis for the remaining applicable

postulated accidents and therefore do

not contribute to the margin of safety

associated with the accident analysis.

Therefore, the proposed changes to the

Maine Yankee Technical Specifications

would not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Wiscasset Public Library, High

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

04578.

Attorney for licensee: Mary Ann

Lynch, Esquire, Maine Yankee Atomic

Power Company, P.O. Box 408,

Wiscasset, ME 04578.

NRC Project Director: Seymour H.

Weiss.

Northeast Nuclear Energy Company

(NNECO), et al., Docket No. 50–423,

Millstone Nuclear Power Station, Unit

No. 3, New London County, Connecticut

Date of amendment request: October

15, 1997.

Description of amendment request:

The proposed change to Technical

Specification 3/4.4.3, Pressurizer, would

replace the pressurizer maximum water

inventory requirement with a

pressurizer maximum indicated level

requirement. The proposed amendment

would also modify the associated Bases

section and make editorial changes.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards
63980 Federal Register / Vol. 62, No. 232 / Wednesday, December 3, 1997 / Notices

consideration, which is presented

below:

NNECO has reviewed the proposed

revision in accordance with 10 CFR

50.92 and has concluded that the

revision does not involve a significant

hazards consideration (SHC). The basis

for this conclusion is that the three

criteria of 10 CFR 50.92(c) are not

satisfied. The proposed revision does

not involve [an] SHC because the

revision would not:

1. Involve a significant increase in the

probability or consequence of an

accident previously evaluated.

The Technical Specification

maximum pressurizer inventory

requirement in Technical Specification

3.4.3 is being changed to use the

numerical value for the Reactor Trip

setpoint on pressurizer high water level

in Technical Specification Section 2.2.

This changes the requirement from a

volume to a level requirement, is

consistent with the Improved Standard

Technical Specifications for

Westinghouse plants, and represents a

more restrictive level requirement than

the current technical specification. The

bases change clarifies that the 89% level

requirement only assures that there is a

steam bubble in the pressurizer. Also,

the bases change states that pressurizer

level is maintained by automatic and

procedural controls to provide

assurance that the design basis analyses

are valid. These changes do not modify

plant operation. Lowering the maximum

level requirement so that it is

numerically consistent with the reactor

trip setpoint, while clarifying the bases

of the requirement, [cannot] involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

Therefore, the proposed revision does

not involve a significant increase in the

probability or consequence of an

accident previously evaluated.

2. Create the possibility of a new or

different kind of accident from any

accident previously evaluated.

There are no hardware modifications

associated with the change. The change

does not modify the way that the plant

is operated. The change modifies

neither accident mitigation nor system

response post-accident.

Therefore, the proposed revision does

not create the possibility of a new or

different kind of accident from any

accident previously evaluated.

3. Involve a significant reduction in a

margin of safety.

The change places a lower maximum

pressurizer level requirement for the

pressurizer. The change imposes the

numerical setpoint value for the reactor

trip on pressurizer high water level as

the restriction on the pressurizer level.

The change to the bases clarifies that the

89% level requirement only ensures the

existence of a steam bubble and not the

validity of the design basis analyses.

The design basis non-LOCA [loss-ofcoolant accident] analyses use the

current programmed pressurizer level

and the LOCA analysis uses 62% level

for full power. Those events that are

analyzed to address pressurizer filling

concerns are initiated assuming a higher

initial pressurizer water level that

accounts for 6% level uncertainty. The

bases change makes it clear that the

pressurizer level required to assure the

validity of the design basis analyses is

maintained by the automatic and

procedural controls and not the less

than or equal to 89% level in the

requirement.

Therefore, the proposed revision does

not involve a significant reduction in a

margin of safety.

In conclusion, bases on the

information provided, it is determined

that the proposed revision does not

involve an SHC.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, Connecticut, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut.

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

P.O. Box 270, Hartford, Connecticut.

NRC Deputy Director: Phillip F.

McKee.

Northeast Nuclear Energy Company

(NNECO), et al., Docket No. 50–423,

Millstone Nuclear Power Station, Unit

No. 3, New London County, Connecticut

Date of amendment request:

November 11, 1997.

Description of amendment request:

The proposed amendment to Technical

Specifications (TS) 3.9.1.2 and 3.9.13

and their Bases will allow crediting

soluble boron for maintaining keffective at less than or equal to 0.95

within the spent fuel pool (SFP) rack

matrix following a seismic event of a

magnitude greater than or equal to an

operating basis earthquake (OBE).

