Federal Register / Vol. 61, No. 168 / Wednesday, August 28, 1996 / Notices 44353

Room T–2B3, 11545 Rockville Pike,

Rockville, Maryland.

A portion of this meeting may be

closed to discuss the proposed NRC

thermal hydraulic research program

budget. Discussion of the impact of

possible budget reduction on continuing

and proposed research contracts, if held

in public session, might result in

premature disclosure of information

which would in turn frustrate the

Commission’s ability to effectively

implement the affected programs.

The agenda for the subject meeting

shall be as follows:

Wednesday, September 18, 1996—8:30

a.m. until the conclusion of business

Thursday, September 19, 1996—8:30

a.m. until the conclusion of business

The Subcommittee will: (1) Begin its

review of the NRC–RES Program to

revise/replace the current suite of NRC–

RES thermal hydraulic codes and (2)

discuss the status of the RES thermal

hydraulic research program and

associated budget. The purpose of this

meeting is to gather information,

analyze relevant issues and facts, and to

formulate proposed positions and

actions, as appropriate, for deliberation

by the full Committee.

Oral statements may be presented by

members of the public with the

concurrence of the Subcommittee

Chairman; written statements will be

accepted and made available to the

Committee. Electronic recordings will

be permitted only during those portions

of the meeting that are open to the

public, and questions may be asked only

by members of the Subcommittee, its

consultants, and staff. Persons desiring

to make oral statements should notify

the cognizant ACRS staff engineer

named below five days prior to the

meeting, if possible, so that appropriate

arrangements can be made.

During the initial portion of the

meeting, the Subcommittee, along with

any of its consultants who may be

present, may exchange preliminary

views regarding matters to be

considered during the balance of the

meeting.

The Subcommittee will then hear

presentations by and hold discussions

with representatives of the NRC staff, its

consultants, and other interested

persons regarding this review.

Further information regarding topics

to be discussed, whether the meeting

has been cancelled or rescheduled, the

scheduling of sessions which are open

to the public, the Chairman’s ruling on

requests for the opportunity to present

oral statements and the time allotted

therefor can be obtained by contacting

the cognizant ACRS staff engineer, Mr.

Paul A. Boehnert (telephone 301/415–

8065) between 7:30 a.m. and 4:15 p.m.

(EDT). Persons planning to attend this

meeting are urged to contact the above

named individual one or two working

days prior to the meeting to be advised

of any potential changes to the agenda,

etc., that may have occurred.

Dated: August 21, 1996.

Sam Duraiswamy,

Chief, Nuclear Reactors Branch.

[FR Doc. 96–21940 Filed 8–27–96; 8:45 am]

BILLING CODE 7590–01–P

Biweekly Notice

Applications and Amendments to

Facility Operating Licenses Involving

No Significant Hazards Considerations

I. Background

Pursuant to Public Law 97-415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 97-415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from August 3,

1996, through August 16, 1996. The last

biweekly notice was published on

August 14, 1996 (61 FR 42274).

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 September 27, 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
44354 Federal Register / Vol. 61, No. 168 / Wednesday, August 28, 1996 / Notices

Domestic Licensing Proceedings’’ in 10

CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555-0001, Attention:

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.

Carolina Power & Light Company, et

al., Docket No. 50-400, Shearon Harris

Nuclear Power Plant, Unit 1, Wake and

Chatham Counties, North Carolina

Date of amendment request: July 19,

1996

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) Section 3/

4.6.2, Containment Spray System, to

extend the surveillance interval for

performance of an air or smoke flow test

through containment spray nozzles from

once per 5 years to once per 10 years.

This change is consistent with the

guidance in NRC Generic Letter 93-05,

‘‘Line Item Technical Specifications

Improvements to Reduce Surveillance

Requirements for Testing During Power

Operations,’’ and NUREG-1366,

‘‘Improvements To Technical

Specifications Surveillance

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 amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed extended testing frequency

of containment spray nozzles will not affect

any initiators of any previously evaluated

accidents or change the manner of operation

for any system or component. The

containment spray system serves a mitigating

function by removing heat and fission

products from a post accident containment

atmosphere. Increasing the surveillance test

interval will not affect the system’s ability to

provide this function. Therefore, there would

be no increase in the probability or

consequences of an accident previously

evaluated.
Federal Register / Vol. 61, No. 168 / Wednesday, August 28, 1996 / Notices 44355

2. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

Since the proposed change affects only a

surveillance frequency, it will not involve

any physical alterations to plant equipment

or alter the manner in which any safetyrelated system performs its function.

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 amendment does not

involve a significant reduction in the margin

of safety.

The proposed change does not affect any

Final Safety Analysis Report (FSAR) Chapter

15 accident analyses or impact the margin of

safety for the containment spray system as

defined in the Bases to the Technical

Specifications. Therefore, the proposed

change does not involve a significant

reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh,

North Carolina 27605

Attorney for licensee: William D.

Johnson, Vice President and Senior

Counsel, Carolina Power & Light

Company, Post Office Box 1551,

Raleigh, North Carolina 27602

NRC Project Director: Eugene V.

Imbro

Commonwealth Edison Company,

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

Cities Nuclear Power Station, Units 1

and 2, Rock Island County, Illinois

Date of amendment request: June 10,

1996

Description of amendment request: To

change the technical specifications to

reflect the transition from General

Electric Company (GE) to Siemens

Power Corporation (SPC) as the fuel

supplier for the Quad Cities Nuclear

Power Station, Units 1 and 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) Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The probability of an evaluated accident is

derived from the probabilities of the

individual precursors to that accident. The

consequences of an evaluated accident are

determined by the operability of plant

systems designed to mitigate those

consequences. Limits will be established

consistent with NRC approved methods to

ensure that fuel performance during normal,

transient, and accident conditions is

acceptable. The proposed Technical

Specifications amendment reflects previously

approved SPC methodology used to analyze

normal operations, including anticipated

operational occurrences (AOOs), and to

determine the potential consequences of

accidents.

Licensing Methods and Models

The proposed amendment is to support

operation with NRC approved fuel and

licensing methods supplied from Siemens

Power Corporation. In accordance with FSAR

Chapter 15, the same accidents and transients

will be analyzed with the new fuel and

methods as were analyzed by GE for GE fuel.

The analysis methods and models are NRC

approved. These approved methods and

models are used to determine the fuel

thermal limits (e.g., LHGR, APLHGR, MCPR).

The SPC core monitoring code enables the

site to monitor keff as well as rod density to

perform the reactivity anomaly surveillance.

This is consistent with GE methodology. The

support systems for minimizing the

consequences of transients and accidents are

not affected by the proposed amendment.

Therefore, the change in licensing analysis

methods and models does not significantly

increase the probability of an accident or the

consequences of an accident previously

identified.

New Fuel Design

The use of ATRIUM 9B fuel at Quad Cities

does not involve a significant increase in the

probability or consequences of any accident

previously evaluated in the FSAR. The

ATRIUM-9B fuel is generically approved for

use as a reload BWR fuel type (Reference:

ANF-89-014(P)(A) Rev. 1 Supplement 1,

General Mechanical Design for Advanced

Nuclear Fuels 9X9-IX and 9X9-9X BWR

Reload Fuel). Limiting postulated

occurrences and normal operation have been

analyzed using NRC-approved methods for

the ATRIUM 9B fuel design to ensure that

safety limits are protected and that

acceptable transient and accident

performance is maintained.

The reload fuel has no adverse impact on

the performance of in-core neutron flux

instrumentation or CRD response. The

ATRIUM-9B fuel design will not adversely

affect performance of neutron

instrumentation nor will it adversely affect

the movement of control blades relative to

the GE fuel. The exterior dimensions of the

ATRIUM-9B fuel have been evaluated by

ComEd; the SPC fuel provides adequate

clearances relative to the GE10 fuel installed

at Quad Cities. Thus, no increased

interactions with the adjacent control blade

and nuclear instrumentation are created.

Additionally, given the above mentioned

overall envelope similarities, no problems are

anticipated with other station equipment

such as the fuel storage racks, the new fuel

inspection stand and the spent fuel pool fuel

preparation machine. Therefore, the

probability of adverse interactions between

the Siemens fuel and components in the core

and fuel handling equipment is not

significantly increased.

The ATRIUM 9B design is neutronically

compatible with the existing fuel types and

core components in the Quad Cities core.

SPC tests have demonstrated that the

ATRIUM-9B fuel design is hydraulically

compatible with the GE9/GE10 fuel. The

bundle pressure drop characteristics of the

ATRIUM 9B bundle are similar to those of

the GE9/GE10 fuel design, hence core

thermal-hydraulic stability characteristics are

not adversely affected by the ATRIUM 9B

design. Cycle stability calculations are

performed by SPC. Therefore, the probability

of thermal hydraulic instability is not

significantly increased.

An evaluation of the Emergency

Procedures is being performed to ensure that

the use of the ATRIUM-9B fuel at Quad Cities

does not alter any assumptions previously

made in evaluating the radiological

consequences of an accident at Quad Cities

Station. Therefore, the radiological

consequences of accidents are not

significantly increased.

Methods approved by the NRC are being

used in the evaluation of fuel performance

during normal and abnormal operating

conditions. The ComEd and SPC methods to

be used for the cycle specific transient

analyses have been previously NRC

approved. The proposed methodologies are

administrative in nature and do not

significantly affect any accident precursors or

accident results; as such, the proposed

incorporation of the SPC methodologies for

Quad Cities does not significantly increase

the probability or consequences of any

previously evaluated accidents. The

description of the fuel is modified to include

the water box design of the NRC approved

ATRIUM-9B fuel. This change is

administrative.

Review of the above concludes that the

probability of occurrence and the

consequences of an accident previously

evaluated in the safety analysis report have

not been significantly increased.

* * * * *

2) Create the possibility of a new or

different kind of accident from any accident

previously evaluated:

Creation of the possibility of a new or

different kind of accident would require the

creation of one or more new precursors of

that accident. New accident precursors may

be created by modifications of the plant

configuration, including changes in

allowable modes of operation.

Licensing Methods and Models

The proposed Technical Specification

amendment reflects previously approved SPC

methodology used to analyze normal

operations, including AOOs, and to

determine the potential consequences of

accidents. In accordance with FSAR Chapter

15, the same accidents and transients will be

analyzed with the new fuel and methods as

were analyzed by GE for GE fuel. As stated

above, the proposed changes do not permit

modes of operation which differ from those

currently permitted; therefore, the possibility

of a new or different kind of accident is not

created. Plant support equipment is not

affected by the proposed changes; therefore,

no new failure modes are created.

New Fuel Design

The basic design concept of a 9x9 fuel pin

array with an internal water box has been

used in various lead assembly programs and

in reload quantities in Europe since 1986.
44356 Federal Register / Vol. 61, No. 168 / Wednesday, August 28, 1996 / Notices

WNP-2 has loaded reload quantities since

1991. Approximately 650 water box

assemblies have been irradiated in the United

States through 1995, with a substantially

higher number being irradiated overseas. The

NRC has reviewed and approved the

ATRIUM-9B fuel design (Reference: ANF-89-

014(P)(A) Rev. 1 Supplement 1, Generic

Mechanical Design for Advanced Nuclear

Fuels 9X9-IX and 9X9-9X BWR Reload Fuel).

