Federal Register / Vol. 60, No. 234 / Wednesday, December 6, 1995 / Notices 62485

Tuesday, December 19, 1995—8:30 A.M.

until 6:00 P.M.

Wednesday, December 20, 1995—8:30

A.M. until 6:00 P.M.

Thursday, December 21, 1995—8:30

A.M. until 6:00 P.M.

During this meeting the Committee

plans to consider the following:

A. Review of NRC’s Programmatic

Approach to Low-Level Waste

Management. The Committee will

conclude its deliberations and issue a

report on the alternatives to the future

course of the NRC’s Low-Level

Radioactive Waste Disposal Program.

B. National Research Council/

National Academy of Science

Committee Report on the Technical

Bases for Yucca Mountain Standards.

The NRC staff will discuss with the

Committee its insights on the subject

report.

C. International Atomic Energy

Agency (IAEA) Activities. The

Committee will meet with a

representative of the IAEA to discuss

relevant waste-related activities.

D. Meeting with the Director, NRC’s

Division of Waste Management, Office

of Nuclear Materials Safety and

Safeguards. The Director will discuss

items of current interest related to the

Division of Waste Management

programs. Among the topics to be

discussed: pilot test of survey and

statistical methodology for site

decommissioning, status of HLW

program, and public comment on

program options for NRC’s LLW

program.

E. ACNW Priorities. The Committee

will review Task Action Plans for the

initial grouping of priority review issues

identified by the Committee.

F. Committee Activities/Future

Agenda. The Committee will consider

topics proposed for future consideration

by the full Committee and Working

Groups. The Committee will also

discuss ACNW-related activities of

individual members.

G. Miscellaneous. The Committee will

discuss miscellaneous matters related to

the conduct of Committee activities and

organizational activities and complete

discussion of matters and specific issues

that were not completed during

previous meetings, as time and

availability of information permit.

Procedures for the conduct of and

participation in ACNW meetings were

published in the Federal Register on

September 27, 1995 (60 FR 49924). In

accordance with these procedures, oral

or written statements may be presented

by members of the public, electronic

recordings will be permitted only

during those portions of the meeting

that are open to the public, and

questions may be asked only by

members of the Committee, its

consultants, and staff. Persons desiring

to make oral statements should notify

the Chief, Nuclear Waste Branch, Mr.

Richard K. Major, as far in advance as

practicable so that appropriate

arrangements can be made to allow the

necessary time during the meeting for

such statements. Use of still, motion

picture, and television cameras during

this meeting may 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 check with Mr. Major if such

rescheduling would result in major

inconvenience.

Further information regarding topics

to be discussed, whether the meeting

has been cancelled 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. EDT.

ACNW meeting notices, meeting

transcripts, and letter reports are now

available on FedWorld from the ‘‘NRC

MAIN MENU.’’ Direct Dial Access

number to FedWorld is (800) 303–9672;

the local direct dial number is 703–321–

3339.

The ACNW meeting dates for

Calendar Year 1996 are provided below:

ACNW Meeting No. and 1996 ACNW

Meeting Dates

81—January 24–26, 1996

82—March 27–29, 1996

83—May 2–4 or May 15–17, 1996 (TBD)

84—June 26–28, 1996

85—August 21–23, 1996

86—September 25–27, 1996

87—October 22–23, 1996

88—December 10–12, 1996

Dated: November 30, 1995.

Andrew L. Bates,

Advisory Committee Management Officer.

[FR Doc. 95–29661 Filed 12–5–95; 8:45 am]

BILLING CODE 7590–01–P

UNITED STATES 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

10, 1995, through November 24, 1995.

The last biweekly notice was published

on November 27, 1995 (60 FR 58395).

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
62486 Federal Register / Vol. 60, No. 234 / Wednesday, December 6, 1995 / Notices

expiration of the 30-day notice period.

However, should circumstances change

during the notice period such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received before

action is taken. Should the Commission

take this action, it will publish in the

Federal Register a notice of issuance

and provide for opportunity for a

hearing after issuance. The Commission

expects that the need to take this action

will occur very infrequently.

Written comments may be submitted

by mail to the Rules Review and

Directives Branch, Division of Freedom

of Information and Publications

Services, Office of Administration, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555, and should cite

the publication date and page number of

this Federal Register notice. Written

comments may also be delivered to

Room 6D22, Two White Flint North,

11545 Rockville Pike, Rockville,

Maryland from 7:30 a.m. to 4:15 p.m.

Federal workdays. Copies of written

comments received may be examined at

the NRC Public Document Room, the

Gelman Building, 2120 L Street, NW.,

Washington, DC. The filing of requests

for a hearing and petitions for leave to

intervene is discussed below.

By January 5, 1996, the licensee may

file a request for a hearing with respect

to issuance of the amendment to the

subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington DC, by

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

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

1-(800) 342–6700). The Western Union

operator should be given Datagram

Identification Number N1023 and the

following message addressed to (Project

Director): petitioner’s name and

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555, and to the attorney for the

licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

factors specified in 10 CFR

2.714(a)(1)(i)-(v) and 2.714(d).
Federal Register / Vol. 60, No. 234 / Wednesday, December 6, 1995 / Notices 62487

For further details with respect to this

action, see the application for

amendment which is available for

public inspection at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, and at the local public

document room for the particular

facility involved.

Arizona Public Service Company, et al.,

Docket Nos. STN 50–528, STN 50–529,

and STN 50–530, Palo Verde Nuclear

Generating Station, Units Nos. 1, 2, and

3, Maricopa County, Arizona

Date of amendments request: May 2,

1995

Description of amendments request:

The proposed change revises the largebreak loss-of-coolant accident (LOCA)

dose consequences. The large-break

LOCA dose calculation is being changed

to include an additional release path

through allowable steam generator tube

leakage to the atmospheric dump valves

(ADVs) or turbine bypass valves (TBVs).

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 probability or consequences of an

accident previously evaluated are not

significantly increased by this change to the

large break LOCA dose consequences. This

change has no effect on the LOCA safety

analysis for emergency core cooling system

performance, which demonstrates

conformance to the acceptance criteria of 10

CFR 50.46, as described in the PVNGS

Updated Final safety Analysis Section 6.3.3.

This change has no effect on structures,

systems or components prior to a LOCA or

any other accident. The new radiological

consequences of the revised large break

LOCA dose calculation are below 10 CFR 100

limits for the exclusion area boundary (EAB)

and low population zone (LPZ), and the 10

CFR 50, Appendix A, GDC 19 limits for the

control room, as shown in Table 1–1, Column

C. The NRC has previously approved changes

to the PVNGS LOCA dose consequences with

the acceptance criteria that the doses are still

within the guidelines set forth in 10 CFR 100

and GDC 19. This acceptance criteria is

described in the Safety Evaluation related to

amendment Nos. 64, 50, and 37 to PVNGS

Units 1, 2, and 3 respectively, dated

September 8, 1992.

The LOCA dose calculation is being

changed to include an additional release path

through allowable steam generator tube

leakage to the ADVs or TBVs. This change is

necessary to reflect a revised calculation

assumption that, following a large break

LOCA, the secondary system pressure would

fall below reactor coolant system pressure

and containment pressure when operators

cooldown the steam generators by using

ADVs or the TBVs (in accordance with the

safety analysis and EOPs [emergency

operating procedures]). It is desirable to use

the ADVs or TBVs to vent secondary system

steam and thus reduce heat input to the

reactor coolant system following a large break

LOCA. No other LOCA analysis assumptions

are being changed, and no changes are being

made to structures, systems, components or

procedures.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

This change has no impact on any

structures, systems, components, or

procedures. The only impact is the revised

radiological consequences of a large break

LOCA to include an additional release path,

as discussed in the response to Standard 1

above. Therefore, the proposed change does

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

This change to the large break LOCA dose

consequences does not involve a significant

reduction in a margin of safety. The new

radiological consequences of the revised

large break LOCA dose calculation are below

10 CFR 100 limits for the EAB and LPZ, and

the 10 CFR 50, Appendix A, GDC 19 limits

for the control room, as described in the

response to Standard 1 above. The NRC has

previously approved changes to the PVNGS

LOCA dose consequences with the

acceptance criteria that the doses are still

within the guidelines set forth in 10 CFR 100

and GDC 19. This acceptance criteria is

described in the Safety Evaluation related to

amendment Nos. 64, 50, and 37 to PVNGS

Units 1, 2, and 3 respectively, dated

September 8, 1992. No equipment

qualification is affected by the new

assumption of a release path through the

secondary system following a large break

LOCA, and no post LOCA radiation zones

will be changed. This change has no impact

on any structures, systems, components, or

procedures.

The NRC staff has reviewed the

licensee’s analysis and, based on that

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendments request

involve no significant hazards

consideration.

Local Public Document Room

location: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona

85004

Attorney for licensee: Nancy C. Loftin,

Esq., Corporate Secretary and Counsel,

Arizona Public Service Company, P.O.

Box 53999, Mail Station 9068, Phoenix,

Arizona 85072–3999

NRC Project Director: William H.

Bateman

Carolina Power & Light Company,

Docket No. 50–261, H. B. Robinson

Steam Electric Plant, Unit No. 2,

Darlington County, South Carolina

Date of amendment request:

November 22, 1995

Description of amendment request:

The current Technical Specifications

(TS) Section 3.3.4.2 describes the

limiting condition during which

components in the Service Water (SW)

system may be inoperable. The TS

Section 3.3.4.2 states, in part, ‘‘During

power operation, the requirements of

3.3.4.1 may be modified to allow any

one of the following components to be

inoperable provided the remaining

systems are in continuous operation.’’

The proposed change will delete the

qualifying statement,’’... provided the

remaining systems are in continuous

operation,’’ from TS Section 3.3.4.2.

Currently, this statement requires the

‘‘remaining systems to be in continuous

operation’’ while allowing one SW loop

header, or one SW pump, or one SW

booster pump to be inoperable for a

period of 24 hours.

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 would remove the

requirement for the remaining SW system

components to be in continuous operation

while one TS-required component is

inoperable. Rather, the remaining

components would remain operable, and no

change would be made in normal system

operation. The SW system provides an

accident mitigation function and is not

involved in accident initiation sequences.

Therefore, the proposed change would not

involve a significant increase in the

probability of an accident previously

evaluated.

The capacity of the SW system is such that

its accident mitigation function can be

performed by operation of a maximum of two

SW pumps, one SW booster pumps, and one

SW header. While a TS-required component

is inoperable, sufficient accident mitigation

capability is provided by the remaining

operable components, rather than requiring

the remaining systems to be in continuous

operation. Therefore, the proposed change

would not cause a significant increase in the

consequences of an accident previously

evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed change would remove the

requirement for the remaining SW system
62488 Federal Register / Vol. 60, No. 234 / Wednesday, December 6, 1995 / Notices

components to be in continuous operation

while one TS-required component is

inoperable. Rather, the remaining

components would remain operable. The

proposed change would not change the

normal operation of the system, nor would

any physical modifications result from the

change. The function and capability of the

SW systems would remain unchanged.

Therefore, the proposed change would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed change does not involve

a significant reduction in the margin of

safety.

The proposed change would remove the

requirement for the remaining SW system

components to be in continuous operation

while one allowed TS-required component is

inoperable. Rather, the remaining TSrequired components would remain operable.

Adequate assurance of operability is

maintained by performance of regular

surveillance testing. Maintaining operable

status rather than placing equipment in

continuous operation does not result in a

change in the ability of the SW system to

perform its intended function, since the

system provides an automatic response to

accident conditions, and the system

possesses adequate capacity to perform its

normal operating function with one allowed

TS-required component inoperable.

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: Hartsville Memorial Library,

147 West College Avenue, Hartsville,

South Carolina 29550

Attorney for licensee: R. E. Jones,

General Counsel, Carolina Power &

Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602

NRC Project Director: David B.