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

NNECO has reviewed the proposed

revision in accordance with 10CFR50.92

and has concluded that the revision

does not involve a significant hazards

consideration (SHC). The basis for this

conclusion is that the three criteria of 10

CFR 50.92(c) are not satisfied. The

proposed revision does not involve [an]

SHC because the revision would not:

1. Involve a significant increase in the

probability or consequence of an

accident previously evaluated.

There is one spent fuel pool accident

condition discussed in Chapter 15 of the

FSAR [final safety analysis report]. The

FSAR discusses a fuel handling accident

which drops a fuel assembly onto the

fuel racks during fuel movement.

Degradation of the Boraflex panels in a

post-seismic condition will have no

effect on the probability of a fuel

assembly drop onto the stored fuel, or

the fuel racks. Changing the way

Boraflex responds to a seismic event

will have no impact on the probability

of a seismic event. A misplaced fuel

assembly can be postulated in the MP3

[Millstone Unit 3] fuel pool as a result

of either equipment malfunction or

operator error. Degradation of the

Boraflex panels will have no effect on

the probability of a fuel misplacement

event. Therefore, the degradation of

Boraflex in a post-seismic condition

does not involve an increase in the

probability of an accident previously

evaluated.

A fuel handling accident could cause

a radioactive release of fission gases,

resulting in dose consequences. This

radioactive release of fission gases is

due to the failure of a certain number of

fuel pins which are postulated to fail

during the fuel handling accident. The

number of fuel pins which are

postulated to fail in this event is not

changed by the degradation of the

Boraflex panels in a post-seismic

condition. There are no criticality issues

with this fuel handling accident for the

reason described next. Although

conservative, should a fuel handling

accident occur during or after a seismic

event, even with no Boraflex credit, the

proposed 1750 ppm [parts per million]

of soluble boron is sufficient to ensure

that K-effective of the SFP is maintained

at less than or equal to 0.95. The 1750

ppm boron requirement also bounds any

criticality concerns for a fuel handling

or dropped load event due to the no

Boraflex assumption. Therefore, this

proposed change does not involve an

increase in the probability or
Federal Register / Vol. 62, No. 232 / Wednesday, December 3, 1997 / Notices 63981

consequences of an accident previously

evaluated.

Therefore, the proposed revision does

not involve a significant increase in the

probability or consequence of an

accident previously evaluated.

2. Create the possibility of a new or

different kind of accident from any

accident previously evaluated.

The change in the way Boraflex

responds to a seismic event with the

presence of 1750 ppm boron does not

create a new accident. The use of

soluble boron in the spent fuel pool is

safe. There is no possibility of a dilution

event during or following a seismic

event up to the magnitude of an SSE

[safe shutdown earthquake]. The

normally filled piping systems in the

vicinity of the spent fuel pool are fire

protection, hot water heating, hot water

preheating, domestic water, and

component cooling. In addition, the roof

drain system piping runs through the

building. An engineering review of

these systems has determined that the

majority of the systems are leak tight

and meet NU’s [Northeast Utilities’]

commitment to seismic II/I criteria for a

seismic event up to and including an

SSE. The analysis was performed

consistent with the original design

criteria for seismic II/I piping as

documented in section 3.9.2 of the

Millstone 3 Safety Evaluation Report

(SER) Number 4.

Portions of fuel building piping

systems that may not be leak tight

following an SSE, and that would not

leak into the spent fuel pool based on

location of the potential leak, are not

possible sources of dilution.

Two lines in the Hot Water Preheating

system will be modified to meet the leak

tight seismic II/I criteria and will not be

possible sources of dilution.

A new pipe support will be added to

the roof drain piping to meet the seismic

II/I criteria. With the new support

installed, one portion of the drain

piping will still not meet leak tight

requirements. The inlet opening on the

roof feeding this portion of the piping

will therefore be capped. Since the

location of the potential cracking in the

drain piping lies above the connection

to the balance of the drain piping, and

the system is not under pressure, water

flowing from other portions of the drain

system will not flow up to and out of

the potentially cracked portion. This

precludes a possible source of dilution.

Non borated water sources that are

connected to the SFP will be isolated

following a seismic event of greater than

or equal to an OBE to prevent dilution.

Therefore there is no possibility of a

SFP boron dilution accident coincident

with or following a seismic event up to

an SSE, and credit for soluble boron is

acceptable to meet the K-effective limit

of 0.95 for the SFP. The crediting of

soluble boron in the spent fuel pool to

control K-effective following a seismic

event does not create a new accident as

boron dilution of the pool can be

prevented by closing and

administratively controlling the opening

of dilution paths to the pool and

initiating routine sampling requirements

on SFP boron. At present the crediting

of soluble boron following a fuel

misplacement event is allowed for [in]

the Millstone 3 [TS]. Analysis has

shown that a seismic event of greater

than an OBE level earthquake can cause

Boraflex damage which can be more

limiting than a fuel misplacement event.