The similarities in fuel design and operation

between GE and SPC, and the previous

Boiling Water Reactor experience with both

vendors’ fuel indicate there would be no new

or different types of accidents for Quad Cities

than have been considered for the existing

fuel. Therefore, the use of ATRIUM-9B fuel

at Quad Cities does not create the possibility

of a new or different kind of accident from

any accident previously evaluated.

* * * * *

3) Involve a significant reduction in the

margin of safety for the following reasons:

The existing margin to safety is provided

by the existing acceptance criteria (e.g.,

10CFR50.46 limits). The proposed Technical

Specification amendment reflects previously

approved SPC methodology used to

demonstrate that the existing acceptance

criteria are satisfied. The revised

methodology has been previously reviewed

and approved by the USNRC for application

to reload cores of GE BWRs. References for

the Licensing Topical Reports which

document this methodology, and include the

Safety Evaluation Reports prepared by the

USNRC, are added to the Reference section

of the Technical Specifications as part of this

amendment.

Licensing Methods and Models

The proposed amendment does not involve

changes to the existing operability criteria.

NRC approved methods and established

limits (implemented in the COLR) ensure

acceptable margin is maintained. The ComEd

and SPC reload methodologies for the

ATRIUM-9B reload design are consistent

with the Technical Specification Bases. The

Limiting Conditions for Operation are taken

into consideration while performing the

cycle specific and generic reload safety

analyses. NRC approved methods are listed

in Section 6 of the Technical Specifications.

Analyses performed with NRC-approved

methodology have demonstrated that fuel

design and licensing criteria will be met

during normal and abnormal operating

conditions. The same margins of safety are

utilized by SPC as GE (e.g., limits on peak

cladding temperature, cladding oxidation,

plastic strain). Therefore, there is not a

significant reduction in the margin of safety.

New Fuel Design

The exterior dimensions of the ATRIUM9B fuel assembly result in equivalent

clearances relative to the GE10B. Thus, no

increased interactions with the adjacent

control blade and nuclear instrumentation

are created. The change does not adversely

impact equipment important to safety;

therefore,the margin of safety is not

significantly reduced.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

requested amendments involve no

significant hazards consideration.

Local Public Document Room

location: Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois 61021

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60603

NRC Project Director: Robert A. Capra

Duke Power Company, Docket Nos. 50-

269, 270 and 50-287, Oconee Nuclear

Station, Units 1, 2 and 3, Oconee

County, South Carolina

Date of amendment request: August

12, 1996

Description of amendment request:

The proposed change would implement

the performance-based containment leak

rate testing provisions of Option B to 10

CFR Part 50 Appendix J for the Type A

(containment) testing program.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The following analysis is presented,

pursuant to 10 CFR 50.91, to demonstrate

that the proposed change will not create a

Significant Hazard Consideration.

1. The proposed change will not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

Containment leak rate testing is not an

initiator of any accident; the proposed

change does not affect reactor operations or

accident analysis, and has no significant

radiological consequences. Therefore, this

proposed change will not involve an increase

in the probability or consequences of any

previously-evaluated accident.

2. The proposed change will not create the

possibility of any new accident not

previously evaluated.

The proposed change does not affect

normal plant operations or configuration, or

change any design basis. The proposed

changes will not affect the response of [the]

containment during a design basis accident.

3. There is no significant reduction in a

margin of safety.

The proposed changes are based on NRCaccepted provisions, and maintain necessary

levels of reliability of containment integrity.

The performance-based approach to leakage

rate testing recognizes that historically good

results of containment testing provide

appropriate assurance of future containment

integrity; this supports the conclusion that

the impact on the health and safety of the

public as a result of extended test intervals

is negligible.

Based on the above, no significant hazards

consideration is created by the proposed

change.

The NRC has reviewed the licensee’s

analysis and, based on thisreview, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Oconee County Library, 501

West South Broad Street, Walhalla,

South Carolina 29691

Attorney for licensee: J. Michael

McGarry, III, Winston and Strawn, 1200

17th Street, NW., Washington, DC 20036

NRC Project Director: Herbert N.

Berkow

Entergy Operations, Inc., Docket No.

50-313, Arkansas Nuclear One, Unit

No. 1, Pope County, Arkansas

Date of amendment request: May 31,

1996

Description of amendment request:

The proposed amendment revises the

surveillance test interval for the reactor

protection system reactor trip breakers,

reactor trip modules, and electronic trip

relays from 1 month to 6 months. In

addition to requesting a change to the

Arkansas Nuclear One, Unit 1 Technical

Specifications, the request also proposes

the same changes to NUREG-1430,

Standard Technical Specifications -

Babcock and Wilcox Plants.

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:

Criterion 1 - Does Not Involve a Significant

Increase in the Probability or Consequences

of an Accident Previously Evaluated.

The accident mitigation features of the

plant are not affected by the proposed test

interval extension. The results of the B&W

Owners Group Topical Report BAW-10167,

Supplement 3, ‘‘Justification for increasing

The Reactor Trip System On-Line Test

Intervals,’’ show that the test interval

extension of the reactor protection system

trip devices is not a significant contributor to

trip system unavailability or the risk of core

damage.

Therefore, this change does not involve a

significant increase in the probability or

consequences of any accident previously

evaluated.

Criterion 2. Does Not Create the Possibility

of a New or Different Kind of Accident from

any Previously Evaluated.

The reactor trip device surveillance test

interval is not, in and of itself, considered to

be an accident initiator. Failure of a trip

device to function is an analyzed condition

and does not constitute a new or different

kind of accident.

Therefore, this change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

Criterion 3. Does Not Involve a Significant

Reduction in the Margin of Safety.

The results of the B&W Owners Group

Topical Report BAW-10167, Supplement 3,

‘‘Justification for Increasing The Reactor Trip
Federal Register / Vol. 61, No. 168 / Wednesday, August 28, 1996 / Notices 44357

System On-Line Test Intervals,’’ show that

the test interval extension of the reactor

protection system trip devices is not a

significant contributor to trip system

unavailability or the risk of core damage. In

addition, the uncertainty analysis contained

in BAW-10167 confirms the robustness of the

results by demonstrating that even with an

order of magnitude change in the failure data,

the incremental increase due to an increased

test interval is insignificant. Entergy

Operations has reviewed BAW-10167 and

found it applicable to ANO-1.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration. Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801 Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

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

20005-3502

NRC Project Director: William D.

Beckner

Entergy Operations, Inc., Docket Nos.

50-313 and 50-368, Arkansas Nuclear

One, Unit Nos. 1 and 2 (ANO-1&2),

Pope County, Arkansas

Date of amendment request: May 9,

1996

Description of amendment request:

The proposed amendment changes the

name of Arkansas Power and Light

Company (AP&L) to Entergy Arkansas,

Inc. in both the Operating License and

the Technical Specifications. AP&L is

licensed to own and possess Arkansas

Nuclear One (ANO). The company

licensed to operate ANO, Entergy

Operations, Inc. is unaffected by this

change. Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does Not Involve a Significant Increase

in the Probability or Consequences of an

Accident Previously Evaluated.

The proposed change documents changing

the legal name of the company. The proposed

change will not affect any other obligations.

The company will continue to own all of the

same assets, will continue to serve the same

customers, and will continue to honor all

existing obligations and commitments.

Therefore, this change does not involve a

significant increase in the probability or

consequences of any accident previously

evaluated.

2. Does Not Create the Possibility of a New

or Different Kind of Accident from any

Previously Evaluated.

The administrative changes in the

operating license requirements do not

involve any change in the design of the plant.

Therefore, this change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

3. Does Not Involve a Significant

Reduction in the Margin of Safety.

The proposed change is administrative in

nature and does not reduce the margin of

safety imposed by any current requirements.

Therefore, this change does not involve a

significant reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration. Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801 Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

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

20005-3502

NRC Project Director: William D.

Beckner

Florida Power and Light Company,

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

Point Plant Units 3 and 4, Dade County,

Florida

Dates of amendment request: July 17,

1996

Description of amendment request:

The licensee proposed to change the

Turkey Point Units 3 and 4 Technical

Specifications (TS) to implement 10

CFR 50, Appendix J, Option B, for

containment leakage testing. Changes

include relocating the details for

containment testing to the ‘‘containment

leakage rate testing program’’ and

adding the requirements of the

containment leakage rate testing

program to TS 6.8.4, which describes

facility programs. Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below.

(1) Operation of the facility in accordance

with the proposed amendments would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed amendments do not involve

a significant increase in the probability or

consequences of an accident previously

evaluated because:

a) These proposed changes are all

consistent with NRC requirements and

guidance for implementation of 10 CFR 50,

Appendix J, Option B.

b) Based on industry and NRC evaluations

performed in support of developing Option

B, these changes potentially result in a minor

increase in the consequences of an accident

previously evaluated due to the expanded

testing intervals. However, the proposed

changes do not result in an increase in the

core damage frequency since the containment

system is used for mitigation purposes only.

c) These changes are 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.

Therefore, these changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

(2) Operation of the facility in accordance

with the proposed amendments would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The use of the modified specifications can

not create the possibility of a new or different

kind of accident from any previously

evaluated since the proposed amendments

will not change the physical plant or the

modes of plant operation defined in the

facility operating license. No new failure

mode is introduced due to the

implementation of a performance-based

program for containment leakage rate testing,

since the proposed changes do not involve

the addition or modification of equipment,

nor do they alter the design or operation of

affected plant systems, structures, or

components.

(3) Operation of the facility in accordance

with the proposed amendments would not

involve a significant reduction in a margin of

safety.

The operating limits and functional

capabilities of the affected systems,

structures, and components are basically

unchanged by the proposed amendments due

to the following reasons:

a) The acceptance criteria for total

integrated containment leakage of 1.0 La is

consistent with the current technical

specifications and is within the design basis

accident assumptions, and therefore does not

reduce the margin of safety.

b) The increase in intervals between leaktest surveillances will not significantly

reduce the margin of safety as shown by

findings in NUREG 1493, ‘‘PerformanceBased Containment Leak-Test Program’’,

which was based on implementation of the

performance-based testing of Option B.

Therefore these 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 50.92(c) are

satisfied.Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration. Local Public Document Room

location: Florida International

University, University Park, Miami,

Florida 33199

Attorney for licensee: J. R. Newman,

Esquire, Morgan, Lewis & Bockius, 1800

M Street, NW., Washington, DC 20036

NRC Project Director: Frederick J.

