Federal Register / Vol. 61, No. 216 / Wednesday, November 6, 1996 / Notices 57481

4. How can the burden of the

information collection be minimized,

including the use of automated

collection techniques or other forms of

information technology?

A copy of the submittal may be

viewed free of charge at the NRC Public

Document Room, 2120 L Street NW,

(lower level), Washington, DC. Members

of the public who are in the

Washington, DC, area can access this

document via modem on the Public

Document Bulletin Board (NRC’s

Advanced Copy Document Library),

NRC subsystem at FedWorld, 703–321–

3339. Members of the public who are

located outside of the Washington, DC,

area can dial FedWorld, 1–800–303–

9672, or use the FedWorld Internet

address: fedworld.gov (Telnet). The

document will be available on the

bulletin board for 30 days after the

signature date of this notice. If

assistance is needed in accessing the

document, please contact the FedWorld

help desk at 703–487–4608. Additional

assistance in locating the document is

available from the NRC Public

Document Room, nationally at 1–800–

397–4209, or within the Washington,

DC, area at 202–634–3273.

Comments and questions should be

directed to the OMB reviewer by

December 6, 1996: Edward Michlovich,

Office of Information and Regulatory

Affairs (3150–0010), NEOB–10202,

Office of Management and Budget,

Washington, DC 20503.

Comments can also be submitted by

phone at (202) 395–3084.

The NRC Clearance Officer is Brenda

Jo. Shelton, (301) 415–7233.

Dated at Rockville, Maryland, this 7th day

of October, 1996.

For the Nuclear Regulatory Commission.

Gerald F. Cranford,

Designated Senior, Official for Information

Resources Management.

[FR Doc. 96–28505 Filed 11–5–96; 8:45 am]

BILLING CODE 7590–01–P

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 October 11,

1996, through October 25, 1996. The last

biweekly notice was published on

October 23, 1996.

Notice Of Consideration Of Issuance Of

Amendments To Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

And Opportunity For A Hearing

The Commission has made a

proposed determination that the

following amendment requests involve

no significant hazards consideration.

Under the Commission’s regulations in

10 CFR 50.92, this means that operation

of the facility in accordance with the

proposed amendment would not (1)

involve a significant increase in the

probability or consequences of an

accident previously evaluated; or (2)

create the possibility of a new or

different kind of accident from any

accident previously evaluated; or (3)

involve a significant reduction in a

margin of safety. The basis for this

proposed determination for each

amendment request is shown below.

The Commission is seeking public

comments on this proposed

determination. Any comments received

within 30 days after the date of

publication of this notice will be

considered in making any final

determination.

Normally, the Commission will not

issue the amendment until the

expiration of the 30-day notice period.

However, should circumstances change

during the notice period such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received before

action is taken. Should the Commission

take this action, it will publish in the

Federal Register a notice of issuance

and provide for opportunity for a

hearing after issuance. The Commission

expects that the need to take this action

will occur very infrequently.

Written comments may be submitted

by mail to the Chief, Rules Review and

Directives Branch, Division of Freedom

of Information and Publications

Services, Office of Administration, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555-0001, and

should cite the publication date and

page number of this Federal Register

notice. Written comments may also be

delivered to Room 6D22, Two White

Flint North, 11545 Rockville Pike,

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

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

written comments received may be

examined at the NRC Public Document

Room, the Gelman Building, 2120 L

Street, NW., Washington, DC. The filing

of requests for a hearing and petitions

for leave to intervene is discussed

below.

By December 6, 1996, the licensee

may file a request for a hearing with

respect to issuance of the amendment to

the subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be
57482 Federal Register / Vol. 61, No. 216 / Wednesday, November 6, 1996 / Notices

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555-0001, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington DC, by

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

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

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

operator should be given Datagram

Identification Number N1023 and the

following message addressed to (Project

Director): petitioner’s name and

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555-0001, and to the attorney for

the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

factors specified in 10 CFR

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

For further details with respect to this

action, see the application for

amendment which is available for

public inspection at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, and at the local public

document room for the particular

facility involved.

Baltimore Gas and Electric Company,

Docket No. 50-317, Calvert Cliffs

Nuclear Power Plant, Unit No. 1,

Calvert County, Maryland

Date of amendment request: October

3, 1996

Description of amendment request:

The proposed amendment changes the

provision for receiving, possessing and

using byproducts, source and special

nuclear material at Calvert Cliffs Unit 1.

Currently, Unit 1 is licensed under 10

CFR Part 30 to receive, possess, and use

100 millicuries of byproduct material

for sample analysis or instrument

calibration, 500 millicuries of byproduct

material in the form of equipment; and

500 millicuries of Sodium-24 for steam

turbine acceptance testing. In addition,

Unit 1 is licensed to receive, possess

and use 100 milligrams each of source

or special nuclear material under 10

CFR Parts 40 and 70. Unit 2 is licensed

under 10 CFR Parts 30, 40, and 70 to

receive, possess, and use in amounts as

required any byproduct, source, or

special nuclear material for sample

analysis or instrument calibration or

associated with radioactive apparatus or

components. This proposed amendment

would change the Unit 1 license to be

consistent with the Unit 2 license by

replacing license conditions 2.B.3 and

2.B.4 with the same wording as Unit 2’s

license condition 2.B.4.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Would not involve a significant increase

in the probability or consequences of an

accident.

Currently, Unit 1 is licensed under 10 CFR

Part 30 to receive, possess, and use 100

millicuries of byproduct material for sample

analysis or instrument calibration, 500

millicuries of byproduct material in the form

of equipment; and 500 millicuries of Sodium24 for steam turbine acceptance testing. Unit

1 is also licensed under 10 CFR parts 40 and

70 to receive, possess, and use 100

milligrams of source or special nuclear

material. Unit 2 is licensed under 10 CFR

Parts 30, 40, and 70 to receive, possess, and

use in amounts as required any byproduct,

source, or special nuclear material for sample

analysis or instrument calibration or

associated with radioactive apparatus or

components. This proposed amendment

would change the Unit 1 license to be

consistent with the Unit 2 license. The

reason for this proposed change is that it is

sometimes necessary to receive and use

byproduct material, sources, or special

nuclear material with different activity

levels, and in different quantities than is

specified by the Unit 1 license.

The current licenses for the two units

allow radioactive materials to be accepted

and used at Unit 2, although these same

materials would not be acceptable for use at

Unit 1. These byproduct, source, and special

nuclear materials are used by the same

people and for the same function in either
Federal Register / Vol. 61, No. 216 / Wednesday, November 6, 1996 / Notices 57483

unit. Training and procedures for handling

radioactive material have been developed

and used at both Units over the last 20 years.

These procedures are adequate to control the

acceptance and use of radioactive material at

Unit 2 and, therefore, adequate to control

radioactive material at Unit 1.

Receiving, possessing, and using

byproduct, source, or special nuclear

material is not related to accident conditions.

Therefore, changing the Unit 1 license

conditions to be the same as the Unit 2

license condition does not involve a

significant increase in the probability or

consequences of an accident.

2. Would not create the possibility of a new

or different type of accident from any

accident previously evaluated.

Procedures and training governing the

acceptance and use of radioactive materials

are the same for both Unit 1 and Unit 2.

These procedures will not be changed as a

result of this license change. In addition,

receiving, possessing, and using radioactive

material is not related to accident conditions.

Therefore, making the Unit 1 license the

same as the Unit 2 license will not create the

possibility of a new or different type of

accident from any accident previously

evaluated.

3. Would not involve a significant

reduction in the margin of safety.

The margin of safety in this case is

exposure to contaminated material or

equipment. Exposure is controlled by

adequate training and procedures.

Radioactive material is received by personnel

assigned to the Radiation Safety Section.

These personnel are trained in receiving and

shipping contaminated material. Once the

material is onsite, it becomes the

responsibility of the radiation protection staff

who are trained in the handling of all levels

of radioactive material. Training and

procedures for handling radioactive materials

have been developed and used over the 20-

year life of the plant, and are currently

deemed adequate for compliance with the

Unit 2 license. Therefore, making the Unit 1

license the same as the Unit 2 license will

not involve a significant reduction in the

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 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: Calvert County Library, Prince

Frederick, Maryland 20678.

Attorney for licensee: Jay E. Silbert,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: S. Singh Bajwa,

Acting Director

Commonwealth Edison Company,

Docket Nos. 50-295 and 50-304, Zion

Nuclear Power Station, Units 1 and 2,

Lake County, Illinois

Date of amendment request:

September 20, 1996

Description of amendment request:

The proposed amendments would add a

footnote to specification 4.3.1.B.4.A.10.a

which refers to a letter that describes

enhancements made to the Combustion

Engineering sleeve installation process.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed amendment continues to

allow the Combustion Engineering sleeves to

be used as an alternate tube repair method for

Zion steam generators, along with the process

enhancements which are described in the

letter identified in the proposed Technical

Specification note. The sleeve configuration,

which was designed and analyzed in

accordance with the criteria of Regulatory

Guide (RG) 1.121 and Section III of the ASME

Code, is unaffected by the enhancements.

Fatigue and stress analyses of the sleeved

tube assemblies as described in the currently

approved Topical Report, CEN-331-P,

Revision 1-P, are unaffected by the

enhancements.

Mechanical testing which has shown that

the structural integrity of the sleeves under

normal, faulted, and upset conditions is

within the acceptable limits and is unaffected

by the enhancements. Leakage rate testing for

the tube sleeves which has demonstrated that

primary to secondary leakage is not expected

during any plant condition is unaffected by

the enhancements. The consequences of

leakage through the sleeved region of the

tube, including the enhancements, is

bounded by the existing steam generator tube

rupture (SGTR) analysis included in the Zion

Updated Final Safety Analysis Report.

The proposed Technical Specification

change reflects enhancements to the

installation and inspection process identified

in Topical Report CEN-331-P, Revision 1-P,

which is currently referenced in the

Technical Specifications. These

enhancements do not increase the probability

or consequences of an accident previously

evaluated. The enhancement which

disallows the installation of the tube plugs

made from Inconel 600 material was done so

based upon industry information and is

addressed by NRC Bulletin 89-01. The use of

the Plus Point Probe, its associated data

acquisition equipment, and improved visual

inspection equipment, are conservative

actions and improve the quality of the

sleeving process. The use of the mechanical

plug in lieu of the welded plug meets the

established design requirements and is

advantageous in the area of dose reduction,

because of reduced time to install. Minor

changes to the sleeve installation equipment

as described in the Topical Report, represent

equipment enhancements and do not alter

the sleeve design or qualification testing.

The proposed Technical Specification

change does not adversely impact any

previously evaluated design basis accident.

Installation of the sleeves, with the described

enhancements, can be used to repair

degraded tubes by returning the condition of

the tubes to their original design basis

condition for tube integrity and leak tightness

during all plant conditions. Therefore, the

currently approved sleeving process with the

described enhancements will not increase the

probability of occurrence of an accident

previously evaluated.