As such, the minimum boron

requirement in the fuel pool will be

increased from 800 ppm to 1750 ppm.

As such, no new accident has been

created because the crediting of boron

following a malfunction/accident has

always been allowed.

Therefore, the proposed revision does

not create the possibility of a new or

different kind of accident from any

accident previously evaluated.

3. Involve a significant reduction in a

margin of safety.

The margin of safety, as defined by

MP3 Technical Specifications, is to

ensure that the K-effective of the MP3

SFP is maintained less than or equal to

0.95 at all times. The proposed change

does not credit soluble boron during

normal operations, but allows crediting

soluble boron at a new higher

concentration for control of K-effective

during malfunction conditions. There is

no reduction in the margin of safety as

the result of the degradation of Boraflex

following a greater than OBE seismic

event, because soluble boron will

compensate for the loss of Boraflex. A

value of 1750 ppm of soluble boron in

the SFP at all times ensures that Keffective of the MP3 SFP is maintained

less than or equal to 0.95 at all times,

including this new malfunction of

degraded Boraflex following a greater

than OBE seismic event.

Eliminating the credit for the

reactivity [hold-down] effect of Boraflex

panels in conjunction with 1750 ppm

boron will have no effect on the

probability of a seismic event. As the

probability of a seismic event has not

changed there is no increase in the

probability of an accident or

malfunction due to a seismic event.

Following a seismic event, operators are

presently required to make inspections

of the plant to determine post seismic

event plant conditions. As a result of

this change, inspections will be required

to review the status of the spent fuel

pool and isolate potential dilution paths

following a seismic event of greater than

or equal to [an] OBE. These actions are

consistent with present guidance in the

seismic response procedure and do not

create an undue burden on the operator.

To compensate for the potential loss of

Boraflex after a seismic event, the SFP

is now required to be [] borated at all

times to at least 1750 ppm to maintain

the proper post seismic K-effective

condition. As such, there is no

mitigation equipment that has to operate

in the spent fuel pool following a

seismic event.

Although the Boraflex in the fuel

racks is assumed to fail in a seismic

event greater than an OBE, the presence

of soluble boron in the fuel pool water

will compensate for the loss of Boraflex.

Surveillance requirements on SFP boron

will ensure that there will be boron

present in the SFP and ensure that the

SFP is not diluted below the minimum

required boron concentration during

normal operation.

As the presence of SFP soluble boron

during and after a seismic event

maintains k-effective less than 0.95

there is no effect on the consequences

of any accidents evaluated. As there are

no new accidents created, there are no

changes in the consequences of

previously analyzed accidents, and

there is no effect on the consequences

of any accident. There is no reduction

in the margin of safety as the result of

the degradation of Boraflex following a

greater than OBE seismic event, because

during normal operations k-effective

remains less than 0.95 without reliance

on soluble boron, and during

malfunction and accident conditions

soluble boron can be used to

compensate for the loss of Boraflex to

maintain K-effective less than 0.95.

Therefore, the proposed revision does

not involve a significant reduction in a

margin of safety.

In conclusion, based on the

information provided, it is determined

that the proposed revision does not

involve an SHC.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, Connecticut, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut.
63982 Federal Register / Vol. 62, No. 232 / Wednesday, December 3, 1997 / Notices

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

P.O. Box 270, Hartford, Connecticut.

NRC Deputy Director: Phillip F.

McKee.

Omaha Public Power District, Docket

No. 50–285, Fort Calhoun Station, Unit

No. 1, Washington County, Nebraska

Date of amendment request: October

3, 1997.

Description of amendment request:

Omaha Public Power District (OPPD)

proposes to change the Fort Calhoun

Station Unit No. 1 Technical

Specifications (TS) by revising TS

Surveillance Requirement 3.9,

‘‘Auxiliary Feedwater System,’’ to

clarify what flow paths are required to

be tested. Additionally, OPPD proposes

to revise the auxiliary feedwater pumps’

surveillance requirements to delete the

specific discharge pressure.

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.