Hebdon
44358 Federal Register / Vol. 61, No. 168 / Wednesday, August 28, 1996 / Notices

Georgia Power Company, Oglethorpe

Power Corporation, Municipal Electric

Authority of Georgia, City of Dalton,

Georgia, Docket Nos. 50-321 and 50-

366, Edwin I. Hatch Nuclear Plant,

Units 1 and 2, Appling County, Georgia

Date of amendment request: May 21,

1996

Description of amendment request:

The proposed change to the condensate

storage tank (CST) level indication

would ensure that the water level is

sufficient to provide 50,000 gallons of

water for core spray makeup to the

reactor pressure vessel.

Technical Specification (TS)

Surveillance Requirement (SR) 3.5.2.2.b

for ECCS - Shutdown states:

‘‘Condensate storage tank (CST) water

level is [greater than or equal to] 12

feet.’’ The corresponding Bases state: ’’...

the CST contains [greater than or equal

to] 150,000 gallons of water, equivalent

to 12 feet, ensures that the CS System

can supply at least 50,000 gallons of

makeup water to the RPV.’’

Subsequent licensee analyses

confirmed that Plant Hatch Units 1 and

2 CST configurations are different; that

is, for both CSTs, a water level of 12 feet

is not equivalent to the required

capacity of 150,000 gallons of water.

Based on these calculations, the correct

level for the Unit 1 CST is 13 feet, and

the correct level for the Unit 2 CST is

15 feet.

The proposed change would revise

Unit 1 and Unit 2 SR 3.5.2.2.b to require

a CST water level of greater than or

equal to 13 feet and greater than or

equal to 15 feet, respectively, to ensure

at least 50,000 gallons of water are

available for core spray (CS) makeup to

the reactor pressure vessel (RPV).

The associated Bases for each unit

will be revised accordingly.

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 change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated, because this

administrative change to the CST water level

does not alter the operation of any plant

system or component. The proposed change

does not involve a physical modification to

any structure, system, or component. The

minimum CST water level for each unit is

being increased to account for the height of

the CS suction standpipe within each CST

and the differences in the Unit 1 and

Unit 2 CST diameters (gallons/ft of water)

as follows:

a. Unit 1 - The proposed minimum water

level is calculated as: CS suction standpipe

height of 9 ft + (50,000 gallons divided by

12,704 gallons/ft) = 12.93 ft or 13 ft.

b. Unit 2 - The proposed minimum water

level is calculated as: CS suction standpipe

height of 10 ft + (50,000 gallons divided by

11,343 gallons/ft) = 14.4 ft or 15 ft.

The revised minimum levels ensure at least

50,000 gallons of water are provided above

the top of the standpipe in each unit’s CST

and are available for CS makeup to the RPV,

as stated in the applicable Bases. The TS

Limiting Conditions for Operation (LCO)

remain unaffected by the proposed change.

2. The proposed TS change does not create

the possibility of a new or different kind of

accident from any accident previously

evaluated. Revising Surveillance

Requirement acceptance criteria does not

result in any physical modification to the

plant or operation of any existing equipment.

3. The proposed TS change does not

involve a significant reduction in a margin of

safety, since this administrative change only

ensures the existing TS Bases are satisfied by

increasing the minimum CST water level

requirement to ensure at least 50,000 gallons

of water are available for CS injection to the

RPV. The proposed change does not involve

a physical modification to any structure,

system or component, and does not modify

the operation of any existing equipment.

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: Appling County Public

Library, 301 City Hall Drive, Baxley,

Georgia 31513

Attorney for licensee: Ernest L. Blake,

Jr., Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037

NRC Project Director: Herbert N.

Berkow

Houston Lighting & Power Company,

City Public Service Board of San

Antonio, Central Power and Light

Company, City of Austin, Texas, Docket

Nos. 50-498 and 50-499, South Texas

Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request: July 8,

1996

Description of amendment request:

The proposed amendment would clarify

that the component cooling water

system surge tank level instrumentation

can be demonstrated operable by

performing a channel calibration test

during any plant mode of operation.

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 to Technical

Specification Surveillance Requirement

4.7.3.b.3 will not effect any accident

initiators or precursors and will not alter the

design assumptions for the systems or

components used to mitigate the

consequences of an accident. Calibration is

performed on level instrumentation of

Component Cooling Water System trains that

are out of service for scheduled maintenance.

Isolation redundancy is provided by

instrumentation associated with the trains

that are in service during the calibration.

Since the surveillance will continue to be

performed at the specified interval, this

proposed change will not increase the

probability of occurrence of an accident

previously evaluated. The surveillance does

not differ from those previously performed;

therefore, there is no impact on the

consequences of an accident previously

evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

Clarifying the surveillance interval for

surge tank level instrumentation does not

involve installation or operation of new or

different kinds of equipment. There is no

change in the procedures as described in the

Technical Specifications. The change only

clarifies the interval at which the subject

calibration will be performed. Therefore, the

proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The specified surveillance will remain as

stated in the Technical Specifications.

Consequently, there is no reduction in the

effectiveness of the surveillance in ensuring

equipment operability. Calibration is

performed on level instrumentation of

Component Cooling Water System trains that

are out of service for scheduled maintenance.

Isolation redundancy is provided by

instrumentation associated with the trains

that are in service during the calibration.

Consequently, clarifying the interval at

which the calibration is performed will have

no significant impact on the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the standards of

10 CFR 50.92(c) are satisfied. Therefore,

the NRC staff proposes to determine that

the request for amendments involves no

significant hazards consideration.

Local Public Document Room

location: Wharton County Junior

College, J. M. Hodges Learning Center,

911 Boling Highway, Wharton, TX

77488

Attorney for licensee: Jack R.

Newman, Esq., Morgan, Lewis &

Bockius, 1800 M Street, N.W.,

Washington, DC 20036-5869

NRC Project Director: William D.

Beckner
Federal Register / Vol. 61, No. 168 / Wednesday, August 28, 1996 / Notices 44359

Houston Lighting & Power Company,

City Public Service Board of San

Antonio, Central Power and Light

Company, City of Austin, Texas, Docket

Nos. 50-498 and 50-499, South Texas

Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request: August 8,

1996

Description of amendment request:

The proposed amendment would allow

the transition from Mode 4 to Mode 3

with the turbine-driven auxiliary

feedwater pump inoperable and allow a

72-hour period after the entry into Mode

3 to complete all necessary operability

testing. Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed change will allow entry into

Mode 3 with an inoperable Turbine Driven

Auxiliary Feedwater pump. Since the

operability test on the Turbine Driven

Auxiliary Feedwater pump can only be

performed once steam pressure is greater

than or equal to 1000 psig, this change will

allow the plant to reach the Mode where

steam pressure greater than or equal to 1000

psig is available to perform the operability

testing on the Turbine Driven Auxiliary

Feedwater pump. The allowance of 72 hours

to complete the surveillance testing will

make the surveillance requirements

consistent with the allowed outage time

already established in the Action Statements.

The proposed change does not affect the

probability of an accident. The Turbine

Driven Auxiliary Feedwater pump is not

assumed to be an initiator of any analyzed

event. The consequences of an accident

previously evaluated remain unchanged by

allowing the pump to be inoperable until

suitable conditions exist to perform the

operability testing. The operability testing

will continue to demonstrate that the Turbine

Driven Auxiliary Feedwater pump will

perform as required prior to entry into Mode

2. This change will not alter assumptions

relative to the mitigation of an accident or

transient event. Therefore, this change will

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 previously evaluated.

This change will not physically alter the

plant (no new or different type of equipment

will be installed). The changes in methods

governing normal plant operation are

consistent with current safety analysis

assumptions. The proposed change will

allow entry into Mode 3 with the Turbine

Driven Auxiliary Feedwater pump inoperable

in order to perform the pump Operability

Test on the turbine driven AFW [Auxiliary

Feedwater] pump once steam pressure is

greater than or equal to 1000 psig. Therefore,

this change does not create the possibility of

a new or different kind of accident from any

accident previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed change will allow entry into

Mode 3 with the Turbine Driven AFW pump

inoperable in order to perform the pump

Operability Test on the turbine driven AFW

pump once steam pressure is greater than or

equal to 1000 psig. This will allow time for

the plant to obtain suitable test conditions

with steam pressure greater than or equal to

1000 psig. The margin of safety is not

affected by this change. The operability

testing will continue to maintain assurance

that the AFW Pumps will perform as

required prior to entry into Mode 2. The

safety analysis assumptions will still be

maintained, thus, no question of safety exists.

Therefore, this change does not involve a

significant reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the standards of

10 CFR 50.92(c) are satisfied. Therefore,

the NRC staff proposes to determine that

the request for amendments involves no

significant hazards consideration.

Local Public Document Room

location: Wharton County Junior

College, J. M. Hodges Learning Center,

911 Boling Highway, Wharton, TX

77488

Attorney for licensee: Jack R.

Newman, Esq., Morgan, Lewis &

Bockius, 1800 M Street, N.W.,

Washington, DC 20036-5869

NRC Project Director: William D.

Beckner

North Atlantic Energy Service

Corporation, Docket No. 50-443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request: June 4,

1996

Description of amendment request:

The proposed amendment would

modify the Seabrook Station, Unit No. 1

Technical Specifications to implement

Option B to 10 CFR Part 50, Appendix

J by referring to Regulatory Guide 1.163,

‘‘Performance-Based Containment LeakTest Program. The following Technical

Specifications would be affected by the

proposed amendment:

1. Definitions: Definition 1.7,

Containment Integrity (Item d.) would

be revised to reflect that leakage rates

would be in accordance with the

Containment Leakage Rate Testing

Program.

2. Limiting Conditions for Operation

and Surveillance Requirements:

a. Containment Integrity: Surveillance

Requirement 4.6.1.1.c would be deleted

because the specific guidance would be

contained in the Containment Leakage

Rate Testing Program.

b. Containment Leakage: Limiting

Condition for Operation 3.6.1.2.a

through 3.6.1.2.c and Surveillance

Requirements 4.6.1.2.a through 4.6.1.2.h

would be revised to replace specific

guidance with a reference to the

Containment Leakage Rate Testing

Program. c. Containment Leakage: The Action

for Limiting Condition for Operation

3.6.1.2 would be revised to include the

equivalent Action as required for

Limiting Condition for Operation 3.6.1.1

when the overall integrated containment

leak rate exceeds 1.0 La. d. Containment Air Locks: Limiting

Conditions for Operation 3.6.1.3.a and

3.6.1.3.b would be deleted and

Surveillance Requirements 4.6.1.3.a and

4.6.1.3.b would be revised to replace

specific guidance with a reference to the

Containment Leakage Rate Testing

Program. The footnote addressing the

exemption to Appendix J regarding

testing the air locks prior to establishing

containment integrity would be

maintained in the Containment Leakage

Rate Testing Program. e. Containment Vessel Structural

Integrity: Surveillance Requirement

4.6.1.6 would be revised to replace

specific guidance with a reference to the

Containment Leakage Rate Testing

Program. f. Containment Ventilation System:

Limiting Condition for Operation

3.6.1.7, Action b. would be revised to

replace specific guidance with a

reference to the Containment Leakage

Rate Testing Program. Surveillance

Requirement 4.6.1.7.1 would be revised

to replace specific guidance with a

reference to the Containment Leakage

Rate Testing Program. g. Containment Enclosure Building:

Limiting Condition for Operation 3.6.5.3

and Surveillance Requirement 4.6.5.3

would be revised to include a reference

to the requirements in the Containment

Leakage Rate Testing Program. 3. Bases: Sections 3/4.6.1.2,

Containment Leakage; 3/4.6.1.7,

Containment Ventilation System; and 3/

4.6.5.3, Containment Enclosure Building

Structural Integrity, would be revised to

reflect the above changes including a

reference to the Containment Leakage

Rate Testing Program. In addition, a

statement would be added to Section 3/

4.6.1.2 to clarify the operability of

containment regarding allowable

leakage rates. 4. Administrative Controls: Section

6.15 would be added to establish a

Containment Leakage Rate Testing

Program, as specified in Regulatory

Guide 1.163, dated September 1995.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the
44360 Federal Register / Vol. 61, No. 168 / Wednesday, August 28, 1996 / Notices

licensee has provided its analysis of the

issue of no significant hazards

consideration. The NRC staff has

reviewed the licensee’s analysis against

the standards of 10 CFR 50.92(c). The

NRC staff’s review is presented below.

