52926 Federal Register / Vol. 60, No. 196 / Wednesday, October 11, 1995 / Notices

RECORD GROUPS CLOSING AND REOPENING—Continued

Cluster title Rg. No. Record group short title Close date Reopen

date

Old Navy ............................................................... 072 Bureau of Aeronautics ......................................... 02/06/96 04/12/96

Old Navy ............................................................... 074 Bureau of Ordnance ............................................ 02/16/96 04/16/96

Old Navy ............................................................... 125 Judge Advocate General (Navy) ......................... 02/28/96 04/30/96

Old Navy ............................................................... 127 U.S. Marine Corps ............................................... 03/21/96 04/30/96

Old Navy ............................................................... 143 Bureau of Supplies and Accounts ....................... 03/13/96 05/10/96

Old Navy ............................................................... 181 Naval Districts and Shore Establishments .......... 03/25/96 05/10/96

Old Navy ............................................................... 313 Naval Operating Forces ...................................... 03/13/96 05/31/96

State/Foreign Relations ......................................... 469 U.S. Foreign Assistance Agencies, 1948–61 ...... 10/24/95 01/08/96

State/Foreign Relations ......................................... 490 Peace Corps ........................................................ 12/04/95 01/10/96

World War I Period Agencies ............................... 002 National War Labor Board (World War One) ...... 04/15/96 08/27/96

World War I Period Agencies ............................... 004 U.S. Food Corporation ........................................ 04/17/96 09/04/96

World War I Period Agencies ............................... 005 U.S. Grain Corporation ........................................ 04/24/96 09/06/96

World War I Period Agencies ............................... 006 U.S. Sugar Equalization Board, Inc. ................... 04/24/96 09/06/96

World War I Period Agencies ............................... 014 U.S. Railroad Administration ............................... 04/29/96 09/11/96

World War I Period Agencies ............................... 061 War Industries Board ........................................... 04/30/96 09/17/96

World War I Period Agencies ............................... 062 Council on National Defense ............................... 05/06/96 09/19/96

World War I Period Agencies ............................... 067 U.S. Fuel Administration ...................................... 05/08/96 09/25/96

World War I Period Agencies ............................... 158 Capital Issues Committee ................................... 05/13/96 09/27/96

World War I Period Agencies ............................... 182 War Trade Board ................................................. 05/17/96 10/03/96

World War I Period Agencies ............................... 194 War Minerals Relief Commission ........................ 05/24/96 10/07/96

[FR Doc. 95–24962 Filed 10–10–95; 8:45 am]

BILLING CODE 7515–01–P
Federal Register / Vol. 60, No. 196 / Wednesday, October 11, 1995 / Notices 52927

NUCLEAR REGULATORY

COMMISSION

Biweekly Notice

Applications and Amendments to

Facility Operating LicensesInvolving

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 September

16, through September 28, 1995. The

last biweekly notice was published on

Septmeber 27, 1995 (60 FR 49929).

Notice Of Consideration Of Issuance Of

Amendments To Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

And Opportunity For A Hearing

The Commission has made a

proposed determination that the

following amendment requests involve

no significant hazards consideration.

Under the Commission’s regulations in

10 CFR 50.92, this means that operation

of the facility in accordance with the

proposed amendment would not (1)

involve a significant increase in the

probability or consequences of an

accident previously evaluated; or (2)

create the possibility of a new or

different kind of accident from any

accident previously evaluated; or (3)

involve a significant reduction in a

margin of safety. The basis for this

proposed determination for each

amendment request is shown below.

The Commission is seeking public

comments on this proposed

determination. Any comments received

within 30 days after the date of

publication of this notice will be

considered in making any final

determination.

Normally, the Commission will not

issue the amendment until the

expiration of the 30-day notice period.

However, should circumstances change

during the notice period such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received before

action is taken. Should the Commission

take this action, it will publish in the

Federal Register a notice of issuance

and provide for opportunity for a

hearing after issuance. The Commission

expects that the need to take this action

will occur very infrequently.

Written comments may be submitted

by mail to the 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 November 10, 1995, 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
52928 Federal Register / Vol. 60, No. 196 / Wednesday, October 11, 1995 / Notices

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).

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.

Connecticut Yankee Atomic Power

Company, Docket No. 50-213, Haddam

Neck Plant, Middlesex County,

Connecticut

Date of amendment request: August

10, 1995

Description of amendment request:

The proposed amendment will add a

footnote to Technical Specification (TS)

Section 3/4.4.3, ‘‘Pressurizer,’’ 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.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

...The proposed change does not involve an

SHC because the change would not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The design basis accidents analyzed in

Mode 3 are steam line break, control rod

withdrawal from subcritical, boron dilution

and control rod ejection. Of these four

analyzed accidents, the relaxing of the

pressurizer level requirement can only

impact the steam line break accident

analyses. The initial pressurizer level can

impact the timing of the safety injection

signal and the subsequent boron addition

from the HPSI [high pressure safety injection]

system. The proposed change requires that

the boron concentration be equal to the Mode

5 required concentration in order for the

pressurizer level to be higher than the current

requirement. The Mode 5 boron

concentration ensures that there is sufficient

negative reactivity in the core due to boron

that a steam line break from this condition

would not need the boron addition from the

HPSI system and would be bounded by the

design basis analyses. Thus the proposed

change cannot increase the probability or

consequences of the design basis accidents.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

The proposed change only modifies the

Mode 3 pressurizer level requirement. This

change does not impact the lower bound but

provides flexibility to the plant operators in

the maximum pressurizer level. The upper

limit still provides margin to pressurizer

overfill. This cannot cause an accident nor

introduce a new type of malfunction. The

modified level would allow for a higher

initial pressurizer level in Mode 3. This

higher level is already used in the accident

analyses which result in an increase in

pressurizer level. Therefore, the change does

not modify the plant’s response to accidents.

3. Involve a significant reduction in the

margin of safety.

The proposed change is consistent with or

bounded by the design basis analyses. The

higher shutdown margin required in order to

relax the upper bound of the pressurizer level

assures that a steam line break from these

conditions is bounded by the design basis

analyses. Therefore, the proposed change

cannot 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

Duke Power Company, Docket Nos. 50-

269, 50-270 and 50-287, Oconee

Nuclear Station, Units 1, 2 and 3,

Oconee County, South Carolina

Date of amendment request:

September 1, 1995

Description of amendment request:

Generic Letter 88-16 provided guidance

on removing cycle-specific parameters

which are calculated using NRCapproved methodologies from the

Technical Specifications (TS). The

parameters are replaced in the TS with

a reference to a named report which

contains the parameters, and a

requirement that the parameters remain

within the limits specified in the report.

The proposed changes incorporate NRCapproved methodologies, approved

revisions to previously approved

methodologies, or republished versions

of previously approved methodologies

into Section 6.9.2 of the Oconee TS. The

limits to which these methodologies are

applied are 1) Axial Power Imbalance

Protective Limits and Variable Low RCS

Pressure Protective Limits, 2) Reactor

Protective System Trip Setting Limits

for the Flux/Flow/Imbalance and

Variable Low Reactor Coolant System

Pressure Trip functions, and 3) Power

Imbalance Limits. Since the proposed

changes only incorporate NRC-approved

methodologies into the TS, the licensee

proposed that the changes are

administrative in nature and can be
Federal Register / Vol. 60, No. 196 / Wednesday, October 11, 1995 / Notices 52929

assumed to have no impact, or potential

impact, on the health and safety of the

public.

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 will not create a

significant hazards consideration, as defined

by 10 CRF 50.92, because:

1) The proposed changes will not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes are administrative

in nature, and do not affect any system,

procedure, or manipulation of any equipment

which could affect the probability or

consequences of any accident.

2) The proposed changes will not create

the possibility of any new or different kind

of accident from any accident previously

evaluated.

The proposed changes are administrative

in nature, and cannot introduce any new

failure mode or transient which could create

any accident.

3) The proposed changes will not involve

a significant reduction in a margin of safety.

The proposed changes are administrative

in nature, and will not affect any operating

parameters or limits which could result in a

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Oconee County Library, 501

West South Broad Street, Walhalla,

South Carolina 29691

Attorney for licensee: J. Michael

McGarry, III, Winston and Strawn, 1200

17th Street, NW., Washington, DC 20036

NRC Project Director: Herbert N.

Berkow

Entergy Operations, Inc., et al., Docket

No. 50-416, Grand Gulf Nuclear

Station, Unit 1, Claiborne County,

Mississippi

Date of amendment request:

November 9, 1994, as supplemented by

letter dated August 4, 1995

Description of amendment request:

This supplement revises the licensee’s

November 9, 1994, application by

updating the request to reflect

implementation of the Improved

Standard Technical Specifications on

March 20, 1995, and by deleting the

request for a definition of the term

RECENTLY IRRADIATED FUEL. The

proposed amendment revises those

specifications associated with various

engineered safety feature systems

following a design basis fuel handling

accident. The proposed changes affect

conditions where irradiated fuel is

handled in the primary or secondary

containment and when fuel is handled

over the reactor vessel with fuel in the

vessel. These changes are based on a

recent re-analysis of the fuel handling

accident for Grand Gulf Nuclear Station

(GGNS).

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed changes do not

significantly increase the probability or

consequences of an accident previously

evaluated.

A new term to describe irradiated fuel is

used to establish operational conditions

where specific activities represent situations

where significant radioactive releases can be

postulated. These operational conditions are

consistent with the design basis analysis.

Because the equipment affected by the

revised operational conditions is not

considered an initiator to any previously

analyzed accident, inoperability of the

equipment cannot increase the probability of

any previously evaluated accident. The

proposed requirements in conjunction with

existing administrative controls on light

loads, bounds the conditions of the current

design basis fuel handling accident analysis

which concludes that the radiological

consequences are within the acceptance

criteria of NUREG 0800, Section 15.7.4 and

General Design Criteria 19. Therefore, the

proposed changes do not significantly

increase the probability or 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 changes would not create

the possibility of a new or different kind of

accident from any previous analyzed.

