27334 Federal Register / Vol. 60, No. 99 / Tuesday, May 23, 1995 / Notices

fellowships. The Board will meet in

Boston, Massachusetts on June 15, 1995

from 9 a.m. to 4 p.m. The meeting of the

Board is open to the public. The agenda

includes discussion of the Institute’s

plans and priorities for program years

1995 and 1996; the status of new Board

member nominations; and other

relevant Institute matters. Records are

kept of all Board proceedings and are

available for public inspection at the

National Institute for Literacy, 800

Connecticut Avenue, NW, Suite 200,

Washington, DC 20006 from 8:30 a.m. to

5 p.m.

Dated: May 17, 1995.

Andrew J. Hartman,

Executive Director, National Institute for

Literacy.

[FR Doc. 95–12536 Filed 5–22–95; 8:45 am]

BILLING CODE 6055–01–M

NUCLEAR REGULATORY

COMMISSION

Biweekly Notice; Applications and

Amendments to Facility Operating

Licenses Involving No Significant

Hazards Considerations

I. Background

Pursuant to Public Law 97–415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 97–415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from May 1, 1995,

through May 12, 1995. The last

biweekly notice was published on May

10, 1995 (60 FR 24904).

Notice of Consideration of Issuance of

Amendments to Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a

proposed determination that the

following amendment requests involve

no significant hazards consideration.

Under the Commission’s regulations in

10 CFR 50.92, this means that operation

of the facility in accordance with the

proposed amendment would not (1)

involve a significant increase in the

probability or consequences of an

accident previously evaluated; or (2)

create the possibility of a new or

different kind of accident from any

accident previously evaluated; or (3)

involve a significant reduction in a

margin of safety. The basis for this

proposed determination for each

amendment request is shown below.

The Commission is seeking public

comments on this proposed

determination. Any comments received

within 30 days after the date of

publication of this notice will be

considered in making any final

determination.

Normally, the Commission will not

issue the amendment until the

expiration of the 30-day notice period.

However, should circumstances change

during the notice period such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received before

action is taken. Should the Commission

take this action, it will publish in the

Federal Register a notice of issuance

and provide for opportunity for a

hearing after issuance. The Commission

expects that the need to take this action

will occur very infrequently.

Written comments may be submitted

by mail to the Rules Review and

Directives Branch, Division of Freedom

of Information and Publications

Services, Office of Administration, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555, and should cite

the publication date and page number of

this Federal Register notice. Written

comments may also be delivered to

Room 6D22, Two White Flint North,

11545 Rockville Pike, Rockville,

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

Federal workdays. Copies of written

comments received may be examined at

the NRC Public Document Room, the

Gelman Building, 2120 L Street, NW.,

Washington, DC. The filing of requests

for a hearing and petitions for leave to

intervene is discussed below.

By June 23, 1995, the licensee may

file a request for a hearing with respect

to issuance of the amendment to the

subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) The nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the
Federal Register / Vol. 60, No. 99 / Tuesday, May 23, 1995 / Notices 27335

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington DC, by

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

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

1–(800) 342–6700). The Western Union

operator should be given Datagram

Identification Number N1023 and the

following message addressed to (Project

Director): petitioner’s name and

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555, and to the attorney for the

licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

factors specified in 10 CFR

2.714(a)(1)(i)–(v) and 2.714(d).

For further details with respect to this

action, see the application for

amendment which is available for

public inspection at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, and at the local public

document room for the particular

facility involved.

Arizona Public Service Company, et al.,

Docket Nos. STN 50–528, STN 50–529,

and STN 50–530, Palo Verde Nuclear

Generating Station, Units 1, 2, and 3,

Maricopa County, Arizona

Date of application for amendments:

April 6, 1995.

Brief description of amendments: The

proposed amendment involves changes

in personnel titles, implementation of

line item improvements delineated in

Generic Letter 93–07, ‘‘Modification of

the Technical Specification

Administrative Control Requirements

for Emergency and Security Plans,’’

changes in the Plant Review Board, and

miscellaneous minor changes.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) The proposed changes do not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

These changes involve (1) minor changes

in the organization of PVNGS, (2) line item

improvements recommended by the NRC, or

(3) clarification or corrections to existing

specifications. It is expected that the

organizational changes will have a positive

effect on the conduct of plant operations and

safety-related work. Functions which are

necessary to operate the facility safely and in

accordance with the operating licenses,

remain in the new organization. The line

item improvements to the Technical

Specifications will not affect the safe

operation of the plant and continue to ensure

proper control of administrative activities.

The proposed changes will not affect the

operation of structures, systems and

components, and will not reduce

programmatic controls such that plant safety

would be affected. Therefore, the proposed

changes do not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

(2) The proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

analyzed.

The proposed changes will not affect the

operation of structures, systems and

components, and will not reduce

programmatic controls such that plant safety

would be affected. The changes in the

organization and as a result of line item

improvements will continue to provide

necessary oversight and control of

administrative processes. Therefore, the

proposed changes do not create the

possibility of a new or different kind of

accident from any previously evaluated.

(3) The proposed changes do not involve

a significant reduction in a margin of safety.

These changes are administrative and will

not diminish any organizational or

administrative controls currently in place.

The proposed changes will not affect the

operation of structures, systems and

components, and will not reduce

programmatic controls such that plant safety

would be affected. Therefore, the proposed

changes do not involve a significant

reduction in a margin of safety.

Local Public Document Room

location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona

85004.

Attorney for licensee: Nancy C. Loftin,

Esq., Corporate Secretary and Counsel,

Arizona Public Service Company, P.O.

Box 53999, Mail Station 9068, Phoenix,

Arizona 85072–3999.

NRC Project Director: William H.

Bateman.

Arizona Public Service Company, et al.,

Docket Nos. STN 50–528, STN 50–529,

and STN 50–530, Palo Verde Nuclear

Generating Station, Unit Nos. 1, 2, and

3, Maricopa County, Arizona

Date of amendment requests: April

18, 1995.

Description of amendment requests:

The proposed Technical Specification

amendments would revise the

surveillance requirements for Technical

Specification 3/4.4.4, ‘‘Steam

Generators,’’ and the associated Bases.

These amendments would allow the

installation of tube sleeves as an

alternative to plugging defective steam

generator tubes.

Basis for proposed no significant

hazards consideration determination:
27336 Federal Register / Vol. 60, No. 99 / Tuesday, May 23, 1995 / Notices

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed amendment to permit the

use of steam generator tube sleeves as an

alternative to tube plugging is a safe and

effective repair procedure that does not

require removing a tube from service.

Mechanical strength, corrosion resistance,

installation methods, and inservice

inspection techniques of sleeves have been

shown to meet NRC acceptance criteria.