A. The changes do not involve a significant

increase in the probability or consequences

of an accident previously evaluated (10 CFR

50.92(c)(1)) because the proposed changes

merely revise the testing criteria for

containment penetrations. The revised

criteria will be based on the guidance in

Regulatory Guide 1.163, ‘‘Performance-Based

Containment Leak-Test Program.’’

This guidance allows for the use of relaxed

testing frequencies for containment

penetrations that have performed

satisfactorily on a historical basis.

To support consideration of Option B to

Appendix J, the NRC staff reviewed the

potential impact of performance-based

testing frequencies for containment

penetrations. The NRC staff review is

documented in NUREG-1493 ‘‘PerformanceBased Containment Leak-Test Program.’’ One

of the staff’s conclusions was that reducing

the frequency of Type A tests (Integrated

Leak Rate Tests) from three per 10 years to

one per 10 years leads to a marginal increase

in risk. For Type B and C testing (Local Leak

Rate Tests), the change in testing frequency

will not have significant impact since, under

existing requirements, leakage contributes

less than 0.1 percent of the overall accident

risk. The use of a performance-based testing

program will continue to provide assurance

that the accident analysis assumptions

remain bounding.

B. The changes do not create the possibility

of a new or different kind of accident from

any accident previously evaluated (10 CFR

50.92(c)(2)) because they do not affect the

manner by which the facility is operated or

involve changes to structures, systems, or

components that affect the operational

characteristics of the facility. The changes

merely revise the testing criteria for the

containment penetrations, and establish a

Containment Leakage Rate Testing Program

to ensure that the performance history of

each penetration is satisfactory prior to

changing any test frequency. Since there is

no change to the facility or the way in which

the facility is operated, there is no possibility

of creating a new or different kind of accident

than previously analyzed.

C. The changes do not involve a significant

reduction in a margin of safety (10 CFR

50.92(c)(3)). During the development of 10

CFR Part 50, Appendix J, Option B, the NRC

staff determined the reduction in safety

associated with the implementation of the

performance-based testing program. The staff

concluded that reducing the frequency of

Type A tests (Integrated Leak Rate Tests)

from three per 10 years to one per 10 years

would have an imperceptible impact upon

risk. For Type B and C testing (Local Leak

Rate Tests), the change in testing frequency

will not have significant impact since this

leakage contributes less than 0.1 percent of

the overall risk based on the existing

regulations. The use of Option B will have

minimal impact on the radiological release

rates since most penetration leakage is well

below the specified limits. The staff noted

that the accident risk is relatively insensitive

to containment leakage rate because accident

risk is dominated by accident sequences that

result in failure of or bypass of the

containment. The use of a performance-based

testing program will continue to provide

assurance that the accident analysis

assumptions remain bounding.

Based on this review, it appears that

the three standards of 10 CFR 50.92(c)

are satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration. Local Public Document Room

location: Exeter Public Library,

Founders Park, Exeter, NH 03833

Attorney for licensee: Lillian M.

Cuoco, Esquire, Northeast Utilities

Service Company, Post Office Box 270,

Hartford CT 06141-0270

NRC Project Director: Phillip F.

McKee

Omaha Public Power District, Docket

No. 50-285, Fort Calhoun Station, Unit

No. 1, Washington County, Nebraska

Date of amendment request: May 17,

1996

Description of amendment request:

The proposed amendment would revise

the technical specifications (TS) to

relocate the operability requirements for

shock suppressors (snubbers) from the

TS to the Updated Safety Analysis

Report (USAR) and incorporate snubber

examination and testing requirements

into TS 3.3.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed change will relocate

operability requirements for shock

suppressors (snubbers) from the Technical

Specifications (TS) to the Updated Safety

Analysis Report (USAR) and/or plant

procedures. On July 16, 1993, the NRC issued

a Final Policy Statement on Technical

Specification Improvements for Nuclear

Power Reactors. The Final Policy Statement

contains four criteria which can be used to

determine which constraints on the design

and operation of nuclear power plants are

appropriate for inclusion in TS. The NRC has

incorporated these criteria into 10 CFR 50.36,

‘‘Technical specifications.’’ Snubbers do not

meet any of the four criteria for inclusion as

a Limiting Condition for Operations within

the TS, and therefore it is proposed that these

requirements be relocated from the TS. The

proposed change would not reduce or revise

any of the current requirements for snubber

operability, only relocate the requirements.

Any changes to the requirements contained

in the USAR and/or plant procedures can be

made without NRC approval only when the

changes meet the criteria of 10 CFR 50.59.

Changes to the snubber operability

requirements that do not meet the criteria of

10 CFR 50.59 must be approved by the NRC

by license amendment. Therefore, the

relocation of the requirements on snubber

operability from the TS to the USAR does not

increase the probability or consequences of

any accident previously analyzed.

The proposed change also deletes sections

of the TS which are redundant or in conflict

with the American Society of Mechanical

Engineers (ASME) Boiler and Pressure Vessel

Code. Snubbers are required to be examined

and tested in accordance with ASME Section

XI by 10 CFR 50.55a. The proposed change

will ensure that the TS implement ASME

Section XI examination and testing

requirements for snubbers in accordance

with 10 CFR 50.55a. Where differences

between the deleted sections of the TS and

ASME Section XI requirements exist, the

Section XI requirements are similar or more

conservative than the TS. For example,

although the functional test sample size

differs between the methodologies, both

ensure that a very high percentage of the

snubbers in the plant are operable within

acceptance limits. Therefore, the proposed

revision does not reduce the effectiveness of

snubber examination and testing.

The proposed change would not reduce the

operability requirements, acceptance criteria,

or examination and testing of snubbers.

Therefore, the proposed change would not

increase the probability or consequences of

an accident previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

There will be no physical alterations to the

plant configuration, changes to setpoint

values, or changes to the implementation of

setpoints or limits as a result of this proposed

change.

The proposed change deletes duplicate or

conflicting requirements between the TS and

the ASME Section XI. In these areas, the

proposed deletions would remove the TS

requirements and testing would be

conducted in accordance with ASME Section

XI as directed by 10 CFR 50.55a. Although

the requirements of ASME Section XI differ

from the TS in some cases, the differences do

not decrease the effectiveness of testing and

examination as compared to the TS

requirements. Other areas, such as snubber

operability requirements and service life

monitoring, which are presently addressed

by TS, but are not covered under ASME

Section XI, will be maintained in the USAR

so that these requirements cannot be deleted

without NRC approval.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any previously

evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed change does not reduce the

operability, examination, or testing

requirements for snubbers. Snubbers will still

be required to meet the requirements of

ASME Section XI and 10 CFR 50.55a except

where specific written relief has been granted

by the NRC. Therefore, the proposed change

does not involve a significant reduction in a

margin of safety.
Federal Register / Vol. 61, No. 168 / Wednesday, August 28, 1996 / Notices 44361

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration. Local Public Document Room

location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska

68102

Attorney for licensee: Perry D.

Robinson, Winston & Strawn, 1400 L

Street, N.W., Washington, DC 20005-

3502

NRC Project Director: William H.

Bateman

Omaha Public Power District, Docket

No. 50-285, Fort Calhoun Station, Unit

No. 1, Washington County, Nebraska

Date of amendment request: May 20,

1996

Description of amendment request:

The proposed amendment would revise

the technical specifications (TS) to

clarify surveillance test requirements of

TS 3.1, Tables 3-1, 3-2, 3-3, 3-3A, and

3-5.

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 changes to the Table of Contents are

administrative in nature to reflect the

removal of incore instrumentation

(Specification 2.10.3) from the TS by

Amendment 167 and for consistency.

Amendment 169 inadvertently reinserted

incore instrumentation back into the Table of

Contents.

The change to Specification 2.1.7(1)b is

necessary because the requirement to test the

signal to alarm meter relay located in

Specification 3.1, Table 3-3, Item 6 is being

deleted. The test, which verifies the high and

low pressurizer level alarm settings and the

pressurizer heater cutout function is

unnecessary. Operating experience has

shown that a shiftly pressurizer level

verification as proposed for Specification 3.1,

Table 3-3, Item 6.a is sufficient to detect any

level deviation and verify that operation is

within safety analyses assumptions. The

level alarms serve as early warning devices

but do not provide an accident mitigation

function. Replacing the monthly test with a

channel check is in accordance with NUREG1432, Combustion Engineering (CE),

Standard Technical Specifications (STS),

Surveillance Requirement (SR) 3.3.11.1 (post

accident monitoring instrumentation). The

monthly channel check supplements the

shiftly level verification.

The Basis of Specification 3.1 is revised to

clarify expectations regarding a channel

check of channels that are normally off scale

when the surveillance is required. In this

situation, the channel check only verifies that

they are off scale in the same direction. Off

scale low current loop channels are verified

to be reading at the bottom of the range and

not failed downscale. These statements are

taken from the Bases of CE STS SR 3.3.4.1

Engineered Safety Features Actuation System

(ESFAS) Instrumentation (Analog).

In addition, the Basis of Specification 3.1

is revised to clarify that power operated relief

valve (PORV) actuation is not required

during the channel functional test of the

PORV low temperature setpoint (Table 3-3,

Item 18.a). PORV actuation is not required

because it could depressurize the reactor

coolant system. This clarification is modeled

after a similar statement from the Bases of SR

3.4.12.6 (Low Temperature Overpressure

Protection (LTOP) System) of the CE STS.

Changing Specification 3.1, Tables 3-1, 3-

2, 3-3, and 3-3A by using defined terms to

enable the Surveillance Method to match the

Surveillance Function is an administrative

change designed to simplify the tables.

Removal of the extraneous text does not alter

the surveillance because the defined terms

are equivalent in meaning to the deleted text.