The new term to describe irradiated fuel is

used to establish operational conditions

where specific activities represent situations

where significant radioactive releases can be

postulated. These operational conditions are

consistent with the design basis analysis. The

proposed changes do not introduce any new

modes of plant operation and do not involve

physical modification of the plant. Therefore,

the proposed changes do not create the

possibility of a new or different kind of

accident from any previous analyzed.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously analyzed.

3. The proposed changes do not involve a

significant reduction in a margin of safety.

The new term to describe irradiated fuel is

used to establish operational conditions

where specific activities represent situations

where significant radioactive releases can be

postulated. These operational conditions are

consistent with the design basis analysis and

are established such that the radiological

consequences are at or below the current

GGNS licensing limit. Safety margins and

analytical conservatisms have been evaluated

and are well understood. Substantial margins

are retained to ensure that the analysis

adequately bounds all postulated event

scenarios. The proposed change only

eliminates the excess margin from the

analysis. 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 NUREG 0800,

Section 15.7.4. With respect to the control

room personnel doses, the margin of safety is

the difference between the 10 CFR 100 limits

and the licensing limit defined by 10 CFR 50,

Appendix A, Criterion 19 (GDC 19). Excess

margin is the difference between the

postulated doses and the corresponding

licensing limit.

The proposed applicability continues to

ensure that the

whole-body and thyroid dose 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.

Therefore, the proposed changes do not

result in a significant reduction in a margin

of safety.

Based on the above evaluation, operation

in accordance with the proposed amendment

involves no significant hazards

considerations.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Judge George W. Armstrong

Library, 220 S. Commerce Street,

Natchez, MS 39120

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, N.W., 12th Floor,

Washington, DC 20005-3502

NRC Project Director: William D.

Beckner

Entergy Operations, Inc., Docket No.

50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of amendment request: July 19,

1995

Description of amendment request:

The proposed amendment reduces

requirements associated with the

exercise frequency of control element

assemblies from once per 31 days to

once per 92 days.

Basis for proposed no significant

hazards consideration determination:
52930 Federal Register / Vol. 60, No. 196 / Wednesday, October 11, 1995 / Notices

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does not Involve a Significant Increase

in the Probability or Consequences of an

Accident Previously Evaluated.

Changing the frequency of the control

element assemblies (CEA) exercise test

surveillance introduces no new failure

mechanism for the system, so the

consequences of a postulated stuck CEA are

no different than those previously evaluated.

As explained in NUREG-1366,

‘‘Improvements to Technical Specifications

Surveillance Requirements,’’ the purpose of

this test is to identify immovable CEAs.

NUREG-1366 goes on to explain that the

majority of CEA problems are identified

during the performance of startup physics

testing and during CEA withdrawal for

startup, not during the exercise test. The

incidence of electrical malfunctions which

will still allow CEA insertion is much greater

than the incidence of mechanically bound

CEAs. As stated in NUREG-1366, there has

only been one incidence of multiple CEAs

failing to fully insert upon a reactor trip

(Point Beach Nuclear Plant, May 1985) and

in this case the two affected CEAs partially

inserted. Based on this history, simply

reducing the test frequency will not increase

the probability of a stuck CEA.

Therefore, this change does not involve a

significant increase in the probability or

consequences of any accident previously

evaluated.

2. Does Not Create the Possibility of a New

or Different Kind of Accident from any

Previously Evaluated.

Because the proposed change does not alter

the design, configuration, or method of

operation of the plant, it does not create the

possibility of a new or different kind of

accident from any previously evaluated.

3. Does Not Involve a Significant

Reduction in the Margin of Safety.

The proposed change does not alter the

acceptance criteria of any surveillance

requirements, alter any assumptions used in

accident analysis, change any actuation

setpoints, nor allow operations in any

configuration not previously evaluated. This

change in surveillance frequency is based on

a satisfactory operating history of CEAs.

Additionally, the number of problems

created by this test when compared with the

number of problems identified by this test

indicate that reducing the test frequency will

have no adverse impact on the continued safe

operation of the unit.

Therefore, this change does not involve a

significant reduction in the margin of safety.

Therefore, based upon the reasoning

presented above and the previous discussion

of the amendment request, Entergy

Operations had determined that the

requested change does not involve a

significant hazards consideration.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

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

20005-3502

NRC Project Director: William D.

Beckner

Florida Power and Light Company,

Docket Nos. 50-250 and 50-251, Turkey

Point Plant Units 3 and 4, Dade County,

Florida

Date of amendment request:

September 11, 1995

Description of amendment request:

The licensee proposes to change Turkey

Point Units 3 and 4 Technical

Specifications (TS) to incorporate lineitem improvements to Specifications 3/

4.8.1, ‘‘Electrical Power Systems-A.C.

Sources,’’ and the associated BASES.

The licensee stated that the proposed

changes are consistent with the

guidance provided by the NRC in GL 93-

05, ‘‘Line-Item Technical Specifications

Improvements to Reduce Surveillance

Requirements for Testing During Power

Operation,’’ and the corresponding

recommendations contained in NUREG1366, ‘‘Improvements to Technical

Specifications Surveillance

Requirements.’’

In addition, line-item improvements

are proposed following the guidance in

GL 94-01, ‘‘Removal of Accelerated

Testing and Special Reporting

Requirements for Emergency Diesel

Generators.’’ The implementation of a

maintenance program for monitoring

and maintaining Emergency Diesel

Generator (EDG) performance for Turkey

Point Units 3 and 4, consistent with the

provisions of 10 CFR 50.65

‘‘Requirements for Monitoring the

Effectiveness of Maintenance at Nuclear

Power Plants’’ and the associated

guidance of Regulatory Guide (RG)

1.160 will be met by FPL within 90 days

following issuance of the proposed

amendments.

The licensee also requested to revise

the current wording used in the Turkey

Point Units 3 and 4 TS to require testing

of remaining required diesel generators

‘‘[i]f the diesel generator became

inoperable due to any cause other than

planned preventative maintenance...’’.

The licensee requested that TS 3.8.1.1,

ACTION statements b. and c. be

amended such that the word

’preventative’ is deleted. Deleting this

wording will reduce unnecessary testing

of diesel generators as a result of

planned corrective maintenance.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) Operation of the facility in accordance

with the proposed amendments would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The license amendments proposed for

Turkey Point Units 3 and 4 will incorporate

line-item Technical Specification (TS)

improvements for Emergency Diesel

Generators (EDG) pursuant to guidance

provided in Generic Letters (GL) 93-05 and

94-01. The EDGs are not accident initiators,

the proposed TS changes do not involve any

assumptions relative to accident initiators in

the plant safety analyses, and therefore the

proposed amendments will not impact the

probability of occurrence for accidents

previously analyzed.

The EDG line-item TS improvements

associated with GL 93-05 are based on

recommendations designed to remove

unwarranted requirements for testing during

power operation and other factors that are

counter-productive to safety in terms of

equipment degradation and availability.

These recommendations resulted from a

comprehensive study of industry-wide EDG

surveillance requirements and subsequent

findings reported by the NRC in NUREG1366. The proposed amendments are

consistent with the guidance of GL 93-05 for

implementing such recommendations as well

as contemporary licensing actions by the

NRC on other light water reactors.

Similarly, GL 94-01 provides guidance for

a line-item TS improvement that will remove

accelerated testing requirements from the TS

provided that the licensee commits to a

maintenance program for monitoring and

maintaining EDG performance that includes

the applicable provisions of the maintenance

rule (10 CFR 50.65). Such a program will

further assure EDG availability. Since the

availability of EDGs is assumed in certain

success paths for mitigating analyzed

accidents, an improvement in EDG

availability will enhance accident mitigation

capabilities.

Therefore, operation of the facility in

accordance with the proposed amendments

would not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

(2) Operation of the facility in accordance

with the proposed amendments would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed amendments incorporate

line-item TS and other improvements to EDG

surveillance testing requirements, and will

not change the physical plant or the modes

of plant operation defined in the Facility

License. The changes do not involve the

addition or modification of equipment, nor

do they alter the design or methods of

operation of plant systems. Plant

configurations that are prohibited by TS will
Federal Register / Vol. 60, No. 196 / Wednesday, October 11, 1995 / Notices 52931

not be created by the amendments. Therefore,

operation of the facility in accordance with

the proposed amendment would not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

(3) Operation of the facility in accordance

with the proposed amendments would not

involve a significant reduction in a margin of

safety.

The proposed amendments are designed to

improve EDG availability by eliminating

unwarranted surveillance testing. The

currently specified surveillance intervals are

not changed, except to delete the requirement

for accelerated testing under certain

circumstances. The proposed changes do not

otherwise alter the basis for any Technical

Specification that is related to the

establishment of, or the maintenance of a

nuclear safety margin. Therefore, operation of

the facility in accordance with the proposed

amendment would not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Florida International

University, University Park, Miami,

Florida 33199

Attorney for licensee: J. R. Newman,

Esquire, Morgan, Lewis & Bockius, 1800

M Street, NW., Washington, DC 20036

NRC Project Director: David B.

Matthews

Maine Yankee Atomic Power Company,

Docket No. 50-309, Maine Yankee

Atomic Power Station, Lincoln County,

Maine

Date of amendment request: July 24,

1995

Description of amendment request:

The proposed amendment would

modify Technical Specification 3.6.C to

allow up to 7 days to restore low

pressure safety injection (LPSI) pump

subsystem operability, and up to 24

hours to restore safety injection tank

(SIT) operability.

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

staff’s review is presented below:

1. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated. The LPSI system is

designed primarily to mitigate the

consequences of a large loss-of-coolant

accident (LOCA). Inoperable LPSI

components are not accident initiators in any

accident previously evaluated, and the

proposed change does not affect any of the

assumptions relative to accident initiators in

the plant’s safety analysis. Probabilistic

safety analysis (PSA) methods were used to

fully evaluate the extension of the LPSI

system allowed outage time (AOT). The

licensee asserts that the results of these

analyses show no significant increase in the

consequences of an accident previously

evaluated. The SITs were designed to

mitigate the consequences of a LOCA. The

proposed amendment does not affect any of

the assumptions used in the deterministic

LOCA analysis. Probabilistic safety analysis

methods were used to fully evaluate the

effect of the SIT allowable outage time

(AOT). The licensee asserts that the results of

these analyses show no significant increase

in the consequences of an accident

previously evaluated. Thus, there is no

significant increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated. The proposed

amendment does not change the design,

physical configuration, or modes of operation

of the plant. Plant configurations that are

prohibited by TS will not be created by this

proposed amendment. Thus, the proposed

amendment does not create the possibility or

consequences of an accident previously

evaluated.