Therefore, these proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The implementation of the enhancements

to the proposed sleeving process will not

affect the plant design basis. The current

stress and fatigue analyses of the repair

identified in Topical Report CEN-331-P,

Revision 1-P, has shown the ASME Code and

RG 1.121 allowable values are met and are

unaffected by the described enhancements.

The current sleeving design, with the

described enhancements, will continue to

maintain overall tube bundle structural

integrity and leak tightness at a level

consistent with that of the originally

supplied tubing. Leak and mechanical testing

of the sleeves, are unaffected by the proposed

enhancements and continue to support the

conclusions that the sleeve retains both

structural integrity and leak tightness during

all operating and accident conditions. Repair

of a tube with a sleeve, utilizing the

described enhancements, does not provide a

mechanism that results in an accident

outside of the area affected by the sleeve.

The described change to implement the

cited enhancements will not create a new or

different type of accident. The change only

reflects enhancements to the currently

approved installation/inspection process

and, would not change or impact any

hypothetical accident previously discussed.

Use of improved Non-Destructive

Examination, data acquisition and visual

inspection equipment improves the quality of

the sleeving process and has no negative

effect on the margin of safety. The

elimination of the use of the Inconel 600 plug

also improves the margin of safety.

Any hypothetical accident as a result of

potential tube or sleeve degradation in the

repaired portion of the tube is bounded by

the existing SGTR analysis. The sleeve

design, including described enhancements,

does not affect any other component, or affect

any location on the tube outside of the

immediate area repaired.

Therefore, the proposed changes do not

create the possibility of a new or different

type of accident from any accident

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.
57484 Federal Register / Vol. 61, No. 216 / Wednesday, November 6, 1996 / Notices

The currently approved sleeving repair of

degraded steam generator tubes has been

shown by analysis to restore the integrity of

the tube bundle to its original design basis

condition. By implementing the described

enhancements, the consistent quality of the

upper sleeve weld has increased thereby

reducing the potential for rework and

reducing the potential for leaving a weld

indication in service.

The proposed change does not involve a

reduction in the margin of safety. The change

reflects enhancements to the installation/

inspection processes which are currently

referenced in the Technical Specifications.

These enhancements would not have any

adverse effects on the previously evaluated

design transient or accident analyses. The

enhancements represent acceptable industry

standards.

Therefore, the proposed changes do not

involve a significant reduction in the margin

of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

requested amendments involve no

significant hazards consideration.

Local Public Document Room

location: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois

60085

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60603

NRC Project Director: Robert A. Capra

Entergy Operations Inc., Docket No. 50-

382, Waterford Steam Electric Station,

Unit 3, St. Charles Parish, Louisiana

Date of amendment request: August

21, 1996

Description of amendment request:

The proposed amendment will modify

Containment Penetrations Nos. 53 and

65 design by modifying the design of

instrumentation lines for Containment

Vacuum Relief (CVR) system that pass

through these containment penetrations.

The proposed change will correct the

error in previously docketed

information that was used by NRC

during licensing process.

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 will not increase the

probability of previously analyzed accidents.

The proposed change seeks to clearly

document the design and licensing bases for

acceptance of the CVR sensing instrument

lines. The proposed change to the monitoring

lines will provide greater assurance that

containment integrity will be maintained

following a LOCA concurrent with a single

active failure. The design change to the nonessential monitoring line will reduce the

potential bypass leakage from penetrations 53

and 65 by adding a redundant automatic

containment isolation valve on penetration

53 and isolating the non-essential instrument

line on penetration 65. This design change

can be performed at power without violating

any license/regulatory requirements that

ensure containment integrity is maintained.

There is no change in the function of the

instrumentation. The only difference is that

CVR-IDPT-5017B and C non-safety

differential transmitters that monitor the CVR

system will be sensing containment pressure

from penetration 53. If the non-essential line

coming from penetration 53 becomes

inoperable, containment to annulus

differential pressure can be obtained from

alternate instrumentation. The essential

sensing line that actuates the CVR system to

protect containment within design vacuum

pressure is not affected by the design change.

Adding a redundant automatic

containment isolation valve in penetration

53’s non-essential instrument line instead of

the excess flow check valve and isolating the

non-essential line in penetration 65’s will

significantly reduce the potential bypass

leakage. The proposed change will credit the

essential instrument lines as a closed system

outside containment. The appropriate testing

and acceptance criteria will be applied to

ensure that any leakage associated with these

potential bypass leakage paths, will not

exceed the limits used in the Waterford 3

safety analysis or result in a significant

increase in analyzed dose consequences.

Therefore, the proposed change will not

involve significant increase in the probability

or consequences of any accident previously

evaluated.

The proposed change will credit the

essential sensing lines outside containment

as a closed system and will not affect the

plant or the manner in which the plant [is]

operated.

The failure modes associated with

containment isolation remain unchanged as a

result of the design change to the nonessential monitoring lines. The function of

the non-safety instrumentation is not

affected. The only difference is that all of the

non-safety instrumentation will be sensing

containment pressure from penetration 53.

However, if the non-essential line coming

from penetration 53 becomes inoperable,

containment pressure can be obtained from

alternate instrumentation. Adding a

redundant automatic containment isolation

valve in series with CVR 401A in the nonessential instrument line ensures

containment isolation following a LOCA with

a concurrent a single active failure.

Therefore, the proposed change will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The addition of a redundant automatic

containment isolation valve in series with

CVR 401A in the non-essential instrument

line breaching penetration 53 ensures

containment isolation postulating a single

active failure on a Containment Isolation

Actuation Signal (CIAS). While this

modification is performed, administrative

controls will require containment integrity to

be maintained by a seismic Category 1,

ASME Section III, Class 2, passive

containment isolation device.

The essential CVR instrument sensing lines

form a seismically qualified, closed system

outside containment which is designed for

pressure equal to or greater than

containment. The instrument cabinets C3A(B) are seismic Category I and safety

related. The instruments are Safety Class 1E

and have a static pressure rating of 1000 psig.

These lines meet the criteria of BTP CSB 6-

3 for crediting a closed system as a leakage

boundary to preclude bypass leakage by

being designed, fabricated, erected, and

tested to standards commensurate with the

safety function to be performed. The

proposed change will apply the appropriate

testing and acceptance criteria to ensure that

any leakage associated with these potential

bypass leakage paths, will not exceed the

limits used in the Waterford 3 safety analysis

or result in a significant increase in analyzed

dose consequences. Therefore, the proposed

change will not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, LA 70122

Attorney for licensee: N.S. Reynolds,

Esq., Winston & Strawn 1400 L Street

N.W., Washington, D.C. 20005-3502

NRC Project Director: William D.

Beckner

GPU Nuclear Corporation, et al.,

Docket No. 50-219, Oyster Creek

Nuclear Generating Station, Ocean

County, New Jersey

Date of amendment request: October

4, 1996 (TSCR No. 250)

Description of amendment request:

The proposed Technical Specification

(TS) change reflects a change in the

Safety Limit Minimum Critical Power

Ratio (SLMCPR) and as a result, a

change in the operating Minimum

Critical Power Ratio limit.

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 derivation of the revised SLMCPR for

Oyster Creek for incorporation into the TS,
Federal Register / Vol. 61, No. 216 / Wednesday, November 6, 1996 / Notices 57485

and its use to determine cycle-specific

thermal limits, have been performed using

NRC-approved methods. Additionally,

interim implementing procedures, which

incorporate cycle-specific parameters, have

been used. Based on the use of these

calculations, the revised SLMCPR will not

increase the probability or consequences 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

transition boiling if the limit is not violated.

The new SLMCPR preserves the existing

margin to transition boiling and fuel damage

in the event of a postulated accident. The

probability of fuel damage is not increased.

Revising the operating MCPR limit for

stability will ensure that adequate margin is

retained to the SLMCPR.

Therefore, the proposed TS change does

not involve an increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed TS changes do not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The MCPR Safety Limit 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. The stability

MCPR limit ensures an adequate operating

MCPR margin to the SLMCPR. These revised

limits cannot create the possibility of any

new type of accident. The new SLMCPR has

been calculated using NRC-approved

methods. Additionally, interim procedures,

which incorporate cycle-specific parameters,

have been used. 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 changes do not involve

a significant reduction in a margin of safety.

The margin of safety as defined in the TS

Bases will remain the same. The new

SLMCPR is calculated using NRC-approved

methods which are in accordance with the

current fuel design and licensing criteria.

Additionally, interim implementing

procedures, which incorporate cycle-specific

parameters, have been used. The MCPR

Safety Limit remains high enough to ensure

that greater than 99.9% of all fuel rods in the

core will avoid transition boiling if the limit

is not violated, thereby preserving fuel

cladding integrity. The revised stability

MCPR limit retains the existing margin to the

SLMCPR. 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: Ocean County Library,

Reference Department, 101 Washington

Street, Toms River, NJ 08753

Attorney for licensee: Ernest L. Blake,

Jr., Esquire. Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: John F. Stolz

GPU Nuclear Corporation, et al.,

Docket No. 50-219, Oyster Creek

Nuclear Generating Station, Ocean

County, New Jersey

Date of amendment request: October

10, 1996 (TSCR No. 203)

Description of amendment request:

The proposed Technical Specification

revision will extend the instrumentation

surveillances for Condenser Low

Vacuum, High Temperature Main

Steamline Tunnel, Recirculation Flow,

and Reactor Coolant Leakage.

Additionally, the change will extend the

equipment tests/operability checks for

Containment Vent and Purge Isolation,

Electromagnetic Relief Valve

Operability, and Drywell to Torus

Leakage Test. The above change

extensions conform with the 24 month

refueling interval.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Operation of the facility in accordance

with the proposed amendment would not

involve a significant increase in the

probability of occurrence or the

consequences of an accident previously

evaluated.

The proposed amendment extends the

period between successive refueling interval

surveillance(s) to once every 24 months for

those surveillance(s) evaluated herein. The

proposed surveillance interval changes do

not involve any change to the actual

surveillance requirements, nor does it

involve any change to the limits and

restrictions on plant operations. The

reliability of systems and components relied

upon to prevent or mitigate the consequences

of accidents previously evaluated is not

degraded by the proposed change to the

surveillance interval. Assurance of system

and equipment availability is maintained.

This change does not involve any change to

system or equipment configuration.

Therefore, this change does not increase the

probability of occurrence or the

consequences of an accident previously

evaluated.

Operation of the facility in accordance

with the proposed amendment would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed amendment extends the

period between successive refueling interval

surveillance(s) to once every 24 months for

those surveillance(s) evaluated herein. The

proposed surveillance interval changes do

not involve any change to the actual

surveillance requirements, nor does it

involve any change to the limits and

restrictions on plant operation. This change

does not involve any change to system or

equipment configuration. Therefore, this

change is unrelated to the possibility of

creating a new or different kind of accident

from any previously evaluated.

Operation of the facility in accordance

with the proposed amendment would not

involve a significant reduction in a margin of

safety.