Matthews

Connecticut Yankee Atomic Power

Company, Docket No. 50–213, Haddam

Neck Plant, Middlesex County,

Connecticut

Date of amendment request: October

24, 1995

Description of amendment request:

The proposed amendment will increase

the trip setpoints and allowable values

for the low power block (P–7).

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:

In accordance with 10CFR50.92, CYAPCO

has reviewed the proposed change and has

concluded that it does not involve a

significant hazards consideration (SHC). The

basis for this conclusion is that the three

criteria of 10CFR50.92(c) are not

compromised. The proposed change does not

involve an SHC because the change would

not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change will relax the power

level values for the P–7 interlock by 2

percent. This change affects both the P–7 and

P–7N interlocks. The P–7 interlock affects

reactor trips on 1) low flow in more than one

reactor coolant loop, 2) reactor coolant pump

bus under voltage, 3) more than one reactor

coolant pump breaker open, 4) main steam

line isolation valve closure, 5) turbine trip,

and 6) variable low pressure. The P–7

interlock automatically blocks these reactor

trips on decreasing power and automatically

unblocks these reactor trips on increasing

power. The P–7N interlock affects the reactor

trip on wide range, neutron flux, high startup

rate. P–7N automatically enables this reactor

trip on decreasing power level and

automatically blocks this reactor trip on

increasing power level. The Applicable

Modes requirement and Action Statements

for the P–7 interlock and the reactor trips

associated with both P–7 and P–7N in the

Instrumentation Channel and Surveillance

Requirements of Technical Specification 3/

4.3.1 are being changed by 2 percent to be

consistent with the change to P–7. The

interlock setpoint cannot cause an accident.

Also, the proposed 2 percent increase in the

power level still results in a power level well

below the power level at which the P–7

interlocked reactor trips are required for

accident mitigation, as well as maintaining

the high startup rate trip enabled at a higher

power level. This proposed power level is

consistent with the technical specification

requirement prior to the conversion to

standard format technical specifications and

is also consistent with the Standard

Westinghouse technical specification value.

Therefore, the proposed change can neither

increase the consequences of the design basis

accident nor the probability of occurrence of

the design basis accidents.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed change only modifies the

power level for the P–7 and P–7N interlocks.

The proposed setpoint is a power level at

which stable plant conditions are easier to

maintain while transferring the power supply

for the reactor coolant pumps between offsite

power and the main generator. The setpoint

is also well below the power level for which

the reactor protection afforded by the trips

that are bypassed by P–7 is needed. This

cannot create the possibility of a new or

different kind of accident from any

previously analyzed.

3. Involve a significant reduction in a

margin of safety.

The proposed change maintains the power

level for the P–7 interlock below the power

level for which the reactor trips that are

blocked by the P–7 interlock are required. It

also raises the power level to a value at

which it is easier to maintain stable plant

conditions. This will reduce the likelihood of

an automatic reactor trip during the

transferring of power for the reactor coolant

pumps between offsite power and the main

generator. The proposed change will result in

the high startup rate reactor trip being

enabled at a higher power level. This is

conservative since it expands the range of

coverage for the trip. Therefore, the proposed

change does not impact the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Russell Library, 123 Broad

Street, Middletown, CT 06457.

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

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

NRC Project Director: Phillip F.

McKee

Connecticut Yankee Atomic Power

Company, Docket No. 50–213, Haddam

Neck Plant, Middlesex County,

Connecticut

Date of amendment request:

November 1, 1995

Description of amendment request:

The proposed amendment will modifiy

Surveillance Requirement 4.6.3.2,

‘‘Containment Isolation Valves,’’ (CIVs)

to change the surveillance interval from

at least once per 18 months to at least

once per refueling interval.

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:

CYAPCO has reviewed the proposed

change in accordance with 10CFR50.92 and

concluded that the change does not involve

a significant hazards consideration (SHC).

The basis for this conclusion is that the three

criteria of 10CFR50.92(c) are not

compromised. The proposed change does not

involve an SHC because the change would

not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change to Surveillance

Requirement 4.6.3.2 of the Haddam Neck

Plant Technical Specifications extends the

frequency for verifying that each CIV actuates

to its required position in response to a safety

injection actuation test signal. The proposal

would extend the frequency from at least

once per 18 months to at least once per
Federal Register / Vol. 60, No. 234 / Wednesday, December 6, 1995 / Notices 62489

refueling interval (24 months + 25% as

allowed by Technical Specification 4.0.2).

The proposed change to Surveillance

Requirement 4.6.3.2 does not alter the intent

or method by which the surveillance is

conducted, does not involve any physical

changes to the plant, does not alter the way

any structure, system, or component

functions, and does not modify the manner

in which the plant is operated.

Additional assurance of CIV operability is

provided by Surveillance Requirement

4.6.3.3. Surveillance Requirement 4.6.3.3

requires the confirmation of the mechanical

operability of the CIVs by the inservice

inspection program. The proposed change

does not modify these requirements.

Equipment performance over the last four

operating cycles was evaluated to determine

the impact of extending the frequency of

Surveillance Requirement 4.6.3.2. This

evaluation included a review of surveillance

results, preventive maintenance records, and

corrective maintenance records. It has been

concluded that the CIVs are highly reliable,

and that there is no indication that the

proposed extension could cause deterioration

in valve condition or performance.

As such, the proposed change to the

frequency of Surveillance Requirement

4.6.3.2 will not degrade the ability of the

CIVs to perform their safety function.

Based on the above, the proposed change

to Surveillance Requirement 4.6.3.2 of the

Haddam Neck Plant Technical Specifications

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

previously evaluated.

The proposed change to Surveillance

Requirement 4.6.3.2 of the Haddam Neck

Plant Technical Specifications extends the

frequency for verifying that each CIV actuates

to its required position in response to a safety

injection actuation test signal. The proposal

would extend the frequency from at least

once per 18 months to at least once per

refueling interval (24 months + 25% as

allowed by Technical Specification 4.0.2).

The proposed change does not alter the

intent or method by which the surveillance

is conducted, does not involve any physical

changes to the plant, does not alter the way

any structure, system, or component

functions, and does not modify the manner

in which the plant is operated. As such, the

proposed change in the frequency of

Surveillance Requirement 4.6.3.2 will not

degrade the ability of the CIVs to perform

their safety function.

Based on the above, the proposed change

to Surveillance Requirement 4.6.3.2 of the

Haddam Neck Plant Technical Specifications

will not create the possibility of a new or

different kind of accident from any

previously evaluated.

3. Involve a significant reduction in a

margin of safety.

The proposed change to Surveillance

Requirement 4.6.3.2 of the Haddam Neck

Plant Technical Specifications extends the

frequency for verifying that each CIV actuates

to its required position in response to a safety

injection actuation test signal. The proposal

would extend the frequency from at least

once per 18 months to at least once per

refueling interval (24 months + 25% as

allowed by Technical Specification Section

4.0.2).

The proposed change does not alter the

intent or method by which the surveillance

is conducted, does not involve any physical

changes to the plant, does not alter the way

any structure, system, or component

functions, and does not modify the manner

in which the plant is operated. As such, the

proposed change in the frequency of

Surveillance Requirement 4.6.3.2 will not

degrade the ability of the CIVs to perform

their safety function.

Additional assurance of the operability of

the CIVs is provided by Surveillance

Requirement 4.6.3.3.

Equipment performance over the last four

operating cycles was evaluated to determine

the impact of extending the frequency of

Surveillance Requirement 4.6.3.2. This

evaluation included a review of surveillance

results, preventive maintenance records, and

corrective maintenance records. It has been

concluded that the CIVs are highly reliable,

and that there is no indication that the

proposed extension could cause deterioration

in valve condition or performance.

Based on the above, the proposed change

to Surveillance Requirement 4.6.3.2 of the

Haddam Neck Plant Technical Specifications

does not involve a significant reduction in

the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Russell Library, 123 Broad

Street, Middletown, CT 06457.

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

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

NRC Project Director: Phillip F.

McKee

Consumers Power Company, Docket

No. 50–155, Big Rock Point Plant,

Charlevoix County, Michigan

Date of amendment request:

November 8, 1995, as supplemented

November 17, 1995

Description of amendment request:

The proposed amendment would

remove the prescriptive Type A

containment leakage test rate frequency

of 40 plus or minus 10 months and add

a reference to perform containment

leakage rate tests in accordance with the

criteria specified in Appendix J of 10

CFR Part 50 as amended by approved

exemptions. In addition, the proposed

amendment would revise the test

pressure for Type B and C testing to

correct a typographical error.

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:

Leakage test rate frequency

1) The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

This change is administrative in nature and

does not impact plant systems, structures or

components. The proposed change will allow

the facility’s technical specifications to be

revised to allow containment sphere leakage

testing in accordance with Appendix J to 10

CFR Part 50 as modified by approved

exemptions.

2) The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

This change is administrative in nature and

does not impact plant syst ems, structures or

components. The proposed change will allow

the facility’s technical specifications to be

revised to allow containment sphere leakage

testing in accordance with Appendix J to 10

CFR Part 50 as modified by approved

exemptions.

3)The proposed change does not involve a

significant reduction in a margin of safety.

This change is administrative in nature and

does not impact plant systems, structures or

components. The underlying purpose of

Appendix J is still achieved. Appendix J

states that the leakage test requirements

provide for periodic verification testing of the

leak tightness integrity of the primary reactor

containment. The appendix further states

that the purpose of the tests is to assure that

leakage through the primary containment

shall not exceed the allowable leakage rate

values as specified in the technical

specifications or associated bases. As stated

previously, for Big Rock Point and a large

percentage of other plants, the Appendix J

Type B and C testing programs provide the

most significant and meaningful assessment

of containment leak tightness. The testing

history and structural capability of the

containment establish that there is significant

assurance that the extended interval between

Type A tests will not adversely impact the

integrity of the containment.

Test pressure revision

As stated in the technical specification

change request, this revision is being

performed to be consistent with accident

pressure, Pa, used for Big Rock Point. 20 psig

is a typographical error. 23 psig has always

been used for these tests.

The proposed change does not:

1) involve a significant increase in the

probability or consequences of an accident

previously evaluated.

This change is administrative in nature and

does not impact plant systems, structures or

components. The proposed change will allow

the facility’s technical specifications to be

revised to reflect current containment sphere

leakage testing in accordance with Appendix

J to 10 CFR Part 50.
62490 Federal Register / Vol. 60, No. 234 / Wednesday, December 6, 1995 / Notices

2) create the possibility of a new or

different kind of accident from any accident

previously evaluated.

This change is administrative in nature and

does not impact plant systems, structures or

components. The proposed change will allow

the facility’s technical specifications to be

revised to reflect current containment sphere

leakage testing in accordance with Appendix

J to 10 CFR Part 50.

3) involve a significant reduction in a

margin of safety.

This change is administrative in nature and

does not impact plant systems, structures or

components.

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: North Central Michigan

College, 1515 Howard Street, Petoskey,

Michigan 49770

Attorney for licensee: Judd L. Bacon,

Esquire, Consumers Power Company,

212 West Michigan Avenue, Jackson,

Michigan 49201

NRC Project Director: Brian E. Holian,

Acting

Georgia Power Company, Oglethorpe

Power Corporation, Municipal Electric

Authority of Georgia, City of Dalton,

Georgia, Docket Nos. 50–424 and 50–

425, Vogtle Electric Generating Plant,

Units 1 and 2, Burke County, Georgia

Date of amendment request: October

16, 1995

Description of amendment request:

Appendix J of 10 CFR Part 50, ‘‘Primary

Reactor Containment Leakage Testing

for Water-Cooled Power Reactors,’’ has

recently been revised to include Option

B. This option allows the

implementation of a performance based

Type B and C testing program. The

proposed change will add a footnote to

Technical Specification (TS) 4.6.1.2.d

stating that the Type B and C tests

scheduled for Unit 1 refueling outage

Cycle 6 (1R6) will be conducted in

accordance with Option B and using the

guidance of Regulatory Guide 1.163,

Revision 0. This option is being

incorporated into the licensee’s request

to implement the improved TS.