Analytical verifications were performed

using design and operating transient

parameters selected to envelope loads

imposed during normal operating and

accident conditions. Fatigue and stress

analysis of sleeved tube assemblies were

completed in accordance with the

requirements of Section III of the ASME

Code. The results of qualification testing,

analysis and plant operating experience at

other facilities demonstrates that the sleeving

process is an acceptable means of

maintaining steam generator tube integrity.

The sleeve configuration has been designed

and analyzed in accordance with the

structural margins specified in Regulatory

Guide (RG) 1.121. Furthermore, the installed

sleeve will be monitored through periodic

inspections on a sample basis with eddy

current techniques. A sleeve-specific

plugging margin, per the recommendations of

RG 1.121, has been specified with

appropriate allowances for NDE

(nondestructive examination) uncertainty

and defect growth rate.

The consequences of accidents previously

analyzed are not increased as a result of

sleeving activities. The hypothetical failure

of the sleeve would be bounded by the

current steam generator tube rupture analysis

contained in the PVNGS (Palo Verde Nuclear

Generating Station) UFSAR (updated final

safety analysis report). Due to the slight

reduction in diameter caused by the sleeve

wall thickness, it is expected that the primary

release rates would be less than assumed for

the steam generator tube rupture analysis,

and therefore would result in lower total

primary fluid mass release to the secondary

system. Additionally, further conservatism is

introduced if the break were postulated to

occur at a location on the tube higher than

the location where a sleeve is installed. The

overall effect would be reduced steam

generator tube rupture release rates. The

minimal reduction in flow area associated

with a tube sleeve has no significant affect on

steam generator performance with respect to

heat transfer or system flow resistance and

pressure drop. The installation of sleeves

rather than plugging also maintains a greater

heat transfer surface in the steam generator.

In any case, the impacts are bounded by

evaluations which demonstrate the

acceptability of tube plugging which totally

removes the tube from service. Therefore, in

comparison to plugging, tube sleeving is

considered a significant improvement with

respect to steam generator performance. The

cumulative impact of multiple sleeved tubes

was evaluated to ensure the effects remain

within the analytical design bases.

Recent industry experience with forced

shutdown events associated with tube

failures at sleeve junctions was assessed by

ASP and ABB–CE. The root cause of these

events has been attributed to the lack of

proper post-installation stress relief and/or

the imposition of high stresses due the tube

growth restrictions at locked tube support.

The material and design of the PVNGS steam

generator supports minimizes the potential

for locked supports. The tube supports are of

eggcrate design and are constructed of ferritic

stainless steel. The large flow area in the

eggcrate design provides better irrigation and

reduces the potential for steam blanketing,

therefore, the tube-to-tube support crevices

are less likely to be blocked by crud, boiler

water deposits and corrosion products. Since

the support material is type 409 ferritic

stainless steel, it is not susceptible to

magnetite corrosion which has resulted in

denting and lockup at plants with carbon

steel supports. These conclusions have been

substantiated via tube pull activities

conducted in PVNGS Unit 2. Although ABB–

CE does not require post-weld heat treatment

in all applications, APS will require that a

post-weld stress relief be conducted for all

sleeve installations.

APS has incorporated an integrated leakage

monitoring program, utilizing equipment,

procedure upgrades and administrative

shutdown limits significantly lower than

Technical Specification requirements. The

program is designed to provide plant

operators with the ability to detect and

respond to changes in primary-to-secondary

leakage and shutdown the unit prior to a

significant leak or steam generator tube

rupture, should sleeve or tube degradation

exceed expected values. The program is

designed to reduce the probability of steam

generator tube rupture events.

Therefore, based on the above, the

proposed amendment does 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

analyzed.

A sleeved steam generator tube performs

the same function in the same passive

manner as an unsleeved steam generator

tube. Tube sleeves are designed, qualified,

and maintained under the stress and pressure

limits of Section III of the ASME Code and

Regulatory Guide 1.121.

The installation of the sleeve, including

weld and welder qualification and

nondestructive examination (NDE), meets or

exceeds the requirements of ASME Section

XI. Three types of NDE are conducted.

Ultrasonic Testing (UT) is performed to

verify adequacy of the tube to sleeve weld

assuring proper fusion. Eddy current testing

(ET) is performed following each installation

to establish baseline data for each sleeve in

order to monitor future degradation of the

primary to secondary pressure boundary.

Visual inspections may be performed to

verify or ascertain the mechanical and

structural condition of a weld. Critical

conditions which are checked include weld

width and completeness, and the absence of

visibly noticeable indications such as cracks,

pits, and burn through.

ABB-Combustion Engineering Inc., Report

CEN–613–P, ‘‘Arizona Public Service Co.,

Palo Verde Units 1, 2, and 3, Steam Generator

Tube Repair Using Leak Tight Sleeves,’’

Revision 01, January 1995, demonstrates that

the repair of degraded steam generator tubes

using tube sleeves will result in tube bundle

integrity consistent with the original design

basis. An extensive analysis and corrosion

and mechanical test programs were

undertaken to prove the adequacy of tube

sleeve repair. The proposed amendments

have no significant effect on the

configuration of the plant, and the change

does not effect the way in which the plant

is operated. Based upon the results of the

analytical and test programs described in the

ABB Combustion Engineering Inc. report, the

tube sleeve fulfills its intended function and

meets or exceeds established design criteria.

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.

Evaluation of the sleeved tubes indicates

no detrimental effects on the sleeve-tube

assembly resulting from reactor system flow,

coolant chemistries, or thermal and pressure

conditions. Structural analyses of the sleevetube assembly, using demonstrated margins

of safety, have established sleeve-tube

integrity under normal and accident

conditions. Structural analyses have been

performed for sleeves which span the tube at

the top of the tubesheet and which span the

flow distribution plate or eggcrate support.

Mechanical testing has been performed to

support the analyses. Corrosion testing of

typical sleeve-tube assemblies has been

completed and reveals no evidence of sleeve

or tube corrosion considered detrimental

under anticipated service conditions.

Based upon the testing and analyses

performed, the installation of tube sleeves

will not result in a significant reduction in

a margin of safety.

Steam generator tube integrity is

maintained under the same limits for sleeved

tubes as for unsleeved tubes, i.e., Section III

of the ASME Code and Regulatory Guide

1.121. The portions of the installed sleeve

assembly which represents the reactor

coolant pressure boundary can be monitored

for the initiation and progression of sleeve/

tube wall degradation, thus satisfying the

requirements of Regulatory Guide 1.83. The

degradation limit at which a sleeve/tube

boundary is considered inoperable has been

analyzed in accordance with Regulatory

Guide 1.121 and is specified. Eddy current

detectability of flaws has been verified by

ABB Combustion Engineering. The Technical

Specifications continue to require monitoring

and restriction of primary to secondary

system leakage through the steam generators.