A change to TS 3.9(2) is proposed to

delete the specific discharge pressure

specified for the Auxiliary Feedwater

(AFW) pumps’ surveillance. The

developed head of the motor-driven and

steam turbine-driven AFW pumps is

verified quarterly. These tests are in

addition to those required by TS 3.3,

which implements ASME Section XI

Inservice Testing (IST) to evaluate a

pump’s performance against its pump

curve to determine operability. The IST

program is controlled by TS 3.3, and

requires that testing of ASME Code

Class 1, Class 2, and Class 3 pumps

shall be performed in accordance with

Section XI of the ASME Boiler and

Pressure Vessel Code, as required by 10

CFR 50.55a(g), except where specific

written relief has been granted by the

NRC. Therefore, the proposed change

does not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

The proposed changes to TS 3.9(4)

and the Basis Section only clarify the

AFW flow paths that are required to be

tested. The proposed change follows the

recommendations of NUREG–0635,

‘‘Generic Evaluation of Feedwater

Transients and Small Break Loss-ofCoolant Accidents in Combustion

Engineering Designed Operating

Plants,’’ Recommendation GS–6(2). No

physical changes are proposed,

information is being added to clarify the

testing required to meet the

recommendations of NUREG–0635,

therefore these proposed changes do not

involve a significant increase in the

probability or consequences of an

accident previously evaluated.

2. The proposed change does not

create the possibility of a new or

different kind of accident from any

accident previously evaluated.

There will be no physical alterations

to the plant configuration or changes in

operating modes. The proposed change

to delete the specific discharge pressure

of the AFW pumps from the TS is

consistent with the ASME Code Section

XI requirements that are controlled by

TS 3.3. Testing requirements of TS 3.3

require testing of ASME Code Class 1,

Class 2, and Class 3 pumps in

accordance with Section XI of the

ASME Boiler and Pressure Vessel Code,

as required by 10 CFR 50.55a(g), except

where specific written relief has been

granted by the NRC. The clarifications

being provided to describe the flow

paths only provide additional

information for testing required to meet

the recommendation of NUREG–0635.

Therefore, the proposed changes do

not create the possibility of a new or

different kind of accident from any

previously evaluated.

3. The proposed change does not

involve a significant reduction in a

margin of safety.

The proposed changes will not result

in any physical alterations to the plant

configuration or changes to the

application of setpoints or limits.

Therefore, the proposed change does not

involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska

68102.

Attorney for licensee: Perry D.

Robinson, Winston & Strawn, 1400 L

Street, N.W., Washington, DC 20005–

3502.

NRC Project Director: William H.

Bateman.

Public Service Electric & Gas Company,

Docket Nos. 50–272 and 50–311, Salem

Nuclear Generating Station, Unit Nos. 1

and 2, Salem County, New Jersey

Date of amendment request:

November 4, 1997.

Description of amendment request:

The amendments would change the

Emergency Diesel Generator (EDG)

Technical Specification (TS) 3/4.8.1 to

(1) delete 18-month surveillance

requirement 4.8.1.1.2.d.1, and (2)

eliminate the accelerated testing

requirement of Table 4.8–1. Both

changes have been approved on other

nuclear power facilities.

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 deleting the

requirement for an 18 month diesel

inspection is consistent with the

improved Standard Technical

Specifications (NUREG–1433) and does

not result in any changes to the existing

plant design. The Salem preventive

maintenance program utilizes diesel

generator performance history,

engineering analyses and

manufacturer’s recommendations as

appropriate for determining diesel

generator inspection requirements. The

Technical Specifications will continue

to contain surveillance requirements

that demonstrate the functional

capability of the diesel generators. The

change does not impact the ability of the

diesel generators or the AC electrical

power sources to perform their function,

nor result in a significant increase in the

consequences of any accident

previously evaluated. The diesel

generators will continue as designed.

PSE&G has implemented the

provisions of the maintenance rule for

EDG’s, including the appropriate

regulatory guidance. This provides a

program which assures EDG

performance. The elements of this

program include the performance of

detailed root cause analysis of

individual failures, effective corrective

actions taken in response to individual

failures, and implementation of

preventive maintenance consistent with

the Maintenance Rule. Additionally, the

proposed changes (elimination of

accelerated diesel generator testing

requirements of TS 4.8.1.1.a in lieu of

monthly testing and deletion of special
Federal Register / Vol. 62, No. 232 / Wednesday, December 3, 1997 / Notices 63983

reporting requirements for diesel

failures), do not delete the surveillance

requirements but rather set their

frequency at every 31 days. Monitoring

the effectiveness of EDG maintenance

and continuing surveillance testing will

ensure that the diesel generators will

perform their intended functions and

will minimize failures. As is noted in

the recommendations of GL [Generic

Letter] 94–01, because PSE&G is

monitoring and maintaining EDG

performance in accordance with the

provisions of 10 CFR 50.65, there is no

longer a need for special reporting

requirements.

Since the changes do not affect the

assurance of diesel generator reliability

or operability as discussed above, there

is no significant increase in the

probability or consequences of any

accident previously evaluated.