The reordering of several items in the

tables into a Check-Test-Calibrate sequence

adds consistency to the tables. Text revisions

in the Channel Description or Surveillance

Function columns of Tables 3-1 and 3-2 add

clarity and/or consistency. Footnote No. 1 in

Table 3-1 concerning the bistable trip tester

was deleted because it is unnecessary.

The Surveillance Function of Table 3-1,

Item 1.c (Power Range Safety Channels) is

being changed to ‘‘Test’’ from ‘‘Calibrate and

Test.’’ It is not necessary for Item 1.c to

require both because Item 1.b already

requires the power range safety channel

adjustment (calibration) to be performed

daily. As stated in the Basis of Specification

3.1, ‘‘The minimum calibration frequencies

of once-per-day for the power range safety

channels, ...are considered adequate.’’ To

further clarify the issue, the Basis of

Specification 3.1 is being revised to note that

the daily calibration is a heat balance

adjustment only.

Changing Table 3-1, Item 4 (Thermal

Margin/Low Pressure (TM/LP)) to use the

defined term CHANNEL CALIBRATION will

allow OPPD to relax the current TM/LP

calibration requirements with a negligible

impact on safety. Calibration of the

temperature input and pressure input will

still require calibration to known standards

(i.e., resistance and pressure), but will allow

the calibrations to be done separately instead

of coincidently. The channel functional test

that follows the channel calibration verifies

proper function of the TM/LP circuitry.

Removing the word ‘‘Instruments’’ from

the Channel Description of Table 3-2, Item 14

makes the Channel Description consistent

with the Surveillance Method. Table 3-2,

Item 14 is not intended to verify safety

injection tank (SIT) instrumentation

operability but rather that the parameters

level and pressure are within limits. Generic

Letter (GL) 93-05, Item 7.4, states that the

operability of SIT instrumentation is not

directly related to the capability of a SIT to

perform its safety function. GL 93-05

concludes that the surveillance should only

confirm that the parameters defining SIT

operability are within their specified limits.

Items 22 & 24 are being added to Table 3-

2 to clearly state the requirement for testing

manual actuation of the Engineered Safety

Features (ESF) channels for Off-site Power

Low Signal (OPLS) and Auxiliary Feedwater.

Although testing manual actuation of these

channels is done via the existing

Specifications, the requirement to do so is

not clearly stated. Reordering Table 3-2, Item

23 into a Check-Test-Calibrate Surveillance

Frequency sequence adds clarity and

consistency.

The addition of Footnote No. 7 to Table 3-

2 clarifies that the refueling frequency ESF

channel functional test pertains to the

backup channels such as derived circuits and

equipment that cannot be tested when the

plant is at power. Operating certain relays

during power operation could cause plant

transients or equipment damage.

The revisions to Table 3-3, Item 6, clarify

that pressurizer level is the parameter to be

verified and not the pressurizer level

instruments. The revision to Item 6.a is

consistent with CE STS SR 3.4.9.1

(pressurizer water level). Reordering Item 6

into a Check-Test-Calibrate Surveillance

Function sequence makes Item 6 consistent

with the ordering of the other items in Table

3-3. The requirement to test the signal to

alarm meter relay currently located in

Specification 3.1, Table 3-3, Item 6.c is

unnecessary. Operating experience has

shown that a shiftly pressurizer level

verification as proposed for Specification 3.1,

Table 3-3, Item 6.a is sufficient to detect any

level deviation and verify that operation is

within safety analyses assumptions. Thus,

the monthly ‘‘Test’’ requirement will be

replaced with a ‘‘Check’’ to supplement the

less formal but more frequent shiftly level

verification of Item 6.a.

Table 3-3, Items 21 (PORV Operation &

Acoustic Position Indication Channel) and 23

(Safety Valve Acoustic Position Indication

Channel) should be revised to a channel

functional test from a channel/circuit check.

An oscillator and installed impactors are

used to generate noise signals and therefore,

this surveillance is more accurately described

as a channel functional test rather than a

channel check.

Table 3-3, Items 21 and 22 (PORV Block

Valve Operation & Position Indication)

should have the requirement to verify

operation on the emergency power supply

deleted. Permanent Class 1E power supplies

the PORV and PORV Block Valve. Therefore,

verification of PORV or PORV Block Valve

operability while powered from the

emergency power supply system provides no

additional benefit. (Operability of the

emergency power supply system is tested in

accordance with Specification 3.7.) The

proposed revision is in accordance with the

exception for plants with a permanent Class

1E power supply to these valves as stated in

CE STS, SR 3.4.11.4.

Deletion of the requirement of TS 3.2,

Table 3-5, Item 15, to test spent fuel pool

surveillance coupons for a change in

hardness corrects an oversight in the

Application for Amendment dated December

7, 1992.

As stated in the Safety Evaluation Report

enclosed with Amendment 155, ‘‘Each
44362 Federal Register / Vol. 61, No. 168 / Wednesday, August 28, 1996 / Notices

coupon, upon its removal from the mounting

jacket, will be analyzed according to the

following tests:

visual observation and photography

neutron attenuation

dimensional measurements (length, width,

and thickness)

weight and specific gravity.’’

The tests listed above are sufficient to

detect degradation of the Boral— material

and do not require that the surveillance

coupons be tested for hardness.

Based on the above discussion, the

proposed changes clarify and standardize

existing surveillance requirements, remove

redundant requirements, correct minor

oversights from previous amendment

requests or are in accordance with CE STS.

Thus, none of the requested changes 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 revisions will not result in

any physical alterations to the plant

configuration, changes to setpoint values, or

changes to the application of setpoints or

limits. 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 change does not involve

a significant reduction in a margin of safety.

The proposed changes clarify existing

surveillance requirements, remove redundant

requirements, correct minor oversights from

previous amendment requests or are in

accordance with CE STS. Thus, none of the

requested changes involves a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska

68102

Attorney for licensee: Perry D.

Robinson, Winston & Strawn, 1400 L

Street, N.W., Washington, DC 20005-

3502

NRC Project Director: William H.

Bateman

Pennsylvania Power and Light

Company, Docket No. 50-388

Susquehanna Steam Electric Station,

Unit 2, Luzerne County, Pennsylvania

Date of amendment request: May 20,

1996, as supplemented by letter dated

July 25, 1996

Description of amendment request:

This amendment request would modify

the Technical Specifications for the unit

by: changing the Minimum Critical

Power Ratio safety limit values, adding

a reference to reflect the use of the ANFB Critical Power Correlation, and

modifying the associated Bases. 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 change to the ANFB correlation and

corresponding MCPR Safety Limits does not

physically change the plant systems,

structures, or components. Thus, the

probability of an event evaluated in the SAR

is not increased. The acceptance criterion for

the MCPR Safety Limit (i.e., 99.9% of the fuel

rods expected to avoid boiling transition) is

not changed. Only the methodology used to

demonstrate compliance is changed.

Therefore, the consequences of anticipated

operational occurrences (which must show

the Safety Limit is not violated) are not

changed. Results of incorporating this change

will not significantly increase 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.

As stated above, this methodology change

does not impact the acceptance criteria for

the MCPR Safety Limits and does not

physically change the plant systems,

structures, or components. Since no changes

to the physical plant are being made, this

change does not create the possibility of a

new or different kind of accident from any

accident previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

A cycle specific MCPR Safety Limit

analysis was performed by SPC [Siemens

Power Corporation]. This analysis used NRC

approved methods described in the SPC

report: ANF-524(P)(A), Revision 2 and

Supplement 1, Revision 2. The MCPR Safety

Limit value is calculated such that at least

99.9% of the fuel rods are expected to avoid

boiling transition during normal operation or

anticipated operation occurrences. Both the

existing analysis using XN-3 and the new

analysis using ANFB utilize NRC approved

methods to accomplish this same objective.

Therefore, the change to an ANFB based

Safety Limit 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: Osterhout Free Library,

Reference Department, 71 South

Franklin Street, Wilkes-Barre, PA 18701

Attorney for licensee: Jay Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street NW.,

Washington, DC 20037

NRC Project Director: John F. Stolz

TU Electric Company, Docket Nos. 50-445

and 50-446, Comanche Peak Steam Electric

Station (CPSES), Units 1 and 2, Somervell

County, Texas

Date of amendment request: July 31,

1996

Brief description of amendments:

Based on analyses of the core

configuration and expected operation

for CPSES Unit 1, Cycle 6, the proposed

amendments would revise core safety

limit curves and Overtemperature N-16

reactor trip setpoints. In addition, the

TU Electric Small Break LOCA Topical

Report on the Core Operating Limits

Report Technical Specification is

incorporated. The topical report change

is applicable to both Units.

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.a. Revision to the Unit 1 Core Safety

Limits

Analyses of reactor core safety limits are

required as part of reload calculations for

each cycle. TU Electric has performed the

analyses of the Unit 1, Cycle 6 core

configuration to determine the reactor core

safety limits. The methodologies and safety

analysis values result in new operating

curves which, in general, permit plant

operation over a similar range of acceptable

conditions. This change means that if a

transient were to occur with the plant

operating at the limits of the new curve, a

different temperature and power level might

be attained than if the plant were operating

within the bounds of the old curves.

However, since the new curves were

developed using NRC approved

methodologies which are wholly consistent

with and do not represent a change in the

Technical Specification BASES for safety

limits, all applicable postulated transients

will continue to be properly mitigated. As a

result, there will be no significant increase in

the consequences, as determined by accident

analyses, of any accident previously

evaluated.

1.b. Revision to Unit 1 Overtemperature N16 Reactor Trip Setpoints, Parameters and

Coefficients

As a result of changes discussed, the

Overtemperature N-16 reactor trip setpoint

has been recalculated. These trip setpoints

help ensure that the core safety limits are

maintained and that all applicable limits of

the safety analysis are met.

Based on the calculations performed, the

safety analysis value for Overtemperature N16 reactor trip setpoint has changed. This

essentially means if a transient were to occur,

the actual temperature and power level

achievable prior to initiating a reactor trip

could be slightly higher. However, the

analyses performed show that, using the TU

Electric methodologies, all applicable limits

of the safety analysis are met. This setpoint
Federal Register / Vol. 61, No. 168 / Wednesday, August 28, 1996 / Notices 44363

provides a trip function which allows the

mitigation of postulated accidents and has no

impact on accident initiation. Therefore, the

changes in safety analysis values do not

involve an increase in the probability of an

accident and, based on satisfying all

applicable safety analysis limits, there is no

significant increase in the consequences of

any accident previously evaluated.

In addition, sufficient operating margin has

been maintained in the overtemperature

setpoint such that the risk of turbine

runbacks or reactor trips due to upper

plenum flow anomalies or other operational

transients will be minimized, thus reducing

potential challenges to the plant safety

systems.

1.c. Incorporation of TU Electric Small

Break LOCA Topical Report, RXE-95-0001-P.