A conservative integrated leakage program

employed by APS provides reasonable

assurance than an orderly unit shutdown will
Federal Register / Vol. 60, No. 99 / Tuesday, May 23, 1995 / Notices 27337

occur prior to a significant increase in

leakage due to failure of a sleeved or

unsleeved tube. The minimal reduction in

reactor coolant system flow, due to sleeving,

is considered to have an insignificant impact

on steam generator operation during normal

operation or accident conditions and is

bounded by tube plugging evaluations.

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 that

review, it appears that the three

standards of § 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment requests

involve no significant hazards

consideration.

Local Public Document Room

location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona

85004.

Attorney for licensees: Nancy C.

Loftin, Esq., Corporate Secretary and

Counsel, Arizona Public Service

Company, P.O. Box 53999, Mail Station

9068, Phoenix, Arizona 85072–3999.

NRC Project Director: William H.

Bateman.

Commonwealth Edison Company,

Docket Nos. 50–237 and 50–249,

Dresden Nuclear Power Station, Units 2

and 3, Grundy County, Illinois, Docket

Nos. 50–254 and 50–265, Quad Cities

Nuclear Power Station, Units 1 and 2,

Rock Island County, Illinois

Date of application for amendment

request: February 16, 1993, as

supplemented by letter dated May 2,

1995.

Description of amendment request: As

a result of findings by a Diagnostic

Evaluation Team inspection performed

by the NRC staff at the Dresden Nuclear

Power Station in 1987, Commonwealth

Edison Company (ComEd, the licensee)

made a decision that both the Dresden

Nuclear Power Station and sister site

Quad Cities Nuclear Power Station,

needed attention focused on the existing

custom Technical Specifications (TS)

used.

The licensee made the decision to

initiate a Technical Specification

Upgrade Program (TSUP) for both

Dresden and Quad Cities. The licensee

evaluated the current TS for both

Dresden and Quad Cities against the

Standard Technical Specifications (STS)

contained in NUREG–0123, ‘‘Standard

Technical Specifications General

Electric Plants BWR/4.’’ The licensee’s

evaluation identified numerous

potential improvements such as

clarifying requirements, changing TS to

make them more understandable and to

eliminate interpretation, and deleting

requirements that are no longer

considered current with industry

practice. As a result of the evaluation,

ComEd has elected to upgrade both the

Dresden and Quad Cities TS to the STS

contained in NUREG–0123.

The TSUP for Dresden and Quad

Cities is not a complete adaption of the

STS. The TSUP focuses on (1)

integrating additional information such

as equipment operability requirements

during shutdown conditions, (2)

clarifying requirements such as limiting

conditions for operations and action

statements utilizing STS terminology,

(3) deleting superseded requirements

and modifications to the TS based on

the licensee’s responses to Generic

Letters (GL), and (4) relocating specific

items to more appropriate TS locations.

The February 16, 1993, and May 2,

1995, applications proposed to upgrade

only Section 3/4.10 (Refueling

Operations) of the Dresden and Quad

Cities TS.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated because:

In general, the proposed amendment

represents the conversion of current

requirements to a more generic format, or the

addition of requirements which are based on

the current safety analysis. Implementation

of these changes will provide increased

reliability of equipment assumed to operate

in the current safety analysis, or provide

continued assurance that specified

parameters remain within their acceptance

limits, and as such, will not significantly

increase the probability or consequences of a

previously evaluated accident.

Some of the proposed changes represent

minor curtailments of the current

requirements which are based on generic

guidance or previously approved provisions

for other stations. The proposed amendment

for Dresden and Quad Cities Station’s

Technical Specification Section 3/4.10 are

based on STS guidelines or later operating

BWR plant’s NRC accepted changes. Any

deviations from STS requirements do not

significantly increase the probability or

consequences of any previously evaluated

accidents for Dresden or Quad Cities

Stations. The proposed amendment is

consistent with the current safety analyses

and has been previously determined to

represent sufficient requirements for the

assurance and reliability of equipment

assumed to operate in the safety analysis, or

provide continued assurance that specified

parameters remain within their acceptance

limits. As such, these changes will not

significantly increase the probability or

consequences of a previously evaluated

accident.

The associated systems that make up the

Refueling Systems are not assumed in any

safety analysis to initiate any accident

sequence for Dresden or Quad Cities Stations;

therefore, the probability of any accident

previously evaluated is not increased by the

proposed amendment. In addition, the

proposed surveillance requirements for the

proposed amendments to these systems are

generally more prescriptive than the current

requirements specified within the Technical

Specifications. The additional surveillance

requirements improve the reliability and

availability of all affected systems and

therefore, reduce the consequences of any

accident previously evaluated as the

probability of the systems outlined within

Section 3/4.10 of the proposed Technical

Specifications, performing its intended

function is increased by the additional

surveillances.

Create the possibility of a new or different

kind of accident from any previously

evaluated because:

In general, the proposed amendment

represents the conversion of current

requirements to a more generic format, or the

addition of requirements which are based on

the current safety analysis. Others represent

minor curtailments of the current

requirements which are based on generic

guidance or previously approved provisions

for other stations. These changes do not

involve revisions to the design of the station.

Some of the changes may involve revision in

the operation of the station; however, these

provide additional restrictions which are in

accordance with the current safety analysis,

or are to provide for additional testing or

surveillances which will not introduce new

failure mechanisms beyond those already

considered in the current safety analyses.

The proposed amendment for Dresden and

Quad Cities Station’s Technical Specification

Section 3/4.10 is based on STS guidelines or

later operating BWR plants’ NRC accepted

changes. The proposed amendment has been

reviewed for acceptability at the Dresden and

Quad Cities Nuclear Power Stations

considering similarity of system or

component design versus the STS or later

operating BWRs. Any deviations from STS

requirements do not create the possibility of

a new or different kind of accident

previously evaluated for Dresden or Quad

Cities Stations. No new modes of operation

are introduced by the proposed changes,

considering the acceptable operational modes

in present specifications, the STS, or later

operating BWRs. Surveillance requirements

are changed to reflect improvements in

technique, frequency of performance or

operating experience at later plants. Proposed

changes to action statements in many places

add requirements that are not in the present

technical specifications or adopt

requirements that have been used

successfully at other operating BWRs with

designs similar to Dresden and Quad Cities.

The proposed changes maintain at least the

present level of operability. Therefore, the

proposed changes do not create the

possibility of a new or different kind of

accident from any previously evaluated.