The proposed amendment extends the

period between successive refueling interval

surveillance(s) to once every 24 months (+/

-25% or 30 months) for the surveillances

evaluated herein. The proposed surveillance

interval changes do not involve any change

to the actual surveillance requirements, nor

does it involve any change to the limits and

restrictions on plant operation. The

reliability of systems and components is not

degraded by the proposed change to the

surveillance interval. Assurance of system

and equipment availability is maintained.

Therefore, it is concluded that operation of

the facility in accordance with the proposed

amendment does not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Ocean County Library,

Reference Department, 101 Washington

Street, Toms River, NJ 08753

Attorney for licensee: Ernest L. Blake,

Jr., Esquire. Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: John F. Stolz

GPU Nuclear Corporation, et al.,

Docket No. 50-219, Oyster Creek

Nuclear Generating Station, Ocean

County, New Jersey

Date of application for amendment:

October 10, 1996 (TSCR No. 243)

Description of amendment: The

proposed Technical Specification (TS)

will change the trip setting for TS Table

3.1.1 Item G.3, Automatic

Depressurization System (ADS) by

clarification of the functional

requirement to provide an interlock

permissive which ensures that a source

of cooling water is available via the Core

Spray System prior to depressurization.

This will be accomplished by replacing

the present interlock description ‘‘AC

Voltage’’ with core spray booster pump

differential pressure, as the permissive

required for initiation of ADS. A

corresponding surveillance requirement

is being added to TS Table 4.1.1 which

reflects the need to test and calibrate the

core spray booster pump differential

pressure switches pursuant to existing
57486 Federal Register / Vol. 61, No. 216 / Wednesday, November 6, 1996 / Notices

plant procedures. Additionally, allowed

outage time (AOT) is addressed in the

footnote ‘‘i’’ for the differential pressure

switches based upon the currently

designed ADS logic trains and footnote

‘‘h’’ to parallel the ‘‘Low-Low Reactor

Water Level’’ and ‘‘High Drywell

Pressure’’ AOTs associated with

Standard 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:

Operation of the facility in accordance

with the proposed amendment would not

involve a significant increase in the

probability of occurrence or the

consequences of an accident previously

evaluated.

The implementation of this TSCR does not

involve an increase in the probability of

occurrence or the consequences of an

accident previously evaluated, as no plant

modifications are proposed by the change

request, and no changes in instrument set or

reset setpoints are required in order to

implement the change. This change serves to

clarify and to incorporate the ‘‘as-built’’ ADS

system logic parameter (core spray booster

pump differential pressure) as the functional

permissive required for initiation of ADS.

This ‘‘interlock’’ permissive compares closely

with that of the BWR [boiling-water reactor]

STS [Standard Technical Specifications]

requirement to monitor core spray discharge

pressure for initiation of ADS. In addition,

the AOTs for the ADS initiation signals are

being revised to align with the AOTs

provided for such signals in the STS. The

performance and function of the Automatic

Depressurization System is unchanged by

this request. However, by implementation of

the change the specific functions of the ADS

as-built d/p permissives would then be

clearly identified in and controlled by T.S.

Table 3.1.1, ‘‘Protective Instrumentation

Requirements,’’ including the associated

surveillance requirements as shown on the

revised T.S. Table 4.1.1.

Operation of the facility in accordance

with the proposed amendment would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The implementation of this TSCR does not

impact upon the operation of the facility, and

would not create the possibility of a new or

different kind of accident from any

previously evaluated because no plant

modifications are proposed by this change

request, and no changes in instrument set or

reset setpoints are required in order to

implement the change. This change clarifies

the technical specifications by incorporating

the ‘‘as-built’’ ADS system logic parameter

(core spray booster pump differential

pressure) as the functional permissive

required for initiation of ADS. This

‘‘interlock’’ permissive compares closely

with that of the BWR STS requirement to

monitor core spay discharge pressure. The

revised AOTs for ADS initiation signals are

also being changed to conform with those

allowed by and provided in the STS. The

performance and function of the Automatic

Depressurization System (ADS) is unchanged

by this request.

OC plant surveillance procedures for both

ADS and the Core Spray system presently

incorporate the calibration requirements and

both the set and reset setpoints calculated for

the core spray booster pump d/p switch

permissive to the ADS initiation logic.

Hence, a new or different kind of accident

from any previously evaluated is not created.

Operation of the facility in accordance

with the proposed amendment would not

involve a significant reduction in a margin of

safety.

The implementation of this TSCR does not

involve a reduction in the margin of safety

for operation of the ADS or the Core Spray

system. The Technical Specification Bases

which presently define the margin of safety

are not impacted as the core spray booster

pump d/p ‘‘interlock’’ permissives are not

described in the specifications for ‘‘Protective

Instrumentation Requirements’’ or its

surveillance requirements. In addition, the

margin of safety for ADS initiation is not

reduced by this TSCR because the required

system response is not affected by the

proposed changes as no plant modifications

are required which could create a potential

impact upon the margins of safety previously

established.

The revision of AOTs associated with ADS

actuation signals by extension form 72 hours

to 4 days is consistent with that presently

provided in the STS. This does not decrease

the margin of safety associated with

availability of ADS as placement of the

initiation signals into the ‘‘tripped

condition’’ maintains the operability of the

ADS trip systems while in the automatic

mode. Additionally, the Bases for STS

Specification 3.1 provides justifications for

AOTs using the GE [General Electric]

reliability analyses referenced therein and

therefore 4 days is both justified and

conservative. The margin of safety with

respect to the instrument channels ability to

perform its intended actuation function is not

impacted; therefore, there is no reduction in

the margin of safety.

Lastly, the surveillance frequency for the

new surveillance interval created on Table

4.1.1 for the d/p [s]witches is consistent with

that established in Reference 2 of the Bases

for Technical Specification 4.1. Therefore,

there is no reduction in the margin of safety

as a result of this change request.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Ocean County Library,

Reference Department, 101 Washington

Street, Toms River, NJ 08753

Attorney for licensee: Ernest L. Blake,

Jr., Esquire. Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: John F. Stolz

Illinois Power Company and Soyland

Power Cooperative, Inc., Docket No. 50-

461, Clinton Power Station, Unit No. 1,

DeWitt County, Illinois

Date of amendment request: October

17, 1996

Description of amendment request:

The proposed amendment would revise

Facility Operating License NPF-62 to

acknowledge the transfer of Soyland

Power Cooperative’s 13.21% minority

ownership interest in the Clinton Power

Station to Illinova Power Marketing,

Inc., the unregulated power marketing

affiliate of Illinois Power, and a wholly

owned subsidiary of Illinova

Corporation.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration which is presented below:

1. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated because it merely

revises the Operating License to indicate the

transfer of a minority ownership interest to

the corporate parent of the majority owner

and licensee. This proposed amendment

represents an administrative rather than

operational change and, therefore, has no

impact on accidents previously evaluated.

2. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated because Illinois Power

will continue to be the operator of Clinton

Power Station, and further, there will be no

change to the plant’s physical configuration

or operating philosophy as a result of this

proposed amendment.

3. The proposed amendment does not

involve a significant reduction in the margin

of safety because it is only an administrative

change and will have no impact on any

margin of safety related to the design or

operation of the facility.

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: Vespasian Warner Public

Library, 120 West Johnson Street,

Clinton, Illinois 61727

Attorney for licensee: Leah Manning

Stetzner, Vice President, General

Counsel, and Corporate Secretary, 500

South 27th Street, Decatur, Illinois

62525

NRC Project Director: Gail H. Marcus
Federal Register / Vol. 61, No. 216 / Wednesday, November 6, 1996 / Notices 57487

Maine Yankee Atomic Power Company,

Docket No. 50-309, Maine Yankee

Atomic Power Station, Lincoln County,

Maine

Date of amendment request:

September 13, 1996

Description of amendment request:

The proposed amendment would revise

the Maine Yankee containment testing

technical specification (TS 4.4) to

implement 10 CFR Part 50, Appendix J,

Option B, by referring to Regulatory

Guide 1.163, ‘‘Performance-Based

Containment Leakage-Test Program’’

dated September 1995.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. This amendment request does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated, because the proposed

changes to the Technical Specification do not

affect the assumption, parameters or results

of any FSAR accident analysis.[...] These

changes potentially result in a minor increase

in the consequences of an accident

previously evaluated due to the increased

testing intervals. However, the proposed

changes do not result in an increase in the

probability of an accident previously

identified since the containment system is

used for mitigation purposes only. The

changes are also expected to result in

increased attention to components with poor

leakage test history as part of the

performance-based nature of Option B such

that the marginally increased consequences

from the expanded testing intervals may be

further reduced or negated. The addition of

the ’’...[as modified by approved]

exemptions’’ phrase is an administrative

change. Any specific exemptions from the

requirements of Appendix J will continue to

require a submittal under 10 CFR 50.12 and

subsequent review and approval by the NRC

prior to implementation. Therefore, these

changes do not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

2. Operation of Maine Yankee in

accordance with the proposed changes does

not create the possibility of a new or different

kind of accident from any accident

previously evaluated. The proposed changes

do not involve a physical alteration of the

plant (i.e., no new or different type of

equipment will be installed) nor alter the

function of the containment system. The

changes only provide for additional time

between leakage tests and an increase in the

test pressure value equal to the containment

design pressure which bounds the

containment peak accident pressure. Thus,

these changes do not create the possibility of

a new or different kind of accident from any

previously evaluated.

3. Operation of Maine Yankee in

accordance with the proposed changes does

not involve a significant reduction in a

margin of safety. The proposed changes do

not alter the manner in which safety limits,

limiting safety system setpoints, or limiting

conditions for operation are determined. The

changes are expected to result in an

increased focus on components

demonstrating poor leakage test history

without excessive testing of components

which continue to demonstrate good test

history. Therefore, these changes do not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration. Local

Public Document Room location:

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 04011NRC Deputy

Director: John A. Zwolinski

Maine Yankee Atomic Power Company,

Docket No. 50-309, Maine

YankeeAtomic Power Station, Lincoln

County, Maine

Date of amendment request:

September 13, 1996, as supplemented

September 25, 1996

Description of amendment request:

The proposed amendment would revise

TS 5.5.B to eliminate references to the

Vice President (YNSD) and designate

the President, Maine Yankee, as the

responsible official for matters related to

the composition, review and audit

responsibilities, authority and

recordkeeping responsibilities of the

Nuclear Safety Audit and Review

(NSAR) Committee. Minor editorial

changes are also proposed.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration which is presented below.

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed change is administrative in

nature and will not have a direct effect on the

physical plant or the maintenance of the

physical plant. The audit and review

functions of the NSAR Committee will

continue to be required. The proposed

changes will not, of themselves, decrease the

effectiveness of these functions. This

authority and responsibility realignment will

continue to assure that NSAR Committee has

direct access to a level of management

necessary to perform their audit and review

functions.