3. The proposed amendment does not

involve a significant reduction in a margin of

safety. The proposed amendment does not

affect the limiting conditions for operation or

the bases used in the deterministic analyses

to establish the margin of safety. The licensee

asserts that PSA methods were used to

evaluate these changes and demonstrate that

the changes are either risk neutral or risk

beneficial. Thus, the proposed amendment

does not involve a significant reduction in a

margin of safety.

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 this

amendment request involves no

significant hazards determination.

Local Public Document Room

location: Wiscasset Public Library, High

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

04578

Attorney for licensee: Mary Ann

Lynch, Esquire, Maine Yankee Atomic

Power Company, 329 Bath Road,

Brunswick, ME 04011

NRC Project Director: Phillip F.

McKee

Maine Yankee Atomic Power Company,

Docket No. 50-309, Maine Yankee

Atomic Power Station, Lincoln County,

Maine

Date of amendment request: August 8,

1995

Description of amendment request:

The proposed amendment would

modify the definition of Transthermal

(Condition 4), Hot Shutdown (Condition

5), and Hot Standby (Condition 6)

reactor operating conditions. The

Transthermal and Hot Shutdown

conditions are modified to establish an

applicable range of subcriticality and be

consistent with other Definitions. The

wording of Hot Standby is modified to

remove reference to control rod

position, consistent with NUREG-1432,

Standard Technical Specifications for

Combustion Engineering Plants,

Revision 1 dated April 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. The NRC staff has

reviewed the licensee’s analysis against

the standards of 10 CFR 50.92(c). The

staff’s review is presented below:

1. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated. The changes to these

Definitions are administrative in nature. The

Transthermal and Hot Shutdown conditions

are changed by adding ‘‘at least’’ to establish

a range of subcriticality. The current

Definitions for the Transthermal and Hot

Shutdown conditions set one minimum

value for subcriticality; the change to these

two Definitions would allow a range of

values for subcriticality. All values of

subcriticality that may be established by this

change are below the current Definitions

(more subcritical). The change to the wording

of Hot Standby removes confusion about the

Conditions during which control rods may be

withdrawn and is consistent with current

NRC guidance. All current plant analyses,

requirements and acceptance criteria on

subcriticality conditions remain in effect.

The changes to these Definitions have no

impact on event probabililty. Thus, the

proposed amendment does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated. The proposed

amendment clarifies the subject Definitions.

Limits on subcriticality requirements are

unaffected, as are reactivity transients

previously evaluated. Plant procedures

currently require that minimum values for

subcriticality be established. All values of

subcriticality that may be established by this

change are below the current Definitions

(more subcritical). Further, the change to the

wording of Hot Standbyis consistent with

current NRC guidance. Thus, the proposed

amendment does not create the possibility of

a new or different kind of accident from any

accident previously evaluated.

3. The proposed amendment does not

involve a significant reduction in a margin of

safety. Adding the words ‘‘at least’’ to the

Transthermal and Hot Shutdown conditions

establishes a range of subcriticality to the
52932 Federal Register / Vol. 60, No. 196 / Wednesday, October 11, 1995 / Notices

Definitions for these terms. All values of

subcriticality are below (more subcritical)

than the current value, thus the margin of

safety is increased. All current plant

analyses, requirements and acceptance

criteria on subcriticality conditions remain in

effect. The change to the wording of Hot

Standby removes confusion about the

Conditions during which control rods may be

withdrawn and is consistent with current

NRC guidance. Thus, there is no significant

reduction in a margin of safety.

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 this

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Wiscasset Public Library, High

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

04578

Attorney for licensee: Mary Ann

Lynch, Esquire, Maine Yankee Atomic

Power Company, 329 Bath Road,

Brunswick, ME 04011

NRC Project Director: Phillip F.

McKee

Maine Yankee Atomic Power Company,

Docket No. 50-309, Maine Yankee

Atomic Power Station, Lincoln County,

Maine

Date of amendment request: August

30, 1995

Description of amendment request:

The proposed amendment would

change Technical Specification (TS)

1.3.A, Reactor Core, to allow the use of

fuel rods clad with zirconium alloy,

rather than restrict fuel rod cladding to

Zircaloy-4. In addition, the fuel

enrichment limit described in this

specification would be changed to more

closely agree with the wording found in

NUREG-1432, ‘‘Standard Technical

Specifications for Combustion

Engineering Plants,’’ dated April 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. The NRC staff has

reviewed the licensee’s analysis against

the standards of 10 CFR 50.92(c). The

staff’s review is presented below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an acident previously

evaluated. Maine Yankee (MY) reload cores

containing fuel rods clad with zirconium

alloy and having higher fuel enrichments

will be analyzed using NRC-approved

methods and applicable acceptance criteria.

In addition, the impact of fuel assembly

design changes on fuel storage will be

analyzed using NRC-approved methods and

acceptance criteria. Compliance with the

acceptance criteria for the applicable analysis

for a given core design must be determined

for each core prior to reloading. The material

used to clad the fuel and the fuel enrichment

are only two of the factors considered in this

determinination. The application of

approved methods ensures that all

appropriate variables are addressed and their

acceptance criteria satisfied. 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 determination of compliance

with the acceptance criteria of the approved

safety evaluation for any given core reload

design is performed for each MY reload core

prior to loading. In addition, determination

of compliance with the acceptance criteria of

the approved safety evaluation for fuel

storage is performed for each core prior to

receipt of the fuel. The use of approved

methods and their acceptance criteria

ensures that new or different accidents will

not be encountered by the use of fuel rods

clad with zirconium alloy and having higher

fuel enrichments. Further, the proposed

change does not involve any altertions to

plant equipment that would affect any

operational modes or accident precursors.

Finally, the proposed change does not

involve, or require secondary involvement of,

any equipment important to safety. Thus 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.

Maine Yankee reload cores containing fuel

rods clad with zirconium alloy and having

higher fuel enrichments will be analyzed

using NRC-approved methods and applicable

acceptance criteria. Safety evaluations

performed for each core reload ensure that

the core design meets appropriate safety

assessment acceptance criteria. In addition,

the impact of fuel assembly design changes

on fuel storage also will be analyzed using

NRC-approved methods and aceptance

criteria. Application of the approved

methods ensures that the requirements of MY

TS 1.1, Fuel Storage, are achieved. Because

these requirements are not changed, the

margin of safety remains the same. Thus

there is no significant reduction in a margin

of safety.

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 this

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Wiscasset Public Library, High

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

04578

Attorney for licensee: Mary Ann

Lynch, Esquire, Maine Yankee Atomic

Power Company, 329 Bath Road,

Brunswick, ME 04011

NRC Project Director: Phillip F.

McKee

Maine Yankee Atomic Power Company,

Docket No. 50-309, Maine

YankeeAtomic Power Station, Lincoln

County, Maine

Date of amendment request: August

31, 1995

Description of amendment request:

The proposed amendment would

relocate fire protection requirements

from the Maine Yankee (MY) Atomic

Power Station Technical Specifications

(TS) to other, licensee-controlled

documents. The proposed amendment

is consistent with the guidance of U.S.

NRC Generic Letters 86-10,

Implementation of Fire Protection

Requirements, and 88-12, Removal of

Fire Protection Requirements from the

Technical Specifications.

Basis for proposed no significant

hazards consideration etermination: As

required by 10 CFR 50.91(a), the

licensee has provided its analysis if 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:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated. The proposed change is

administrative and consistent with the

guidance provided by the U.S. NRC.

Removing fire protection requirements from

the TS does not affect any fire protection

equipment, or involve any physical

modifications to plant structures, systems or

components. The proposed change is not

associated with accident initiation or

mitigation and cannot affect the probability

of occurrence of an accident, or increase the

consequences of an accident. The licensee’s

fire protection plan contains the relocated

requirements.

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 introduces

no new mode of plant operation, does not

involve physical modification of any

structure, system or component, and does not

affect the function, operation or surveillance

requirements of any equipment necessary for

safe operation or shutdown. Further, the

proposed change does not involve any

change to equipment setpoints or operating

parameters. The proposed change is

administrative in nature. Existing plant fire

protection equipment requirements are

retained. Thus, the proposed change does not

create the possibility for 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.

No margins of safety established by system or

component design, or verified by testing to

ensure operability of fire protection systems

or components, are affected. Fire protection

requirements currently found in the TS will

be relocated in their entirety to the Maine

Yankee Fire Protection Plan. Any future
Federal Register / Vol. 60, No. 196 / Wednesday, October 11, 1995 / Notices 52933

changes to the Plan will be evaluated in

accordance with the requirements of 10 CFR

50.59, Changes, tests and experiments. Thus

the proposed change does not involve a

significant reduction in a margin of safety.

Based on this review, it appears that

the three standards of 10 CFR 50.92(c)

are satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Wiscasset Public Library, High

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

04578

Attorney for licensee: Mary Ann

Lynch, Esquire, Maine Yankee Atomic

Power Company, 329 Bath Road,

Brunswick, ME 04011

NRC Project Director: Phillip F.

McKee

Northeast Nuclear Energy Company, et

al., Docket No. 50-336, Millstone

Nuclear Power Station, Unit Nos. 2,

New London, Connecticut

Date of amendment request:

September 11, 1995

Description of amendment request:

The proposed changes affect Technical

Specification Sections 3.4.8 and 3.9.9,

Tables 2.2-1, 3.3-3, 3.3-5 and 3.3-8, and

Bases Sections 3/4.2.1, 3/4.4.8 and 3/

4.11.2.1. These changes combine several

different administrative changes which

will correct typographical errors,

provide clarifications, or make editorial

changes.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration (SHC), which is presented

below:

Pursuant to 10CFR50.92, NNECO has

reviewed the proposed changes. NNECO

concludes that these changes do not involve

a significant hazards consideration since the

proposed change satisfies the criteria in

10CFR50.92(c). That is, the proposed changes

do not:

1. Involve a significant increase in the

probability or consequences of an accident

previously analyzed.