The associated systems that make up the

Refueling Systems are not assumed in any
27338 Federal Register / Vol. 60, No. 99 / Tuesday, May 23, 1995 / Notices

safety analysis to initiate any accident

sequence for Dresden or Quad Cities Stations.

In addition, the proposed surveillance

requirements for affected systems associated

with the Refueling Systems are generally

more prescriptive than the current

requirements specified within the Technical

Specifications; therefore, the proposed

changes do not create the possibility of a new

or different kind of accident from any

previously evaluated.

Involve a significant reduction in the

margin of safety because:

In general, the proposed amendment

represents the conversion of current

requirements to a more generic format, or the

addition of requirements which are based on

the current safety analysis. Others represent

minor curtailments of the current

requirements which are based on generic

guidance or previously approved provisions

for other stations. Some of the later

individual items may introduce minor

reductions in the margin of safety when

compared to the current requirements.

However, other individual changes are the

adoption of new requirements which will

provide significant enhancement of the

reliability of the equipment assumed to

operate in the safety analysis, or provide

enhanced assurance that specified

parameters remain with their acceptance

limits. These enhancements compensate for

the individual minor reductions, such that

taken together, the proposed changes will not

significantly reduce the margin of safety.

The proposed amendment to Technical

Specification Section 3/4.10 implements

present requirements, or the intent of present

requirements in accordance with the

guidelines set forth in the STS. Any

deviations from STS requirements do not

significantly reduce the margin of safety for

Dresden or Quad Cities Stations. The

proposed changes are intended to improve

readability, usability, and the understanding

of technical specification requirements while

maintaining acceptable levels of safe

operation. The proposed changes have been

evaluated and found to be acceptable for use

at Dresden and Quad Cities based on system

design, safety analysis requirements and

operational performance. Since the proposed

changes are based on NRC accepted

provisions at other operating plants that are

applicable at Dresden and Quad Cities and

maintain necessary levels of system,

component or parameter (reliability), the

proposed changes do not involve a

significant reduction in the margin of safety.

The proposed amendment for Dresden and

Quad Cities Stations will not reduce the

availability of systems associated with the

Refueling Systems when required to mitigate

accident conditions; therefore, the proposed

changes do not involve a significant

reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: For Dresden, Morris Area

Public Library District, 604 Liberty

Street, Morris, Illinois 60450; for Quad

Cities, Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois

61021.

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, et al., Docket

Nos. 50–413 and 50–414, Catawba

Nuclear Station, Units 1 and 2, York

County, South Carolina

Date of amendment request:

September 19, 1994, as supplemented

by letter dated April 26, 1995.

Description of amendment request:

The amendments would change the

Technical Specifications (TS) to

increase the enrichment limits for fuel

stored in the fuel pools and establish

restricted loading patterns and

associated burnup criteria for qualifying

fuel in the spent fuel pools. In addition,

several administrative changes have

been included in order to provide

clarity to the TS and bring them more

in line with the Standard Technical

Specifications format. These changes are

as follows:

(1) The TS index is changed to add TS

3/4.9.12 and 3/4.9.13, Tables 3.9–1 and

3.9–2 and Figure 3.9–1.

(2) TS 3/4.9.12, Spent Fuel Pool (SFP)

Boron Concentration, is added to

establish a boron concentration limit

and to establish a Limiting Condition for

Operation (LCO) for all modes of

operation and to allow the numerical

value of the limit to be specified in the

Core Operating Limits Report (COLR).

(3) TS 3/4.9.13, Tables 3.9–1 and 3.9–

2 and Figure 3.9–1 are being added to

establish restricted loading patterns for

spent fuel storage and associated

burnup criteria.

(4) Corresponding BASES for TSs 3/

4.9.12 and 3/4.9.13 are added to explain

the basis for each LCO, Action

Statement, and Surveillance

Requirement covered by the subject TSs.

(5) TS 5.6, Fuel Storage, is changed to

reflect limits for criticality analysis for

fuel storage.

(6) TS 6.9, Reporting Requirements, is

changed to reflect the inclusion of the

SFP boron concentration limit values in

the COLR as established by TS 3/4.9.12.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

There is no increase in the probability or

consequences of an accident in the new fuel

vault since the only credible accidents for

this area are criticality accidents and it has

been shown that calculated, worst case Keff

for this area is (less than or equal to) 0.95

under all conditions.

There is no increase in the probability of

a fuel drop accident in the Spent Fuel

Storage Pool since the mass of an assembly

will not be affected by the increase in fuel

enrichment. The likelihood of other

accidents, previously evaluated and

described in Section 9.1.2 of the FSAR (Final

Safety Analysis Report), is also not affected

by the proposed changes. In fact, it could be

postulated that since the increase in fuel

enrichment will allow for extended fuel

cycles, there will be a decrease in fuel

movement and the probability of an accident

may likewise be decreased. There is also no

increase in the consequences of a fuel drop

accident in the Spent Fuel Pool since the

fission product inventory of individual fuel

assemblies will not change significantly as a

result of increased initial enrichment. In

addition, no change to safety related systems

is being made.

Therefore, the consequences of a fuel

rupture accident remain unchanged. In

addition, it has been shown that Keff is (less

than or equal to) 0.95, under all conditions.

Therefore, the consequences of a criticality

accident in the Spent Fuel Pool remain

unchanged as well.

2. The proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes do not create the

possibility of a new or different kind of

accident since fuel handling accidents (fuel

drop and misplacement) are not new or

different kinds of accidents. Fuel handling

accidents are already discussed in the FSAR

for fuel with enrichments up to 4.0 weight %.

As described in Section IV.9 of Attachment

IV, additional analyses have been performed

for fuel with enrichment up to 5.00 weight

%. Worst case misloading accidents

associated with the new loading patterns

were evaluated. It was shown that the

negative reactivity provided by soluble boron

maintains Keff (less than or equal to) 0.95.

3. The proposed changes do not involve a

significant reduction in the margin of safety.

The proposed change does not involve a

significant reduction in the margin of safety

since, in all cases, a Keff [less than or equal

to] 0.95 is being maintained. Criticality

analyses have been performed which show

that the new fuel storage vault will remain

subcritical under a variety of moderation

conditions, from fully flooded to optimum

moderation. As discussed above, the Spent

Fuel Pool will remain sufficiently subcritical

during any fuel misplacement accident.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are
Federal Register / Vol. 60, No. 99 / Tuesday, May 23, 1995 / Notices 27339

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: York County Library, 138 East

Black Street, Rock Hill, South Carolina

29730.

Attorney for licensee: Mr. Albert Carr,

Duke Power Company, 422 South

Church Street, Charlotte, North Carolina

28242.