Since, the proposed change will not

adversely effect the audit and review

functions of the NSARC and since the

proposed change will not have a direct effect

on the physical plant or maintenance of the

physical plant, the proposed change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed change is administrative in

nature and does not introduce any new

structures, systems, or components into the

plant design. This change continues to

ensure that the NSAR Committee reports to

a management level such that there is

sufficient authority and organizational

freedom to execute their audit and review

functions. Consequently, an unbiased

oversight of the programs and procedures is

not compromised by this proposed change.

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 the margin of

safety.

The proposed change realigns the authority

and responsibility relationship of the NSAR

Committee. The NSAR Committee will

continue to maintain effective oversight of

programs and procedures. The proposed

change will continue to ensure that the

NSAR Committee is sufficiently independent

from cost and schedule when opposed to

safety considerations. Therefore, the

proposed change does not involve a

significant reduction in the 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 04011NRC Deputy

Director: John A. Zwolinski

Northeast Nuclear Energy Company

(NNECO), Docket No. 50-245, Millstone

Nuclear Power Station, Unit 1, New

London County, Connecticut

Date of amendment request: July 2,

1996

Description of amendment request:

The proposed amendment incorporates

limiting conditions for operation and

surveillance requirements for the safety/

relief valve (SRV) electrical lift design

modification. The proposed amendment

also makes clarification and editorial

changes, as well as revising the

associated Bases section.
57488 Federal Register / Vol. 61, No. 216 / Wednesday, November 6, 1996 / Notices

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:

Pursuant to 10CFR50.92, NNECO has

reviewed the proposed change and concludes

that the change does not involve a significant

hazards consideration (SHC) since the

proposed change satisfies the criteria in 10

CFR 50.92(c). That is, the proposed change

does not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The safety relief valves are considered for

two analyzed accidents, an overpressure

transient (such as MSIV [main steam

isolation valve] closure with flux scram) and

an inadvertent SRV opening.

The new technical specifications do not

affect normal operation, therefore, they

cannot increase the probability of an

overpressure event. Since the mechanical

function will not be affected by the new

equipment, the new LCOs [limiting

conditions for operation], or the new

surveillance requirements, there is no

adverse affect on the consequences of an

overpressure event. The SRVs will be

expected to lift mechanically. If they do not

open at the design setpoints, the electrical

actuation, which has the same setpoints, will

cause the valves to open less than 400

milliseconds later.

Sufficient redundancy and diversity is

established for the electrical lift by the use

of two sensors in a two-out-of-two-taken-once

configuration. Therefore, the failure of any

single component cannot result in an

inadvertent opening of an SRV. The only

proposed surveillance performed while at

power is the daily instrument check. This

surveillance does not require the

manipulation of any controls and, as such,

cannot affect the probability of an accident.

Therefore, based on the above, the

proposed change to the Technical

Specifications does not involve a significant

increase in the probability or consequences

of any previously evaluated accident.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

None of the proposed new LCOs or

surveillance requirements has a potential for

creating a new or different kind of accident.

Expanding the LCO and surveillance

requirements to address both the mechanical

actuation and the pressure sensor lift does

not change the type of action that these

valves are expected to perform, nor does it

change the initial ‘‘as-left’’ requirements for

the valves. Plant operating parameters have

also not changed.

Therefore, this change will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. Involve a significant reduction in a

margin of safety.

The margin of safety previously analyzed

for the SRVs was based on the current

nominal setpoints and allowable percent

drift. The electrical lift system improves the

confidence that the SRVs will lift within the

specified range. The setpoint uncertainty of

the electrical lift system is similar to the drift

allowed for the mechanical lift in the

Technical Specifications. All existing

functions that may actuate the SRVs (safety,

manual, or automatic lift) remain unaffected.

The design of the pressure transmitters,

combined with the logic configuration,

minimizes the possibility of inadvertently

opening the SRVs.

Therefore, this change has no impact 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: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360, and the Waterford

Library, ATTN: Vince Juliano, 49 Rope

Ferry Road, Waterford, CT 06385

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

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

NRC Project Director: Phillip F.

McKee

Philadelphia Electric Company, Docket

Nos. 50-352 and 50-353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of amendment request: March

29, 1996

Description of amendment request:

The proposed Technical Specifications

(TS) changes would revise TS

Surveillance Requirement (SR)

4.5.1.d.2.b to delete the requirement to

perform in-situ functional testing of the

Automatic Depressurization System

(ADS) valves once every 24-months as

part of start-up testing activities.

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 proposed TS change does not involve

any physical changes to plant structures,

systems, or components (SSC). The ADS will

continue to function as designed. The ADS

is an Emergency Core Cooling System (ECCS)

designed to mitigate the consequences of an

accident, and therefore, can not contribute to

the initiation of any accident. The ADS

utilizes five (5) of the 14 main steam line

SRVs as the primary method for

depressurizing the reactor pressure vessel to

permit low pressure core cooling capability

in the event of a small break Loss-of-CoolantAccident (LOCA) if the high pressure cooling

systems (i.e., High Pressure Coolant Injection

(HPCI) and Reactor Core Isolation Cooling

(RCIC) systems) fail to maintain adequate

reactor vessel water level.

Deleting the TS SR to perform the in-situ

testing of the ADS/SRVs during start-up, as

proposed, should reduce the probability of an

inadvertent opening of an SRV as discussed

in Section 15.1.4 of the LGS Updated Final

Safety Analysis Report (UFSAR) since

deleting this testing requirement will

eliminate a known initiator of SRV pilot

leakage and subsequent erosion. This

proposed TS change will have a tendency to

increase, rather than decrease, the reliability

of the ADS/SRVs by eliminating the in-situ

ADS functional start-up testing. The

probability of the ADS/SRVs to open on

demand has been demonstrated to be

extremely high and is not measurably

improved through the in-situ ADS functional

start-up testing.

This proposed TS change will not increase

the probability of occurrence of a

malfunction of any plant equipment

important to safety. Alternate testing

methods at LGS, Units 1 and 2, and at the

off-site test facility, adequately demonstrate

proper ADS valve operation and assure that

the valves will continue to function as

designed. Existing surveillance testing and

inspections of the ADS/SRVs at LGS verify

that the ADS initiation logic, solenoid valve

operation, pneumatic gas supply integrity

and air operator assembly (including pilot

rod) will operate as designed. Offsite testing

verifies pilot disc operation, setpoint

calibration and main valve disc operation.

Deleting the in-situ testing requirement, as

proposed, will reduce the probability of

inflating SRV leakage which should reduce

the probability of an inadvertent SRV

opening. It has been documented throughout

the BWR industry that pilot disc leakage

leads to pilot disc and rod erosion, which can

ultimately result in an inadvertent opening of

an SRV. Therefore, any SRV pilot leakage

that can be eliminated would reduce the

probability of occurrence of a malfunction of

that SRV.

Deleting the ADS/SRV in-situ functional

test will in no way increase any

consequences of a malfunction of plant

equipment important to safety. The

consequences of a malfunction of an ADS/

SRV as discussed in the LGS UFSAR remain

unchanged.

In addition, eliminating a known initiator

of SRV leakage, as proposed in this TS

change, would help to reduce operator

workarounds in the form of suppression pool

cooling and letdown operation activities. As

a result, this will reduce the unnecessary

operation of the Residual Heat Removal

(RHR) and Residual Heat Removal Service

Water (RHRSW) systems.

Therefore, the proposed TS change does

not involve an increase in the probability or

consequences of an accident previously

evaluated.
Federal Register / Vol. 61, No. 216 / Wednesday, November 6, 1996 / Notices 57489

2. The proposed TS change does not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

This proposed TS change does not involve

any physical changes to plant SSC. The

design and operation of the ADS/SRVs is not

changed from that currently described in the

Safety Analysis Report (SAR). The ADS will

continue to function as designed to mitigate

the consequences of an accident. No changes

of any kind are being made to the valves,

auxiliary components, or ADS logic. Deleting

the requirement to perform the ADS in-situ

functional test during plant start-up as

proposed in this TS Change Request reduces

the likelihood of a SRV developing a leak and

degrading throughout the subsequent

operating cycle. There is no possibility that

implementing this proposed TS change

would create a different type of malfunction

to the ADS/SRVs than any previously

evaluated.

Eliminating the requirement to perform the

in-situ testing of the ADS/SRVs during startup activities, does not create a new or

different type of accident than any previously

evaluated. There is no accident scenario

associated with testing the ADS/SRVs other

than the inadvertent opening of a relief valve

which is currently discussed in Section

15.1.4 of the LGS UFSAR. This proposed TS

change does not alter the conclusions

described in the UFSAR regarding an

inadvertent opening of an SRV. No new or

different type of accident will be created as

a result of this proposed TS change.

Therefore, the proposed TS change does

not create the possibility of a new or different

kind of accident from any previously

evaluated.

3. The proposed TS change does not

involve a significant reduction in a margin of

safety.

The proposed TS change does not involve

any physical changes to plant SSC. The

design and functional requirements of the

ADS will not change. The ADS will still

function as designed to mitigate the

consequences of an accident.

This proposed TS change involves deleting

the requirement to perform in-situ functional

testing of the ADS/SRVs during start-up

activities. This testing imposes an

unnecessary challenge on the ADS/SRVs and

has been linked to SRV degradation (e.g.,

pilot valve and/or main valve leakage). This

proposed TS change should reduce SRV

leakage and improve ADS/SRV reliability by

reducing the potential for spurious SRV

actuation. The LGS TS Bases do not identify

specific testing requirements for ADS. ADS

operability can be readily demonstrated with

extremely high confidence by the existing

additional surveillance tests and inspections

performed for the ADS. There will be no

reduction in any margin of safety resulting

from this proposed TS change.

Therefore, the proposed TS change does

not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464

Attorney for licensee: J. W. Durham,

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

Counsel, Philadelphia Electric

Company, 2301 Market Street,

Philadelphia, PA 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: August 8,

1996

Description of amendment request:

The proposed Technical Specifications

(TS) changes would revise TS Sections

3/4.3.1, ‘‘Reactor Protection System

Instrumentation,’’ 3/4.3.2, ‘‘Isolation

Actuation Instrumentation,’’ 3/4.3.3,

‘‘Emergency Core Cooling System

Actuation Instrumentation,’’ and the

associated TS Bases Sections 3/4.3.1

and 3/4.3.2 to eliminate selected

response time testing requirements.

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) changes do not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

The proposed TS changes do not make any

physical alterations or modifications to the

plant systems or equipment. The proposed

changes do not affect the capability of the

associated systems to perform their intended

functions within their required response

times, nor do the proposed changes adversely

impact the operation of any plant equipment.

The affected plant systems will continue to

function as designed. Elimination of the

response time testing requirements as

proposed by this TS change for selected

components in RPS Instrumentation,

Isolation Actuation System Instrumentation,

and ECCS Actuation Instrumentation will not

adversely affect the operation of these

components.