The proposed changes are administrative

in nature and do not result in changes to

plant configuration, operation, accident

mitigation, or analysis assumptions. Thus, it

cannot increase the probability or

consequence of an accident.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

The proposed changes are administrative

in nature and do not result in changes to

plant configuration, operation, accident

mitigation, or analysis assumptions. The

intent and application of the proposed

specification will not change. Therefore, the

proposal does not create the possibility of a

new or different kind of accident from any

previously analyzed.

3. Involve a significant reduction in the

margin of safety.

Since the proposed change[s] are

administrative in nature and do not result in

changes to plant configuration, operation,

accident mitigation, or analysis assumptions,

there is no reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 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

Northern States Power Company,

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

Island Nuclear Generating Plant, Unit

Nos. 1 and 2, Goodhue County,

Minnesota

Date of amendment requests: July 17,

1995

Description of amendment requests:

The proposed amendments would

revise the Prairie Island Radiological

Effluent Technical Specifications and

other sections relating to radiological

controls to conform to NUREG-1431,

Standard Technical Specifications,

Westinghouse Plants, Revision 1, and

Generic Letter 89-01, ‘‘Implementation

of Programmatic Controls for

Radiological Effluent Technical

Specifications in the Administrative

Controls Section of the Technical

Specifications and the Relocation of

Procedural Details of RETS to the Offsite

Dose Calculation Manual or to the

Process Control Program.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed amendment will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated. The proposed changes are

administrative in nature and alter only the

format and location of programmatic controls

and procedural details relative to radioactive

effluents, radiological environmental

monitoring, radioactive source leakage

testing, solid radioactive wastes, and

associated reporting requirements. Existing

Technical Specifications containing

procedural details on radioactive effluents,

radiological environmental monitoring,

radioactive source leakage testing, explosive

gas monitoring, storage tank radioactive

content limits, solid radioactive wastes and

associated reporting requirements are being

relocated to the Offsite Dose Calculation

Manual, Process Control Program or other

new programs as appropriate. Compliance

with applicable regulatory requirements will

continue to be maintained. In addition, the

proposed changes do not alter the conditions

or the assumptions in any of the previous

accident analyses. Since the previous

accident analyses remain bonding, the

radiological consequences previously

evaluated are not adversely affected by the

proposed changes.

Therefore, the probability or consequences

of an accident previously evaluated are not

affected by any of the proposed amendments.

2. The proposed amendment will not

create the possibility of a new or different

kind of accident from any accident

previously analyzed.

The proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated. The proposed changes do not

involve any change to the configuration or

method of operation of any plant equipment.

Accordingly, no new failure modes have

been defined for any plant system or

component important to safety nor has any

new limiting single failure been identified as

a result of the proposed changes. Also, there

will be no change in types or increase in the

amounts of any effluents released offsite.

Therefore, the possibility of a new or

different kind of accident from any accident

previously evaluated would not be created.

3. The proposed amendment will not

involve a significant reduction in the margin

of safety.

The proposed changes do not involve a

significant reduction in a margin of safety.

The proposed changes do not involve any

actual change in the methodology used in the

control of radioactive effluents, radioactive

sources, solid radioactive wastes, or

radiological environmental monitoring.

These changes are considered administrative

in nature and provide for the relocation of

procedural details outside of the technical

specifications but add appropriate

administrative controls to provide continued

assurance of compliance to applicable

regulatory requirements. These proposed

changes also comply with the guidance

contained in Generic Letter 89-01 and the

Standard Technical Specifications.

Therefore, it can be concluded a significant

reduction in the margin of safety would not

be involved.

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 requests involve no

significant hazards consideration.
52934 Federal Register / Vol. 60, No. 196 / Wednesday, October 11, 1995 / Notices

Local Public Document Room

location: Minneapolis Public Library,

Technology and Science Department,

300 Nicollet Mall, Minneapolis,

Minnesota 55401

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts, and Trowbridge,

2300 N Street, NW, Washington, DC

20037

NRC Project Director: John N. Hannon

Philadelphia Electric Company, Docket

No. 50-352, Limerick Generating

Station, Unit 1, Montgomery County,

Pennsylvania

Date of amendment request: June 19,

1995

Description of amendment request:

The proposed amendment would revise

Technical Specification Section 2.1,

‘‘Safety Limits,’’ to change the

Minimum Critical Power Ratio Safety

Limit due to the use of General Electric

13 fuel product line.

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 Technical Specifications

(TS) change does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

The derivation of the revised GE13

[General Electric] Minimum Critical Power

Ratio (MCPR) Safety Limit for incorporation

into the Technical Specifications, and its use

to determine cycle-specific thermal limits

have been performed using NRC-approved

methods within the existing design and

licensing basis, and cannot increase the

probability or severity of an accident.

The basis of the MCPR Safety Limit

calculation is to ensure that greater than

99.9% of all fuel rods in the core avoid

boiling transition if the limit is not violated.

The new MCPR Safety Limit preserves the

existing margin to transition boiling and fuel

damage in the event of a postulated accident.

All design bases of the MCPR Safety Limit

calculation apply to GE13 fuel in the same

manner that they have applied to previous

fuel designs. The probability of fuel damage

is not increased.

Therefore, the proposed TS change does

not involve an increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed TS change does not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The MCPR Safety Limit for the GE13 fuel

design is a Technical Specification numerical

value, designed to ensure that fuel damage

from transition boiling does not occur as a

result of the limiting postulated accident. It

cannot create the possibility of any new type

of accident. The new Minimum Critical

Power Ratio (MCPR) Safety Limit is

calculated using NRC-approved methods and

has the same calculational basis as the MCPR

Safety Limit for other GE fuel designs

currently used at LGS [Limerick Generating

Station] Unit 1.

Therefore, the proposed TS change does

not create the possibility of a new or different

kind of accident, from any accident

previously evaluated.

3. The proposed TS change does not

involve a significant reduction in a margin of

safety.

The following TS Bases were reviewed for

potential reduction in the margin of safety:

2.1 ‘‘Safety Limits’’

3/4.2.1 ‘‘Average Planar Linear Heat

Generation Rate’’

3/4.2.3 ‘‘Minimum Critical Power Ratio’’

3/4.2.4 ‘‘Linear Heat Generation Rate’’

3/4.4.1 ‘‘Recirculation System’’

3/4.9 ‘‘Refueling Operations’’

The margin of safety as defined in the TS

Bases will remain the same. The new

Minimum Critical Power Ratio (MCPR)

Safety Limit is calculated using NRC

approved methods which are in accordance

with the current fuel design and licensing

criteria. The MCPR Safety Limit for GE13 fuel

remains high enough to ensure that greater

than 99.9% of all fuel rods in the core will

avoid boiling transition if the limit is not

violated, thereby preserving the fuel cladding

integrity.

Therefore, the proposed TS change does

not involve a reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania

19464.

Attorney for licensee: J. W. Durham,

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

Counsel, Philadelphia Electric

Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101

NRC Project Director: John F. Stolz

Philadelphia Electric Company, Docket

Nos. 50-352 and 50-353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of amendment request:

September 14, 1995

Description of amendment request:

The amendments change the Technical

Specifications (TS) by removing the

Reactor Enclosure and Refueling Area

Secondary Containment Isolation Valve

Tables 3.6.5.2.1-1 and 3.6.5.2.2-1 from

TS in accordance with NRC Generic

Letter (GL) 91-08, ‘‘Removal of

Component Lists from Technical

Specifications.’’

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed TS changes do not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes will remove

component tables from TS. The component

lists will be retained in licensee controlled

documents (UFSAR [Updated Final Safety

Analysis Report] and a plant procedure)

which will be maintained under the

requirements of TS Administrative Controls

Section 6.0 and the provisions of 10 CFR

50.59. Since any changes to licensee

controlled documents are required to be

evaluated per 10 CFR 50.59, no increase

(significant or insignificant) in the

probability or consequences of an accident

previously evaluated will be allowed.

In addition, these proposed changes will

not affect any equipment important to safety,

in structure or operation. These changes will

not alter operation of process variables,

structures, systems, or components as

described in the safety analysis and licensing

basis. The changes will not increase the

probability or consequences of occurrence of

a malfunction of equipment important to

safety previously evaluated in the SAR

[Safety Analysis Report].

2. The proposed TS changes do not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes will not alter the

plant configuration or change the methods

governing normal plant operation. The

changes will not impose different operating

requirements and adequate control of

information will be retained. The changes

will not alter assumptions made in the safety

analysis and licensing basis. Since the

proposed changes cannot cause an accident,

and the plant response to the design basis

events is unchanged, the changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed TS changes do not involve

a significant reduction in a margin of safety.

The proposed changes to remove the

component tables from TS have been

performed under the guidance of NRC GL 91-

08. The component lists will be retained in

licensee controlled documents (UFSAR and a

plant procedure) which will be maintained

under the requirements of TS Administrative

Controls Section 6.0 and the provisions of 10

CFR 50.59. These changes will not reduce the

margin of safety since they have no impact

on any safety analysis assumptions. Since

any future changes to the removed tables will

be evaluated under the requirements of 10

CFR 50.59, no reduction (significant or

insignificant) in a margin of safety will be

allowed. Therefore, the proposed TS changes

do not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are
Federal Register / Vol. 60, No. 196 / Wednesday, October 11, 1995 / Notices 52935

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania

19464.

Attorney for licensee: J. W. Durham,

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

Counsel, Philadelphia Electric

Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101

NRC Project Director: John F. Stolz

Rochester Gas and Electric

Corporation, Docket No. 50-244, R. E.