NRC Project Director: Herbert N.

Berkow.

Florida Power and Light Company,

Docket Nos. 50–250 and 50–251, Turkey

Point Plant Units 3 and 4, Dade County,

Florida

Date of amendment request: March

30, 1995, and supplemented May 5,

1995.

Description of amendment request:

The licensee proposes to change Turkey

Point Units 3 and 4 Technical

Specifications (TS) by separation of the

24-hour emergency diesel generator

(EDG) run from the hot restart EDG test.

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 TS changes would revise the

EDG surveillance criteria to allow the EDG

hot-start test with full ESF load acceptance

to be performed separately and

independently from the 24-hour EDG run.

The proposed SRs (surveillance

requirements) would continue to

demonstrate that the objectives of these two

tests are met. Specifically, the EDGs are

shown to be: (1) Capable of starting and

running continuously at full load capability

for an interval not less than 24 hours, and (2)

capable of restarting from a full load

temperature condition. The proposed

changes would not affect the EDGs’ ability to

support mitigation of the consequences of

any previously evaluated accident.

Additionally, the proposed changes to the

SRs do not affect the initiating assumptions

or progression of any accident sequence.

Therefore, operation of the facility would

not involve a significant increase in the

probability or consequences of an accident

previously analyzed.

(2) Operation of the facility in accordance

with the proposed amendments would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed TS SR changes do not

require any physical changes to the plant or

equipment, and do not impact any design or

functional requirements of the EDGs. The

proposed changes do not create any plant

configurations which are prohibited by the

TS. The proposed changes continue to meet

the EDG test objectives associated with

demonstrating EDG operability.

Therefore, 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.

(3) Operation of the facility in accordance

with the proposed amendments would not

involve a significant reduction in a margin of

safety.

The proposed TS SR changes do not

require any physical changes to the plant or

equipment and do not impact any design or

functional requirements of the EDGs.

Surveillance testing in accordance with the

proposed TS will continue to demonstrate

the ability of the EDGs to perform their

intended function of providing electrical

power to mitigate design basis transients,

consistent with the plant safety analyses.

Therefore, operation of the facility in

accordance with the proposed amendments

would not involve a reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of § 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Florida International

University, University Park, Miami,

Florida 33199.

Attorney for licensee: J. R. Newman,

Esquire, Morgan, Lewis & Bockius, 1800

M Street, NW., Washington, DC 20036.

NRC Project Director: David B.

Matthews.

Omaha Public Power District, Docket

No. 50–285, Fort Calhoun Station, Unit

No. 1, Washington County, Nebraska

Date of amendment request: April 7,

1995.

Description of amendment request:

The proposed amendment would revise

the technical specifications (TS) to

relocate the axial power distribution

limits to the Core Operating Limits

Report (COLR).

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 relocates the cyclespecific Axial Power Distribution (APD)

limits contained in Figure 1–2 of the

Technical Specifications (TS), to the Core

Operating Limits Report (COLR). This change

is consistent with the NRC recommendations

of Generic Letter 88–16, and will not modify

the methodology used in generating the

limits nor the manner in which they are

implemented. The methodology used to

determine the APD limits is reviewed and

approved by the NRC in accordance with TS

5.9.5. The APD limits will continue to be

determined by analyzing the same postulated

events as previously analyzed. The plant will

continue to operate within the limits

specified in the COLR and will take the same

remedial actions if the APD limit is exceeded

as required by the current TS. 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 only relocates

the APD figure from the TS to the COLR

consistent with NRC Generic Letter 88–16.

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.

As indicated above, the implementation of

the APD into the COLR, consistent with the

guidance of NRC Generic Letter 88–16, makes

use of the existing safety analysis

methodologies and the resulting limits and

setpoints for plant operation. Additionally,

the safety analysis acceptance criteria for

operations with the proposed change have

not changed from that use in the current

reload analysis. Therefore, the margin of

safety is not reduced due to the relocation of

the APD from the TS and implementation in

the COLR.

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: LeBoeuf, Lamb,

Leiby, and MacRae, 1875 Connecticut

Avenue, NW., Washington, DC 20009–

5728.

NRC Project Director: William

Bateman.
27340 Federal Register / Vol. 60, No. 99 / Tuesday, May 23, 1995 / Notices

Pacific Gas and Electric Company,

Docket Nos. 50–275 and 50–323, Diablo

Canyon Nuclear Power Plant, Unit Nos.

1 and 2, San Luis Obispo County,

California

Date of amendment requests: April

19, 1995 (Reference LAR 95–03).

Description of amendment requests:

The proposed amendments would

revise the combined Technical

Specifications (TS) for the Diablo

Canyon Nuclear Power Plant, Unit Nos.

1 and 2 to revise TS 3/4.8.1.1, ‘‘A.C.

Sources, Operating.’’ The specific TS

changes proposed are as follows:

(1) TS 4.8.1.1.2b.8), emergency diesel

generator (EDG) 24-hour load run and

hot restart surveillance, would be

revised to delete the requirement to

perform TS 4.8.1.1.2b.5)b), loss of offsite

power (LOOP) load sequencing

surveillance within 5 minutes following

the 24-hour test.

(2) New TS 4.8.1.1.2e. would be

added to perform an EDG hot restart test

within 5 minutes of shutting down the

EDG after the EDG has operated for at

least 2 hours at a load of greater than or

equal to 2484 kW.

(3) TS 4.8.1.1.2b.8), TS 4.8.1.1.2e.,

and footnote ‘‘*’’ on page 3/4 8–5 would

be changed to be cycle-specific with the

new TS requirements effective for Units

1 and 2, Cycle 8 and after.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

Demonstrating emergency diesel generator

(EDG) hot restart capability without

sequencing loss of offsite power (LOOP)

loads does not invalidate or reduce the

effectiveness of the hot restart test, since

normal operating temperatures are achieved

prior to the hot restart test. Sequencing the

LOOP loads does not contribute to verifying

that the EDG will start from normal operating

temperatures. The proposed TS 4.8.1.1.[2]e

may be performed in any plant condition

since performance of this new surveillance

will have no adverse effect on plant

operations. The reliability of the EDGs is not

affected by the proposed changes.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes do not involve any

physical alterations to the plant. The

proposed changes will not have any adverse

effect on the ability of the EDGs to perform

their required safety function.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed changes do not involve a

significant reduction in a margin of safety.

The proposed changes will not alter any

accident analysis assumptions, initial

conditions, or results. Consequently, the

proposed changes do not have any effect on

the margin of safety. The proposed changes

to the surveillance requirements would

continue to demonstrate the ability of the

EDGs to perform their intended safety

function.