The supporting analysis provided in

NEDO-32291, demonstrates that response

time testing is redundant to other TS

required testing. NEDO-32291 demonstrated

that these other required tests (i.e., channel

checks, channel calibrations, channel

functional tests, and logic system functional

tests), in conjunction with actions taken in

response to NRC Bulletin 90-01 and NRCB

90-01, Supplement 1, are sufficient to

identify failure modes or degradation in

instrument response times, and ensure

operation of the associated systems within

acceptable limits. There are no known failure

modes that can be detected by response time

testing that cannot also be detected by other

TS required testing. The continued

application of other existing TS required

testing such as channel checks, channel

calibrations, channel functional tests, and

logic system functional tests, ensures that the

response times for these systems will be

maintained within the acceptance limits. The

capability of these systems to perform their

intended functions within their required

response times is not adversely impacted by

this proposed TS change. NEDO-32291

evaluated the potential failure modes of the

affected instrumentation loops which could

impact the instrument loop response times.

Industry operating experience was also

reviewed to identify failures that affect

response times and how they are detected.

The failure modes identified were evaluated

to determine if other TS required

surveillances and actions taken in response

to NRC Bulletin 90-01, and NRCB 90-01,

Supplement 1, would detect any effects on

response time. There are no failures [sic]

[failure] modes identified that can be

detected by response time testing that cannot

also be detected by other TS required testing.

PECO Energy has confirmed the

applicability of the generic evaluation

provided in NEDO-32291 to LGS, Units 1 and

2. By letter dated December 28, 1994, the

NRC concluded that response time testing

can be eliminated from the TS for the

selected instrumentation identified in NEDO32291, with certain provisions, and that

NEDO-32291 can be referenced in license

amendment requests.

Therefore, the proposed TS changes do not

involve an increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed TS changes do not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed TS changes do not involve

any physical changes to plant systems or

equipment. The proposed changes apply only

to the testing requirements for the selected

components involved and do not result in

any physical modifications to these

components, or to other plant system

components. Elimination of the response

time testing requirements as proposed by this

TS change for selected components in RPS

Instrumentation, Isolation Actuation System

Instrumentation, and ECCS Actuation

Instrumentation will not adversely affect the

operation of these components. These

components will continue to function as

designed. Consequently, no new failure

modes are introduced as a result of the

proposed TS changes.

Eliminating the response time testing

requirements as proposed, does not create a

new or different type of accident than any

previously evaluated. No new or different

type of accident will be created as a result

of this proposed TS change.

NEDO-32291 demonstrates that other

required tests (i.e., channel checks, channel

calibrations, channel functional tests, and

logic system functional tests), in conjunction

with actions taken in response to NRC

Bulletin 90-01 and NRCB 90-01, Supplement
57490 Federal Register / Vol. 61, No. 216 / Wednesday, November 6, 1996 / Notices

1, are sufficient to identify failure modes or

degradation in instrument response times,

and ensure operation of the associated

systems within acceptable limits. There are

no known failure modes that can be detected

by response time testing that cannot also be

detected by other TS required testing, and

therefore, response time testing for the

selected components is redundant to the

other TS required testing.

Therefore, the proposed TS changes do not

create the possibility of a new or different

kind of accident from any previously

evaluated.

3. The proposed TS changes do not involve

a significant reduction in a margin of safety.

The proposed TS changes do not involve

any physical changes to plant systems or

equipment. The proposed TS changes do not

affect the capability of the associated systems

or equipment from performing their intended

functions. The systems involved will

continue to respond within their allowed

response times. Elimination of the response

time testing requirements are based on the

evaluation provided in NEDO-32291 which

demonstrates that response time degradation

can be detected by other TS required testing.

The evaluation concluded that other TS

required tests (i.e., channel checks, channel

calibrations, channel functional tests, and

logic system functional tests), in conjunction

with actions taken in response to NRC

Bulletin 90-01 and NRCB 90-01, Supplement

1, are sufficient to identify failure modes or

degradation in instrument response times,

and ensure operation of the associated

systems within acceptable limits.

In addition, although not specifically

evaluated, the proposed TS changes will

provide an improvement to plant safety and

operation by reducing the time safety systems

are unavailable, reducing the potential for

safety system actuations, reducing plant

operating and shutdown risk, limiting

radiation exposure to plant personnel, and

eliminating the diversion of key personnel to

conduct unnecessary testing. Therefore,

PECO Energy considers that the proposed TS

changes will result in an overall increase in

the margin of safety and that the changes do

not constitute an unreviewed safety question.

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

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, PA 19464

Attorney for licensee: J. W. Durham,

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

Counsel, Philadelphia Electric

Company, 2301 Market Street,

Philadelphia, PA 19101

NRC Project Director: John F. Stolz

Philadelphia Electric Company, Docket

No. 50-353, Limerick Generating

Station, Unit 2, Montgomery County,

Pennsylvania

Date of amendment request: August 1,

1996

Description of amendment request:

The proposed Technical Specifications

(TS) changes would revise TS Section 3/

4.4.6 (i.e., Figure 3.4.6.1-1) to reflect the

addition of two hydrotest curves,

effective for 6.5 and 8.5 Effective Full

Power Years (EFPY), to the existing

Pressure-Temperature Operating Limit

(PTOL) curves for LGS Unit 2.

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 proposed Technical Specification (TS)

change includes Pressure-Temperature

Operating Limit (PTOL) curves which were

conservatively generated in accordance with

the fracture toughness requirements of

10CFR50, Appendix G. The Adjusted

Reference Temperatures to the initial nil

ductility reference temperatures (RTNDT)

used to evaluate the pressure/temperature

limits for the beltline materials were based

on Regulatory Guide 1.99, Revision 2. Future

analyses of the Reactor Pressure Vessel (RPV)

surveillance capsule contents and future

revisions to the PTOL curve as required,

ensure that the reactor pressure boundary

will behave in a non-brittle manner during

plant testing, startup, and operation

throughout the life of the plant. The current

schedule for removal of the surveillance

specimens from Limerick Generating Station

(LGS) Unit 2 RPV is during 2R05. The

proposed change does not impact the existing

PTOL curves for 10 Effective Full Power

Years (EFPY), currently shown in the LGS

Unit 2 TS. The proposed change only

provides additional information (i.e., two

new curves) related to the RPV condition

following 6.5 and 8.5 EFPY, in order to

facilitate hydrostatic testing performed after

2R04 and 2R05, respectively. The added

PTOL curves are established in compliance

with the methodology used to calculate the

predicted irradiation effects on vessel

beltline materials as documented in the LGS

Updated Final Safety Analysis Report

(UFSAR). There are no physical changes to

the plant being introduced by the added

PTOL curves.

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 proposed Technical Specification (TS)

change includes Pressure-Temperature

Operating Limit (PTOL) curves which were

conservatively generated in accordance with

the fracture toughness requirements of

10CFR50, Appendix G. The Adjusted

Reference Temperatures to the initial nil

ductility reference temperatures (RTNDT)

used to evaluate the pressure/temperature

limits for the beltline materials were based

on Regulatory Guide 1.99, Revision 2. The

proposed changes do not impact the existing

PTOL curves for 10 Effective Full Power

Years (EFPY), currently shown in the TS.

They only provide additional information

(i.e., two new curves) related to the reactor

pressure vessel condition for 6.5 and 8.5

EFPY, in order to facilitate hydrostatic testing

performed after 2R04 and 2R05, respectively.

The added PTOL curves are established in

compliance with the previous methodology

used to calculate the predicted irradiation

effects on vessel beltline materials as

documented in the LGS [Updated Final

Safety Analysis Report] UFSAR. The

proposed TS change does not involve any

physical changes to safety-related equipment.

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 proposed change to Technical

Specifications (TS) does not reduce the

margin of safety as defined in the Bases for

any TS. The added Pressure-Temperature

Operating Limit (PTOL) curves for 6.5 and

8.5 Effective Full Power Years (EFPY)

corresponding to 2R04 and 2R05,

respectively, have been calculated in

accordance with the existing methodology

used to calculate the PTOL curves currently

existing in the LGS Unit 2 TS (i.e., complying

with the requirements of 10CFR50 Appendix

G, and Regulatory Guide 1.99, Revision 2)

and will more closely reflect the actual

required reactor pressure vessel condition at

the time in which the hydrotest is performed.

Therefore, the margin of safety is not

affected.

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, PA 19464

Attorney for licensee: J. W. Durham,

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

Counsel, Philadelphia Electric

Company, 2301 Market Street,

Philadelphia, PA 19101

NRC Project Director: John F. Stolz
Federal Register / Vol. 61, No. 216 / Wednesday, November 6, 1996 / Notices 57491

Philadelphia Electric Company, Docket

No. 50-353, Limerick Generating

Station, Unit 2, Montgomery County,

Pennsylvania

Date of amendment request: August 5,

1996

Description of amendment request:

The proposed Technical Specifications

(TS) changes would revise TS Section

2.1 and its associated TS Basis to reflect

the change in the Minimum Critical

Power Ratio (MCPR) Safety Limit due to

the plant specific evaluation performed

by General Electric Co. (GE), for LGS

Unit 2 Cycle 4.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed Technical Specifications

(TS) change does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

The revised Minimum Critical Power Ratio

(MCPR) Safety Limit for LGS Unit 2

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

transition boiling 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.

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 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 NRCapproved methods and is based on LGS Unit

2 Cycle 4 specific inputs.

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 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 remains high

enough to ensure that greater than 99.9% of

all fuel rods in the core will avoid transition

boiling 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, PA 19464

Attorney for licensee: J. W. Durham,

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

Counsel, Philadelphia Electric

Company, 2301 Market Street,

Philadelphia, PA 19101

NRC Project Director: John F. Stolz

Southern Nuclear Operating Company,

Inc., Docket Nos. 50-348 and 50-364,

Joseph M. Farley Nuclear Plant, Units

1 and 2, Houston County, Alabama

Date of amendments request:

September 30, 1996

Description of amendments request:

The proposed amendments would

revise Technical Specifications (TSs) 3/

4.1.1, 3/4.1.3, 3.1.3.6, 3.2.1, 3/4.2.2, and

3.2.3 and associated Bases to remove

certain cycle-specific parameter limits

from the TSs and relocate them to the

Core Operating Limits Report (COLR).

These changes result from NRC Generic

Letter (GL) 88-16, dated October 4, 1988,

which provided guidance to licensees

on requests for removal of the values of

cycle-specific parameter limits from the

TSs. The licensee’s proposed

amendments are consistent with the GL.

The COLR has been included in the

Definitions section of the TSs. The

definition notes that it is the unitspecific document that provides these

limits for the current operating reload

cycle. The values of these cycle-specific

parameter limits are to be determined in

accordance with TS 6.9.1.11. This TS

requires that the core operating limits be

determined for each reload cycle in

accordance with the referenced NRCapproved methodology for these limits

and consistent with the applicable

limits of the safety analysis. The COLR

shall be provided to the NRC upon

issuance.In addition, the above TS

changes would produce administrative

changes to the TS Table of Contents.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The removal of cycle-specific core

operating limits from the FNP [Farley

Nuclear Plant] Technical Specifications has

no influence or impact on the probability or

consequences of a Design Basis Accident

(DBA) occurrence. The cycle-specific core

operating limits, although not in Technical

Specifications, will be followed in the

operation of FNP. The proposed amendment

retains the same required actions to be taken

when or if limits are exceeded as stipulated

by current Technical Specifications. In

addition, the associated surveillance

requirements are not altered by the proposed

changes.