Ginna Nuclear Power Plant, Wayne

County, New York

Date of amendment request: May 26,

1995

Brief description of amendment: The

proposed amendment would represent a

full conversion from the current

Technical Specifications (TSs) to a set of

TS based on NUREG-1431, ‘‘Standard

Technical Specifications, Westinghouse

Plants,’’ Revision 0, dated September

1993, together with approved travellers

used in the issuance of Revision 1,

dated April 1995. NUREG-1431 was

developed through working groups

composed of NRC staff members and

industry representatives and has been

endorsed by the staff as part of an

industry-wide initiative to standardize

and improve the TSs. As part of this

submittal, the licensee has applied the

criteria contained in the Commission’s

Final Policy Statement on Technical

Specification Improvements for Nuclear

Power Reactors of July 22, 1993, to the

current Ginna TSs, and using NUREG1431 as a basis, developed a proposed

set of improved TSs for Ginna.Date of

publication of individual notice in

Federal Register: September 26, 1995

(60 FR 49636)

Expiration date of individual notice:

October 26, 1995

Local Public Document Room

location: Rochester Public Library, 115

South Avenue, Rochester, New York

14610

Tennessee Valley Authority, Docket No.

50-296, Browns Ferry Nuclear Plant,

Unit 3, Limestone County, Alabama

Date of amendment request:

September 13, 1995 (TS 368)

Description of amendment request:

The proposed amendment deletes

requirements for daily checks for certain

instruments that do not have

indications, and provides editorial

changes.

Basis for proposed no significant

hazards consideration determination:As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

A. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed changes are administrative

in nature and correct errors that were

introduced by previous changes to the TSs.

These changes do not affect any of the design

basis accidents nor do they involve an

increase in the probability or consequences

of an accident previously evaluated.

B. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed changes are administrative

in nature. These changes do not change the

operation or function of the affected

instrumentation. The deletion of the RCIC

and HPCI instrument checks reflects the

actual installed configuration of this

instrumentation (no indication) and the

change to Table 4.2.C corrects the referenced

note for the SRM Upscale function.

Therefore, the possibility for an accident or

malfunction of a different type than any

evaluated previously is not created by this

change.

C. The proposed amendment does not

involve a significant reduction in a margin of

safety.

The proposed changes are administrative

in nature. The proposed changes to TS Tables

4.2.B and 4.2.C do not affect any acceptable

limit of operation, instrument setpoint, or

analysis assumption in the TS or Bases.

Therefore, this change does not reduce the

margin of safety as defined in the basis for

any TS.

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: Athens Public Library, South

Street, Athens, Alabama 35611

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET llH,

Knoxville, Tennessee 37902

NRC Project Director: Frederick J.

Hebdon

TU Electric Company, Docket Nos. 50-

445 and 50-446, Comanche Peak Steam

Electric Station, Units 1 and 2,

Somervell County, Texas

Date of amendment request: August

15, 1995

Brief description of amendments: The

proposed amendment would relocate

the Shutdown Margin limits from the

Technical Specifications (TSs) to the

Core Operating Limits Report. The

proposed changes are consistent with

the intent of Generic Letter (GL) 88-16

which provides guidelines for the

removal of cycle-specific parameter

limits from the TSs.

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. Do the proposed changes involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

The proposed changes remove cyclespecific parameter limits from the Technical

Specifications, add them to the list of limits

contained in the Core Operating Limits

Report (COLR), and revise the Administrative

Controls section of the Technical

Specifications. The changes do not, by

themselves, alter any of the parameter limits.

The changes are administrative in nature and

have no adverse effect on the probability of

an accident or on the consequences of an

accident previously evaluated. The removal

of parameter limits from the Technical

Specifications does not eliminate the

requirement to comply with the parameter

limits.

The parameter limits in the COLR may be

revised without prior NRC approval.

However, Specification 6.9.1.6c continues to

ensure that the parameter limits are

developed using NRC-approved

methodologies and that applicable limits of

the safety analyses are met. While future

changes to the COLR parameter limits could

result in event consequences which are either

slightly less or slightly more severe than the

consequences for the same event using the

present parameter limits, the differences

would not be significant and would be

bounded by the requirement of specification

6.9.1.6c to meet the applicable limits in the

safety analysis.

Based on the above, removal of the

parameter limits from the Technical

Specifications and the addition of these

limits the list of limits in the COLR, thus

allowing revision of the parameter limits

without prior NRC approval, has no

significant effect on the probability or

consequences of an accident previously

evaluated.

2. Do the proposed changes create the

possibility of a new or different kind of

accident from any accident previously

evaluated?

The proposed changes remove certain

parameter limits from the Technical

Specifications and add these limits to the list

of limits in the COLR, removing the

requirement for prior NRC approval of

revisions to those parameters. The changes

do not add new hardware or change plant

operations and therefore cannot initiate an

event nor cause an analyzed event to progress

differently. Thus, the possibility of a new or

different kind of accident is not created.

3. Do the proposed changes involve a

significant reduction in a margin of safety?

The margin of safety, as it relates to a

parameter limit, is the difference between the
52936 Federal Register / Vol. 60, No. 196 / Wednesday, October 11, 1995 / Notices

acceptance criterion for that parameter and

its failure value. The proposed changes do

not affect the failure values for any system.

Through the accident analyses, all relevant

event acceptance criteria (as described in the

NRC-approved analysis methodologies) are

shown to be satisfied; therefore, there is no

impact on an event acceptance criteria.

Because neither the failure values nor the

acceptance criteria are affected, the proposed

change has no effect on 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: University of Texas at

Arlington Library, Government

Publications/Maps, 702 College, P.O.

Box 19497, Arlington, TX 76019

Attorney for licensee: George L. Edgar,

Esq., Morgan, Lewis and Bockius, 1800

M Street, N.W., Washington, DC 20036

NRC Project Director: William D.

Beckner

Wisconsin Public Service Corporation,

Docket No. 50-305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of amendment request:

September 19, 1995

Description of amendment request:

The proposed amendment would make

administrative changes to the Kewaunee

Nuclear Power Plant (KNPP) Technical

Specifications (TS) to improve their

clarity and consistency. The proposed

amendment includes changes to reflect

revisions to 10 CFR Part 20, and changes

to correct minor typographical and

format inconsistencies as part of an

ongoing effort to convert the TS to the

WordPerfect format.

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 were reviewed in

accordance with the provisions of 10 CFR

50.92 to show no significant hazards exist.

The proposed changes will not:

1. involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The likelihood that an accident will occur

is neither increased or decreased by these TS

changes. These TS changes will not impact

the function or method of operation of plant

equipment. Thus, there is not a significant

increase in the probability of a previously

analyzed accident due to these changes. No

systems, equipment, or components are

affected by the proposed changes. Thus, the

consequences of the malfunction of

equipment important to safety previously

evaluated in the Updated Safety Analysis

Report (USAR) are not increased by these

changes.

The proposed changes are administrative

in nature and, therefore, have no impact on

accident initiators or plant equipment, and

thus, do not affect the probabilities or

consequences of an accident.

2. create the possibility of a new or

different kind of accident from any accident

previously evaluated.

Operation of the facility in accordance

with the proposed TS changes would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed changes do not involve

changes to the physical plant or operations.

Since these administrative changes do not

contribute to accident initiation, they do not

produce a new accident scenario or produce

a new type of equipment malfunction. Also,

these changes do not alter any existing

accident scenarios; they do not affect

equipment or its operation, and thus, do not

create the possibility of a new or different

kind of accident.

3. involve a significant reduction in the

margin of safety.

Operation of the facility in accordance

with the proposed TS would not involve a

significant reduction in a margin of safety.

The proposed changes do not affect plant

equipment or operation. Safety limits and

limiting safety system settings are not

affected by these proposed changes.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of Wisconsin,

Cofrin Library, 2420 Nicolet Drive,

Green Bay, Wisconsin 54311-7001.

Attorney for licensee: Bradley D.

Jackson, Esq., Foley and Lardner, P. O.

Box 1497, Madison, Wisconsin 53701-

1497.

NRC Project Director: Gail H. Marcus

Wolf Creek Nuclear Operating

Corporation, Docket No. 50-482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request:

September 14, 1995

Description of amendment request:

The proposed amendment would revise

Technical Specification 3/4.5.5 to

increase the outage time allowed for

adjusting the boron concentration of the

refueling water storage tank (RWST)

from 1 hour to 8 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 increase in the RWST allowed outage

time does not alter the plant configuration or

operation. The potential for the RWST boron

concentration to be outside the technical

specification limits is small because the

RWST and its contents are not involved with

normal plant operation and are not subject to

process variations associated with plant

operation.

The potential causes of boron

concentration deviation have been evaluated

with the conclusion that any deviation in

RWST boron concentration would not be

expected to increase significantly during the

proposed 7 hour allowed outage time

increase.

Therefore, the proposed change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

Increasing the RWST allowed outage time

from 1 hour to 8 hours for reasons directly

related to boron concentration does not

require physical alteration to any plant

system and does not change the method by

which any safety related system performs its

functions. 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.

Increasing the RWST allowed outage time

for reasons directly related to boron

concentration does not affect any accident

analysis assumptions, initial conditions, or

results. The margins of safety reflected in the

Wolf Creek Generating Station Technical

Specifications are not compromised by the 7

hour allowed outage time increase.

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

locations: Emporia State University,

William Allen White Library, 1200

Commercial Street, Emporia, Kansas

66801 and Washburn University School

of Law Library, Topeka, Kansas 66621

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts and Trowbridge,

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

20037

NRC Project Director: William H.

Bateman
Federal Register / Vol. 60, No. 196 / Wednesday, October 11, 1995 / Notices 52937

Previously Published Notices Of

Consideration Of Issuance Of

Amendments To Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

And Opportunity For A Hearing

The following notices were previously

published as separate individual

notices. The notice content was the

same as above. They were published as

individual notices either because time

did not allow the Commission to wait

for this biweekly notice or because the

action involved exigent circumstances.

They are repeated here because the

biweekly notice lists all amendments

issued or proposed to be issued

involving no significant hazards

consideration.

For details, see the individual notice

in the Federal Register on the day and

page cited. This notice does not extend

the notice period of the original notice.

Commonwealth Edison Company,

Docket Nos. STN 50-454 and STN 50-

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

Ogle County, Illinois, Docket Nos. STN

50-456 and STN 50-457, Braidwood

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

Illinois

Date of amendment request:

September 1, 1995

Description of amendment request:

The proposed amendments would

revise the present voltage-based repair

criteria in the Byron 1 and Braidwood

1 Technical Specifications (TSs). These

proposed revisions would raise the

lower voltage limit from its present

value of 1.0 volt to 3.0 volts; there

would no longer be an upper voltage

limit.