Therefore, the proposed changes do not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of § 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment requests

involve no significant hazards

consideration.

Local Public Document Room

location: California Polytechnic State

University, Robert E. Kennedy Library,

Government Documents and Maps

Department, San Luis Obispo, California

93407.

Attorney for licensee: Christopher J.

Warner, Esq., Pacific Gas and Electric

Company, PO Box 7442, San Francisco,

California 94120.

NRC Project Director: William H.

Bateman.

Philadelphia Electric Company, Public

Service Electric and Gas Company,

Delmarva Power and Light Company,

and Atlantic City Electric Company,

Docket No. 50–278, Peach Bottom

Atomic Power Station, Unit No. 3, York

County, Pennsylvania

Date of application for amendment:

November 21, 1994.

Description of amendment request:

The proposed change would extend the

Type A test (i.e., Containment Integrated

Leak Rate Test (CILRT)) interval on a

one-time basis.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed Technical Specifications

(TS) change does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

The accidents which are potentially

adversely impacted by the proposed change

are any Loss of Coolant Accident (LOCA)

inside primary containment as described in

the PBAPS, Units 2 and 3 UFSAR.

The proposed change increases the

surveillance interval of the 10 CFR part 50,

appendix J Type A test (i.e., Containment

Integrated Leakage Rate Test (CILRT)) from

46 months to 70 months. This test is

performed to determine that the total leakage

from containment does not exceed the

maximum allowable primary containment

leakage rate (i.e., designated La) at a

calculated peak containment internal

pressure (Pa), as defined in 10 CFR part 50,

appendix J. The primary containment limits

the leakage of radioactive material during

and following design bases accidents in order

to comply with the offsite does limits

specified in 10 CFR part 100. Accordingly,

the primary containment is not an accident

initiator. It is an accident mitigator. No

physical or operational changes to the

containment structure, plant systems, or

components would be made as a result of the

proposed change. Therefore, the probability

of occurrence of an accident previously

evaluated is not increased.

The failure effects that are potentially

created by the proposed one-time TS change

have been considered. The relevant

components important to safety which are

potentially affected are the containment

structure, plant systems, and containment

penetrations. There are no physical or

operational changes to any plant equipment

associated with the proposed TS change.

Therefore, the probability or consequences of

a malfunction of equipment important to

safety is not increased.

The proposed change introduces the

possibility that primary containment leakage

in excess of the allowable value (i.e., La)

would remain undetected during the

proposed 24 month extension of the interval

between the Type A tests. The types of

mechanisms which would cause degradation

of the primary containment can be

categorized into two types. These are: (1)

Degradation due to work which is performed

as part of a modification or maintenance

activity on a component or system (i.e.,

activity-based), or; (2) degradation resulting

from a time-based failure mechanism.

A review of the history of the PBAPS, Unit

3 CILRT results was performed to evaluate

the risk of activity-based and time-based

degradation. This review has determined that

the potential for a time-based and activitybased failure is minimal. The PBAPS LLRT

program would identify most types of

penetration leakage. The LLRT program

involves measurement of leakage from Type

B and Type C primary containment

penetrations as defined in 10 CFR part 50,

appendix J.

The 10 CFR part 50, appendix J, Type B

tests are intended to detect local leaks and to

measure leakage across pressure containing

or leakage-limiting boundaries other than

values, such as containment penetrations

incorporating resilient seals, gaskets,

expansion bellows, flexible seal assemblies,

door operating mechanism penetrations that

are part of the containment system, doors,

and hatches. 10 CFR part 50, appendix J,

Type C testing is intended to measure reactor

system primary containment isolation valve

leakage rates. The frequency of the Type B

and Type C testing is not being altered by the
Federal Register / Vol. 60, No. 99 / Tuesday, May 23, 1995 / Notices 27341

proposed TS change. The acceptance

criterion for Type B and Type C leakage is

0.6 La (i.e., 0.3% wt/day) which, when

compared to the Type A test acceptance

criterion of 0.75 La (i.e., 0.375% wt/day), is

a significant portion of the Type A test

allowable leakage.

The proposed TS change only extends the

interval between two consecutive Type A

tests. The Type B and Type C tests will be

performed as required. The Type B and Type

C tests will continue to be used to confirm

that the containment isolation valves and

penetrations have not degraded. Containment

system components that would not be tested

are the containment structure itself and

small-diameter instrumentation lines. Timebased degradation of any of the

instrumentation lines would not likely be

identified by faulty instrument indication or

during instrument calibrations that will be

performed during the PBAPS, Unit 3

refueling outage 10. In examining the

potential for a time-based failure mechanism

that could cause significant degradation of

the containment structure, we concluded that

the risk, if any, of such a mechanism is small

since the design requirements and fabrication

specifications established for the

containment structure are in themselves

adequate to ensure containment leak tight

integrity.

Based on the above evaluation, we have

concluded that the proposed TS change will

have a negligible impact on the consequences

of any accident previously evaluated.

Although this review concluded that the

risk of undetected primary containment

degradation is not increased, the Individual

Plan Examination (IPE) for PBAPS, Units 2

and 3, was also reviewed in order to access

the impact of exceeding the primary

containment allowable leakage rate, if a nonmechanistic activity type (i.e., time-based)

failure were to occur. The IPE included an

evaluation of the effect of various

containment leakage sizes under different

scenarios. The IPE results showed that a

containment leakage rate of 35% wt/day

would represent less than a 5% increase in

risk to the public of being exposed to

radiation. This evaluation was based on a

study performed by Oak Ridge National

Laboratory for light water reactors that

evaluated the impact of leakage rates on

public risk. As stated earlier, the current

value of La for PBAPS, Unit 3, is 0.5% wt/

day, which is significantly less than the 35%

wt/day discussed in the IPE evaluation.

Therefore, the proposed TS change

involving a one-time extension of the Type

A test interval and performing the Type A

test after the second appendix J 10-year

service period will not involve an increase in

the probability or consequences of an

accident previously evaluated.

2. The proposed TS change does not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed change is an increase of a

surveillance test interval and does not make

any physical or operational changes to

existing plant systems or components.

Primary containment acts as an accident

mitigator not initiator. Therefore, the

possibility of a different type of accident than

any previously evaluated or the possibility of

a different type of equipment malfunction is

not introduced.

Therefore, the proposed TS change does

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed TS change does not

involve a significant reduction in a margin of

safety.