2. The proposed change does not

create the possibility of a new or

different kind of accident from any

accident previously analyzed.

This request does not result in any

change to the plant design or does it

involve a significant change in current

plant operation. The diesel generators

are inspected utilizing diesel generator

operating history, engineering analyses

and manufacturer’s recommendations as

appropriate, and the remaining

surveillance requirements continue to

demonstrate the functional capability of

the diesel generators.

Changing the surveillance of

frequency of TS 4.8.1.2.a to 31 days the

existing frequency as determined by

Table 4.8–1, does not create a new or

different kind of accident. Deleting of

special reporting requirements,

appropriate in light of the monitoring

and maintenance in conformance with

10 CFR 50.65, and reliance on the

reporting requirements of 10 CFR 50.72

and 10 CFR 50.73, does not create the

possibility of a new or different kind of

accident.

The proposed changes do not result in

any change to the plant design nor do

they involve a significant change in

current plant design. No new failure

modes will be introduced. Therefore,

the proposed changes will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The proposed change does not

involve a significant reduction in a

margin of safety.

The proposed request does not

adversely impact the reliability of the

diesel generators. As stated above, the

diesel generator operating history,

engineering analyses and the

manufacturer’s recommendations will

be utilized as appropriate to perform

diesel generator inspections.

Additionally, other Technical

Specification surveillance requirements

will continue to demonstrate the

functional capability of the diesel

generators. The diesel generators will

continue to perform their design

functions.

Noting the monitoring and

maintenance being performed in

conformance with 10 CFR 50.65,

revision of the frequency of surveillance

testing of 4.8.1.1.2.a does not adversely

impact the reliability of the diesel

generators. Deletion of the special

reporting requirements of 4.8.1.1.4 does

not impact the operability or the

reliability of the diesel generators.

This request does not involve an

adverse impact on diesel generator

operation or reliability. Since the diesel

generator function is not affected by the

proposed change, this request 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: Salem Free Public library, 112

West Broadway, Salem, NJ 08079.

Attorney for licensee: Jeffrie J. Keenan,

Esquire, Nuclear Business Unit—N21,

P.O. Box 236, Hancocks Bridge, NJ

08038.

NRC Project Director: John F. Stolz.

Southern Nuclear Operating Company,

Inc, Docket Nos. 50–348 and 50–364,

Joseph M. Farley Nuclear Plant, Units 1

and 2, Houston County, Alabama

Date of amendments request: October

16, 1997.

Description of amendments request:

The proposed amendments would

revise the Farley Nuclear Plant (FNP)

Units 1 and 2 Technical Specifications

(TS) to increase the allowable number of

charging pumps capable of injecting

into the reactor coolant system (RCS)

when the temperature of one or more of

the RCS cold legs is 180°F or less. The

amendments would also modify the

FNP TS to allow a maximum of two

charging pumps to be capable of

injecting into the RCS during pump

swap operations.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed changes do not

involve a significant increase in the

probability or consequences of an

accident previously evaluated.

The proposed changes to TS 3.1.2.3

allow two charging pumps to be capable

of injecting into the reactor coolant

system (RCS) for a period not to exceed

15 minutes while RCS cold leg

temperature is at or below 180 degrees

F. The intent is to allow the operator to

start a second pump long enough to

ensure that it operates properly and

then to promptly secure the pump that

was originally running. This order of

pump operation will allow seal

injection flow to be maintained to the

RCS pumps number one seal

continuously, thus preventing loss of

pressure to the seals and maintaining

filtered water flow through the seals.

The proposed revised bases address the

potential for [an] RCS mass addition

transient. Guidance is given to prevent

the charging pump swap from being

conducted while the RCS is in a

condition conducive to an overpressure

transient. The RCS should be in a non

water solid condition and the residual

heat removal (RHR) relief valves must

be operable or the RCS must be vented

while the pump swap evolution is in

progress. The proposed revision to TS

3.1.2.3 allows 15 minutes to have two

pumps capable of injecting into the

RCS, although two pumps will be

running only momentarily, the

remaining time is needed to perform the

charging pump circuit breaker racking

operations needed to render one of the

two pumps incapable of injecting into

the RCS. The proposed actions

statement 3.1.2.3b directs that

immediate action be taken to render all

but one pump inoperable should the

allotted 15 minutes be exceeded. This

action is more appropriate than is

currently specified. These proposed

changes include sufficient controls to

prevent an RCS overpressurization

event.

Therefore, the proposed TS changes

do not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

2. The proposed changes do not create

the possibility of a new or different kind

of accident from any accident

previously evaluated.