However, the improved TS are not

scheduled to become effective until after

the Unit 1 refueling outage 1R6.

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 change does not involve a

significant increase in probability or

consequences of an accident previously

evaluated. The proposed change does not

involve a change to structures, systems, or

components which would affect the

probability or consequences of an accident

previously evaluated in the Vogtle Electric

Generating Plant (VEGP) Final Safety

Analysis Report (FSAR). The proposed

change only provides a mechanism within

the Technical Specifications for

implementing a performance-based method

of determining the frequency for leak rate

testing which has been approved by the NRC

via a revision to 10 CFR 50, Appendix J.

2. The proposed change will not create the

possibility of a new or different kind of

accident from any accident previously

analyzed. The amendment will not change

the design, configuration, or method of plant

operation. It only allows for the

implementation of Option B of 10 CFR 50,

Appendix J for Unit 1 refueling outage 1R6

without violating the plant Technical

Specifications.

3. Operation of VEGP, Unit 1, in

accordance with the proposed change will

not involve a significant reduction in the

margin of safety. The proposed change does

not affect a safety limit, an LCO [limiting

condition for operation], or the way plant

equipment is operated. The NRC is aware

that changes similar to this proposed change

are required in order to implement Option B

of 10 CFR 50, Appendix J. In fact, the staff

indicates in Paragraph V.B. of Appendix J

that Option B or parts thereof may be

adopted by a licensee 30 days after the rule

becomes effective by submitting notification

of its implementing plan and a request for

revision to Technical Specifications. Since

the NRC has approved the provision for

performance-based testing and must approve

this Technical Specification[] change before

the performance-based Option B can be

implemented, the margin of safety will not be

significantly reduced.

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: Burke County Public Library,

412 Fourth Street, Waynesboro, Georgia

30830

Attorney for licensee: Mr. Arthur H.

Domby, Troutman Sanders,

NationsBank Plaza, Suite 5200, 600

Peachtree Street, NE., Atlanta, Georgia

30308

NRC Project Director: Herbert N.

Berkow

Gulf States Utilities Company, Cajun

Electric Power Cooperative, and

Entergy Operations, Inc., Docket No.

50–458, River Bend Station, Unit 1,

West Feliciana Parish, Louisiana

Date of amendment request: May 30,

1995 (noticed in the Federal Register

July 5, 1995, (60 FR 35080) as

supplemented by letter dated November

20, 1995

Description of amendment request:

The proposed amendment would revise

the Technical Specifications as follows:

1. The Surveillance Frequency for the

drywell bypass test is changed from 18

months to 10 years with an increased

testing frequency required if

performance degrades.

2. The following changes are

requested for the drywell air lock

testing: (a) the leakage rate surveillance

is moved from the air lock Limiting

Condition for Operation (LCO) to the

drywell LCO, (b) the requirement for the

air lock to meet a specific overall

leakage limit is deleted, (c) the Note that

an inoperable air lock door does not

invalidate the previous air lock leakage

test is deleted, (d) the Note which

required that the air lock leakage test at

3 psid be preceded by pressurizing the

air lock to 19.2 psid is moved to the

bases, and (e) the Surveillance

Frequency for the air lock leakage test

and interlock test is changed from 18

months to 24 months.

3. The Actions Notes in the drywell

air lock LCO and the drywell isolation

valve LCO that identifies that the

Actions required by the drywell LCO

must be taken when the drywell bypass

leakage limit is not met is deleted.

4. The requirement for the drywell air

lock seal leakage rate to meet a specific

leakage limit is deleted.

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 for River Bend Station

(RBS) and Grand Gulf Nuclear Station

(GGNS), which is presented below:

I. The proposed change does not

significantly increase the probability or

consequences of an accident previously

evaluated.

The requested changes are either

administrative changes which clarify the

format of the requirement or change the

requirement to match the design bases of the

plant, a change which relocates the

requirement to the Technical Specification

Bases, or a change in surveillance interval.

Each of these types of change are discussed

below:

1. The administrative changes clarify the

format of the requirement or change

therequirement to match the design bases of

the plant. Clarifying administrative format of
Federal Register / Vol. 60, No. 234 / Wednesday, December 6, 1995 / Notices 62491

the Technical Specifications does not result

in any changes to the Technical Specification

requirements and, as a result, does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated. Also, changing the

requirements of the Technical Specifications

to more closely match the design bases of the

plant will continue to assure that the plant

will respond as assumed in the accident

analyses and, as a result, does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed changes relocate

information to the Technical Specification

Bases. In the Technical Specifications Bases

the relocated information will be maintained

in accordance with 10 CFR 50.59 and subject

to the change control provisions in Chapter

5 of Technical Specifications. Since any

changes to the Technical Specifications Bases

will be evaluated per the requirements of 10

CFR 50.59, no increase (significant or

insignificant) in the probability or

consequences of an accident previously

evaluated will be allowed. Therefore, this

change does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

3. The proposed changes in frequency for

the drywell bypass leakage and drywell air

lock surveillances will continue to ensure

that no paths exist through passive drywell

boundary components that would permit

gross leakage from the drywell to the primary

containment air space and result in

bypassing the primary containment pressuresuppression feature beyond the design basis

limit. The Mark III primary containment

system satisfies General Design Criterion 16

of Appendix A to 10 CFR Part 50. Maximum

drywell bypass leakage was determined

previously by reviewing the full range of

postulated primary system break sizes. The

limiting case was a primary system small

break loss of coolant accident (LOCA) and

yielded a design allowable drywell bypass

leakage rate limit of approximately 35,000

scfm for GGNS and 46,000 scfm (the

Technical Specification limit is based on a

lower limit of 40,110 scfm) for RBS. The

Technical Specifications acceptable limit for

the bypass leakage following a surveillance is

less than 10% of this design basis value. The

most recent bypass leakage value was

approximately 2.5% for GGNS and .91% for

RBS of the design allowable leakage rate limit

for the limiting event. EOI is committed to

maintaining programmatic and oversight

controls that ensure that drywell bypass

leakage remains a small fraction of the design

allowable leakage limit.

The drywell is typically exposed to

essentially 0 psig during normal plant

operation and 3 psig during drywell bypass

leak rate testing. These pressures are

considerably lower than the structural

integrity test pressure and are less likely to

initiate a crack or cause an existing crack to

grow. Visual inspections of the accessible

drywell surfaces that have been performed

since the structural integrity tests have not

revealed the presence of additional cracking

or other abnormalities. Therefore, additional

cracking of the drywell structure is not

expected due to testing or operation and,

similar to the justification for the ten year 10

CFR 50 Appendix J Type A test interval, it

is not considered credible for the passive

drywell structure to begin to leak sufficiently

to impact the design drywell bypass leakage

limit.

The primary containment’s ability to

perform its safety function is fairly

insensitive to the amount of drywell leakage,

thereby providing a margin to loss of the

drywell safety function that is not normally

available for safety systems. This

insensitivity is demonstrated by the

extremely high limiting event design basis

allowable leakage for the drywell (e.g., 35,000

scfm for GGNS and 46,000 scfm for RBS).

The limiting leakage is almost an order of

magnitude higher for other events.

Additionally, an even higher allowable

leakage can be realistically accommodated by

the primary containment due to the margins

in the containment design. Because of the

margins available, it will take valves in

multiple penetration flow paths leaking

excessively to cause the primary containment

to fail as a result of overpressurization, the

probability that drywell isolation valve

leakage will result in primary containment

failure due to excessive drywell leakage is

not considered significant and this drywell/

primary containment failure mode is not

considered credible.

The proposed Technical Specification

changes have no significant impact on the

GGNS Individual Plant Examination (IPE) or

the RBS IPE conducted per NRC Generic

Letter 88–20. The IPEs considered

overpressurization failure of primary

containment as part of the primary

containment performance assessment. Due to

the magnitude of acceptable drywell leakage

and the extremely low probabilities of

achieving such leakage, primary containment

failure due to preexisting excessive drywell

leakage was considered a non significant

contributor to primary containment failure.

Primary containment overpressurization

failure can occur with or without preexisting

excessive drywell leakage in a severe

accident. This is due to physical phenomena

associated with potentially extreme

environmental conditions inside primary

containment following a severe accident.

However, the calculated frequency of such

extreme conditions is very small. The

proposed changes do not impact the IPE

evaluated phenomena causing primary

containment overpressurization failure nor

significantly increase the probability that the

drywell has preexisting excessive leakage

and therefore would not contribute to these

accident scenarios.

For the reasons discussed above, the

proposed changes do not have any significant

risk impact to accidents previously evaluated

and do not significantly increase the

consequences of an accident previously

evaluated. Additionally, drywell leakage is

not the initiator of any accident evaluated;

therefore, changes in the frequency of the

surveillance for drywell leakage does not

increase the probability of any accident

evaluated.

Therefore, 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 requested changes are either

administrative changes which clarify the

format of the requirement or change the

requirement to match the design bases of the

plant, a change which relocates the

requirement to the Technical Specification

Bases, or a change in surveillance interval.

Each of these types of change are discussed

below:

1. The administrative changes in the

Technical Specification requirements do not

involve a physical alteration of the plant (no

new or different type of equipment will be

installed) nor does it change the methods

governing normal plant operation. Thus, this

change does not create the possibility of a

new or different kind of accident from any

accident previously evaluated.

2. The proposed relocation of requirements

does not involve a physical alteration of the

plant (no new or different type of equipment

will be installed) nor does it change the

methods governing normal plant operation.

The proposed change will not impose or

eliminate any requirements. Adequate

control of the information will be maintained

in the Technical Specification Bases. Thus,

the change proposed does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The proposed change modifies the

surveillance frequency for drywell bypass

leakage and drywell air lock surveillances.

The changes only impact the test frequency

and do not result in any change in the

response of the equipment to an accident.

The changes do not alter equipment design

or capabilities. The changes do not present

any new or additional failure mechanisms.

The drywell is passive in nature and the

surveillance will continue to verify that its

integrity has not deteriorated. Therefore, the

proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

III. The proposed change does not involve

a significant reduction in a margin of safety.

The requested changes are either

administrative changes which clarify the

format of the requirement or change the

requirement to match the design bases of the

plant, a change which relocates the

requirement to the Technical Specification

Bases, or a change in surveillance interval.

Each of these types of changes are discussed

below:

1. The administrative changes in the

Technical Specification requirements do not

involve a physical alteration of the plant (no

new or different type of equipment will be

installed) nor does it change the methods

governing normal plant operation. Thus, this

change does not cause a significant reduction

in the margin of safety.
62492 Federal Register / Vol. 60, No. 234 / Wednesday, December 6, 1995 / Notices

2. The relocation of requirements will not

reduce a margin of safety because it has no

impact on any safety analysis assumptions.

In addition, the requirements to be

transferred from the Technical Specifications

to the Technical Specifications Bases are the

same as the existing Technical

Specifications. Since any future changes to

these requirements in the Technical

Specifications Bases will be evaluated per the

requirements of 10 CFR 50.59, no reduction

(significant or insignificant) in a margin of

safety will be allowed.

3. The proposed change modifies the

surveillance frequency for drywell bypass

leakage and associated air lock surveillances.

Reliability of drywell integrity is evidenced

by the measured leakage rate during past

drywell bypass leakage surveillances.

Appropriate design basis assumptions will be

upheld, even when combined with the

complementary bypass leakage surveillances

as proposed. Drywell integrity will continue

to be tested by means of the proposed

periodic drywell bypass leakage test,

performance of the drywell air lock door

latching and interlock mechanism

surveillance, and performance of additional

surveillances including excercising of

drywell isolation valves. The combination of

these surveillances will provide adequate

assurance that drywell bypass leakage will

not exceed the design basis limit. Margins of

safety would not be reduced unless leakage

rates exceeded the design allowable drywell

bypass leakage limit. Therefore, the proposed

change does not cause a significant reduction

in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Government Documents

Department, Louisiana State University,

Baton Rouge, LA 70803

Attorney for licensee: Mark

Wetterhahn, Esq., Winston & Strawn,

1400 L Street, N.W., Washington, D.C.

20005

NRC Project Director: William D.