The total primary containment leakage rate

ensures that the total containment leakage

volume will not exceed the value assumed in

the safety analyses at the peak accident

pressure. As an added conservatism, the

measured overall leakage rate is further

limited to less than or equal to 0.75 La during

performance of periodic tests to account for

possible degradation of the containment

leakage barriers between leakage tests. There

is the potential that containment degradation

could remain undetected during the

proposed 24 month surveillance interval

extension and result in the containment

leakage exceeding this allowable value

assumed in safety analysis. A review of the

history of the PBAPS, Unit 3 CILRT results

was performed to evaluate the risk of

activity-based and time-based degradation.

This review has determined that the potential

for a time-based and activity-based failure is

minimal. The PBAPS LLRT program would

identify most types of penetration leakage.

The LLRT program involves measurement of

leakage from Type B and Type C primary

containment penetrations as defined in 10

CFR part 50, appendix J.

The 10 CFR part 50, appendix J, Type B

tests are intended to detect local leaks and to

measure leakage across pressure containing

or leakage-limiting boundaries other than

valves, such as containment penetrations

incorporating resilient seals, gaskets,

expansion bellows, flexible seal assemblies,

door operating mechanism penetrations that

are part of the containment system, doors,

and hatches. 10 CFR part 50, appendix J,

Type C testing is intended to measure reactor

system primary containment isolation valve

leakage rates. The frequency of the Type B

and Type C testing is not being altered by the

proposed TS change.

Therefore, we have concluded that the

proposed extended test interval would not

result in a non-detectable PBAPS, Unit 3

primary containment leakage rate in excess of

the allowable value (i.e., 0.5% wt/day)

established by the TS and 10 CFR part 50,

appendix J.

Therefore, the proposed TS change does

not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Education

Building, Walnut Street and

Commonwealth Avenue, Box 1601,

Harrisburg, Pennsylvania 17105.

Attorney for Licensee: J.W. Durham,

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

Counsel, Philadelphia Electric

Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101.

NRC Project Director: John F. Stolz.

Public Service Electric & Gas Company,

Docket No. 50–272, Salem Nuclear

Generating Station, Unit No. 1, Salem

County, New Jersey

Date of amendment request: April 4,

1995.

Description of amendment request:

The amendment would provide a onetime interval extension for the Type A

test required by 10 CFR part 50,

appendix J. The extension would allow

the test to be conducted during the

thirteenth refueling outage, rather than

the twelfth refueling outage.

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. Will not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

The proposed change will provide a onetime exemption from 10 CFR part 50,

appendix J Section III.D.1(a) leak rate test

schedule requirement. This change will

allow for a one-time test interval for Type A

Integrated Leak Rate Tests (ILRTs) of 65+/

¥10 months.

Leak rate testing is not an initiating event

in any accident, therefore, this proposed

change does not involve a significant

increase in the probability of a previously

evaluated accident.

Type A tests are capable of detecting both

local leak paths and gross containment

failure paths. The history at Salem

Generating Station Unit 1 (SGS1)

demonstrates that Type B and C Local Leak

Rate Tests (LLRTs) have consistently

detected any excessive local leakages. SGS1

has passed all of its ILRTs with significant

margin.

Administrtive controls govern the

maintenance and testing of containment

penetrations such that the probability of

excessive penetration leakage due to

improper maintenance or valve misalignment

is very low. Following any maintenance that

could affect the leakage characteristics of any

containment penetration, an LLRT is

performed to ensure acceptable leakage

levels. Following any LLRT on a containment

isolation valve, an independent valve

alignment check is performed before

declaring the penetration OPERABLE.

Therefore, Type A testing is not necessary to

ensure acceptable leakage rates through

containment penetrations.

While Type A testing is not necessary to

ensure acceptable leakage rates through
27342 Federal Register / Vol. 60, No. 99 / Tuesday, May 23, 1995 / Notices

containment penetrations, Type A testing is

necessary to demonstrate that there are no

gross containment failures. Structural failure

of the containment is considered to be a very

unlikely event, and in fact, since SGS1 has

been in operation, it has never failed a Type

A ILRT. Therefore, a one-time exemption

increasing the interval for performing an

ILRT does not result in a significant decrease

in the confidence in the leak tightness of the

containment structure.

Therefore, this proposed change does not

result in a significant increase of the

probability or consequences of any

previously evaluated accident.

2. Will not create the possibility of a new

or different kind of accident from any

accident previously evaluated.

This proposed change allows a one-time

interval of 65+/¥10 months for the next

ILRT. The method of performing the test is

not changed. No new accident modes are

created by extending the testing intervals. No

safety-related equipment or safety functions

are altered as a result of this change. A onetime extension of the ILRT test interval has

no influence on, nor does it contribute in any

way to, the possibility of a new or different

kind of accident or malfunction from those

previously analyzed.

3. Will not involve a significant reduction

in a margin of safety.

The purpose of the existing schedule of

ILRTs is to ensure that the release of

radioactive materials will be restricted to

those leak paths and leak rates assumed in

accident analyses. The relaxed schedule for

ILRTs does not allow for relaxation of Type

B and C LLRTs. Therefore, methods for

detecting local containment leak paths and

leak rates are unaffected by this proposed

change. Given that the test history for ILRTs

shows no failure during plant life, a one-time

increase of the test interval does not lead to

a significant probability of creating a new

leakage path or increased leakage rates, and

the margin of safety inherent in existing

accident analyses is maintained. Therefore,

this change does not involve a significant

reduction kin 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: Salem Free Public Library, 112

West Broadway, Salem, New Jersey

08079.

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW, Washington,

DC 20005–3502.

NRC Project Director: John F. Stolz.

Public Service Electric & Gas Company,

Docket No. 50–272, Salem Nuclear

Generating Station, Unit No. 1, Salem

County, New Jersey

Date of amendment request: May 4,

1995.

Description of amendment request:

The amendment would authorize a onetime extension for the Type A test

(overall integrated containment leakage

rate) that is required by 10 CFR part 50,

appendix J. The current Technical

Specification would require that this

test be conducted by July 7, 1995. The

amendment would allow this test to be

conducted by November 30, 1995.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Will not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

The proposed change involves no

hardware changes, no changes to the

operation of any systems or components, and

no changes to existing structures. This

change is temporary, allowing a one-time

extension of a specific surveillance

requirement for cycle 12 to allow

surveillance testing to coincide with the

twelfth refueling outage. The proposed

surveillance interval extension is short and

will not result in any significant reduction in

structural reliability nor will the extension

affect the ability of the structure in

performing its intended functions. to

preclude the possibility of an undetected

containment failure/leakage at a valve or

penetration seal, Type ‘‘B’’ and ‘‘C’’ tests will

continue to be performed as required by the

Technical Specifications. Therefore, this

change will not involve a significant increase

in the probability or consequences of any

accidents previously evaluated.

2. Will not create the possibility of a new

or different kind of accident from any

previously evaluated.