Each accident analysis addressed in the

FNP FSAR [Final Safety Analysis Report]

will be examined with respect to changes in

cycle-dependent parameters, which are

obtained from application of the NRCapproved reload design methodologies, to

ensure that the transient evaluation of new

reloads are bounded by previously accepted

analyses. This examination, which will be

performed per requirements of 10 CFR 50.59,

ensures that future reloads will not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

As stated earlier, the removal of the cyclespecific variables has no influence or impact,

nor does it contribute in any way to the

probability or consequences of an accident.

No safety-related equipment, safety function,

or plant operation will be altered as a result

of this proposed change. The cycle-specific

variables are calculated using the NRCapproved methods and submitted to the NRC

to allow the Staff to continue to trend the

values of these limits. The Technical

Specifications will continue to require

operation within the required core operating

limits and appropriate actions will be taken

when or if limits are exceeded. Therefore, the

proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The proposed changes do not result in

a significant reduction in the margin of

safety.

The margin of safety is not affected by the

removal of cycle-specific core operating

limits from the Technical Specifications. The

margin of safety presently provided by

current Technical Specifications remains

unchanged. Appropriate measures exist to

control the values of these cycle-specific

limits. The proposed amendment continues

to require operation within the core limits, as

obtained from the NRC-approved reload

design methodologies. The required actions

to be taken or if limits are violated remain

unchanged.

The development of the limits for future

reloads will continue to conform to those
57492 Federal Register / Vol. 61, No. 216 / Wednesday, November 6, 1996 / Notices

methods described in NRC-approved

documentation. In addition, each future

reload involves a 10 CFR 50.59 safety review

to assure that operation of FNP within the

cycle-specific limits will not involve a

significant reduction in [the] margin of

safety. Therefore, the proposed changes are

administrative in nature and do not impact

the operation of FNP in a manner that

involves a reduction to the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Houston-Love Memorial

Library, 212 W. Burdeshaw Street, Post

Office Box 1369, Dothan, Alabama

36302

Attorney for licensee: M. Stanford

Blanton, Esq., Balch and Bingham, Post

Office Box 306, 1710 Sixth Avenue

North, Birmingham, Alabama 35201

NRC Project Director: Herbert N.

Berkow

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.

Northeast Nuclear Energy Company,

Docket No. 50-245, Millstone

NuclearPower Station, Unit No. 1, New

London County, Connecticut

Date of amendment request: August

29, 1996

Description of amendment request:

The proposed amendment would

modify the applicability requirements

for certain radiation monitors so that the

radiation monitors are required to be

operable only when secondary

containment integrity is required to be

operable; delineate when secondary

containment integrity is required;

modify standby gas treatment

operability requirements; make editorial

corrections to clarify the configuration

of the radiation monitors; and revise the

associated Bases section.

Date of publication of individual

notice in Federal Register: October 17,

1996 (61 FR 54242)

Expiration date of individual notice:

November 18, 1996

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, Connecticut, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut

Wisconsin Public Service Corporation,

Docket No. 50-305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of application for amendment:

September 6, 1996

Brief description of amendment

request: The proposed amendment

would change Technical Specification

(TS) requirements related to steam

generator tubes to allow a laser-welded

repair of Westinghouse hybrid

expansion joint (HEJ) sleeved steam

generator tubes. Date of individual

notice in Federal Register: October 15,

1996 (61 FR 53769)

Expiration date of individual notice:

November 14, 1996

Local Public Document Room

location: University of Wisconsin,

Cofrin Library, 2420 Nicolet Drive,

Green Bay, Wisconsin

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:

June 17, 1996

Brief description of amendments: The

amendment modifies the technical

specifications (TS) to change (1) the

reference method for calculating dose

conversion factors (DCFs) to be used in

dose calculations, and (2) the upper and

lower limits for operating pressurizer

pressure to account for new instrument

uncertainties and to reduce the allowed

operating band.

Date of issuance: October 23, 1996

Effective date: October 23, 1996, to be

implemented within 45 days of issuance

Amendment Nos.: Unit 1 - 109; Unit

2 - 101; Unit 3 - 81

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

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: September 11, 1996 (61 FR

47963). The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

October 23, 1996.No significant hazards

consideration comments received: No.

Local Public Document Room

location: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona

85004

Baltimore Gas and Electric Company,

Docket Nos. 50-317 and 50-318, Calvert

Cliffs Nuclear Power Plant, Unit Nos. 1

and 2, Calvert County, Maryland

Date of application for amendments:

June 9, 1995
Federal Register / Vol. 61, No. 216 / Wednesday, November 6, 1996 / Notices 57493

Brief description of amendments: The

amendments implement changes to

radiological effluent Technical

Specifications in accordance with

Generic Letter 89-01 ‘‘Implementation of

Programmatic for Radiological Effluent

Technical Specification in the

Administrative Controls Section of the

Technical Specifications and Relocation

of Procedural Details of RETS to the

Offsite Dose Calculation Manual or to

the Process Control Program.’’

Date of issuance: October 18, 1996

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment Nos.: 217 and 194

Facility Operating License Nos. DPR53 and DPR-69: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: July 15, 1995 (60 FR 35062)

The Commission’s related evaluation of

these amendments is contained in a

Safety Evaluation dated October 18,

1996.No significant hazards

consideration comments received: No

Local Public Document Room

location: Calvert County Library, Prince

Frederick, Maryland 20678.

Boston Edison Company, Docket No.

50-293, Pilgrim Nuclear Power Station,

Plymouth County, Massachusetts

Date of application for amendment:

May 1, 1996

Brief description of amendment: The

proposed amendment will reflect the

implementation of 10 CFR Part 50

Appendix J, Option B at the Pilgrim

Nuclear Power Station.

Date of issuance: October 4, 1996

Effective date: October 4, 1996

Amendment No.: 167

Facility Operating License No. DPR35: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 5, 1996 (61 FR 28606)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated October 4, 1996. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts

02360.

Carolina Power & Light Company, et

al., Docket No. 50-325, Brunswick

Steam Electric Plant, Unit 1, Brunswick

County, North Carolina

Date of amendment request: April 8,

1996, as supplemented on July 30, 1996,

October 4, 1996, October 8, 1996, and

October 16, 1996.

Brief description of amendment: The

amendment changes the Technical

Specifications to (1) reflect the use of a

new type of fuel (GE13) and (2) modify

the minimum critical power ratio safety

limit and the standby liquid control

system sodium pentaborate limits to

accommodate the GE13 fuel.

Date of issuance: October 17, 1996

Effective date: October 17, 1996

Amendment No.: 182

Facility Operating License No. DPR71: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: August 14, 1996 (61 FR 42276)

which superseded a Federal Register

notice published on June 5, 1996 (61 FR

28607) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

October 17, 1996.No significant hazards

consideration comments received: No.

Local Public Document Room

location: University of North Carolina at

Wilmington, William Madison Randall

Library, 601 S. College Road,

Wilmington, North Carolina 28403-

3297.

Consumers Power Company, Docket

No. 50-255, Palisades Plant, Van Buren

County, Michigan

Date of application for amendment:

January 5, 1996, as supplemented July

12, 1996

Brief description of amendment: The

amendment revises the shutdown

cooling (SDC) requirement to allow one

train of the SDC system to be rendered

inoperable for testing or maintenance

provided that a filled refueling cavity is

available to provide backup decay heat

removal capability in the event that the

operating train of SDC becomes

inoperable.

Date of issuance: October 10, 1996

Effective date: October 10, 1996

Amendment No.: 173

Facility Operating License No. DPR20. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 28, 1996 (61 FR 44348)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated October 10, 1996.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Van Wylen Library, Hope

College, Holland, Michigan 49423

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:

August 8, 1996

Brief description of amendments: The

amendments revise the Technical

Specifications, Section 6.9.1.9, to

reference updated or recently approved

topical reports used to calculate cyclespecific limits contained in the Core

Operating Limits Report.

Date of issuance: October 24, 1996

Effective date: As of the date of

issuance to be implemented within 30

days

Amendment Nos.: 154 and 146

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

the Technical Specifications.

Date of initial notice in Federal

Register: September 11, 1996 (61 FR

47977) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

October 24, 1996.No significant hazards

consideration comments received: No

Local Public Document Room

location: York County Library, 138 East

Black Street, Rock Hill, South Carolina

29730

Entergy Operations, Inc., Docket Nos.

50-313 and 50-368, Arkansas Nuclear

One, Unit Nos. 1 and 2, Pope County,

Arkansas

Date of amendment request: May 9,

1996

Brief description of amendments: The

amendments revised the name from

Arkansas Power & Light Company to

Entergy Arkansas, Inc.

Date of issuance: October 23, 1996

Effective date: October 23, 1996

Amendment Nos.: 187 and 177

Facility Operating License Nos. DPR51 and NPF-6. Amendments revised the

Technical Specifications and the

licenses.

Date of initial notice in Federal

Register: August 28, 1996 (61 FR 44357)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated October 23, 1996.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Entergy Operations, Inc., System

Energy Resources, Inc., South

Mississippi Electric Power Association,

and Entergy Mississippi, Inc., Docket

No. 50-416, Grand Gulf Nuclear

Station, Unit 1, Claiborne County,

Mississippi

Date of application for amendment:

May 8, 1996, as supplemented by letters

dated July 18 and September 19, 1996

Brief description of amendment: The

amendment modified the frequency

requirements in Surveillance

Requirement 3.6.1.3.5 of the Technical

Specifications, on the leakage rate

testing for each containment purge
57494 Federal Register / Vol. 61, No. 216 / Wednesday, November 6, 1996 / Notices

isolation valve with resilient seals, to

place these purge valves on a

performance basis in accordance with

Appendix J of 10 CFR Part 50, as

modified by any exemptions to

Appendix J. In addition, the purge

valves would be required to be leakage

rate tested every 36 months with at least

two pairs tested every 18 months and,

if any purge valve fails to meet the

leakage rate acceptance criterion, all

remaining valves must be tested within

92 days (i.e., a quarter of a year) if not

successfully tested within the previous

92 days.

Date if issuance: October 18, 1996

Effective date: October 18, 1996

Amendment No.: 128

Facility Operating License No. NPF29. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: June 5, 1996 (61 FR 28614)

The additional information contained in

the supplemental letters dated July 18

and September 19, 1996, revised the

proposed amendment in the application

of May 8, 1996; however, the revisions

were within the scope of the initial

notice and did not affect the staff’s

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated October 18,

1996.No significant hazards

consideration comments received: No

Local Public Document Room

location: Judge George W. Armstrong

Library, 220 S. Commerce Street,

Natchez, MS 39120.