Beckner

Gulf States Utilities Company, Cajun

Electric Power Cooperative, and

Entergy Operations, Inc., Docket No.

50–458, River Bend Station, Unit 1,

West Feliciana Parish, Louisiana

Date of amendment request: October

26, 1995

Description of amendment request:

The proposed amendment would revise

the technical specifications for sixteen

editorial changes and would delete the

requirement for a program to prevent

and detect Asiatic Clams (Corbicula) in

the service water system (SWS). The

editorial changes covers such things as

removing systems or components that

do not exist in the River Bend Station,

correcting typographical errors,

correcting to be consistent with the

writers guide for Improved Technical

Specifications, adding descriptions for

systems to make them clear, and

wording changes to be consistent with

approved facility operations. The

Corbicula program is no longer needed

because the facility has been modified

and SWS no longer takes water from the

Mississippi River; source of the larvae

and infestation.

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:

EDITORIAL CHANGES

The purposed changes involves

reformatting, renumbering and rewording of

the existing Technical Specifications. The

reformatting, renumbering and rewording

process involves no technical changes to

existing Technical Specifications. As such,

these changes are administrative in nature

and do not impact initiators of analyzed

events or assumed mitigation of accident or

transient events. Therefore, these changes do

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed changes do not involve a

physical alteration of the plant (no new or

different type of equipment will be installed)

or changes in methods governing normal

plant operation. The proposed changes will

not impose or eliminate any new or different

requirements. Thus, these changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed changes will not reduce a

margin of safety because they have no impact

on any safety analysis assumptions. These

changes are administrative in nature. As

such, no question of safety is involved, and

the changes do not involve a significant

reduction in a margin of safety.

CORBICULA PROGRAM

The proposed change deletes the program

associated with the prevention and detection

of Asiatic Clams (Corbicula) based upon

improvements to the non-safety related

Normal Service Water System (SWS). The

source of makeup water to the SWS is no

longer the Mississippi River, which is the

source of Asiatic Clams. Demineralized water

or well water is used eliminating the source

of asiatic clams. To prevent biofouling SWS

is treated with chlorine/bromine. This

program is not considered as an initiator for

any previously evaluated accident. Therefore,

the proposed change will not increase the

probability or consequences of any accident

previously evaluated.

The proposed change introduces no new

mode of plant operation and it does not

involve a physical modification to the plant.

The possibility of the SES becoming

contaminated by any other means is highly

unlikely since it is a ‘‘closed-loop’’ system.

Therefore it does not create the possibility of

a new or different kind of accident from any

accident previously evaluated.

Prevention of Asiatic Clam infestation in

the SWS and associated safety-related

equipment is ensured by the ‘‘closed-loop’’

design of the SWS. Post Refuel Outage (RF–

4) inspections of the safety-related heat

exchangers that interface with the ‘‘closedloop’’ SWS have shown no evidence of clam

infestations. Therefore, the change does not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Government Documents

Department, Louisiana State University,

Baton Rouge, Louisiana 70803

Attorney for licensee: Mark

Wetterhahn, Esq., Winston & Strawn,

1400 L Street, N.W., Washington, D.C.

20005

NRC Project Director: William D.

Beckner

Gulf States Utilities Company, Cajun

Electric Power Cooperative, and

Entergy Operations, Inc., Docket No.

50–458, River Bend Station, Unit 1,

West Feliciana Parish, Louisiana

Date of amendment request:

November 20, 1995

Description of amendment request:

The proposed amendment would revise

the technical specifications to eliminate

the response time testing requirements

for selected Reactor Protection System

Instrumentation.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The purpose of the proposed Technical

Specification (TS) change is to eliminate

response time testing requirements for

selected components in the Reactor

Protection System (RPS). The Boiling Water

reactors Owners’ Group (BWROG) has

completed an evaluating which demonstrates

that response time testing is redundant to the

other TS-required testing. These other tests,

in conjunction with actions taken in response

to NRC Bulletin 90–01, ‘‘Loss of Fill-Oil in

Transmitters Manufactured by Rosemount,’’

and Supplement 1, are sufficient to identify

failure modes or degradation in instrument

response times and ensure operation of the

associated systems within acceptable limits.

There are no known failure modes that can

be detected by response time testing that

cannot also be detected by the other TSrequired testing. This evaluation was
Federal Register / Vol. 60, No. 234 / Wednesday, December 6, 1995 / Notices 62493

documented in NEDO–32291, ‘‘System

Analyses for Elimination of Selected

Response Time Testing Requirements,’’

January 1994. Entergy Operations, Inc. (EOI)

has confirmed the applicability of this

evaluation to River Bend Station (RBS). In

addition EOI will complete the actions

identified in the NRC staff’s safety evaluation

of NEDO–32291.

Because of the continued application of

other existing TS-required tests such as

channel calibration, channel checks, channel

functional tests, and logic system functional

tests, the response time of these systems will

be maintained within the acceptance limits

assumed in plant safety analyses and

required for successful mitigation of an

initiating event. The proposed changes do

not affect the capability of the associated

systems to perform their intended function

within their required response time, nor do

the proposed changes themselves affect the

operation of any equipment. As a result, EOI

has concluded that the proposed changes do

not involve a significant increase in the

probability or the consequences of an

accident previously evaluated.

The proposed changes only apply to the

testing requirements for the components

identified above and do not result in any

physical change to these or other components

or their operation. As a result, no new failure

modes are introduced. Therefore, the

proposed changes do not create the

possibility of a new or different kind of

accident from any accidents previously

evaluated.

The current TS-required response times are

based on the maximum allowable values as

assumed in the plant safety analyses. These

analyses conservatively establish the margin

of safety. As described above, the proposed

changes do not affect the capability of the

associated systems to perform their intended

function within the allowed response time

used as the basis for the plant safety analyses.

The potential failure modes for the

components within the scope of this request

were evaluated for impact on instrument

response time. This evaluation confirmed

that, with the exception of loss of fill-oil of

Rosemount transmitters, the remaining TSrequired testing is sufficient to identify

failure modes or degradation in instrument

response times and ensure operation of the

instrument within the scope of this request

is within acceptable limits. The actions taken

in response to NRC Bulletin 90–09 and

Supplement 1 are adequate to identify loss of

fill-oil failures of Rosemount transmitters. As

a result, it has been concluded that plant and

systems response to an initiating event will

remain in compliance with the assumptions

of the safety analysis.

Further, although not explicitly evaluated,

the proposed changes will provide an

improvement to plant safety and operation by

reducing the time safety systems are

unavailable, reducing the potential for safety

system actuations, reducing plant shutdown

risk, limiting radiation exposure to plant

personnel, and eliminating the diversion of

key personnel resources to conduct

unnecessary testing. Therefore, EOI has

concluded that this request will result in an

overall increase in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Government Documents

Department, Louisiana State University,

Baton Rouge, LA 70803

Attorney for licensee: Mark

Wetterhahn, Esq., Winston & Strawn,

1400 L Street, N.W., Washington, D.C.

20005

NRC Project Director: William D.

Beckner

North Atlantic Energy Service

Corporation, Docket No. 50–443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request:

September 22, 1995

Description of amendment request:

The proposed amendment would

modify a requirement of the Seabrook

Station, Unit No. 1 Technical

Specifications. Specifically, the

proposed amendment would change the

ACTION referenced in Table 3.3–3,

Engineered Safety Features Actuation

System Instrumentation, for Functional

Unit 8.b, Automatic Switchover to

Containment Sump/RWST Level LowLow. The ACTION requirement would

be changed to ACTION 15 from

ACTION 18. ACTION 15 requires an

inoperable channel to be placed in

bypass (with no time limit specified)

while ACTION 18 requires an

inoperable channel to be placed in the

tripped condition within 6 hours.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration. The NRC staff has

reviewed the licensee’s analysis against

the standards of 10 CFR 50.92(c). The

NRC staff’s review is presented below.

A. The change does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated (10 CFR 50.92(c)(1)) because the

proposed change would result in an

inoperable Functional Unit 8.b. protective

channel being placed in the bypassed

condition vice tripped condition. Functional

Unit 8.b. is not involved in any accident

initiation sequence; therefore, the probability

of a previously-analyzed accident is not

increased. Placing an inoperable Functional

Unit 8.b. in bypass vice trip reduces the

probability of premature opening of the

containment building sump isolation valves

thereby reducing the potential for increasing

the consequences of a previously-analyzed

accident. Thus, the consequences of a

previously-analyzed accident is not

increased.

B. The change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated (10 CFR 50.92(c)(2)) because the

change does not reduce the minimum

required number of channels of

instrumentation to be operable. The change

does not alter the function of or affect the

failure modes of Functional Unit 8.b.

instrumentation channels. The proposed

change does not otherwise affect the manner

by which the facility is operated, and it does

not involve any changes to equipment or

features which affect the operational

characteristics of the facility.

C. The change does not involve a

significant reduction in a margin of safety (10

CFR 50.92(c)(3)) because the change does not

reduce the minimum required number of

channels of instrumentation to be operable,

and it does not involve any changes to

equipment or features which affect the

operational characteristics of the facility.

Therefore, the protection previously

provided remains unchanged.

Based on this review, it appears that

the three standards of 10 CFR 50.92(c)

are satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Exeter Public Library,

Founders Park, Exeter, NH 03833.

Attorney for licensee: Lillian M.

Cuoco, Esquire, Northeast Utilities

Service Company, Post Office Box 270,

Hartford CT 06141–0270.

NRC Project Director: Phillip F.

McKee

Northeast Nuclear Energy Company, et

al., Docket No. 50–336, Millstone

Nuclear Power Station, Unit No. 2, New

London, Connecticut

Date of amendment request: May 26,

1995, supplemented and revised

October 20, 1995.

Description of amendment request:

The proposed changes would modify TS

3.8.1.1., ‘‘Electrical Power Systems, A.C.

Sources, Operating,’’ TS 3.8.1.2,

‘‘Electrical Power Systems, Shutdown,’’

TS 3.8.2.2, ‘‘Electrical Power Systems,

A.C. Distribution - Shutdown,’’ and TS

3.8.2.4, ‘‘Electrical Power Systems, D.C.

Distribution - Shutdown,’’ to provide

operational flexibility as well as

consistency between action statements

and to eliminate certain surveillance

requirements that are not applicable in

Modes 5 or 6.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration (SHC), which is presented

below:
62494 Federal Register / Vol. 60, No. 234 / Wednesday, December 6, 1995 / Notices

In accordance with 10 CFR 50.92, NNECO

has reviewed the proposed changes and has

concluded that they do not involve an SHC.

The basis for this conclusion is that the three

criteria of 10 CFR 50.92(c) are not

compromised. The proposed changes do not

involve an SHC because the change would

not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change to Surveillance

Requirement 4.8.1.1.1 is being made because

presently, the surveillance requirement for

demonstrating offsite sources are operable

states that ‘‘two’’ independent circuits are

required. The surveillance requirement is

referenced for both operating and shutdown

modes. While it is accurate for operating

modes, it is inconsistent with the limiting

condition for operation for shutdown. The

proposed change is safe because it renders

the surveillance requirement consistent with

the applicable limiting condition for

operation (i.e., operating or shutdown) and

eliminates a potential source of confusion.

The change to Surveillance Requirement

4.8.1.2 and Technical Specification 3.8.2.2

merely clarifies the diesel generator

surveillance and operability requirements for

Modes 5 and 6 and renders action statements

for related technical specification sections

consistent with and appropriate for

operational Modes 5 and 6.

Regarding diesel generator surveillance

requirements, automatic A.C. power for LNP

events in Modes 5 and 6 is not required. This

is validated by the fact that the undervoltage

sensors are only required to be operable in

Modes 1, 2 and 3 to meet technical

specifications. Because the undervoltage

sensors provide the logic that results in

actuation of the sequencer, it follows that the

sequencer need not be operable in Modes 5

and 6. Accordingly, the sequencer is not

required to support operability of the

available diesel generator in Modes 5 and 6.