The Braidwood 1 TSs were revised by

License Amendment No. 54, issued on

August 18, 1994, to add voltage-based

repair criteria to the existing steam

generator (SG) tube repair criteria. The

Byron 1 TSs were revised in a similar

manner by License Amendment No. 66,

issued on October 24, 1994.

The voltage-based repair criteria in

the subject TSs are applicable only to a

specific type of SG tube degradation

which is predominantly axially-oriented

outer diameter stress corrosion cracking

(ODSCC). This particular form of SG

tube degradation occurs entirely within

the intersections of the SG tubes with

the tube support plates (TSPs).

The present voltage values for the

ODSCC repair criteria are based on the

assumption of a ‘‘free span’’ exposure of

the SG tube flaw; i.e., no credit is given

for any constraint against burst or

leakage, which may be provided by the

presence of the TSPs. This approach is,

in turn, based on the assumption that

under postulated accident conditions,

the TSPs may be displaced sufficiently

by blowdown hydrodynamic loads such

that a SG tube flaw which was fully

confined within the thickness of the

TSP prior to the accident would then be

fully exposed. This approach was first

advanced by the NRC staff in a draft

generic letter issued on August 12, 1994,

which was subsequently modified

slightly and issued as Generic letter (GL)

95-05, ‘‘Voltage-Based Repair Criteria

For Westinghouse Steam Generator

Tubes Affected by Outside Diameter

Stress Corrosion Cracking,’’ dated

August 3, 1995. The previous license

amendments related to the issue of

ODSCC were based to a large extent on

the draft generic letter cited above.

The fundamental difference between

the pending proposal to raise the lower

voltage repair limit to 3.0 volts and the

methodology contained in GL 95-05, is

that the licensee proposes to install

certain modifications to the SG internal

structures, thereby limiting to a small

value, the maximum displacement of

the TSPs under accident conditions.

The proposed structural modifications

consist of expanding a limited number

of SG tubes only on the hot leg side of

the TSP, at each of the intersections of

the tubes with the TSPs. The purpose of

this approach would be to greatly

reduce the probability of SG tube burst

under postulated accident conditions by

several orders of magnitude. There

would be a negligible impact on the

primary-to-secondary SG tube leakage

under accident conditions.

While the voltage-based repair criteria

for ODSCC flaws are applicable only to

Byron 1 and Braidwood 1, the pending

request for license amendments

involves all four units in that both

stations have a common set of TSs. Date

of publication of individual notice in

Federal Register: September 27, 1995

(60 FR 49963)

Expiration date of individual notice:

October 27, 1995

Local Public Document Room

location: For Byron, the Byron Public

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

Box 434, Byron, Illinois 61010; for

Braidwood, the Wilmington Public

Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481

Commonwealth Edison Company,

Docket Nos. 50-237 and 50-249,

Dresden Nuclear Power Station,

Grundy County, Illinois

Date of amendment request:

September 1, 1995

Description of amendment request:

The proposed amendment would

upgrade the Dresden TS to the standard

Technical Specifications (STS)

contained in NUREG-0123. The

Technical Specification Upgrade

Program (TSUP) is not a complete

adaption of the STS. The TS upgrade

focuses on (1) integrating additional

information such as equipment

operability requirements during

shutdown conditions, (2) clarifying

requirements such as limiting

conditions for operation and action

statements utilizing STS terminology,

(3) deleting superseded requirements

and modifications to the TS based on

the licensee’s responses to Generic

Letters (GL), and (4) relocating specific

items to more appropriate TS locations.

The September 1, 1995, application

proposed to upgrade only Section 6.0

(Administrative Controls) of the

Dresden TS.Date of publication of

individual notice in Federal Register:

September 20, 1995 (60 FR 48728)

Expiration date of individual notice:

October 20, 1995

Local Public Document Room

location: Morris Area Public Library

District, 604 Liberty Street, Morris,

Illinois 60450

Duquesne Light Company, et al., Docket

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

Power Station, Unit Nos. 1 and 2,

Shippingport, Pennsylvania

Date of amendment request:

September 13, 1995

Brief description of amendment

request: The proposed amendments

would revise the Administrative

Controls section and the Bases section

of the Beaver Valley Power Station, Unit

Nos. 1 and 2 (BVPS-1 and BVPS-2),

technical specifications to be 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 BVPS2 License Amendment No. 70 which

were issued June 12, 1995.Date of

publication of individual notice in

Federal Register: September 22, 1995

(60 FR 49292)

Expiration date of individual notice:

October 23, 1995

Local Public Document Room

location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa,

Pennsylvania 15001
52938 Federal Register / Vol. 60, No. 196 / Wednesday, October 11, 1995 / Notices

PECO Energy Company, Public Service

Electric and Gas Company, Delmarva

Power and Light Company, and

Atlantic City Electric Company, Docket

No. 50-278, Peach Bottom Atomic

Power Station, Unit No. 3, York County,

Pennsylvania

Date of amendment request:

September 1, 1995

Brief description of amendment

request: The proposed amendment

would delete License Condition 2.C.(5)

from Facility Operating License DPR-56

which restricts power levels to no less

than seventy percent in the coastdown

condition.

Date of publication of individual

notice in Federal Register: September

19, 1995 (60 FR 48530)

Expiration date of individual notice:

October 18, 1995

Local Public Document Room

location: Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Education

Building, Walnut Street and

Commonwealth Avenue, Box 1601,

Harrisburg, Pennsylvania 17105

Notice Of Issuance Of Amendments To

Facility Operating Licenses

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application

complies with the standards and

requirements of the Atomic Energy Act

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

Commission’s rules and regulations.

The Commission has made appropriate

findings as required by the Act and the

Commission’s rules and regulations in

10 CFR Chapter I, which are set forth in

the license amendment.

Notice of Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for A Hearing in

connection with these actions was

published in the Federal Register as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the applications for

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

the Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, the Gelman Building, 2120 L

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

local public document rooms for the

particular facilities involved.

Arizona Public Service Company, et al.,

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

and STN 50-530, Palo Verde Nuclear

Generating Station, Units 1, 2, and 3,

Maricopa County, Arizona

Date of application for amendments:

December 7, 1994, as supplemented by

letter dated August 1, 1995.

Brief description of amendments: The

amendments change Note 5 to Table 4.3-

1 of Technical Specification 3/4.3.1 to

allow verification of the shapeannealing matrix elements used in the

core protection calculators. This

provides the option of using generic

shape-annealing matrix elements in the

core protection calculators. Presently,

cycle-specific shape-annealing elements

are determined during startup testing

after each core reload. Use of a generic

shape-annealing matrix eliminates

several hours of critical path work

during startup after a refueling outage.

Date of issuance: September 20, 1995

Effective date: September 20, 1995

Amendment Nos.: Unit 1 -

Amendment No. 100; Unit 2 -

Amendment No. 88; Unit 3 -

Amendment No. 71

Facility Operating License Nos. NPF41, NPF-51, and NPF-74: The

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 4, 1995 (60 FR 495).

The August 1, 1995, supplemental letter

provided clarifying information and did

not change the original no significant

hazards consideration determination.

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated September 20,

1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona

85004

Commonwealth Edison Company,

Docket Nos. 50-237 and 50-249,

Dresden Nuclear Power Station, Units 2

and 3, Grundy County, Illinois Docket

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

Nuclear Power Station, Units 1 and 2,

Rock Island County, Illinois

Date of application for amendments:

March 26, 1993, as supplemented May

15, 1995

Brief description of amendments:

These amendments upgrade the current

custom Technical Specifications (TS)

for Dresden and Quad Cities to the

Standard Technical Specifications

contained in NUREG-0123, ‘‘Standard

Technical Specification General Electric

Plants BWR/4.’’ These amendments

upgrade only Section 3/4.9 (Electrical

Power Systems). These amendments

include the relocation of some TS

requirements to licensee-controlled

documents.

Date of issuance: September 18, 1995

Effective date: Immediately, to be

implemented no later than December

31, 1995, for Dresden Nuclear Power

Station and June 30, 1996, for Quad

Cities Nuclear Power Station.

Amendment Nos.: 138, 132, 160, 156

Facility Operating License Nos. DPR19, DPR-25, DPR-29 and DPR-30. The

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 19, 1994 (59 FR 2864)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated September 18, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: for Dresden, Morris Area

Public Library District, 604 Liberty

Street, Morris, Illinois 60450; for Quad

Cities, Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois 61021

Commonwealth Edison Company,

Docket Nos. 50-237 and 50-249,

Dresden Nuclear Power Station, Units 2

and 3, Grundy County, Illinois, Docket

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

Nuclear Power Station, Units 1 and 2,

Rock Island County, Illinois

Date of application for amendments:

December 8, 1992, as supplemented

September 10, 1993, and May 17, 1995.

Brief description of amendments: This

application upgrades the current custom

Technical Specifications (TS) for

Dresden and Quad Cities to the

Standard Technical Specifications (STS)

contained in NUREG-0123, ‘‘Standard

Technical Specification General Electric

Plants BWR/4.’’ This application

upgrades only Section 3/4.1 (Reactor

Protection System). Date of issuance:
Federal Register / Vol. 60, No. 196 / Wednesday, October 11, 1995 / Notices 52939

September 20, 1995Effective date:

Immediately, to be implemented no

later than December 31, 1995, for

Dresden Station and June 30, 1996, for

Quad Cities Station.

Amendment Nos.: 139, 133, 161, and

157

Facility Operating License Nos. DPR19, DPR-25, DPR-29 and DPR-30. The

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 6, 1995 (60 FR 29872)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated September 20, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: for Dresden, Morris Area

Public Library District, 604 Liberty

Street, Morris, Illinois 60450; for Quad

Cities, Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois 61021

Commonwealth Edison Company,

Docket Nos. 50-237 and 50-249,

Dresden Nuclear Power Station, Units 2

and 3, Grundy County, Illinois Docket

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

Nuclear Power Station, Units 1 and 2,

Rock Island County, Illinois

Date of application for amendments:

September 17, 1993, as supplemented

June 30, 1995.