This proposed change involves no

change to the physical plant. It allows

for a very limited and controlled

operational change. The change

increases the potential for a mass

addition transient while the RCS is [at

or] below 180 degrees F; however,

sufficient controls are proposed to

prevent a cold overpressure event.
63984 Federal Register / Vol. 62, No. 232 / Wednesday, December 3, 1997 / Notices

Therefore, the proposed changes do

not create the possibility of a new or

different kind of accident from any

accident previously evaluated.

3. The proposed changes do not

involve a significant reduction in a

margin of safety.

The proposed change includes

controls sufficient to prevent a

significant reduction in the possibility

or consequences of an accident. The

proposed change specifies that the

pump swap evolution be performed

under conditions that will prevent an

adverse plant transient. In addition, the

proposed revision provides appropriate

operator action that does not currently

exist. This change is consistent with

NUREG 1431, Standard Technical

Specifications-Westinghouse Plants.

Therefore, the proposed changes do

not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Houston-Love Memorial

Library, 212 W. Burdeshaw Street, Post

Office Box 1369, Dothan, Alabama.

Attorney for licensee: M. Stanford

Blanton, Esq., Balch and Bingham, Post

Office Box 306, 1710 Sixth Avenue

North, Birmingham, Alabama.

NRC Project Director: Herbert N.

Berkow.

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.

Florida Power Corporation, et al.,

Docket No. 50–302, Crystal River Unit

No. 3 Nuclear Generating Plant, Citrus

County, Florida

Date of application for amendment:

October 4, 1997.

Brief description of amendment: The

proposed amendment revises the

description of the electrical controls for

Operating Reactor Building

Recirculation System Fan/Cooler

contained in the Final Safety Analysis

Report and Improved Technical

Specification Bases.

Date of publication of individual

notice in the Federal Register:

November 13, 1997 (62 FR 60921).

Expiration date of individual notice:

December 15, 1997.

Local Public Document Room

location: Coastal Region Library, 8619

W. Crystal River, Florida 34428.

Florida Power Corporation, et al.,

Docket No. 50–302, Crystal River Unit

No. 3 Nuclear Generating Plant, Citrus

County, Florida

Date of application for amendment:

October 31, 1997.

Brief description of amendment: The

proposed amendment revises Operating

License No. DPR–72, License Condition

2.C.(5) and deletes the requirement for

installation and testing of flow

indicators in the emergency core cooling

system.

Date of publication of individual

notice in the Federal Register:

November 12, 1997 (62 FR 60733).

Expiration date of individual notice:

December 12, 1997.

Local Public Document Room

location: Coastal Region Library, 8619

W. Crystal River, Florida 34428.

Florida Power Corporation, et al.,

Docket No. 50–302, Crystal River Unit

No. 3 Nuclear Generating Plant, Citrus

County, Florida

Date of application for amendment:

October 31, 1997.

Brief description of amendment: The

proposed amendment involves revisions

to the Crystal River 3 Technical

Specifications (TS) relating to decay

heat removal requirements in Mode 4.

Date of publication of individual

notice in the Federal Register:

November 12, 1997 (62 FR 60735).

Expiration date of individual notice:

December 12, 1997.

Local Public Document Room

location: Coastal Region Library, 8619

W. Crystal River, Florida 34428.

Florida Power Corporation, et al.,

Docket No. 50–302, Crystal River Unit

No. 3 Nuclear Generating Plant, Citrus

County, Florida

Date of application for amendment:

October 31, 1997.

Brief description of amendment: The

proposed amendment involves revisions

to the Crystal River 3 Technical

Specifications (TS) relating to the

methodology for post-loss of coolant

accident boron precipitation prevention.

Date of publication of individual

notice in the Federal Register:

November 12, 1997 (62 FR 60731).

Expiration date of individual notice:

December 12, 1997.

Local Public Document Room

location: Coastal Region Library, 8619

W., Crystal River, Florida 34428.

Notice of Issuance of Amendments to

Facility Operating Licenses

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application

complies with the standards and

requirements of the Atomic Energy Act

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

Commission’s rules and regulations.

The Commission has made appropriate

findings as required by the Act and the

Commission’s rules and regulations in

10 CFR Chapter I, which are set forth in

the license amendment.

Notice of Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing in

connection with these actions was

published in the Federal Register as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the applications for

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

the Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment as indicated. All of these

items are available for public inspection

at the Commission’s Public Document
Federal Register / Vol. 62, No. 232 / Wednesday, December 3, 1997 / Notices 63985

Room, the Gelman Building, 2120 L

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

local public document rooms for the

particular facilities involved.

Duke Energy Corporation, et al., Docket

No. 50–414, Catawba Nuclear Station,

Unit 2, York County, South Carolina

Date of application for amendment:

May 27, 1997.