Further, because SIAS is blocked in Modes

5 and 6, automatic start of the diesel

generator upon receipt of a SIAS is similarly

not required to support operability of the

diesel generator in Modes 5 and 6.

Additionally, operation of the diesel

generator in parallel with the system during

Modes 5 and 6 is not required to perform its

intended safety function. In fact, such

operation may compromise both sources as

the result of a single event.

Since automatic A.C. power is not credited

in the mitigation of Mode 5 and 6 events and

accidents, such as fuel handling accidents,

there is no increase in the probability or

consequences of previously evaluated

accidents.

The action statement in Technical

Specification 3.8.2.2 has been revised to cite

actions that are more appropriate for Modes

5 and 6 for Millstone Unit No. 2. This is due

to the ability to maintain the plant in a safe

condition without needing to automatically

load the diesel generator through the

sequencers in Modes 5 and 6. In addition, the

proposed change is consistent with the CE

Owner’s Group Standard Technical

Specification and with other Millstone Unit

No. 2 action statements. Consequently, there

is no increase in the probability or

consequences of previously evaluated

accidents.

The change to TS 3.8.2.4 merely renders

the action statement consistent with, and

appropriate for, operational Modes 5 and 6.

Since D.C. power is not credited in the

mitigation of Mode 5 and 6 events and

accidents, such as fuel handling accidents,

there is no increase in the probability or

consequences of previously evaluated

accidents.

The action statement in TS 3.8.2.4 has been

revised to cite actions that are more

appropriate for Modes 5 and 6 for Millstone

Unit No. 2. This is due to the ability to

maintain the plant in a safe condition

without D.C. power distribution available in

Modes 5 and 6. In addition, the proposed

change is consistent with the CE Owner’s

Group Standard Technical

Specifications (NUREG–1432) and with

other Millstone Unit No. 2 action statements.

Consequently, there is no increase in the

probability or consequences of previously

evaluated accidents.

2. Create the possibility of a new or

different kind of accident from any

previously evaluated.

The proposed changes do not alter or affect

the design, function, failure mode, or

operation of the plant. Therefore, the

proposed changes do not create the

possibility of a new or different kind of

accident from any previously evaluated.

3. Involve a significant reduction in a

margin of safety.

The proposed changes to the technical

specifications provides greater consistency

between the action statements and clarifies

which surveillance requirements are required

in Modes 5 and 6. Since the diesel generators

are not required to be loaded automatically

in Modes 5 and 6, and since it is part of our

shutdown risk management program to

assure that adequate cooling is able to be

provided, and since the diesel will still be

verified to start and achieve rated speed, the

proposed changes to the technical

specifications do not reduce the margin of

safety.

The proposed change to the TS provides

greater consistency among action statements

during Modes 5 and 6. Since the D.C.

distribution system is not credited in the

mitigation of Mode 5 and 6 events and

accidents, and since it is part of our

shutdown risk management program to

assure that adequate fuel cooling is able to be

provided, the proposed change to the TS

does not reduce the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360.

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

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

NRC Project Director: Phillip F.

McKee

Northeast Nuclear Energy Company, et

al., Docket No. 50–423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of amendment request: June 27,

1995, as supplemented July 21, 1995

Description of amendment request:

The amendment revises the Technical

Specifications (TS) to relocate TS

requirements for the containment purge

exhaust and supply valves, and to

remove a duplicate testing requirement

for the safety injection input from

engineered safety features from the 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:

... The proposed changes do not involve an

SHC [significant hazards consideration]

because the changes would not:

1. Involve a significant increase in the

probability or consequences of an accident

previously analyzed.

The first proposed change relocates the

operability and surveillance requirements for

the containment high range radiation

monitors from Technical Specification

Section 3.3.3 to Technical Specification

Section 3.3.2. The proposed changes are

administrative in nature. The proposed

changes do not alter the way any structure,

system, or component functions and do not

modify the manner in which the plant is

operated and do not involve any physical

changes to the plant.

The second proposed modification will

delete the testing requirement for functional

unit 16, ‘‘Safety Injection Input from ESF,’’

of Table 4.3–1 because the logic circuitry that

processes

the safety injection signals and produces a

reactor trip is tested under functional unit 19

‘‘Automatic Trip and Interlock Logic,’’ and

the testing is performed on a more frequent

basis (i.e., on a monthly staggered bases

versus on an 18-month frequency). In

addition, the same logic testing is

accomplished with an 18-month TADOT of

functional unit 1.a of Table 4.3–2 and with

a monthly staggered actuation logic testing of

functional unit 16 of Table 4.3–2. This testing

ensures that operability of the logic under

functional unit 16 of Table 4.3–1 is verified.

The other tests will continue to verify the

operability of the reactor trip system and that

a reactor trip will be initiated when required.

Therefore, there is no change in the

potential for an increase in the consequences

of an accident previously analyzed.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.
Federal Register / Vol. 60, No. 234 / Wednesday, December 6, 1995 / Notices 62495

The proposed changes do not affect the

operation or response of any plant equipment

or introduce any new failure mechanisms.

The proposed elimination of the testing

requirement line item does not affect the test

results since the logic circuitry that processes

the safety injection signal and produces a

reactor trip will be tested and is tested under

functional unit 19 of Table 4.3–1. As such,

the changes do not create the possibility of

a new or different kind of accident

previously evaluated.

3. Involve a significant reduction in the

margin of safety.

The proposed changes do not have any

adverse impact on the protective boundaries

nor do they affect the consequences of any

accident analyzed. The operability and

surveillance requirements, although

relocated to other technical specifications,

will still ensure that the system (the radiation

monitors) is tested and within limits. The

proposed elimination of the testing

equipment will not change the performance

or operating conditions of the safety systems.

The operable reactor trip system

instrumentation ensures that the assumptions

in the Bases of the Technical Specifications

are not affected and ensures that the margin

of safety is not reduced. Therefore, the

proposed changes do not reduce the margin

of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360.

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

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

NRC Project Director: Phillip F.

McKee

Pacific Gas and Electric Company,

Docket Nos. 50–275 and 50–323, Diablo

Canyon Nuclear Power Plant, Unit Nos.

1 and 2, San Luis Obispo County,

California

Date of amendment requests:

November 14, 1994

Description of amendment requests:

The proposed amendment would revise

the combined Technical Specifications

(TS) for the Diablo Canyon Nuclear

Power Plant, Unit Nos. 1 and 2, for the

slave relay test frequency from quarterly

(Q) to refueling (R). The request would

also remove table notation 4 from Table

4.3–2. The associated Bases would also

be appropriately revised.

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 results of WCAPs 14117 and 13878

demonstrate that slave relays are highly

reliable. The WCAPs also provide guidance

to assure that slave relays remain highly

reliable. The aging assessment concludes that

the age/temperature-related degradation of all

ND relays, and NE relays produced after May

1990, is sufficiently slow such that a

refueling frequency surveillance interval will

not significantly increase the probability of

slave relay failures. Finally, the evaluation of

the interposing slave relays in the emergency

diesel generator start circuitry, control room

ventilation and auxiliary building ventilation

realignments, steam generator blowdown

isolation and radwaste isolation systems has

concluded that based on the tests of the

interposing relays performed during other

equipment testing, reasonable assurance is

provided that failures will be identified if the

associated slave relays are tested on a

refueling frequency.

The removal of table notation 4 from TS

Table 4.3–2 is an administrative change that

eliminates unnecessary redundancy from the

TS and does not affect plant operation.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes do not alter the

performance of the ESFAS mitigation

systems assumed in the plant safety analysis.

Changing the interval for periodically

verifying ESFAS slave relays (assuring

equipment operability) will not create any

new accident initiators or scenarios.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated for DCPP.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed changes do not affect the

total ESFAS response assumed in the safety

analysis since the reliability of the slave

relays will not be significantly affected by the

increased surveillance frequency.

Therefore, the proposed changes do not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment requests

involve no significant hazards

consideration.

Local Public Document Room

location: California Polytechnic State

University, Robert E. Kennedy Library,

Government Documents and Maps

Department, San Luis Obispo, California

93407

Attorney for licensee: Christopher J.

Warner, Esq., Pacific Gas and Electric

Company, P.O. Box 7442, San

Francisco, California 94120

NRC Project Director: William H.

Bateman

South Carolina Electric & Gas

Company, South Carolina Public

Service Authority, Docket No. 50–395,

Virgil C. Summer Nuclear Station, Unit

No. 1, Fairfield County, South Carolina

Date of amendment request: August

18, 1995, as supplemented on November

1, 1995

Description of amendment request:

The proposed amendment would revise

the Operating License and Technical

Specifications to allow for a power

uprate to 2900 MWt. The current

maximum power level is 2775 MWt.

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 probability or consequences of an

accident previously evaluated is not

significantly increased.

Implementation of uprate power operation

does not contribute to any accident evaluated

in the FSAR [Final Safety Analysis Report].

The NSSS [Nuclear Steam Supply System]

Components (RV [reactor vessel], RCPs

[reactor coolant pumps], CRDMs [control rod

drive mechanisms], SGs [steam generators],

and piping) are compatible with the revised

operating conditions. These components

have been reanalyzed and the results show

that ASME [American Society of Mechanical

Engineers] Code requirements remain

satisfied and are within the current Licensing

Basis.

Interfacing Systems which are important to

safety are not adversely impacted and will

continue to perform their design function.

Overall secondary plant performance is not

significantly altered by the proposed

changes.

The revision to the Pressure Temperature

Limits will not adversely impact the RCS

[reactor coolant system] Pressure Boundary.

The length of time these curves will be

applicable, due to increased neutron fluence,

is being reduced. Before the 13 Effective Full

Power Years have elapsed, new curves will

be generated to reflect the analysis of the

specimen capsule and will be derived

utilizing NRC approved methodology.

Therefore, since the Reactor Coolant

pressure boundary integrity and system

functions are not adversely impacted, the

probability of occurrence of an accident

evaluated in the VCSNS [Virgil C. Summer

Nuclear Station] FSAR will be no greater

than the original design basis of the plant.

An extensive analysis has been performed

to evaluate the consequences of the following

accident types currently evaluated in the

VCSNS FSAR:
62496 Federal Register / Vol. 60, No. 234 / Wednesday, December 6, 1995 / Notices

- Non-LOCA [loss-of-coolant accident]

Events

- Large Break and Small Break LOCA

- Steam Generator Tube Rupture

With the [delta]75 SGs and revised

operating conditions, the calculated results

(i.e., DNBR [departure from nucleate boiling

ratio], Primary and Secondary System

Pressure, Peak Clad Temperature, Metal

Water Reaction, Challenge to Long Term

Cooling, Environmental Conditions Inside

and Outside containment, etc.) for the

accidents are similar to those currently

reported in the VCSNS FSAR and remain

within applicable Regulatory Acceptance

Criteria. Select results (i.e., Containment

Pressure during a Steam Line Break,

Minimum DNBR for Rod Withdrawal from

Subcritical, etc.) are slightly more limiting

than those currently reported in the FSAR

due to the use of the assumed operating

conditions with the [delta]75 SGs and in

some cases, use of an uprated core power of

2900 MWt. However, in all cases, the

calculated results do not challenge the

integrity of the primary/secondary/

containment pressure boundary and remain

within the regulatory acceptance criteria

applied to VCSNS’s current licensing basis.

Given that calculated radiological

consequences are not significantly higher

than current FSAR results and remain well

within 10 CFR 100 limits, it is concluded that

the consequences of an accident previously

evaluated in the FSAR are not significantly

increased.

2. The proposed license amendment does

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

Uprate power operation will not introduce

any new accident initiator mechanisms.

Structural integrity of the RCS is maintained

during all plant conditions through

compliance with the ASME code and 10 CFR

50 Appendix G requirements. Design

requirements of auxiliary systems are met

with the RSGs [replacement steam

generators] and uprate power operation. No

new failure modes or limiting single failures

have been identified. Since the safety and

design requirements continue to be met and

the integrity of the reactor coolant system

pressure boundary is not challenged, no new

accident scenarios have been created.