Brief description of amendments: This

application upgrades the current custom

Technical Specifications (TS) for

Dresden and Quad Cities to the

Standard Technical Specifications (STS)

contained in NUREG-0123, ‘‘Standard

Technical Specification General Electric

Plants BWR/4.’’ This application

upgrades only Section 3/4.6.

Date of issuance: September 21, 1995

Effective date: Immediately, to be

implemented no later than December

31, 1995, for Dresden Station and June

30, 1996, for Quad Cities Station.

Amendment Nos.: 140, 134, 162, and

158

Facility Operating License Nos. DPR19, DPR-25, DPR-29 and DPR-30. The

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 19, 1995 (60 FR 37087)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated September 21, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: for Dresden, Morris Area

Public Library District, 604 Liberty

Street, Morris, Illinois 60450; for Quad

Cities, Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois 61021

Commonwealth Edison Company,

Docket Nos. 50-373 and 50-374, LaSalle

County Station, Units 1 and 2, LaSalle

County, Illinois

Date of application for amendments:

April 11, 1995

Brief description of amendments: The

amendments allow a one-time extension

of specific LaSalle, Units 1 and 2, 18-

month Technical Specification

Surveillance Requirements to allow

surveillance testing to coincide with the

LaSalle, Unit 1, seventh refueling outage

(L1R07). The shutdown for L1R07 has

been rescheduled from September 1995

until early 1996. The proposed

extensions apply to calibrations and

functional testing of isolation actuation

instrumentation, emergency core

cooling system actuation

instrumentation, and recirculation

pump trip actuation instrumentation;

leakage testing of reactor coolant system

isolation valves; inspection of fire-rated

seals; functional testing of mechanical

snubbers; inspections of emergency

diesel generators; and testing of

batteries, battery chargers, and other

electrical components.

Date of issuance: September 27, 1995

Effective date: Immediately, to be

implemented within 30 days.

Amendment Nos.: 106 and 92

Facility Operating License Nos. NPF11 and NPF-18: The amendments

revised the Facility Operating Licenses.

Date of initial notice in Federal

Register: July 5, 1995 (60 FR 35066) The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated September 27, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Jacobs Memorial Library,

Illinois Valley Community College,

Oglesby, Illinois 61348.

Duke Power Company, et al., Docket

Nos. 50-413 and 50-414, Catawba

Nuclear Station, Units 1 and 2, York

County, South Carolina

Date of application for amendments:

June 17, 1993, as supplemented July 5,

1995

Brief description of amendments: The

amendments revise Technical

Specification Section 5.3.1 ‘‘Fuel

Assemblies’’ in accordance with Generic

Letter 90-02, Supplement 1,

‘‘Alternative Requirements For Fuel

Assemblies in The Design Features

Section of Technical Specifications.’’

Date of issuance: September 18, 1995

Effective date: As of the date of

issuance to be implemented within 30

days from the date of issuance

Amendment Nos.: 135 and 129

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

the Technical Specifications.

Date of initial notice in Federal

Register: July 21, 1993 (58 FR 39048)

and ReNoticed August 16, 1995 (60 FR

42601) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

September 18, 1995. No significant

hazards consideration comments

received: No

Local Public Document Room

location: York County Library, 138 East

Black Street, Rock Hill, South Carolina

29730

Duquesne Light Company, et al., Docket

No. 50-334, Beaver Valley Power

Station, Unit No. 1, Shippingport,

Pennsylvania

Date of application for amendment:

July 11, 1995

Brief description of amendment: This

amendment revised the required area of

the reactor coolant system overpressure

protection system vent from 3.14 square

inches to 2.07 square inches which is

equal to the relief area of a single poweroperated relief valve.

Date of issuance: September 26, 1995

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment No.: 193

Facility Operating License No. DPR66. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 16, 1995 (60 FR 42603)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 26, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa,

Pennsylvania 15001

Duquesne Light Company, et al., Docket

No. 50-412, Beaver Valley

PowerStation, Unit 2, Shippingport,

Pennsylvania

Date of application for amendment:

July 24, 1995

Brief description of amendment: This

amendment revises TS 3/4.4.11, ‘‘Relief

Valves,’’ and associated Bases to make

Unit 2 TS 3/4.4.11 consistent with Unit

1 TS 3/4.4.11 which was revised by Unit

1 License Amendment No. 187 issued

on May 15, 1995. The amendment

generally reflects the guidance provided

in NRC Generic Letter 90-06 and in the

NRC’s Improved Standard Technical

Specifications (NUREG-1431).

Date of issuance: September 18, 1995
52940 Federal Register / Vol. 60, No. 196 / Wednesday, October 11, 1995 / Notices

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment No.: 76

Facility Operating License No. NPF73: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 16, 1995 (60 FR 42604)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 18, 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.

50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of application for amendment:

March 17, 1995

Brief description of amendment: The

amendment revises requirements

associated with the frequency of

containment post-entry visual

inspections.

Date of issuance: September 15, 1995

Effective date: September 15, 1995

Amendment No.: 162

Facility Operating License No. NPF-6.

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 19, 1995 (60 FR 37089)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 15, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Entergy Operations, Inc., Docket No.

50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of application for amendment:

October 27, 1993

Brief description of amendment: The

amendment relocated reactor incore

detector requirements from the TSs to

the safety analysis report.

Date of issuance: September 15, 1995

Effective date: September 15, 1995

Amendment No.: 163

Facility Operating License No. NPF-6.

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: December 8, 1993 (58 FR

64606) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

September 15, 1995. No significant

hazards consideration comments

received: No

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Entergy Operations, Inc., Docket No.

50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of application for amendment:

March 17, 1995

Brief description of amendment: The

amendment transfers requirements for

cycle specific core operating limits from

the Technical Specifications to the Core

Operating Limits Report. Additionally, a

reference to a statistical methodology for

determining uncertainties is being

changed to reference a methodology that

was recently approved by the NRC.

Date of issuance: September 19, 1995

Effective date: September 19, 1995

Amendment No.: 164

Facility Operating License No. NPF-6.

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 19, 1995 (60 FR 37088)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 19, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Entergy Operations, Inc., Docket No.

50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of application for amendment:

April 4, 1995, as supplemented August

25, 1995

Brief description of amendment: The

amendment provides a one-time

extension of the reactor coolant pump

flywheel inservice inspection.

Date of issuance: September 22, 1995

Effective date: September 22, 1995

Amendment No.: 165

Facility Operating License No. NPF-6.

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 5, 1995 (60 FR 35069) The

August 25, 1995, submittal did not

change the original no significant

hazards consideration

determination.The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

September 22, 1995. No significant

hazards consideration comments

received: No

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Entergy Operations, Inc., Docket No.

50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of application for amendment:

May 19, 1995 as supplemented July 21,

1995.

Brief description of amendment: The

amendment revises the specifications to

permit the containment personnel

airlock doors to remain open during fuel

handling.

Date of issuance: September 28, 1995

Effective date: September 28, 1995

Amendment No.: 166

Facility Operating License No. NPF-6.

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 2, 1995 (60 FR 39437)

The July 22, 1995, supplement provided

clarifying information and did not

change the original no significant

hazards consideration determination.

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 28, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Entergy Operations, Inc., Docket No.

50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of application for amendment:

April 4, 1995, as

supplementedSeptember 28, 1995

Brief description of amendment: The

amendment removes the requirement to

maintain water level 23 feet above

irradiated fuel assemblies in the reactor

while latching and unlatching control

element assemblies.

Date of issuance: September 28, 1995

Effective date: September 28, 1995

Amendment No.: 167

Facility Operating License No. NPF-6.

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 16, 1995 (60 FR 42604)

The September 28, 1995, submittal

provided clarifying information and did

not change the original no significant

hazards consideration determination.

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 28, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801
Federal Register / Vol. 60, No. 196 / Wednesday, October 11, 1995 / Notices 52941

Entergy Operations, Inc., Docket No.

50-382, Waterford Steam Electric

Station, Unit 3, St. Charles Parish,

Louisiana

Date of amendment request: June 22,

1994, as supplemented by letters dated

June 28, 1995 and August 22, 1995

Brief description of amendment: The

amendment changes the Appendix A

TSs by increasing the control room

radiation monitor setpoint (CRRMS) to a

fixed value of 5.45E-6 micro curies per

cubic centimeters instead of being set at

two times the background.

Date of issuance: September 27, 1995

Effective date: Septembe 27, 1995

Amendment No.: 114

Facility Operating License No. NPF38. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 3, 1994 (59 FR 39586)

The June 28, 1995 and August 22, 1995,

letters provided clarifying information

that did not change the originial

proposed no significant hazards

consideration determination. The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated September 27, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: University of New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, LA 70122

GPU Nuclear Corporation, et al.,

Docket No. 50-289, Three Mile Island

Nuclear Station, Unit No. 1, Dauphin

County, Pennsylvania

Date of application for amendment:

August 11, 1995

Brief description of amendment: The

amendment removes the Technical

Specifications for the Makeup,

Purification, and Chemical Addition

Systems from the Technical

Specifications (Section 3.2) and

relocates the pertinent design

information, including tank volume and

boron concentrations, to the TMI-1

Updated Final Safety Analysis Report.

Date of issuance: September 19, 1995

Effective date: September 19, 1995

Amendment No.: 196

Facility Operating License No. DPR50. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 18, 1995 (60 FR 43172)

The Commission’s related evaluation of

this amendment is contained in a Safety

Evaluation dated September 19, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Law/Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Walnut

Street and Commonwealth Avenue, Box

1601, Harrisburg, PA 17105

Illinois Power Company and Soyland

Power Cooperative, Inc., Docket No. 50-

461, Clinton Power Station, Unit No. 1,

DeWitt County, Illinois

Date of application for amendment:

June 9, 1995

Brief description of amendment: The

amendment modifies Technical

Specification 4.1, ‘‘Site Location,’’ to

incorporate a description of the

exclusion area boundary. The change is

necessary to ensure the content of the

technical specifications conform to

Section 182 of the Atomic Energy Act of

1954.