Brief description of amendment: The

amendment deletes references to steam

generator tube sleeving and repair

criteria that will not be used for the

Westinghouse Model D5 steam

generators in use at Catawba Unit 2.

Also, unused paragraph numbers have

been deleted and a typographical error

has been corrected.

Date of issuance: November 13, 1997.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 154.

Facility Operating License No. NPF–

52: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 18, 1997 (62 FR 33122).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated November 13,

1997.

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 No. 50–

382, Waterford Steam Electric Station,

Unit 3, St. Charles Parish, Louisiana

Date of amendment request: October

7, 1997.

Brief description of amendment: The

amendment changes the Appendix A

Technical Specifications (TSs) by

modifying TS 3.3.3.7.3, and

Surveillance Requirements (SR)

4.3.3.7.3 for the broad range gas

detection system. Also it makes some

changes to the Bases in section 3/4.3.3.7

to incorporate information associated

with the existing toxic gas monitors.

Date of issuance: November 14, 1997.

Effective date: November 14, 1997, to

be implemented within 60 days.

Amendment No.: 135.

Facility Operating License No. NPF–

38: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 15, 1997 (62 FR

53660).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated November 14,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, LA 70122.

Entergy Operations, Inc., Docket No. 50–

382, Waterford Steam Electric Station,

Unit 3, St. Charles Parish, Louisiana

Date of amendment request: February

6, 1996.

Brief description of amendment: The

proposed change will amend the

Allowable Values of parameters in Table

3.3–4 of Waterford Steam Electric

Station, Unit 3, (Waterford 3) Technical

Specifications (TSs) to make it

consistent with the identical parameters

in Table 2.2–1 of TSs for Waterford 3.

The proposed change will add Mode 4

to surveillance requirements of Table

4.3–2, Item 5.c (Safety Injection System

Automatic Actuation Logic) that was

inadvertently removed. Finally, the

proposed change removes a reference to

TS 3.3.3.2 in Surveillance Requirements

TS 4.10.2.2 and 4.10.4.2 since Incore

Detectors has been removed from the

TSs.

Date of issuance: November 20, 1997.

Effective date: November 20, 1997, to

be implemented within 60 days.

Amendment No.: 136.

Facility Operating License No. NPF–

38: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 5, 1996 (61 FR 28615).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated November 20,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, LA 70122.

Houston Lighting & Power Company,

City Public Service Board of San

Antonio, Central Power and Light

Company, City of Austin, Texas, Docket

Nos. 50–498 and 50–499, South Texas

Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request: August

23, 1996, as supplemented by letters

dated October 1 and 15, 1996, and

January 28, 1997.

Brief description of amendments: The

amendments reflect the approval of the

transfer of the authority to operate

South Texas Project, Units 1 and 2,

under the licenses to a new operating

company, South Texas Project Nuclear

Operating Company.

Date of issuance: November 17, 1997.

Effective date: November 17, 1997.

Amendment Nos.: Unit 1—

Amendment No. 93; Unit 2—

Amendment No. 80.

Facility Operating License Nos. NPF–

76 and NPF–80. The amendments

revised the Technical Specifications and

the operating licenses.

Date of initial notice in Federal

Register: November 7, 1996 (61 FR

57719).

The additional information contained

in the supplemental letter dated January

28, 1997, was clarifying in nature and

thus, it was within the scope of the

initial notice and did not affect the

staff’s proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated November 17,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Wharton County Junior

College, J. M. Hodges Learning Center,

911 Boling Highway, Wharton, TX

77488.

Northeast Nuclear Energy Company, et

al., Docket No. 50–423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of application for amendment:

May 5, 1997.

Brief description of amendment:

Technical Specification Surveillance

4.8.4.1 requires periodic testing of lower

voltage circuit breakers for all

containment penetration conductor

overcurrent protective devices. The

amendment modifies the requirements

for determining the operability of lower

voltage circuit breakers by using the

manufacturer’s curve of current versus

time to test delay trip elements, clarifies

the use of two pole in series testing, and

expands the Bases description of the

testing.

Date of issuance: November 14, 1997.

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment No.: 153.

Facility Operating License No. NPF–

49: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 4, 1997 (62 FR 30637).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated November 14,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,
63986 Federal Register / Vol. 62, No. 232 / Wednesday, December 3, 1997 / Notices

Norwich, Connecticut 06360, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut 06385.

Northern States Power Company,

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

Island Nuclear Generating Plant, Unit

Nos. 1 and 2, Goodhue County,

Minnesota

Date of application for amendments:

May 15, 1997, as supplemented August

29, October 20, October 24, and October

28, 1997.