TU Electric has submitted the topical

report ‘‘Small Break Loss of Coolant Accident

Analysis Methodology,’’ RXE-95-001-P and

plans to use the report to support Unit 1

Cycle 6. In order to accomplish this activity,

it is necessary to include the topical report

in the list of NRC-approved methodologies in

Technical Specification 6.9.1.6b. Use of this

topical report is contingent upon NRC

approval; therefore, inclusion of this report

in Section 6 of the Technical Specifications

is administrative in nature and does not

change the probability or consequences of an

accident.

2. The proposed changes involve the use of

revised safety analysis values and the

calculation of new reactor core safety limits

and reactor trip setpoints. As such, the

changes play an important role in the

analysis of postulated accidents but none of

the changes effect plant hardware or the

operation of plant systems in a way that

could initiate an accident. Therefore, the

proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. In reviewing and approving the methods

used for safety analyses and calculations, the

NRC has approved the safety analysis limits

which establish the margin of safety to be

maintained. While the actual impact on

safety is discussed in response to question 1,

the impact on margin of safety is discussed

below:

3.a.

Revision to the Unit 1 Reactor Core

Safety Limits

The TU Electric reload analysis methods

have been used to determine new reactor

core safety limits. All applicable safety

analysis limits have been met. The methods

used are wholly consistent with Technical

Specification BASES 2.1 which is the bases

for the safety limits. In particular, the curves

assure that for Unit 1, Cycle 6, the calculated

DNBR is no less than the safety analysis limit

and the average enthalpy at the vessel exit is

less than the enthalpy of saturated liquid.

The acceptance criteria remains valid and

continues to be satisfied; therefore, no change

in a margin of safety occurs.

3.b. Revision to Unit 1 Overtemperature N16 Reactor Trip Setpoints, Parameters and

Coefficients

Because the reactor core safety limits for

CPSES Unit 1, Cycle 6 are recalculated, the

Reactor Trip System instrumentation setpoint

values for the Overtemperature N-16 reactor

trip setpoint which protect the reactor core

safety limits must also be recalculated. The

Overtemperature N-16 reactor trip setpoint

helps prevent the core and Reactor Coolant

System from exceeding their safety limits

during normal operation and design basis

anticipated operational occurrences. The

most relevant design basis analysis in

Chapter 15 of the CPSES Final Safety

Analysis Report (FSAR) which is affected by

the change in the safety analysis value for the

CPSES Unit 1 Overtemperature N-16 reactor

trip setpoint is the Uncontrolled Rod Cluster

Control Assembly Bank Withdrawal at Power

(FSAR Section 15.4.2). This event has been

re-analyzed with the revised safety analysis

value for the Overtemperature N-16 reactor

trip setpoint to demonstrate compliance with

event specific acceptance criteria. Because all

event acceptance criteria are satisfied, there

is no degradation in a margin of safety.

The nominal Reactor Trip System

instrumentation setpoints values for the

Overtemperature N-16 reactor trip setpoint

(Technical Specification Table 2.2-1) are

determined based on a statistical

combination of all of the uncertainties in the

channels to arrive at a total uncertainty. The

total uncertainty plus additional margin is

applied in a conservative direction to the

safety analysis trip setpoint value to arrive at

the nominal and allowable values presented

in Technical Specification Table 2.2-1.

Meeting the requirements of Technical

Specification Table 2.2-1 assures that the

Overtemperature N-16 reactor trip setpoint

assumed in the safety analyses remains valid.

The CPSES Unit 1, Cycle 6 Overtemperature

N-16 reactor trip setpoint is different from

previous cycles which provides more

operational flexibility to withstand mild

transients without initiating automatic

protective actions. Although the setpoint is

different, the Reactor Trip System

instrumentation setpoint values for the

Overtemperature N-16 reactor trip setpoint

are consistent with the safety analysis

assumption which has been analytically

demonstrated to be adequate to meet the

applicable event acceptance criteria. Thus,

there is no reduction in a margin of safety.

3.c. Revise 6.9.1.6b to include Topical

Report RXE-95-001-P, ‘‘Small Break Loss of

Coolant Accident Methodology’’

TU Electric has submitted the topical

report ‘‘Small Break Loss of Coolant Accident

Analysis Methodology,’’ RXE-95-001-P and

plans to use the report to support Unit 1

Cycle 6. In order to accomplish this activity,

it is necessary to include the topical report

in the list of NRC-approved methodologies in

Technical Specification 6.9.1.6b. Use of this

topical report is contingent upon NRC

approval; therefore, inclusion of this report

in Section 6 of the Technical Specifications

is administrative in nature and does not

reduce the margin of safety.

Using the NRC approved TU Electric

methods, the reactor core safety limits are

determined such that all applicable limits of

the safety analyses are met. Because the

applicable event acceptance criteria continue

to be met, there is no significant reduction in

the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration. Local Public Document Room

location: University of Texas at

Arlington Library, Government

Publications/Maps, 702 College, P.O.

Box 19497, Arlington, TX 76019 Attorney for licensee: George L. Edgar,

Esq., Morgan, Lewis and Bockius, 1800

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

NRC Project Director: William D.

Beckner

TU Electric Company, Docket Nos. 50-

445 and 50-446, Comanche Peak Steam

Electric Station (CPSES), Units 1 and 2,

Somervell County, Texas

Date of amendment request: July 31,

1996

Brief description of amendments: The

proposed amendments would revise the

Technical Specifications by (1) changing

the battery charger ratings; (2) by

clarifying the meaning of the term

‘‘associated inverter’’; and by (3)

deleting the protection channel and the

vital bus ratings for the instrument

busses identified for Mode 1 through 4.

These changes are associated with a

plant modification in which the

inverters and battery chargers are being

replaced and an installed spare inverter

is being added for each safety train.

These changes are equally applicable to

CPSES Units 1 and 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. DO THE PROPOSED CHANGES

INVOLVE A SIGNIFICANT INCREASE IN

THE PROBABILITY OR CONSEQUENCES

OF AN ACCIDENT PREVIOUSLY

EVALUATED?

CHANGE TO IDENTIFY BATTERY

CHARGER RATINGS

The first proposed change replaces the test

amperes with the design value for the

replacement battery charger and allows a

voltage range (greater than or equal to 130

volts) instead of a single value. The intent of

the surveillance requirement or the

surveillance frequency is not changed. The

replacement inverters and battery chargers

will continue to provide the capacity needed

to perform the required safety functions. The

revised surveillance will continue to assure

that the battery chargers are capable of

performing as designed. Therefore this

change does not impact the probability or the

consequences of an accident previously

evaluated.

CLARIFICATION TO DEFINE

ASSOCIATED INVERTER

The second proposed change adds a foot

note to clarify the term ‘‘associated inverter’’

by describing it as, ’’... the dedicated inverter

or installed spare inverter.’’ Also the Bases
44364 Federal Register / Vol. 61, No. 168 / Wednesday, August 28, 1996 / Notices

for this specification is revised to reflect the

basis for this change. This change allows use

of an installed spare inverter (for each train)

having the capability to energize the

Instrument Bus for the protection channel or

the vital bus. Procedural controls and

interlocks ensure that the spare is available

to feed only one of the protection channel or

vital bus Instrument Bus at a time, in the

event the dedicated inverter is not available.

Procedural controls and interlocks also

ensure that the installed spare inverter is fed

from the same power source as that of the

dedicated inverter not in service and whose

loads are being fed by the spare inverter. This

proposed design only allows the spare

inverter for a safety train to be manually

aligned to replace only one of the four

inverters in that train at a time.

The installation of a spare inverter for each

train and the associated design configuration

increases the availability of energized

Instrument Bus for the protection channel

and vital bus. These changes do not involve

an increase in the probability or

consequences of an accident previously

evaluated.

DELETION OF THE PROTECTION

CHANNEL AND VITAL BUS RATINGS FOR

INSTRUMENT BUS

The third proposed change deletes

specifying of the protection channel and vital

bus KVA ratings for the Instrument Bus. The

ratings of inverter that feeds these instrument

buses are being described in other Licensing

Bases Documents or Design Basis Documents.

There is no change proposed to the intent of

the action statements.

This is considered an administrative

change and does not impact the probability

or consequences of an accident previously

evaluated.

2. DO THE PROPOSED CHANGES

CREATE THE POSSIBILITY OF A NEW OR

DIFFERENT KIND OF ACCIDENT FROM

ANY ACCIDENT PREVIOUSLY

EVALUATED?

CHANGE TO IDENTIFY BATTERY

CHARGER RATINGS

Replacing the inverters and battery

chargers and changing the parameters of the

battery charger surveillance test to match the

replacement chargers does not alter the

functional modes of this portion of the design

and does not result in any new failure modes.

As such, it does not create the possibility of

a new or different accident from any

previously evaluated.

CLARIFICATION TO DEFINE

ASSOCIATED INVERTER

The second proposed change allows use of

an installed spare inverter for each train to

energize the one of the Instrument Bus for the

protection channel and vital bus at a time for

the respective safety train while its dedicated

inverter is not available. The spare inverter

is such that it has the capability to support

the maximum load for the protection channel

or vital bus. Manually aligning the installed

inverter to replace on[e] of the dedicated

inverters is essentially equivalent to a repair

activity which replaces a faulted inverter

with a new inverter. In addition, procedural

controls and interlocks are provided to

ensure the proper alignment of the installed

spare when it is used. The proposed changes

do not create the possibility of a new or

different accident from any previously

evaluated.

DELETION OF THE PROTECTION

CHANNEL AND VITAL BUS RATINGS FOR

INSTRUMENT BUS

The third proposed change as discussed

earlier does not change intent of the

Technical Specifications action statements.

This is an administrative change which does

not introduce new failure modes and has no

new or different accidents from any

previously evaluated are created.

3. DO THE PROPOSED CHANGES

INVOLVE A SIGNIFICANT REDUCTION IN

MARGIN OF SAFETY?

The relevant Technical Specification

sections proposed for changes: (1) ensure that

the battery charger is capable of charging the

battery by performing the surveillance at 18

month frequency; (2) establish operability

requirements of the Instrument Bus for the

protection channel and vital bus in MODES

1 through 6; and (3) identify the actions

required for not meeting item 2.

These proposed changes do not alter the

intent of the above requirements; however

replacement of the currently installed

inverters with inverters which are expected

to be more reliable and available and the

addition of a spare inverter per safety train

to energize Instrument Bus for protection

channel and vital bus does increase the

reliability of the instrument busses for the

train. Allowing credit for this spare inverter

in meeting the operability requirements of

Instrument Bus for the protection channel

and vital bus, minimize potential plant

shutdowns due to non-energized instrument

from its dedicated inverter. These changes do

not involve a significant reduction in margin

of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration. Local Public Document Room

location: University of Texas at

Arlington Library, Government

Publications/Maps, 702 College, P.O.

Box 19497, Arlington, TX 76019 Attorney for licensee: George L. Edgar,

Esq., Morgan, Lewis and Bockius, 1800

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

NRC Project Director: William D.