Therefore, the types of accidents defined in

the FSAR continue to represent the credible

spectrum of events to be analyzed which

determine safe plant operation.

3. The proposed license amendment does

not involve a significant reduction in a

margin of safety.

Although uprate power operation will

require changes to the VCSNS Technical

Specifications, the proposed changes are

supported by extensive LOCA, NON-LOCA

and SGTR [steam generator tube rupture]

analyses. These analyses show acceptable

consequences with margin to the applicable

regulatory limits. All equipment required to

function during accident conditions has been

shown to remain qualified and thus will

perform their design function, and all

components remain in compliance with the

codes and standards in effect when VCSNS

was originally licensed (with the exception of

the replacement steam generators which use

the 1986 ASME Code Section III Edition).

Low Temperature Overpressure transients

which could challenge RCS structural

integrity are not impacted by the revision to

the Pressure Temperature Limitations

Curves. The curves are not directly impacted,

the changes do not reduce any margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Fairfield County Library, 300

Washington Street, Winnsboro, SC

29180

Attorney for licensee: Randolph R.

Mahan, South Carolina Electric & Gas

Company, Post Office Box 764,

Columbia, South Carolina 29218

NRC Project Director: Frederick J.

Hebdon

The Cleveland Electric Illuminating

Company, Centerior Service Company,

Duquesne Light Company, Ohio Edison

Company, Pennsylvania Power

Company, Toledo Edison Company,

Docket No. 50–440, Perry Nuclear

Power Plant, Unit No. 1, Lake County,

Ohio

Date of amendment request: August

29, 1995

Description of amendment request:

The proposed amendment would revise

the Technical Specifications for

allowable values and trip setpoints for

selected plant process instrumentation.

The new allowable values/setpoints are

in accordance with the instrument

setpoint methodology accepted by the

NRC staff in a letter dated July 18, 1995.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration which is presented below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed revised Trip Setpoints and

Allowable Values are more conservative than

those currently approved in the Technical

Specifications. Therefore, any proposed

system or component actuations will occur

earlier, resulting in a more conservative plant

response. Thus, the proposed change does

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed change to the Technical

Specifications does not introduce any new

components nor does it modify the design of

any existing components. Other than making

Trip Setpoints and Allowable Values of

existing instrumentation more conservative,

the change does not affect the design or

function of any plant system, structure, or

component, nor does it change the way plant

systems are operated. Thus, the possibility of

a new or different kind of accident

previously evaluated is not created.

3. The proposed change does not result in

a significant reduction in the margin of

safety.

Since the proposed revised Trip Setpoints

and Allowable Values are more conservative

than the existing values, the margin of safety

would be increased by issuance of the

changes. Thus, the proposed change does not

result in a significant reduction in the margin

of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts & Trowbridge,

2300 N Street, NW., Washington, DC

20037

NRC Project Director: Gail H. Marcus

The Cleveland Electric Illuminating

Company, Centerior Service Company,

Duquesne Light Company, Ohio Edison

Company, Pennsylvania Power

Company, Toledo Edison Company,

Docket No. 50–440, Perry Nuclear

Power Plant, Unit No. 1, Lake County,

Ohio

Date of amendment request:

November 2, 1995

Description of amendment request:

The proposed amendment would revise

the Technical Specifications to allow

120 volt AC buses EV–1–A and EV–1–

B to be energized from either their

normal inverter power supply or from

their alternate power supply.

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:

These buses are not used as the initiator of

any analyzed accidents. Therefore, the

probability of any previously evaluated

accident has not increased. If an accident

were to occur while the buses are supplied

from the alternate power supply, there would
Federal Register / Vol. 60, No. 234 / Wednesday, December 6, 1995 / Notices 62497

be no change in the analyzed accident

scenario since even in the event of a loss of

offsite power event, the safety functions

would be completed. Thus, the consequences

of any previously evaluated accident have

not increased.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any previously evaluated:

The proposed change introduces no new

mode of plant operation and it does not

involve physical modification to the plant.

Therefore, it does not create the possibility of

a new or different kind of accident from any

accident previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety:

This change does not involve a significant

reduction in a margin of safety since the

proposed change maintains a safety related,

diesel-backed power supply to these buses

whether the power is supplied from the

inverters or from the alternate power supply.

If a loss of offsite power event were to occur

while the buses were supplied from the

alternate power source, the safety functions

being performed by components supplied

from these buses would occur. Thus, there

has been no reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts & Trowbridge,

2300 N Street, NW., Washington, DC

20037

NRC Project Director: Gail H. Marcus

The Cleveland Electric Illuminating

Company, Centerior Service Company,

Duquesne Light Company, Ohio Edison

Company, Pennsylvania Power

Company, Toledo Edison Company,

Docket No. 50–440, Perry Nuclear

Power Plant, Unit No. 1, Lake County,

Ohio

Date of amendment request:

November 2, 1995

Description of amendment request:

The proposed amendment to the Perry

Nuclear Power Plant Technical

Specifications revises those

specifications associated with handling

irradiated fuel in Primary Containment

and the Fuel Handling Building, and

selected specifications associated with

CORE ALTERATIONS. Specifically,

analysis identifies that only

—recently— irradiated fuel contains

sufficient fission products to require

OPERABILITY of accident mitigation

features to meet the accident analysis

assumptions. Analyses also show that

accident mitigation features such as

building INTEGRITY and engineered

safety feature (ESF) ventilation systems

are not required for CORE

ALTERATION events.

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 requirements are imposed

during specific activities which can be

postulated to result in significant radioactive

releases. The proposed APPLICABILITY

requirements are consistent with either the

original design basis analyses or with revised

analyses performed to support this proposed

amendment. Because the equipment

controlled by the revised Specifications is

not considered an initiator to any previously

analyzed accident, inoperability of the

equipment cannot increase the probability of

any previously evaluated accident.

Consistent with the original design

basis analysis, the reanalysis concludes

that radiological consequences of the

fuel handling accident are well within

the 10 CFR 100.11 limits, as defined by

acceptance criteria in Standard Review

Plan Section 15.7.4. The reanalysis has

previously been submitted to the

Nuclear Regulatory Commission for

review, and NRC confirmatory

calculations reached consistent results

(reference NRC Safety Evaluation for

License Amendment No. 35). The

results of the CORE ALTERATION

events other than the fuel handling

accident remain unchanged from the

original design basis, which showed

that these events do not result in fuel

cladding integrity damage or radioactive

releases. Therefore, the proposed

changes do not significantly increase the

consequences of any previously

evaluated accident.

Based on the above, the proposed changes

do not significantly increase 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

evaluated.

The proposed requirements are imposed

when specific activities represent situations

where significant radioactive releases can be

postulated. The proposed APPLICABILITY

requirements are consistent with design basis

analyses. The proposed changes do not

introduce any new modes of plant operation

and do not involve physical modifications to

the plant. Therefore, the proposed changes

do not create the possibility of a new or

different kind of accidident from any

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed change imposes controls to

ensure that during performance of activities

which represent situations where radioactive

releases are postulated, the radiological

consequences are at or below the established

licensing limit. Safety margins and analytical

conservatisms have been evaluated and are

well understood. Substantial conservatism is

retained to ensure that the analysis

adequately bounds all postulated event

scenarios. The current margin of safety is

retained.

Specifically, the margin of safety for the

fuel handling accident is the difference

between the 10 CFR 100 limits and the

licensing limit defined by the Standard

Review Plan (NUREG 0800), Section 15.7.4.

The licensing limit is defined by the

Standard Review Plan as being —well

within— the 10 CFR 100 limits, with ‘‘well

within’’ defined as 25% of the 10 CFR 100

limits for the fuel handling accident. Excess

margin is the difference between the

postulated doses and the corresponding

licensing limit. In the NRCs initial licensing

review of the Perry Nuclear Power Plant

(NUREG–0887, Section 15.3.3), the NRC

accepted the design and analyses based on

the results of the analyses being well within

the guideline values of 10 CFR 100.

The proposed APPLICABILITY

requirements continue to ensure that the

whole-body and thyroid doses at the

exclusion area and low population zone

boundaries as well as control room doses are

at or below the corresponding licensing limit.

The margin of safety is unchanged; therefore,

the proposed changes do not involve a

significant reduction in a margin of safety.

The margin of safety for the CORE

ALTERATION events other than the fuel

handling accident discussed above also

remains the same as in the original design

basis analyses, since the proposed changes

do not impact on the Technical Specification

requirements for systems needed to prevent

or mitigate such CORE ALTERATION events.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts & Trowbridge,

2300 N Street, NW., Washington, DC

20037

NRC Project Director: Gail H. Marcus

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
62498 Federal Register / Vol. 60, No. 234 / Wednesday, December 6, 1995 / Notices

requirements of the Atomic Energy Act

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

Commission’s rules and regulations.

The Commission has made appropriate

findings as required by the Act and the

Commission’s rules and regulations in

10 CFR Chapter I, which are set forth in

the license amendment.

Notice of Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for A Hearing in

connection with these actions was

published in the Federal Register as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the applications for

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

the Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, the Gelman Building, 2120 L

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

local public document rooms for the

particular facilities involved.

Commonwealth Edison Company,

Docket Nos. 50–295 and 50–304, Zion

Nuclear Power Station, Units 1 and 2,

Lake County, Illinois

Date of application for amendments:

October 6, 1995, and supplemented

November 20, 1995

Brief description of amendments: The

amendments revise the Technical

Specifications by incorporating a new

acceptance criterion for steam generator

tubes with degradation in the tubesheet

roll expansion region.

Date of issuance: November 21, 1995

Effective date: November 21, 1995

Amendment Nos.: 172 and 159

Facility Operating License Nos. DPR–

39 and DPR–48: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: October 16, 1995 (60 FR

53648) The supplemental letter

provided clarifying information that did

not affect the initial proposed no

significant hazards consideration

determination. The Commission’s

related evaluation of the amendments is

contained in a Safety Evaluation dated

November 21, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois

60085

Connecticut Yankee Atomic Power

Company, Docket No. 50–213, Haddam

Neck Plant, Middlesex County,

Connecticut

Date of application for amendment:

August 10, 1995

Brief description of amendment: The

amendment revises the Haddam Neck

Technical Specification Section 3/4.4.3,

‘‘Pressurizer,’’ to add a footnote to allow

the pressurizer level to be controlled,

outside of the programmed level,

between 25 to 50 percent, plus or minus

5 percent in Mode 3 when the reactor

coolant system is borated to the required

Mode 5 concentrations.

Date of Issuance: November 14, 1995

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment No.: 186

Facility Operating License No. DPR–

61. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 11, 1995 (60 FR

52928) The Commission’s related

evaluation of this amendment is

contained in a Safety Evaluation dated

November 14, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Russell Library, 123 Broad

Street, Middletown, CT 06457

Duquesne Light Company, et al., Docket

Nos. 50–334 and 50–412, Beaver Valley

Power Station, Unit Nos. 1 and 2,

Shippingport, Pennsylvania

Date of application for amendments:

September 13, 1995, as supplemented

October 16,1995

Brief description of amendments:

These amendments revise the

Administrative Controls section of the

BVPS–1 and BVPS–2 TSs to make them

consistent with the requirements of the

Offsite Dose Calculation Manual

(ODCM). The ODCM was recently

updated to reflect the radioactive liquid

and gaseous effluent release limits and

the liquid holdup tank activity limit of

BVPS–1 License Amendment No. 188

and BVPS–2 License Amendment No.

70 which were issued June 12, 1995.

Date of issuance: November 21, 1995

Effective date: As of the date of

issuance, to be implemented within 10

days.