Date of issuance: September 14, 1995

Effective date: September 14, 1995

Amendment No.: 101

Facility Operating License No. NPF62: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: July 19, 1995 (60 FR 37093)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 14, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: The Vespasian Warner Public

Library, 120 West Johnson Street,

Clinton, Illinois 61727

Power Authority of The State of New

York, Docket No. 50-286, Indian Point

Nuclear Generating Unit No. 3,

Westchester County, New York

Date of application for amendment:

July 21, 1995

Brief description of amendment: The

amendment revised Technical

Specifications Section 6.0

(Administrative Controls) to replace the

title-specific list of members on the

Plant Operating Review Committee

(PORC) with a more general statement of

membership requirements. The scope of

disciplines represented on the PORC

was also expanded to include nuclear

licensing and quality assurance. The

amendment also changed the title

‘‘Resident Manager’’ to ‘‘Site Executive

Officer.’’ This title change was an

administrative change that did not affect

the reporting relationship, authority, or

responsibility of the position.

Date of issuance: September 20, 1995

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 163

Facility Operating License No. DPR64: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 16, 1995 (60 FR 42606)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 20, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10610

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of application for amendment:

April 25, 1994

Brief description of amendment: This

amendment revises TS Section 3.8.1.1,

‘‘A.C. Sources - Operating,’’ TS Section

3.8.1.2, ‘‘A.C. Sources - Shutdown,’’ and

associated Bases, to increase the

required quantity of fuel in the

Emergency Diesel Generator Fuel Oil

Day Tanks from 200 to 360 gallons.

Date of issuance: September 15, 1995

Effective date: As of the date of

issuance and shall be implemented

within 60 days.

Amendment No.: 79

Facility Operating License No. NPF57: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: June 8, 1994 (59 FR 29632)The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated September 15, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey

08070

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of application for amendment:

January 20, 1995

Brief description of amendment: This

amendment changes Technical

Specification (TS) 4.1.3.1.2.b, ‘‘Control

Rods - Surveillance Requirement’’ to

change the required action to be taken

when a control rod becomes immovable

due to excessive friction from ‘‘at least

once per’’ 24 hours to ‘‘within’’ 24

hours.

Date of issuance: September 20, 1995

Effective date: As of its date of

issuance, to be implemented within 60

days.

Amendment No.: 80

Facility Operating License No. NPF57: This amendment revised the

Technical Specifications.
52942 Federal Register / Vol. 60, No. 196 / Wednesday, October 11, 1995 / Notices

Date of initial notice in Federal

Register: August 2, 1995 (60 FR 39452)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 20, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey

08070

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of application for amendment:

January 11, 1995

Brief description of amendment: This

amendment changes Technical

Specification (TS) 3/4.3.8, ‘‘Turbine

Overspeed Protection System,’’

removing these requirements from the

TS and relocating the Bases to the Hope

Creek Updated Final Safety Analysis

Report (UFSAR) and the Surveillance

Requirements to the applicable

surveillance procedures. The Limiting

Conditions for Operation (LCOs) are

eliminated.

Date of issuance: September 25, 1995

Effective date: As of the date of

issuance and shall be implemented

within 60 days.

Amendment No.: 81

Facility Operating License No. NPF57: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: August 2, 1995 (60 FR 39451).

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 25, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey

08070

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of application for amendment:

September 29, 1994

Brief description of amendment: This

amendment changes Technical

Specification (TS) Sections 3/4.3.7.2,

‘‘Seismic Monitoring Instrumentation,’’

and 3/4.3.7.3, ‘‘Meteorological

Instrumentation,’’ to remove the

requirements from the TS and relocate

the appropriate descriptive information

and testing requirements to the Hope

Creek Updated Final Safety Analysis

Report.

Date of issuance: September 25, 1995

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment No.: 82

Facility Operating License No. NPF57: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: August 2, 1995 (60 FR 39449).

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 25, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey

08070

Public Service Electric & Gas Company,

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

Nuclear Generating Station, Unit Nos. 1

and 2, Salem County, New Jersey

Date of application for amendments:

September 20, 1994

Brief description of amendments: The

amendments change the Channel

Functional Test surveillance frequency

for the Manual Reactor Trip Switches

and Reactor Trip Breakers (RTB) and

relocate the RTB maintenance

requirements from the Technical

Specifications to the Salem Updated

Final Safety Analysis Report.

Date of issuance: September 18, 1995

Effective date: Both units, as of the

date of issuance, to be implemented

within 60 days.

Amendment Nos.: 176 and 157

Facility Operating License Nos. DPR70 and DPR-75. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: November 9, 1994 (59 FR

55890 The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

September 18, 1995. No significant

hazards consideration comments

received: No

Local Public Document Room

location: Salem Free Public Library, 112

West Broadway, Salem, New Jersey

08079

Public Service Electric & Gas Company,

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

Nuclear Generating Station, Unit Nos. 1

and 2, Salem County, New Jersey

Date of application for amendments:

January 21, 1994, as supplemented June

28 and September 13, 1994, and April

4, 1995.

Brief description of amendments:

Revised Technical Specifications

3.8.2.3, ‘‘125-Volt D.C. DISTRIBUTION -

OPERATING.’’

Date of issuance: September 19, 1995

Effective date: Both units, as of the

day of issuance and shall be

implemented within 60 days.

Amendment Nos.: 177 and 158

Facility Operating License Nos. DPR70 and DPR-75. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: April 28, 1994 (58 FR 22012)

The June 28 and September 13, 1994,

and April 4, 1995 letters provided

clarifying information that did not

change the scope of the January 21, 1994

application and initial proposed no

significant hazards consideration

determination, nor go beyond the scope

of the Federal Register notice. The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated September 19, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Salem Free Public Library, 112

West Broadway, Salem, New Jersey

08079

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 application for amendment:

June 19, 1995, as supplemented on

August 21, 1995.

Brief description of amendment: The

amendment revises the Technical

Specifications to change the required

test frequency for the reactor building

spray nozzle flow test from once per five

years to once per ten years.

Date of issuance: September 18, 1995

Effective date: September 18, 1995

Amendment No.: 127

Facility Operating License No. NPF12: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: July 19, 1995 (60 FR 37100).

The August 21, 1995 letter provided

supplemental information that did not

change the initial proposed no

significant hazards consideration. The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated September 18, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Fairfield County Library, 300

Washington Street, Winnsboro, SC

29180

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 application for amendment:

July 28, 1995
Federal Register / Vol. 60, No. 196 / Wednesday, October 11, 1995 / Notices 52943

Brief description of amendment: The

amendment revises the Technical

Specifications to exclude the

requirement to perform the slave relay

test of the 36-inch containment purge

supply and exhaust valves on a

quarterly basis while the plant is in

Modes 1, 2, 3, or 4.

Date of issuance: September 18, 1995

Effective date: September 18, 1995

Amendment No.: 128

Facility Operating License No. NPF12. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: August 16, 1995 (60 FR 42608)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 18, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Fairfield County Library, 300

Washington Street, Winnsboro, SC

29180

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 application for amendment:

June 19, 1995, as supplemented on

August 21, 1995.

Brief description of amendment: The

amendment revises the Technical

Specifications to change the required

test frequency for the reactor building

spray nozzle flow test from once per five

years to once per ten years.

Date of issuance: September 18, 1995

Effective date: September 18, 1995

Amendment No.: 129

Facility Operating License No. NPF12. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: July 19, 1995 (60 FR 37100).

The August 21, 1995 letter provided

supplemental information that did not

change the initial proposed no

significant hazards consideration. The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated September 18, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Fairfield County Library, 300

Washington Street, Winnsboro, SC

29180

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 3, 1995

Brief description of amendment: The

amendment revised the Technical

Specifications (TS) to relocate

radiological effluent and radiological

environmental monitoring TS to the

Offsite Dose Calculation Manual or to

the Process Control Program.

Programmatic controls for radioactive

effluent and radiological environmental

monitoring were included in TS 6.8.4.

Date of issuance: September 15, 1995

Effective date: September 15, 1995

Amendment No.: 72

Facility Operating License No. NPF58: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: May 10, 1995 (60 FR 24921)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 15, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081

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:

June 1, 1995

Brief description of amendment: The

amendment revised the Technical

Specifications to make them more

restrictive regarding control rod drive

scram time testing. CRD scram time

testing would be required following

maintenance prior to considering the

CRD operable, and could be performed

at any reactor pressure. Additional

testing would be required when reactor

coolant pressure is greater than or equal

to 950 psig and prior to 40 percent rated

thermal power.

Date of issuance: September 26, 1995

Effective date: September 26, 1995

Amendment No. 73

Facility Operating License No. NPF58: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: August 2, 1995 (60 FR 39452)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 26, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081

Washington Public Power Supply

System, Docket No. 50-397, Nuclear

Project No. 2, Benton County,

Washington

Date of application for amendment:

January 14, 1992, as supplemented by

letters dated February 10, 1995, and

August 16, 1995.

Brief description of amendment: The

amendment revises technical

specification surveillance requirements

regarding demonstration of jet pump

operability and corrects several

administrative discrepancies.

Date of issuance: September 18, 1995

Effective date: September 18, 1995, to

be implemented within 30 days of

issuance

Amendment No.: 141

Facility Operating License No. NPF21: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: May 27, 1992 (57 FR 22272)

and March 29, 1995 (60 FR 16204). The

August 16, 1995, supplemental letter

provided additional clarifying

information and did not change the

initial no significant hazards

consideration determination. The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated September 18,

1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: Richland Public Library, 955

Northgate Street, Richland, Washington

99352

Dated at Rockville, Maryland, this 3rd day

of October 1995.

For the Nuclear Regulatory Commission

Elinor G. Adensam,

Deputy Director, Division of Reactor Projects

- III/IV, Office of Nuclear Reactor Regulation

[Doc. 95–25006 Filed 10–10–95; 8:45 am]

BILLING CODE 7590–01–F

[Docket No. 50–251]

Florida Power and Light Company

(Turkey Point Unit 4); Exemption

I

Florida Power and Light Company

(the licensee) is the holder of Facility

Operating License No. DPR–41, which

authorizes operation of Turkey Point

Unit 4 (the facility), at a steady-state