Beckner

Vermont Yankee Nuclear Power

Corporation, Docket No. 50-271,

Vermont Yankee Nuclear Power

Station, Vernon, Vermont

Date of amendment request: August 9,

1996

Description of amendment request:

The proposed amendment would revise

the Safety Limits for Minimum Critical

Power Ratio (MCPR) based upon a

Vermont Yankee plant and cycle

specific analysis, performed by General

Electric. The revised MCPR Safety

Limits are needed to accommodate

Vermont Yankee’s core design for

upcoming refueling cycle number 19.

Specifically, the MCPR Safety Limits of

1.07 and 1.08 in the Vermont Yankee

Technical Specifications (TS) section

1.1.A are proposed to be increased to

1.10 and 1.12 for two loop and single

loop operation, respectively. Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) The Safety Limit Minimum Critical

Power Ratio (MCPR) is defined to ensure that

during normal operation and Anticipated

Operational Transients (AOTs), at least

99.9% of the fuel rods in the core do not

experience transition boiling. Core MCPR

operating limits are developed to ensure

these Safety Limits are maintained in the

event of the worst case transient. Since the

Safety Limit MCPR will be maintained at all

times, operation under the proposed changes

will ensure at least 99.9% of the fuel rods in

the core do not experience transition boiling

and no significant radiological release will

result. Therefore, this Safety Limit MCPR

change does not affect the probability or

consequences of a previously evaluated

accident.

(2) The proposed changes do not involve

any new modes of operation or any plant

modifications. Establishment and monitoring

of the operating limits will continue as per

established procedure. The proposed changes

to these limits do not result in the creation

of any new precursors to an accident.

Therefore, the proposed change does not

create the possibility of a new or a different

kind of accident from any previously

analyzed.

(3) The Safety Limit MCPR values were

evaluated by General Electric based upon a

cycle specific Vermont Yankee analysis,

using NRC approved methods. The resulting

limits are more conservative than the

previous generic limits and will continue to

assure that at least 99.9% of the fuel rods in

the core do not experience transition boiling

during analyzed transients. This acceptance

criteria ensures the safety design limit of ‘‘no

damage to a nuclear system process barrier

shall result from forces associated with

AOTs.’’ Therefore, the implementation of the

proposed change does not involve a

significant reduction in [a] margin of safety.

The NRC staff has reviewed the

licensee’s analysis. The staff notes that,

although the proposed change does not

involve a plant modification, the reason

for the proposed higher safety limit

MCPRs is the cycle-specific core design

and the local power distribution in the

slightly higher enriched fresh GE-9B

fuel bundles. This new fuel will be

loaded during the September/October

1996 refueling outage. In conjunction

with the proposed safety limit MCPRs

and the core operating limits

determined in accordance with Vermont

Yankee TS 6.7.A.4, the new fuel load

will not involve a significant increase in

the probability or consequences of an
Federal Register / Vol. 61, No. 168 / Wednesday, August 28, 1996 / Notices 44365

accident previously evaluated nor a

significant reduction in a margin of

safety. In addition, the new fuel load

does not create the possibility of a new

or different kind of accident from any

accident previously evaluated. 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: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301

Attorney for licensee: R. K. Gad, III,

Ropes and Gray, One International

Place, Boston, MA 02110-2624

NRC Project Director: Jocelyn A.

Mitchell, Acting Director

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.

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 modify the

definition of ‘‘Core Alteration,’’ and the

limiting condition for operation,

Surveillance conditions and Bases

section associated with Technical

Specification 3.7.C, ‘‘Secondary

Containment.’’

Date of issuance: August 12, 1996

Effective date: August 12, 1996

Amendment No.: 166

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 August 12, 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,

Docket No. 50-261, H. B. Robinson

Steam Electric Plant, Unit No. 2,

Darlington County, South Carolina

Date of application for amendment:

July 17, 1995, as supplemented May 2,

1996, and July 1, 1996.

Brief description of amendment: The

change revises technical specification

(TS) section 3.8 to specify that the spent

fuel building refueling filter fan and at

least one containment purge fan shall be

shown to operate within plus or minus

10 percent of the design flow.

Date of issuance: August 6, 1996

Effective date: August 6, 1996

Amendment No. 172

Facility Operating License No. DPR23. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: September 13, 1995 (60 FR

47615). The May 2, and July 1, 1996,

letters provided clarifying information

that did not affect the proposed no

significant hazards consideration. The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated August 6, 1996. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Hartsville Memorial Library,

147 West College Avenue, Hartsville,

South Carolina 29550

Carolina Power & Light Company,

Docket No. 50-261, H. B. Robinson

Steam Electric Plant, Unit No. 2,

Darlington County, South Carolina

Date of application for amendment:

June 6, 1996

Brief description of amendment: The

amendment revises technical

specifications (TS) Section 4.2.3 to

allow the licensee to defer the ultrasonic

inspection of the reactor coolant pump

flywheel for one operating cycle. Date of issuance: August 9, 1996 Effective date: August 9, 1996 Amendment No. 173

Facility Operating License No. DPR23. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: July 3, 1996 (61 FR 34888) The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated August 9, 1996. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Hartsville Memorial Library,

147 West College Avenue, Hartsville,

South Carolina 29550

Carolina Power & Light Company, et

al., Docket No. 50-400, Shearon Harris

Nuclear Power Plant, Unit 1, Wake and

Chatham Counties, North Carolina

Date of application for amendment:

May 31, 1996 Brief description of amendment: The

amendment revises Technical

Specifications (TS) Table 3.3-7, Seismic

Monitoring Instrumentation, and TS

Table 4.3-4, Seismic Monitoring

Instrumentation Surveillance

Requirements, to correct the location

described for one of the three Triaxial

Peak Accelerograph recorders. Date of issuance: August 7, 1996 Effective date: August 7, 1996

Amendment No. 66

Facility Operating License No. NPF63. Amendment revises the Technical

Specifications

Date of initial notice in Federal

Register: July 3, 1996 (61 FR 34888) The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated August 7, 1996. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh,

North Carolina 27605.

Commonwealth Edison Company,

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

County Station, Units 1 and 2, LaSalle

County, Illinois

Date of application for amendments:

April 16, 1996

Brief description of amendments: The

amendments revise the Technical
44366 Federal Register / Vol. 61, No. 168 / Wednesday, August 28, 1996 / Notices

Specifications (TSs) to eliminate

selected response time testing

requirements based on analyses

performed by the Boiling Water Reactor

Owners’ Group as documented in

NEDO-32291. The affected TS sections

are 3/4.3.1, ‘‘Reactor Protection System

Instrumentation;’’ 3/4.3.2, ‘‘Isolation

Actuation Instrumentation;’’ and 3/

4.3.3, ‘‘Emergency Core Cooling System

Actuation Instrumentation.’’

Date of issuance: August 14, 1996

Effective date: Immediately, to be

implemented within 60 days.

Amendment Nos.: 114 and 99

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

revised the Technical Specifications.

Date of initial notice in Federal

Register: May 22, 1996 (61 FR 25702)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated August 14, 1996. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Jacobs Memorial Library,

Illinois Valley Community College,

Oglesby, Illinois 61348.

Detroit Edison Company, Docket No.

50-341, Fermi-2, Monroe County,

Michigan

Date of application for amendment:

December 21, 1995

Brief description of amendment: The

amendment revises the Technical

Specifications (TS) to implement 10

CFR Part 50, Appendix J - Option B, by

referring to Regulatory Guide 1.163,

‘‘Performance-Based Containment LeakTest Program.’’ Specifically, changes

have been made to TS Section 3/4.6.1.2,

‘‘Primary Containment Leakage,’’ TS 3/

4.6.1.3, ‘‘Primary Containment Air

Locks,’’ TS 3/4.6.1.5, ‘‘Primary

Containment Structural Integrity,’’ TS

6.0, ‘‘Administrative Controls,’’ and

their associated Bases.

Date of issuance: August 8, 1996

Effective date: August 8, 1996, with

full implementation within 45 days.

Amendment No.: 108

Facility Operating License No. NPF43. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: February 28, 1996 (61 FR

7551) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

August 8, 1996. No significant hazards

consideration comments received: No.

Local Public Document Room

location: Monroe County Library

System, 3700 South Custer Road,

Monroe, Michigan 48161

Florida Power and Light Company,

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

Point Plant Units 3 and 4, Dade County,

Florida

Date of application for amendments:

April 19, 1996, and supplements dated

May 10 and May 28, 1996. Brief description of amendments: The

amendment changes the Technical

Specifications to address frequency

extension on a periodic basis, deletes

separate notification requirements for an

inoperable startup transformer, and

allows the operating residual heat

removal loop to be removed from

operation, under certain conditions,

during refueling. Date of Issuance: August 6, 1996 Effective Date: August 6, 1996 Amendment Nos.: 189 and

183Facility Operating Licenses Nos.

DPR-31 and DPR-41: Amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: July 3, 1996 (61 FR 34892) The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated August 6, 1996. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Florida International

University, University Park, Miami,

Florida 33199.

Florida Power and Light Company,

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

Point Plant Units 3 and 4, Dade County,

Florida

Date of application for amendments:

July 26, 1995, and supplemented March

13, May 3, and May 9, 1996. Brief description of amendments:

Change TS 6.9.1.7, Core Operating

Limits Report, resulting from a

reanalysis of the small break loss-ofcoolant accident for the Turkey Point

Units using the NOTRUMP code

including the COSI safety injection (SI)

condensation model.

Date of issuance: August 13, 1996 Effective date: August 13, 1996 Amendment Nos. 190 and 184Facility

Operating Licenses Nos. DPR-31 and

DPR-41: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: September 13, 1995 (60 FR

47618). The supplements dated March

13, May 3, and May 9, 1996 provided

clarifying information that did not

change the initial proposed no

significant hazards consideration

determination. The Commission’s

related evaluation of the amendments is

contained in a Safety Evaluation dated

August 13, 1996. No significant hazards

consideration comments received: No

Local Public Document Room

location: Florida International

University, University Park, Miami,

Florida 33199.

Houston Lighting & Power Company,

City Public Service Board of San

Antonio, Central Power and Light

Company, City of Austin, Texas, Docket

Nos. 50-498 and 50-499, South Texas

Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request: May 1,

1996

Brief description of amendments: The

amendments changed the technical

specifications to implement 10 CFR Part

50, Appendix J, Option B, by referring

to Regulatory Guide 1.163,

‘‘Performance-Based Containment LeakTest Program.’’ Part of the requested

change, that regarding the frequency of

leakage rate testing the normal

containment purge valves and the

supplementary containment purge

valves, was denied.

Date of issuance: August 13, 1996

Effective date: August 13, 1996

Amendment Nos.: 84 and 71

Facility Operating License Nos. NPF76 and NPF-80. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: June 5, 1996 (61 FR 28616)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated August 13, 1996. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Wharton County Junior

College, J. M. Hodges Learning Center,

911 Boling Highway, Wharton, TX

77488

IES Utilities Inc., Docket No. 50-331,

Duane Arnold Energy, Center, Linn

County, Iowa

Date of application for amendment:

November 30, 1995

Brief description of amendment: The

amendment implements the Option I-D

long-term stability solution and removes

the existing SIL-380 Rev. 1-based

specifications. In addition, the

amendment requires a plant scram be

initiated should the plant enter natural

circulation conditions and prohibits

restarting a recirculation pump while in

natural circulation. Finally, this

amendment deletes Technical

Specification (TS) actions and

surveillance requirements related to

core plate differential pressure noise

while in single recirculation pump

operation (SLO).