Amendment Nos.: 194 and 77

Facility Operating License Nos. DPR–

66 and NPF–73: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: September 22, 1995 (60 FR

49292) The October 16, 1995, letter did

not change the initial proposed no

significant hazards consideration

determination or expand the

amendment request beyond the scope of

the September 22, 1995, Federal

Register notice. The Commission’s

related evaluation of the amendments is

contained in a Safety Evaluation dated

November 21, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa,

Pennsylvania 15001

Entergy Operations, Inc., Docket No.

50–382, Waterford Steam Electric

Station, Unit 3, St. Charles Parish,

Louisiana

Date of amendment request: February

14, 1994, as supplemented by letters

dated July 25, August 15, and August

29, 1995

Brief description of amendment: The

amendment changes the Appendix A

Technical Specifications (TSs) to make

them consistent with the revised 10 CFR

Part 20, Standards for Protection

Against Radiation.

Date of issuance: November 17, 1995

Effective date: November 17, 1995

Amendment No.: 116

Facility Operating License No. NPF–

38. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 30, 1994 (59 FR 14888)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated November 17, 1995.

The July 25, August 15, and August 29,

1995 letters provided clarifying

information that did not change the

initial propose no significance hazards

consideration determination.

No significant hazards consideration

comments received: No

Local Public Document Room

location: University of New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, LA 70122

Georgia Power Company, Oglethorpe

Power Corporation, Municipal Electric

Authority of Georgia, City of Dalton,

Georgia, Docket Nos. 50–424 and 50–

425, Vogtle Electric Generating Plant,

Units 1 and 2, Burke County, Georgia

Date of application for amendments:

May 12, 1995, as supplemented by

letters dated July 6 and October 2, 1995.
Federal Register / Vol. 60, No. 234 / Wednesday, December 6, 1995 / Notices 62499

Brief description of amendments: The

amendments revise Technical

Specification Surveillance Requirement

4.6.1.2 to add the provision that 10 CFR

Part 50, Appendix J, applies, except as

modified by NRC-approved exemptions.

Date of issuance: November 17, 1995

Effective date: As of the date of

issuance to be implemented within 30

days

Amendment Nos.: 91 and 69

Facility Operating License Nos. NPF–

68 and NPF–81: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: July 5, 1995 (60 FR 35078) The

July 6 and October 2, 1995, letters

provided clarifying information that did

not change the scope of the May 12,

1995, application and initial proposed

no significant hazards consideration

determination. The Commission’s

related evaluation of the amendments is

contained in a Safety Evaluation dated

November 17, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Burke County Library, 412

Fourth Street, Waynesboro, Georgia

30830

Northeast Nuclear Energy Company,

Docket No. 50–245, Millstone Nuclear

Power Station, Unit 1, New London

County, Connecticut

Date of application for amendment:

July 28, 1995, as supplemented

September 12, October 18, and October

31, 1995.

Brief description of amendment: In

order to support a full-core offload as a

normal end-of-cycle event, the

amendment adds License Condition

2.C(6) and will require that: (1) the

reactor be subcritical for at least 100

hours prior to the start of reactor

refueling operations, (2) the spent fuel

pool bulk temperature be maintained

less than or equal to 140—F, and (3) two

trains of shutdown cooling be operable

during reactor refueling operations.

Date of issuance: November 9, 1995

Effective date: As of the date of

issuance.

Amendment No.: 89

Facility Operating License No. DPR–

21. Amendment revised the license.

Date of initial notice in Federal

Register: August 30, 1995 (60 FR 45180)

The September 12, October 18, and

October 31, 1995, submittals provided

additional information that did not

change the initial proposed no

significant hazards consideration

determination. The Commission’s

related evaluation of the amendment

and Final No Significant Hazards

Consideration Determination are

contained in a Safety Evaluation dated

November 9, 1995.

No significant hazards consideration

comments received: No public

comments received. A request for a

hearing was received from We the

People, the Seacoast Anti-Pollution

League, the New England Coalition on

Nuclear Pollution, and Donald Del Core

of Uncasville, Connecticut.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360

Northeast Nuclear Energy Company, et

al., Docket No. 50–336, Millstone

Nuclear Power Station, Unit No. 2, New

London County, Connecticut

Date of application for amendment:

October 6, 1995, supplemented October

23, November 2, and November 15,

1995.

Brief description of amendment: The

amendment adds footnotes to Action

Statement (AS) 3.8.1.1.a of the

Technical Specification (TS) and its

bases to allow a one-time extension of

the allowed outage time (AOT) for an

inoperable offsite power source from the

current 72 hours to 7 days.

Date of issuance: November 22, 1995

Effective date: As of the date of

issuance, to be implemented within 30

days.

Amendment No.: 192

Facility Operating License No. DPR–

65. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 17, 1995 (60 FR

53812). The October 23, November 2,

and November 15, 1995, letters

provided clarifying information and

slight modifications to the original

request that were not outside the scope

of the original notice and 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 22, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360

Northern States Power Company,

Docket No. 50–282, Prairie Island

Nuclear Generating Plant, Unit No. 1,

Goodhue County, Minnesota

Date of application for amendment:

January 10, 1995, as supplemented

August 9 and September 20, 1995.

Brief description of amendment: The

amendments revise the Prairie Island

event monitoring instrumentation

Technical Specifications and associated

Bases to conform to Standard Technical

Specifications for post-accident

monitoring.

Date of issuance: November 9, 1995

Effective date: November 9, 1995,

with full implementation within 30

days.

Amendment Nos.: 121/114

Facility Operating License No. DPR–

42 and DPR–60. Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: February 15, 1995 (60 FR

8753) The August 9 and September 20,

1995, letters provided updated

Technical Specification pages and

clarifying information in response to

discussions with the staff during various

teleconferences conducted during the

review process. This information was

within the scope of the original

application and did not change the

staff’s initial proposed no significant

hazards consideration determination.

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated November 9, 1995.

No Significant hazards consideration

comments received: No

Local Public Document Room

location: Minneapolis Public Library,

Technology and Science Department,

300 Nicollet Mall, Minneapolis,

Minnesota 55401

Philadelphia Electric Company, Docket

Nos. 50–352 and 50–353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of application for amendments:

December 2, 1994, as supplemented

May 12, 1995.

Brief description of amendments:

These amendments relocate the fire

protection requirements from the

Technical Specifications to the Updated

Final Safety Analysis Report in

accordance with the guidance in

Generic Letter (GL) 86–10,

‘‘Implementation of Fire Protection

Requirements,’’ and GL 88–12,

‘‘Removal of Fire Protection

Requirements from Technical

Specifications.’’

Date of issuance: November 20, 1995

Effective date: As of date of issuance,

both units, to be implemented within 30

days.

Amendment Nos.: 104 and 68

Facility Operating License Nos. NPF–

39 and NPF–85. The amendments

revised the Technical Specifications and

the License.

Date of initial notice in Federal

Register: April 26, 1995 (60 FR 20524)

The supplemental letter provided

clarifying information and did not
62500 Federal Register / Vol. 60, No. 234 / Wednesday, December 6, 1995 / Notices

change the initial proposed no

significant hazards consideration

determination. The Commission’s

related evaluation of the amendments is

contained in a Safety Evaluation dated

November 20, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania

19464

Philadelphia Electric Company, Docket

Nos. 50–352 and 50–353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of application for amendments:

September 14, 1995 and supplemented

by letter dated October 27, 1995

Brief description of amendments:

These amendments revise the technical

specifications by deleting Reactor

Enclosure and Refueling Area

Secondary Containment Isolation Valve

Tables 3.6.5.2.1–1 and 3.6.5.2.2–1, and

references to them, in accordance with

Generic Letter 91–08, ‘‘Removal of

Component lists from Technical

Specifications.’’ The TS have been

modified to state requirements in

general terms that include the

components listed in the tables removed

from the TS.

Date of issuance: November 20, 1995

Effective date: As of date of issuance,

to be implemented within 30 days.

Amendment Nos.: November 20, 1995

Facility Operating License Nos. NPF–

39 and NPF–85. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: October 11, 1995 (60 FR

52934) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

November 20, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania

19464

Tennessee Valley Authority, Docket No.

50–296, Browns Ferry Nuclear Plant,

Unit 3, Limestone County, Alabama

Date of application for amendments:

October 4, 1995 (TS 368)

Brief description of amendment: The

amendment delete requirements for

daily checks for certain instruments that

do not have indications, and provides

editorial changes.

Date of issuance: November 13, 1995

Effective Date: November 13, 1995

Amendment No.: 202

Facility Operating License No. DPR–

68: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 11, 1995 (60 FR

52935) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

November 13, 1995.

No significant hazards consideration

comments received: None

Local Public Document Room

location: Athens Public library, South

Street, Athens, Alabama 35611

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of application for amendments:

August 7, 1995 (TS 95–03)

Brief description of amendments: The

amendments address operation with a

rod urgent failure condition, including

limited operation with one control or

shutdown bank inserted up to 18 steps

below its insertion point. In addition,

the surveillance interval for rod

movement verifications has been

increased from 31 to 92 days.

Date of issuance: November 21, 1995

Effective date: November 21, 1995

Amendment Nos.: 215 and 205

Facility Operating License Nos. DPR–

77 and DPR–79: Amendments revise the

technical specifications.

Date of initial notice in Federal

Register: August 30, 1995 (60 FR 45186)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated November 21, 1995.

No significant hazards consideration

comments received: None

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402

The Cleveland Electric Illuminating

Company, Centerior Service Company,

Duquesne Light Company, Ohio Edison

Company, Pennsylvania Power

Company, Toledo Edison Company,

Docket No. 50–440, Perry Nuclear

Power Plant, Unit No. 1, Lake County,

Ohio

Date of application for amendment:

April 28, 1995

Brief description of amendment: The

amendment removes the license

conditions for the Transamerica Delaval,

Inc. emergency diesel generators

specified by paragraph 2.C.(9) and

defined in Attachment 2 to the

Operating License.

Date of issuance: November 16, 1995

Effective date: November 16, 1995

Amendment No.: 74

Facility Operating License No. NPF–

58: This amendment revises the license.

Date of initial notice in Federal

Register: June 6, 1995 (60 FR 29889)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated November 16, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50–346, Davis-Besse Nuclear Power

Station, Unit No. 1, Ottawa County,

Ohio

Date of application for amendment:

June 23, 1995, and facsimile

transmission dated October 31, 1995

Brief description of amendment: This

amendment relocates TS 3/4.3.3.3,

‘‘Seismic Instrumentation;’’ TS 3/

4.3.3.4, ‘‘Meteorological

Instrumentation;’’ and TS 3/4.4.11,

‘‘Reactor Coolant System Vents;’’ and

the Bases for each of the three sections

from the TS to the Updated Safety

Analysis Report, and eliminates the

special reporting requirements for

inoperable seismic and meteorological

monitoring instrumentation from TS

6.9.2.

Date of issuance: November 14, 1995

Effective date: November 14, 1995, and

shall be implemented not later than 90

days after issuance.

Amendment No.: 201

Facility Operating License No. NPF–3:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 2, 1995 (60 FR 39455)

The October 31, 1995, facsimile

transmission was clarifying in nature

and did not affect the initial no

significant hazards consideration

determination. The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

November 14, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: University of Toledo, William

Carlson Library, Government

Documents Collection, 2801 West

Bancroft Avenue, Toledo, Ohio 43606

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50–346, Davis-Besse Nuclear Power

Station, Unit No. 1, Ottawa County,

Ohio

Date of application for amendment:

June 7, 1995

Brief description of amendment: This

amendment revises Technical

Specification 3/4.9.4, Refueling

Operations - Containment Penetrations;
Federal Register / Vol. 60, No. 234 / Wednesday, December 6, 1995 / Notices 62501

Bases 3/4.9.4, Containment

Penetrations; and Limiting Condition for

Operation (LCO) 3.9.4.b to allow both

doors of the containment personnel

airlock to be open during core

alterations or movement of irradiated

fuel within the containment, provided

that certain specified conditions are

meet. Additional changes revise or

clarify TS LCO 3.9.4.c, TS Action

3.9.4.a, and TS Surveillance

Requirement 4.9.4, and modify the

associated Bases.