Entergy Operations, Inc., System

Energy Resources, Inc., South

Mississippi Electric Power Association,

and Entergy Mississippi, Inc., Docket

No. 50-416, Grand Gulf Nuclear

Station, Unit 1, Claiborne County,

Mississippi

Date of application for amendment:

June 20, 1996, as supplemented by the

letter of September 11, 1996

Brief description of amendment: The

amendment redefined the secondary

containment boundary to allow the

enclosure building to be inoperable

during the upcoming refueling Outage 8

(RFO 8) scheduled to begin in October

1996. The amendment added a

condition to the license that the

enclosure building may be inoperable

during core alterations and movement of

non-recently irradiated fuel (i.e., fuel

that has not occupied part of a critical

reactor core for 12 days) during RFO 8

and the standby gas treatment (SGT)

system may be unable to automatically

start or achieve and maintain the

required vacuum, provided the

following conditions exist:

a. All dampers communicating

between the auxiliary building and the

enclosure building are closed.

b. The access door between the

auxiliary building and the enclosure

building is closed, except when the

access opening is being used for entry

and exit.

c. The SGT system is blocked from

automatic initiation.

d. The SGT system is available for

manual initiation or the actions for

Limiting Condition for Operation 3.6.4.3

in the Technical Specifications for

GGNS are complied with.

The non-recently irradiated fuel is

spent fuel that has decayed at least 12

days after the reactor was shut down for

refueling.

Date of issuance: October 18, 1996

Effective date: October 18, 1996

Amendment No: 129

Facility Operating License No. NPF29. Amendment adds a condition to the

license.

Date of initial notice in Federal

Register: July 17, 1996 (61 FR 37299)

The additional information contained in

the supplemental letter of September 11,

1996, was clarifying in nature and thus,

within the scope of the initial notice

and did not affect the staff’s proposed

no significant hazards consideration

determination. The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

October 18, 1996.No significant hazards

consideration comments received: No

Local Public Document Room

location: Judge George W. Armstrong

Library, 220 S. Commerce Street,

Natchez, MS 39120.

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:

February 22, 1996, and as supplemented

by letters dated July 22 and September

20, 1996

Brief description of amendment: The

amendment revises Clinton Power

Station Technical Specification 3.4.11,

‘‘Reactor Coolant System (RCS) Pressure

and Temperature (P/T) Limits,’’ to

incorporate specific P/T limits for the

bottom head region of the reactor vessel,

separate and apart from the core beltline

region of the reactor vessel.

Date of issuance: October 23, 1996

Effective date: October 23, 1996

Amendment No.: 109

Facility Operating License No. NPF62: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: April 24, 1996 (61 FR 18169)

The letters of July 22 and September 20,

1996, provided clarifying information

and did not alter the staff’s initial

finding that the proposed changes

involve no significant hazards

consideration. The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

October 23, 1996.No significant hazards

consideration comments received: No

Local Public Document Room

location: The Vespasian Warner Public

Library, 120 West Johnson Street,

Clinton, Illinois 61727

Northeast Nuclear Energy Company, et

al., Docket No. 50-336, Millstone

Nuclear Power Station, Unit No. 2, New

London County, Connecticut

Date of application for amendment:

November 21, 1995

Brief description of amendment: The

amendment changes Technical

Specification Section 5.2.2, ‘‘Design

Pressure and Temperature,’’ to clarify

that the reactor containment design

temperature is an equilibrium liner

temperature and not the air temperature.

The supporting Technical Specification

Bases is updated to reflect the change

and to include the main steam line

break accident, in addition to the lossof-coolant accident, as the limiting

events affecting the containment

temperature and pressure.

Date of issuance: October 21, 1996

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment No.: 204

Facility Operating License No. DPR65: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: December 20, 1995 (60 FR

65684) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

October 21, 1996.No significant hazards

consideration comments received: No.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360, and the Waterford

Library, ATTN: Vince Juliano, 49 Rope

Ferry Road, Waterford, CT 06385

Omaha Public Power District, Docket

No. 50-285, Fort Calhoun Station, Unit

No. 1, Washington County, Nebraska

Date of amendment requests: July 15,

1996, as supplemented by letters dated

September 3, 1996, October 22, 1996,

October 23, 1996, and August 23, 1996

Brief description of amendment: The

amendment revises Technical

Specifications (TS) Section 4.3.2 to

allow the use of zircaloy or ZIRLO fuel
Federal Register / Vol. 61, No. 216 / Wednesday, November 6, 1996 / Notices 57495

cladding and to use depleted uranium

as reactor fuel material. The amendment

also changes TS Section 5.9.5 to add

Westinghouse Topical Reports, WCAP12610-P-A, ‘‘VANTAGE + Fuel

Assembly Report,’’ and WCAP-13027-P,

‘‘Westinghouse ECCS Evaluation Model

for Analysis of CE-NSSS,’’ to the list of

approved analytical methods for

determining the core operating limits.

Date of issuance: October 25, 1996

Effective date: October 25, 1996

Amendment No.: 178

Facility Operating License No. DPR40. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 31, 1996 (61 FR 40026)

and August 30, 1996 (61 FR 45995). The

September 3, 1996, October 22, 1996,

and October 24, 1996, supplemental

letters provided additional clarifying

and correcting 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 October 25, 1996.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska

68102

Pacific Gas and Electric Company,

Docket Nos. 50-275 and 50-323, Diablo

Canyon Nuclear Power Plant, Unit Nos.

1 and 2, San Luis Obispo County,

California

Date of application for amendments:

June 7, 1996

Brief description of amendments: The

amendments revised the combined

Technical Specifications (TS) for the

Diablo Canyon Nuclear Power Plant,

Unit Nos. 1 and 2 by revising Technical

Specifications 3/4.9.14.1, ‘‘Spent Fuel

Assembly Storage - Spent Fuel Pool

Region 2,’’ and TS 3/4.9.14.3, ‘‘Spent

Fuel Assembly Storage - Spent Fuel

Pool Region 1,’’ to allow storage of fuel

assemblies in a checkerboard pattern in

Region 2 of the spent fuel pool (SFP).

Date of issuance: October 25, 1996

Effective date: October 25, 1996, to be

implemented within 30 days from date

of issuance.

Amendment Nos.: Unit 1 - 116; Unit

2 - 114

Facility Operating License Nos. DPR80 and DPR-82: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: September 25, 1996 (61 FR

50346) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

October 25, 1996.No significant hazards

consideration comments received: No.

Local Public Document Room

location: California Polytechnic State

University, Robert E. Kennedy Library,

Government Documents and Maps

Department, San Luis Obispo, California

93407

Pacific Gas and Electric Company,

Docket Nos. 50-275 and 50-323, Diablo

Canyon Nuclear Power Plant, Unit Nos.

1 and 2, San Luis Obispo County,

California

Date of application for amendments:

December 19, 1995, as supplemented by

letter dated August 8, 1996.

Brief description of amendments: The

amendments revised the combined

Technical Specifications (TS) for the

Diablo Canyon Power Plant Unit Nos. 1

and 2 to relocate Technical

Specification (TS) 6.5, ‘‘Review and

Audit,’’ 6.8, ‘‘Procedures and

Programs,’’ Sections 6.8.1c., 6.8.1d.,

6.8.2, and 6.8.3, in accordance with

guidance in an NRC letter dated October

25, 1993, from William T. Russell to the

chairpersons of industry owners groups

and the Commission’s Final Policy

Statement on TS Improvements for

Nuclear Power Reactors on relocation of

TS that do not satisfy the retention

criteria. As part of the relocation of TS

6.8.2, TS 6.1.1 would be revised to

require that proposed tests,

experiments, or modifications that affect

nuclear safety be approved by the plant

manager or his designee prior to

implementation.

Date of issuance: October 25, 1996

Effective date: October 25, 1996, to be

implemented within 90 days of

issuance.

Amendment Nos.: Unit 1 - 117; Unit

2 - 115

Facility Operating License Nos. DPR80 and DPR-82: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: January 22, 1996 (61 FR 1633)

The August 8, 1996, supplemental letter

provided additional clarifying

information and did not change the

staff’s initial no significant hazards

consideration determination. The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated October 25, 1996.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: California Polytechnic State

University, Robert E. Kennedy Library,

Government Documents and Maps

Department, San Luis Obispo, California

93407

Pennsylvania Power and Light

Company, Docket Nos. 50-387 and 50-

388 Susquehanna Steam Electric

Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of application for amendments:

May 20 and 28, 1996, as supplemented

by letter dated July 25, 1996

Brief description of amendments:

These amendments, for both units, add

a reference to the ANF-B critical power

correlation to Section 6.9.3.2 of the

Technical Specifications (TSs); change

the values of the minimum critical

power ratio (MCPR) in TS Sections 2.1

and 3.4.1.1.2, and make appropriate

Bases changes. For Unit 1 only, a

reference to ABB licensing methodology

report CENPD-300 (for lead use

assemblies being used in the reactor

core during the upcoming operating

cycle) is added to Section 6.9.3.2.

Date of issuance: October 11, 1996

Effective date: For both units, as of

date of issuance, to be implemented

within 30 days.

Amendment Nos.: 161 and 132

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

revised the Technical Specifications.

Date of initial notice in Federal

Register: Unit 2, August 28, 1996 (61 FR

44362); Unit 1, September 4, 1996 (61

FR 47529)The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

October 11, 1996.No significant hazards

consideration comments received: No

Local Public Document Room

location: Osterhout Free Library,

Reference Department, 71 South

Franklin Street, Wilkes-Barre, PA 18701

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:

March 29, 1996, as supplemented July

12, 1996, and September 6, 1996.

Brief description of amendment: The

proposed amendment would change the

Indian Point 3 Technical Specifications

(TSs) relating to minimum reactor

coolant system (RCS) flow and

maximum RCS average temperature to

make these parameters consistent with

an assumption of 100% helium release

from the boron coating of the integral

fuel burnable absorber rods. The

proposed amendment would also add

limits associated with Departure from

Nucleate Boiling to the IP3 Technical

Specifications TSs.

Date of issuance: October 22, 1996

Effective date: As of the date of

issuance to be implemented within 30

days
57496 Federal Register / Vol. 61, No. 216 / Wednesday, November 6, 1996 / Notices

Amendment No.: 170

Facility Operating License No. DPR64: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 17, 1996 (61 FR 37301)

August 14, 1996 (61 FR 42283)The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated October 22, 1996.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10610.

Southern California Edison Company,

et al., Docket Nos. 50-361 and 50-362,

San Onofre Nuclear Generating Station,

Unit Nos. 2 and 3, San Diego County,

California

Date of application for amendments:

July 19, 1995, as supplemented by

letters dated December 22, 1995, and

March 26, 1996.

Brief description of amendments:

These amendments modify Technical

Specification (TS) 3.3.8, ‘‘Containment

Purge Isolation Signal (CPIS),’’ and TS

3.3.9, ‘‘Control Room Isolation Signal

(CRIS).’’ The revisions are needed to (1)

support the upgrading or replacement of

existing radiation monitoring system

with state-of-the-art equipment that will

provide for greater operational

flexibility and reliability, and (2)

incorporate minor editorial changes to

improve clarity of these TS sections.