Brief description of amendments: The

amendments revise certain Technical

Specification (TS) limitations on reactor

coolant system leakage and steam

generator tube surveillance, and

implement a voltage-based repair

criteria per requirements of NRC

Generic Letter 95–05, ‘‘Voltage-Based

Repair Criteria for Westinghouse Steam

Generator Tubes Affected by Outside

Diameter Stress Corrosion Cracking.’’ In

addition, the amendments correct a

typographical error in TS Section 4.12.c.

Date of issuance: November 18, 1997.

Effective date: November 18, 1997,

with full implementation of the

Technical Specifications within 30

days. License Condition 5 of Appendix

B shall be implemented immediately

upon issuance of the amendments.

Amendment Nos.: 133 and 125.

Facility Operating License Nos. DPR–

42 and DPR–60: Amendments revised

the licenses and the Technical

Specifications.

Date of initial notice in Federal

Register: August 13, 1997 (62 FR

43371).

The August 29, October 20, October

24, and October 28, 1997, supplements

provided clarifying information that did

not change the staff’s initial proposed

no significant hazards consideration

determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated November 18,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Minneapolis Public Library,

Technology and Science Department,

300 Nicollet Mall, Minneapolis,

Minnesota 55401.

Rochester Gas and Electric Corporation,

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

Power Plant, Wayne County, New York

Date of application for amendment:

August 19, 1997, as supplemented

September 17, 1997.

Brief description of amendment: The

amendment revised the Ginna Station

Improved Technical Specifications to

correct an error in the required

accumulator borated water volume

specified in Surveillance Requirement

3.5.1.2.

Date of issuance: November 10, 1997.

Effective date: November 10, 1997.

Amendment No.: 69.

Facility Operating License No. DPR–

18: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 8, 1997 (62 FR 52587).

The September 17, 1997, letter

provided clarifying information that did

not change the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated November 10,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Rochester Public Library, 115

South Avenue, Rochester, New York

14610.

The Cleveland Electric Illuminating

Company, Centerior Service Company,

Duquesne Light Company, Ohio Edison

Company, OES Nuclear, Inc.,

Pennsylvania Power Company, Toledo

Edison Company, Docket No. 50–440,

Perry Nuclear Power Plant, Unit 1, Lake

County, Ohio

Date of application for amendment:

October 24, 1996, as supplemented June

16 and October 2, 1997.

Brief description of amendment: This

amendment revised the minimum

critical power ratio safety limit to reflect

the 10 CFR Part 21 condition reported

by General Electric in their letter to the

NRC dated May 24, 1996.

Date of issuance: November 7, 1997.

Effective date: November 7, 1997.

Amendment No.: 91.

Facility Operating License No. NPF–

58: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: February 12, 1997 (62 FR

6569). The June 16 and October 2, 1997,

supplemental letters provided

additional clarifying information and

did not change the initial no significant

hazards consideration determination.

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated November 7, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081.

Dated at Rockville, Maryland, this 25th day

of November 1997.

For the Nuclear Regulatory Commission.

Elinor G. Adensam,

Acting Director, Division of Reactor Projects—

III/IV, Office of Nuclear Reactor Regulation.

[FR Doc. 97–31522 Filed 12–2–97; 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 13e–1, SEC File No. 270–

255, OMB Control No. 3235–0305; Rule

12g3–2, SEC File No. 270–104, OMB Control

No. 3235–0119; Trust Indenture: Act Rules,

SEC File No. 270–115, OMB Control No.

3235–0132.

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

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Office of Management and Budget

requests for extension of the previously

approved collections of information

discussed below.

‘‘Purchase of Securities by issuer

thereof under the Securities Exchange

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Exchange Act is designed to provide

shareholders and the marketplace with

relevant information concerning issuer

repurchases during a tender offer for its

securities by a third party. Public

companies are the respondents. An

estimated 20 respondents will file

submissions annually at an estimated 13

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burden of 260 hours.

‘‘Securities Exchange Act of 1934—

Rule 12g3–2.’’ Rule 12g3–2 provides an

exemption for certain foreign securities.

It affects approximately 1,800 foreign

issuer respondents at an estimated one

burden hour per response for a total

annual burden of 1,800 hours.

‘‘Requirements as to Form and

Content of Applications, Statements and

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under the Trust Indenture Act of 1939

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complying with requirements under the

TIA. Persons and entities subject to TIA

requirements are the respondents. No

information collection burdens are

imposed directly by these rules so they

are assigned only one burden hour for

administrative convenience.

An agency may not conduct or

sponsor, and a person is not required to

respond to, a collection of information