Date of issuance: November 17, 1995

Effective date: November 17, 1995

Amendment No.: 202

Facility Operating License No. NPF–3.

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 2, 1995 (60 FR 39454)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated November 17, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: University of Toledo, William

Carlson Library, Government

Documents Collection, 2801 West

Bancroft Avenue, Toledo, Ohio 43606

TU Electric Company, Docket Nos. 50–

445 and 50–446, Comanche Peak Steam

Electric Station, Unit Nos. 1 and 2,

Somervell County, Texas

Date of amendment request:

December 6, 1994

Brief description of amendments:

These changes revise Technical

Specifications to allow appropriate

remedial action for high particulate

levels in the diesel generator fuel oil

inventory and other out-of-limit

properties in new diesel generator fuel

oil that has been added to the existing

diesel generator fuel oil storage

inventory.

Date of issuance: November 17, 1995

Effective date: November 17, 1995

Amendment Nos.: Unit 1 -

Amendment No. 43; Unit 2 -

Amendment No. 29

Facility Operating License Nos. NPF–

87 and NPF–89. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: February 1, 1995 (60 FR 6311)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated November 17, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: University of Texas at

Arlington Library, Government

Publications/Maps, 702 College, P.O.

Box 19497, Arlington, Texas 76019.

Union Electric Company, Docket No.

50–483, Callaway Plant, Callaway

County, Missouri

Date of amendment request: January

13, 1995

Brief description of amendment: The

amendment revises Technical

Specifications (TS) 3.3.1 and 3.3.2 to

relocate Tables 3.3–2 and 3.3–5, which

provide the response time limits for the

reactor trip system and the engineered

safety features actuation system

instruments, from the TS to the updated

Final Safety Analysis Report (FSAR).

The amendment also relocates the Bases

discussion for TS 3.3.1 and TS 3.3.2 to

Section 16.3 of the updated FSAR.

Date of issuance: November 22, 1995

Effective date: November 22, 1995, to

be implemented within 30 days of

issuance.

Amendment No.: 104

Facility Operating License No. NPF–

30. The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: February 15, 1995 (60 FR

8741) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

November 22, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

locations: Callaway County Public

Library, 710 Court Street, Fulton,

Missouri 65251

Washington Public Power Supply

System, Docket No. 50–397, Nuclear

Project No. 2, Benton County,

Washington

Date of application for amendment:

June 6, 1995

Brief description of amendment: The

amendment modifies the Index of the

WNP–2 Technical Specifications by

deleting reference to the Bases pages.

Date of issuance: November 24, 1995

Effective date: November 24, 1995

Amendment No.: 143

Facility Operating License No. NPF–

21: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: July 19, 1995 (60 FR 37102)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated November 24, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Richland Public Library, 955

Northgate Street, Richland, Washington

99352

Wisconsin Electric Power Company,

Docket Nos. 50–266 and 50–301, Point

Beach Nuclear Plant, Unit Nos. 1 and

2, Town of Two Creeks, Manitowoc

County, Wisconsin

Date of application for amendments:

September 13, 1995, and October 19,

1995, as supplemented by letter dated

October 25, 1995

Brief description of amendments:

These amendments revise Technical

Specification (TS) Section 15.1,

‘‘Definitions,’’ TS Section 15.3.1.G,

‘‘Operational Limitations’’ (and basis),

and TS Figure 15.2.1–2, ‘‘Reactor Core

Safety Limits, Point Beach Unit 2.’’ The

changes reduce the reactor coolant

system raw measured total flow rate

limit and reflect new reactor core safety

limits for Unit 2.

Date of issuance: November 17, 1995

Effective date: November 17, 1995

Amendment Nos.: 165 and 169

Facility Operating License Nos. DPR–

24 and DPR–27: Amendments revised

the Technical Specifications. Public

comments requested as to proposed no

significant hazards consideration: Yes

(60 FR 54527 dated October 24, 1995).

That notice provided an opportunity to

submit comments on the Commission’s

proposed no significant hazards

consideration determination. No

comments have been received. The

notice also provided for an opportunity

to request a hearing by November 24,

1995, but indicated that if the

Commission makes a final no significant

hazards consideration determination

any such hearing would take place after

issuance of the amendment. The

Commission’s related evaluation of the

amendment, finding of exigent

circumstances, and final determination

of no significant hazards consideration

is contained in a Safety Evaluation

dated November 17, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin

54241

Wolf Creek Nuclear Operating

Corporation, Docket No. 50–482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request:

September 14, 1995

Brief description of amendment: The

amendment revised Technical

Specification 3/4.5.5 to increase the

allowed outage time for adjustment of

boron concentration for the refueling

water storage tank from 1 hour to 8

hours.

Date of issuance: November 13, 1995
62502 Federal Register / Vol. 60, No. 234 / Wednesday, December 6, 1995 / Notices

Effective date: November 13, 1995, to

be implemented within 30 days of

issuance.

Amendment No.: 91

Facility Operating License No. NPF–

42. The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: October 11, 1995 (60 FR

52936) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

November 13, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

locations: Emporia State University,

William Allen White Library, 1200

Commercial Street, Emporia, Kansas

66801 and Washburn University School

of Law Library, Topeka, Kansas 66621

Notice Of Issuance Of Amendments To

Facility Operating Licenses And Final

Determination Of No Significant

Hazards Consideration And

Opportunity For A Hearing (exigent

public announcement or emergency

circumstances)

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application for the

amendment complies with the

standards and requirements of the

Atomic Energy Act of 1954, as amended

(the Act), and the Commission’s rules

and regulations. The Commission has

made appropriate findings as required

by the Act and the Commission’s rules

and regulations in 10 CFR Chapter I,

which are set forth in the license

amendment.

Because of exigent or emergency

circumstances associated with the date

the amendment was needed, there was

not time for the Commission to publish,

for public comment before issuance, its

usual 30-day Notice of Consideration of

Issuance of Amendment, Proposed No

Significant Hazards Consideration

Determination, and Opportunity for a

Hearing.

For exigent circumstances, the

Commission has either issued a Federal

Register notice providing opportunity

for public comment or has used local

media to provide notice to the public in

the area surrounding a licensee’s facility

of the licensee’s application and of the

Commission’s proposed determination

of no significant hazards consideration.

The Commission has provided a

reasonable opportunity for the public to

comment, using its best efforts to make

available to the public means of

communication for the public to

respond quickly, and in the case of

telephone comments, the comments

have been recorded or transcribed as

appropriate and the licensee has been

informed of the public comments.

In circumstances where failure to act

in a timely way would have resulted, for

example, in derating or shutdown of a

nuclear power plant or in prevention of

either resumption of operation or of

increase in power output up to the

plant’s licensed power level, the

Commission may not have had an

opportunity to provide for public

comment on its no significant hazards

consideration determination. In such

case, the license amendment has been

issued without opportunity for

comment. If there has been some time

for public comment but less than 30

days, the Commission may provide an

opportunity for public comment. If

comments have been requested, it is so

stated. In either event, the State has

been consulted by telephone whenever

possible.

Under its regulations, the Commission

may issue and make an amendment

immediately effective, notwithstanding

the pendency before it of a request for

a hearing from any person, in advance

of the holding and completion of any

required hearing, where it has

determined that no significant hazards

consideration is involved.

The Commission has applied the

standards of 10 CFR 50.92 and has made

a final determination that the

amendment involves no significant

hazards consideration. The basis for this

determination is contained in the

documents related to this action.

Accordingly, the amendments have

been issued and made effective as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the application for

amendment, (2) the amendment to

Facility Operating License, and (3) the

Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment, as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, the Gelman Building, 2120 L

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

local public document room for the

particular facility involved.

The Commission is also offering an

opportunity for a hearing with respect to

the issuance of the amendment. By

January 5, 1996, the licensee may file a

request for a hearing with respect to

issuance of the amendment to the

subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.
Federal Register / Vol. 60, No. 234 / Wednesday, December 6, 1995 / Notices 62503

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses. Since the Commission has

made a final determination that the

amendment involves no significant

hazards consideration, if a hearing is

requested, it will not stay the

effectiveness of the amendment. Any

hearing held would take place while the

amendment is in effect.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington, DC, by

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

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

1-(800) 342–6700). The Western Union

operator should be given Datagram

Identification Number N1023 and the

following message addressed to (Project

Director): petitioner’s name and

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555, and to the attorney for the

licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of the

factors specified in 10 CFR

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

Wisconsin Electric Power Company,

Docket Nos. 50–266 and 50–301, Point

Beach Nuclear Plant, Unit Nos. 1 and

2, Town of Two Creeks, Manitowoc

County, Wisconsin

Date of application for amendments:

November 9, 1995, as supplemented by

letters dated November 13, 1995, and

November 16, 1995

Brief description of amendments:

These amendments revise Technical

Specification (TS) Section 15.4.2, ‘‘InService Inspection of Safety Class

Components,’’ to incorporate a new

steam generator tube acceptance

criterion for the Unit 2 steam generators.

This criterion allows tubes that are

degraded or defective in a location

(within the tubesheet) that does not

affect the structural integrity of the tube

to remain in service. The applicable

basis is also changed.

Date of issuance: November 22, 1995

Effective date: November 22, 1995

Amendment Nos.: 166 and 170

Facility Operating License Nos. DPR–

24 and DPR–27. Amendments revised

the Technical Specifications. Public

comments requested as to proposed no

significant hazards consideration: No

The Commission’s related evaluation of

the amendments, finding of emergency

circumstances, and final determination

of no significant hazards consideration

are contained in a Safety Evaluation

dated November 22, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin

54241.

Attorney for licensee: Ernest L. Blake,

Jr., Shaw, Pittman, Potts & Trowbridge,

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

20037

NRC Project Director: Gail H. Marcus

Dated at Rockville, Maryland, this

29th day of November 1995.

For the Nuclear Regulatory Commission

Elinor G. Adensam,

Deputy Director, Division of Reactor Projects

- III/IV, Office of Nuclear Reactor Regulation

[Doc. 95–29540 Filed 12–5–95; 8:45 am]

BILLING CODE 7590–01–F

SECURITIES AND EXCHANGE

COMMISSION

[Rel. No. IC–21547; No. 812–9652]

Southland Life Insurance Company, et

al.

November 29, 1995.

AGENCY: Securities and Exchange

Commission (‘‘Commission’’).

ACTION: Notice of application for an

order pursuant to the Investment

Company Act of 1940 (the ‘‘1940 Act’’).

APPLICANTS: Southland Life Insurance

Company (‘‘Southland’’), Southland

Separate Account A1 (the ‘‘Account’’),

and ING America Equities, Inc. (‘‘ING

Equities’’).

RELEVANT 1940 ACT SECTIONS: Order

requested pursuant to Section 6(c) of the

1940 Act granting exemptions from the

provisions of Sections 26(a)(2)(C) and

27(c)(2) thereof.

SUMMARY OF APPLICATION: Applicants

seek an order permitting the deduction

of mortality and expense risk and

enhanced death benefit charges from the

assets of: (a) The Account in connection

with the offer and sale of certain

variable annuity contracts (‘‘Existing

Contracts’’); (b) the Account in

connection with the issuance of variable

annuity contracts that are substantially

similar in all material respects to the

Existing Contracts (‘‘Future Contracts,’’

together with Existing Contracts, the

‘‘Contracts’’); and (c) any other separate

account established in the future by

Southland in connection with the

issuance of Contracts (‘‘Future

Account’’).

FILING DATE: The application was filed

on June 29, 1995. Applicants have

undertaken to amend the application

during the notice period to make the

representations contained herein.

HEARING OR NOTIFICATION OF HEARING: An

order granting the application will be

issued unless the Commission orders a

hearing. Interested persons may request

a hearing by writing to the Secretary of

the Commission and serving Applicants

with a copy of the request, personally or

by mail. Hearing requests must be

received by the Commission by 5:30

p.m. on December 26, 1995, and must be