Date of issuance: October 8, 1996

Effective date: October 8, 1996, to be

implemented within 30 days of date of

issuance

Amendment Nos.: Unit 2 -

Amendment No. 132; Unit 3 -

Amendment No. 121

Facility Operating License Nos. NPF10 and NPF-15: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: September 27, 1995 (60 FR

49948). The December 22, 1995, and

March 26, 1996, letters provided

additional clarifying information and

did not change the initial no significant

hazards consideration

determination.The Commission’s

related evaluation of the amendments is

contained in a Safety Evaluation dated

October 8, 1996.No significant hazards

consideration comments received: No.

Temporary Local Public Document

Room location: Science Library,

University of California, P. O. Box

19557, Irvine, California 92713

Southern Nuclear Operating Company,

Inc., Docket No. 50-364, Joseph M.

Farley Nuclear Plant, Unit 2, Houston

County, Alabama

Date of amendment request: March

29, 1996, as supplemented by letters

dated June 27, August 29, and

September 16, 1996.

Brief description of amendment: The

amendment changes Technical

Specification 3/4.4.6, ‘‘Steam

Generators’’ and associated Bases to

modify the steam generator repair limit

to clarify that the appropriate method

for determining serviceability for tubes

with outside diameter stress corrosion

cracking at the tube support plate is by

a methodology that more reliably

assesses structural integrity.

Date of issuance: October 11, 1996

Effective date: As of the date of

issuance to be implemented within 30

days

Amendment No.: 115

Facility Operating License No. NPF-8:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: May 22, 1996 (61 FR 25711)

The June 27, August 29, and September

16, 1996, letters provided additional,

clarifying information that did not

change the scope of the March 29, 1996,

application and the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated October 11,

1996. No significant hazards

consideration comments received: No.

Local Public Document Room

location: Houston-Love Memorial

Library, 212 W. Burdeshaw Street, Post

Office Box 1369, Dothan, Alabama

36302

Southern Nuclear Operating Company,

Inc., Docket No. 50-364, Joseph M.

Farley Nuclear Plant, Unit 2, Houston

County, Alabama

Date of amendment request: April 22,

1996, as supplemented by letters dated

May 3, July 15, August 7 and 30, and

September 16, 1996

Brief description of amendment: The

amendment changes reflect the

implementation of a new F* criterion

based on maintaining existing safety

margins for steam generator tube

structural integrity concurrent with

allowances for nondestructive

examination eddy current uncertainty.

Date of issuance: October 11, 1996

Effective date: As of the date of

issuance to be implemented within 30

days

Amendment No.: 116

Facility Operating License No. NPF-8:

Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: May 22, 1996 (61 FR 25713)

The May 3, July 15, August 7 and 30,

and September 16, 1996, letters

provided clarifying information that did

not change the scope of the April 22,

1996, application and the initial

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated October 11,

1996. No significant hazards

consideration comments received: No.

Local Public Document Room

location: Houston-Love Memorial

Library, 212 W. Burdeshaw Street, Post

Office Box 1369, Dothan, Alabama

36302

Tennessee Valley Authority, Docket No.

50-390, Watts Bar Nuclear Plant, Unit

1, Rhea County, Tennessee

Date of application for amendment:

June 29, 1996

Brief description of amendment: The

amendment revises Technical

Specification (TS) Section 5.2.2.f to

delete the sentence, ‘‘The Operations

Manager shall hold or have held an SRO

[Senior Reactor Operator] license on a

similar unit.’’ The revision also

indicates that the Operations

Superintendent will hold a valid SRO

license on this unit.

Date of issuance: October 15, 1996

Effective date: Octber 15, 1996

Amendment No.: 4

Facility Operating License No. NPF90: Amendment revises the TS.

Date of initial notice in Federal

Register: September 11, 1996 (61 FR

47983)The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

October 15, 1996.No significant hazards

consideration comments received: No.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga,

TN 37402

TU Electric Company, Docket Nos. 50-

445 and 50-446, Comanche Peak Steam

Electric Station, Unit Nos. 1 and 2,

Somervell County, Texas

Date of amendment request: July 31,

1996 (TXX-96432) as supplemented by

letters dated August 23 and 27 (TXX96447 and TXX-96451), and September

19, 1996 (TXX-96469).

Brief description of amendments: The

amendments (1) change the acceptance

values for amperes and voltage for the

18 month surveillance test of the battery

chargers; (2) clarify the meaning of the
Federal Register / Vol. 61, No. 216 / Wednesday, November 6, 1996 / Notices 57497

term ‘‘associated inverter’’ used in the

context of energizing 118-Volt AC

Instrument Buses during MODES 1

through 6; and (3) delete the protection

channel and the vital bus ratings for the

118-Volt AC Instrument Buses

identified for MODES 1 through 4.

Date of issuance: October 22, 1996

Effective date: October 22, 1996

Amendment Nos.: Unit 1 -

Amendment No. 53; Unit 2 -

Amendment No. 39

Facility Operating License Nos. NPF87 and NPF-89. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: August 28, 1996 (61 FR 44363)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated October 22, 1996.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: University of Texas at

Arlington Library, Government

Publications/Maps, 702 College, P.O.

Box 19497, Arlington, TX 76019.

Dated at Rockville, Maryland, this 30th day

of October 1996.

For the Nuclear Regulatory Commission

Steven A. Varga,

Director, Division of Reactor Projects - I/

II,Office of Nuclear Reactor Regulation

[Doc. 96-28372 Filed 11-5-96; 8:45 am]

BILLING CODE 7590-01-F

SECURITIES AND EXCHANGE

COMMISSION

Issuer Delisting; Notice of Application

To Withdraw From Listing and

Registration; (The Alpine Group, Inc.,

Common Stock, $0.10 Par Value) File

No. 1–9078

October 31, 1996.

The Alpine Group, Inc. (‘‘Company’’)

has filed an application with the

Securities and Exchange Commission

(‘‘Commission’’), pursuant to Section

12(d) of the Securities Exchange Act of

1934 (‘‘Act’’) and Rule 12d2–2(d)

promulgated thereunder, to withdraw

the above specified security (‘‘Security’’)

from listing and registration on the

American Stock Exchange, Inc.

(‘‘Amex’’).

The reasons alleged in the application

for withdrawing the Security from

listing and registration include the

following:

According to the Company, the Board

of Directors (the ‘‘Board’’) adopted a

resolution as of September 27, 1996 to

withdraw the Security from listing on

the Amex and, instead, to list such

Security on the New York Stock

Exchange (‘‘NYSE’’). The decision of the

Board on this matter followed an

appropriate exploration of means to

enhance stockholder value, and was

based upon the belief that the listing of

the Security on the NYSE will be more

beneficial to its shareholders than

continued listing on the Amex.

Any interested person may, on or

before November 22, 1996, submit by

letter to the Secretary of the Securities

and Exchange Commission, 450 Fifth

Street, NW., Washington, DC 20549,

facts bearing upon whether the

application has been made in

accordance with the rules of the

exchanges and what terms, if any,

should be imposed by the Commission

for the protection of investors. The

Commission, based on the information

submitted to it, will issue an order

granting the application after the date

mentioned above, unless the

Commission determines to order a

hearing on the matter.

For the Commission, by the Division of

Market Regulation, pursuant to delegated

authority.

Jonathan G. Katz,

Secretary.

[FR Doc. 96–28455 Filed 11–5–96; 8:45 am]

BILLING CODE 8010–01–M

[Rel. No. IC–22306; File No. 811–7796]

ILI Endeavor Variable Annuity Account

October 30, 1996.

AGENCY: Securities and Exchange

Commission (‘‘SEC’’ or ‘‘Commission’’).

ACTION: Notice of Application for an

order under the Investment Company

Act of 1940 (‘‘1940 Act’’).

APPLICANT: ILI Endeavor Variable

Annuity Account.

RELEVANT 1940 ACT SECTION: Order

requested under Section 8(f) of the 1940

Act.

SUMMARY OF APPLICATION: Applicant

seeks an order declaring that it has

ceased to be an investment company as

defined by the 1940 Act.

FILING DATE: The application was filed

on July 7, 1996.

HEARING OR NOTIFICATION OF HEARING: An

order granting the application will be

issued unless the Commission orders a

hearing. Interested persons may request

a hearing by writing to the Secretary of

the SEC and serving Applicant with a

copy of the request, in person or by

mail. Hearing requests should be

received by the SEC by 5:30 p.m. on

November 25, 1996, and should be

accompanied by proof of service on

Applicant in the form of an affidavit or,

for lawyers, a certificate of service.

Hearing requests should state the nature

of the requestor’s interest, the reason for

the request, and the issues contested.

Persons may request notification of a

hearing by writing to the Secretary of

the SEC.

ADDRESSES: Secretary, Securities and

Exchange Commission, 450 5th Street,

N.W., Washington, D.C. 20549.

Applicant, Frank A. Camp, Esq.,

International Life Investors Insurance

Company, 4333 Edgewood Road N.E.,

Cedar Rapids, Iowa 52499.

FOR FURTHER INFORMATION CONTACT:

Patrice M. Pitts, Branch Chief, or

Michael Koffler, Law Clerk, Office of

Insurance Products (Division of

Investment Management), at (202) 942–

0670.

SUPPLEMENTARY INFORMATION: Following

is a summary of the application; the

complete application is available for a

fee from the Public Reference Branch of

the SEC.

Applicant’s Representations

1. Applicant, a unit investment trust,

is a separate account of International

Life Investors Insurance Company

(‘‘ILI’’) designed as a funding medium

for variable annuity contracts

(‘‘Contracts’’). On June 14, 1993,

Applicant filed with the Commission a

notification of registration as an

investment company on Form N–8A,

and a registration statement under

Section 8(b) of the 1940 Act and under

the Securities Act of 1933 (File No. 33–

64414) registering an indefinite amount

of securities (i.e., the Contracts). The

registration statement was declared

effective, August 12, 1993, and

Applicant began offering Contracts on

August 12, 1993.

2. The boards of directors of ILI and

AUSA Life Insurance Company (‘‘AUSA

Life’’) authorized the adoption of an

‘‘Assumption Reinsurance Agreement’’

on September 27, 1994. Contractholders

were given the right to reject the

assumption of their Contracts by AUSA

Life, as required by the law of the State

of New York, via a solicitation dated

December 1, 1994. No contractholders

rejected the assumption of their

Contracts pursuant to the terms of the

solicitation.

3. The Assumption Reinsurance

Agreement, dated as of December 31,

1994, providef for the transfer of the in

force variable annuity business of ILI to

AUSA Life, as of January 1, 1995.

Effective January 1, 1995, ILI ceded and

transferred to ASUA Life all variable

insurance contracts issued by ILI in

connection with its variable annuity

business.