Date of issuance: August 7, 1996

Effective date: August 7, 1996

Amendment No.: 215

Facility Operating License No. DPR49: Amendment revised the Technical

Specifications.
Federal Register / Vol. 61, No. 168 / Wednesday, August 28, 1996 / Notices 44367

Date of initial notice in Federal

Register: March 13, 1996 (61 FR 10394)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated August 7, 1996. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Cedar Rapids Public Library,

500 First Street, S. E., Cedar Rapids,

Iowa 52401

IES Utilities Inc., Docket No. 50-331,

Duane Arnold Energy, Center, Linn

County, Iowa

Date of application for amendment:

November 15, 1995, as supplemented

April 9, 1996

Brief description of amendment: The

amendment revises the requirements for

the End of Cycle Recirculation Pump

Trip logic to match more closely the

assumptions applicable to the turbine

trip events for which it was installed.

The surveillance requirements are also

revised, based on those same

assumptions.

Date of issuance: August 8, 1996

Effective date: August 8, 1996

Amendment No.: 216

Facility Operating License No. DPR49: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 22, 1996 (61 FR 1629)

The April 9, 1996, submittal was

clarifying in nature and did not affect

the no significant hazards

determination. The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

August 8, 1996. No significant hazards

consideration comments received: No.

Local Public Document Room

location: Cedar Rapids Public Library,

500 First Street, S. E., Cedar Rapids,

Iowa 52401

IES Utilities Inc., Docket No. 50-331,

Duane Arnold Energy, Center, Linn

County, Iowa

Date of application for amendment:

January 18, 1996

Brief description of amendment: The

amendment revises the setpoint at

which the Reactor Water Cleanup

(RWCU) system isolates, based on

reactor vessel water level. In particular,

the amendment changes the Group 5

isolation from isolating on ‘‘reactor

water level low’’ to ‘‘reactor water level

low-low.’’

Date of issuance: August 8, 1996

Effective date: August 8, 1996, and

shall be implemented prior to startup

from RFO 14.

Amendment No.: 217

Facility Operating License No. DPR49: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 14, 1996 (61 FR

5814) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

August 8, 1996. No significant hazards

consideration comments received: No.

Local Public Document Room

location: Cedar Rapids Public Library,

500 First Street, S. E., Cedar Rapids,

Iowa 52401

Indiana Michigan Power Company,

Docket Nos. 50-315 and 50-316, Donald

C. Cook Nuclear Plant, Unit Nos. 1 and

2, Berrien County, Michigan

Date of application for amendments:

January 12, 1996 (AEP:NRC:1233)

Brief description of amendments: The

amendments modify the Technical

Specifications to delete the surveillance

requirement demonstrating operability

of the emergency power supply for the

pressurizer power operated relief valves

and block valves.

Date of issuance: August 15, 1996

Effective date: August 15, 1996, with

full implementation within 45 days

Amendment Nos.: 211 and 196

Facility Operating License Nos. DPR58 and DPR-74. Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: February 28, 1996 (61 FR

7554) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

August 15, 1996. No significant hazards

consideration comments received: No.

Local Public Document Room

location: Maud Preston Palenske

Memorial Library, 500 Market Street, St.

Joseph, Michigan 49085

Niagara Mohawk Power Corporation,

Docket No. 50-410, Nine Mile Point

Nuclear Station, Unit 2, Oswego

County, New York

Date of application for amendment:

February 7, 1996, as supplemented July

26, 1996.

Brief description of amendment: The

amendment revises the operating

license, TSs and associated Bases to

implement Option B ‘‘PerformanceBased Requirements’’ of Appendix J to

10 CFR Part 50 for Type A, B, and C

leakage rate testing.

Date of issuance: August 13, 1996

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 74

Facility Operating License No. NPF69: Amendment revises the Technical

Specifications and operating license.

Date of initial notice in Federal

Register: May 8, 1996 (61 FR 20849) The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated August 13, 1996. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

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:

July 3, 1996

Brief description of amendment: The

amendment removes, on a one-time

basis during the cycle 13 mid-cycle

offload/reload activities, the Technical

Specification (TS) requirement that the

boron concentration in all filled

portions of the reactor coolant system be

‘‘uniform.’’ The requested change also

adds a footnote indicating that it is

acceptable for the boron concentration

of the water volumes in the steam

generators and the connecting piping to

be as low as 1300 parts per million. The

TS Bases are also updated to reflect the

one-time TS change.

Date of issuance: August 12, 1996

Effective date: As of the date of

issuance, to be implemented within 30

days.

Amendment No.: 201

Facility Operating License No. DPR65. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 11, 1996 (61 FR 36583)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated August 12, 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 Waterford

Library, ATTN: Vince Juliano, 49 Rope

Ferry Road, Waterford, CT 06385

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:

February 29, 1996

Brief description of amendments:

These amendments relocate

Specification 3/4.9.6, ‘‘Refueling

Platform,’’ to the Susquehanna Steam

Electric Station Technical Requirements

Manual, a document which is controlled

under the requirements of 10 CFR 50.59.

Date of issuance: August 13, 1996

Effective date: August 13, 1996

Amendment Nos.: 159 and 130
44368 Federal Register / Vol. 61, No. 168 / Wednesday, August 28, 1996 / Notices

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

revised the Technical Specifications.

Date of initial notice in Federal

Register: April 10, 1996 (61 FR 15992)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated August 13, 1996. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Osterhout Free Library,

Reference Department, 71 South

Franklin Street, Wilkes-Barre,

Pennsylvania 18701.

Tennessee Valley Authority, Docket

Nos. 50-259, 50-260, and 50-296,

Browns Ferry Nuclear Plant, Units 1, 2,

and 3, Limestone County, Alabama

Date of application for amendments:

May 20, 1996 (TS 373) Brief description of amendment: The

amendments incorpore the guidance of

Generic Letter 87-09 in the technical

specifications, allowing a 24-hour delay

in implementing action requirements

due to a missed surveillance

requirement.

Date of issuance: August 5, 1996 Effective Date: August 5, 1996 Amendment Nos.: 230, 245 and 205

Facility Operating License Nos. DPR33, DPR-52 and DPR-68: Amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: June 19, 1996 (61 FR 31185)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated August 5, 1996. No

significant hazards consideration

comments received: None

Local Public Document Room

location: Athens Public library, South

Street, Athens, Alabama 35611

Union Electric Company, Docket No.

50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application for amendment:

May 29, 1996 Brief description of amendment: The

amendment authorizes revision of the

Final Safety Analysis Report (FSAR) to

incorporate a modification to the facility

that will reduce the single failure trip

potential for the main feedwater and

bypass valves. Date of issuance: August 13, 1996 Effective date: August 13, 1996 Amendment No.: 115

Facility Operating License No. NPF30: The amendment revised the Final

Safety Analysis Report. Date of initial notice in Federal

Register: July 3, 1996 (61 FR 34900) The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated August 13, 1996. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Callaway County Public

Library, 710 Court Street, Fulton,

Missouri 65251.

Wisconsin Public Service Corporation,

Docket No. 50-305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of application for amendment:

June 4, 1996

Brief description of amendment: The

amendment revises the Technical

Specifications by reducing the

surveillance test frequencies for the

radiation monitoring system (Table TS

4.1-1) and the control rods (Table TS

4.1-3) in accordance with the guidance

of Generic Letter 93-05, ‘‘Line-Item

Technical Specifications Improvements

to Reduce Surveillance Requirements

for Testing During Power Operation,’’

dated September 27, 1993.

Date of issuance: August 7, 1996

Effective date: August 7, 1996

Amendment No.: 125

Facility Operating License No. DPR43: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 3, 1996 (61 FR 34901) The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated August 7, 1996. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: University of Wisconsin,

Cofrin Library, 2420 Nicolet Drive,

Green Bay, Wisconsin 54311-7001

Wolf Creek Nuclear Operating

Corporation, Docket No. 50-482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request: July 29,

1994, as superseded by letter dated

September 15, 1995, and subsequently

supplemented by letters dated March 8,

1996, April 18, 1996, June 14, 1996, and

July 12, 1996.

Brief description of amendment: The

amendment revises TS 3/4.8.1, ‘‘Electric

Power Systems - A.C. Sources,’’ and its

associated Bases to achieve an overall

improvement in emergency diesel

generator reliability and availability.

Date of issuance: August 9, 1996

Effective date: August 9, 1996, to be

implemented within 90 days of the date

of issuance.

Amendment No.: 101

Facility Operating License No. NPF42. The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: May 22, 1996 (61 FR 25716)

The June 14, 1996, and July 12, 1996,

supplemental letters provided Bases

page changes 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 August 9, 1996. No

significant hazards consideration

comments received: No.

Local Public Document Room

locations: Emporia State University,

William Allen White Library, 1200

Commercial Street, Emporia, Kansas

66801 and Washburn University School

of Law Library, Topeka, Kansas 66621

Dated at Rockville, Maryland, this 21st day

of August 1996.

For the Nuclear Regulatory Commission

Steven A. Varga,

Director, Division of Reactor Projects - I/II

Office of Nuclear Reactor Regulation

[Doc. 96-21813 Filed 8-27-96; 8:45 am]

BILLING CODE 7590-01-F

SECURITIES AND EXCHANGE

COMMISSION

[Investment Company Act Release No.

22160; 811–3925]

Alliance Growth Fund, Inc.; Notice of

Application

August 21, 1996.

AGENCY: Securities and Exchange

Commission (‘‘SEC’’).

ACTION: Notice of Application for

Deregistration under the Investment

Company Act of 1940 (the ‘‘Act’’).

APPLICANT: Alliance Growth Fund, Inc.

RELEVANT ACT SECTION: Section 8(f).

SUMMARY OF APPLICATION: Applicant

requests an order declaring that it has

ceased to be an investment company.

FILING DATES: The application was filed

on July 26, 1996.

HEARING OR NOTIFICATION OF HEARING: An

order granting the application will be

issued unless the SEC orders a hearing.

Interested persons may request a

hearing by writing to the SEC’s

Secretary and serving applicant with a

copy of the request, personally or by

mail. Hearing requests should be

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

September 16, 1996, and should be

accompanied by proof of service on the

applicant, in the form of an affidavit or,

for lawyers, a certificate of service.

Hearing requests should state the nature

of the writer’s interest, the reason for the

request, and the issues contested.

Persons may request notification of a

hearing by writing to the SEC’s

Secretary.

ADDRESSES: Secretary, SEC, 450 Fifth

Street NW., Washington, DC 20549.

Applicant, 1345 Avenue of the

Americas, New York, New York 10105.