Extending the surveillance interval for the

performance of specific testing will not create

the possibility of any new or different kinds

of accident. No changes are required to any

system configurations, plant equipment, or

analyses. Therefore, this change will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Will not involve a significant reduction

in a margin of safety.

The proposed change will not alter any

assumptions, initial conditions, or results of

any accident analyses. The safety limits

assumed in the accident analyses and the

design function of the structure required to

mitigate the consequences of any postulated

accidents will not be changed since only the

surveillance interval is being extended.

Historical performance indicates a high

degree of reliability, and surveillance testing

performed during continued plant operation

will verify that Salem 1 will remain within

analyzed limits. Consequently, the change

does not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Salem Free Public library, 112

West Broadway, Salem, New Jersey

08079.

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW.,

Washington, DC 20005–3502.

NRC Project Director: John F. Stolz.

Public Service Electric & Gas Company,

Docket Nos. 50–272 and 50–311, Salem

Nuclear Generating Station, Unit Nos. 1

and 2, Salem County, New Jersey

Date of amendment request: April 18,

1995.

Description of amendment request:

The amendments would delete the

quarterly leak rate test for the

containment pressure-vacuum relief

valves which is presently required

because of the valves’ resilient seat

material. The resilient valve seat

material will be replaced with a hard

seat (metal to metal) design. The valves

would still remain in the 10 CFR part 50

appendix J, Type C leak rate test

program.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Do not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

The containment pressure/vacuum relief

valves are normally closed, and are used

under administrative control to maintain

containment internal pressure within ¥1.5

psig and +0.3 psig, as required by SGS

Technical Specifications. The pressure/

vacuum relief valves are relied upon for

containment isolation and automatically

close on high containment pressure or high

containment atmosphere radioactivity. The

pressure/vacuum relief system does not affect

the probability of any previously evaluated

accident.

The containment isolation function of the

pressure/vacuum relief valves limits the

consequences of a radiological release inside

containment (i.e., Loss of Coolant Accident).

The proposed changes to eliminate quarterly

pressure drop (leak rate) testing would not

increase the consequences of any previously

evaluated accident. The valve flow

characteristics and closure time requirements

are not affected. The valves will continue to

be subject to the Type C leak rate test criteria

of 10 CFR part 50, appendix J. The deletion

of the augmented quarterly test requirement

is justified by replacement of the resilient
Federal Register / Vol. 60, No. 99 / Tuesday, May 23, 1995 / Notices 27343

valve seat material (which has a history of

degradation and loss of leaktightness) with a

metal to metal seating design.

2. Do not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

Eliminating quarterly leak rate testing

based on improved valve design would not

result in any new or different kind of

accident. The valves would continue to

perform the containment isolation function

consistent with the plant safety analyses, and

would not adversely affect the initiation or

progression of any accident sequence.

(3) Do not involve a significant reduction

in a margin of safety.

This proposal involves replacement of the

existing pressure/vacuum relief valves,

which have resilient seating material, with

valves using a hard seat (metal to metal

design). Based on the improved design and

operating experience of the replacement

valves, augmented quarterly leak rate testing

is no longer necessary or appropriate to

verify leaktightness of the valves. Periodic

leak rate testing will continue to be

performed in accordance with 10 CFR part

450, appendix J. The pressure/vacuum relief

valves will continue to maintain their

containment isolation capability such that no

margin of safety is affected by the proposed

changes.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Salem Free Public library, 112

West Broadway, Salem, New Jersey

08079.

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW.,

Washington, DC 20005–3502.

NRC Project Director: John F. Stolz.

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plants, Units 1 and 2, Hamilton

County, Tennessee

Date of amendment request: May 3,

1995 (TS 93–09).

Descripton of amendment request:

The proposed change would revise the

implementation schedule for

Amendment Nos. 182 and 174 from that

stated in the amendments when they

were approved by the Commission by

letter dated May 24, 1994. As issued, the

amendments reflected the licensee’s

plans to implement the changes for both

units during the Unit 2 Cycle 6 refueling

outage. However, by letter dated August

19, 1994, the licensee requested that

implementation be delayed to 1995.

This request was granted by

Amendment Nos. 188 and 180 for Units

1 and 2 respectively by letter dated

October 17, 1994. By letter dated May 3,

1995, the licensee informed the staff

that evaluation of the design changes

have concluded that significant safety

risks would be involved with

modification activities associated with

installation. Therefore, the licensee has

requested that implementation of the

amendment be changed to specify that

the amendment will be implemented

along with the related plant

modifications, without specifying the

date when the modifications would be

performed. No changes to the technical

specification pages other than those

approved when the amendments were

issued are needed.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has determined that the no

significant hazards consideration exists.

This analysis was provided in the

original submittal for the amendment

from the licensee dated October 1, 1993,

and was used in the preparation of the

amendments. The licensee has

determined that this analysis remains

valid for the proposed revision and that

the changes do not constitute a

significant hazard. The staff previously

issued the proposed finding in the

Federal Register (59 FR 4947 and 59 FR

47182) and there were no public

comments on the finding. This analysis

is reproduced as follows:

TVA has evaluated the proposed technical

specification (TS) change and has determined

that it does not represent a significant

hazards consideration based on criteria

established in 10 CFR 50.92(c). Operation of

Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed revision supports the

implementation of design logic and setpoint

changes to the loss-of-power relaying. This

relaying is designed to ensure adequate

voltage is available to safety-related loads in

order to enhance their operability and

support accident mitigation functions and to

provide for auxiliary feedwater (AFW) pump

starts. The design changes alter relay logic

and delete unnecessary relaying, but do not

change the diesel generator (D/G) start and

load-shedding actuations that result from

loss-of-power conditions. Therefore, no new

actuations or functions have been created;

and because the existing and proposed

functions provide for accident mitigation

considerations that are not the source of an

accident, the probability of an accident is not

increased. The deletion of the 6.9-kilovolt

shutdown board normal-feeder undervoltage

relays actually reduces the potential for

inadvertent shutdown board blackouts as a

result of short-duration voltage transients or

instrument failures.

The setpoints and time delays for loss-ofpower functions have been modified based

on the guidelines developed by the Electrical

Distribution System Clearinghouse as

evaluated and determined through detailed

analysis by TVA. This design is documented

in TVA Calculations SQN–EEB–MS–TI06–

0008, 27DAT, and DS–1–2 and is available

for NRC review at the SQN site. The assigned

values are conservative settings that will

ensure adequate voltage is supplied to safetyrelated loads for accident mitigation and

safety functions under normal, degraded, and

loss-of-offsite-power voltage conditions with

appropriate time delays to prevent damage to

electrical loads and minimize premature or

unnecessary actuations. The identification of

loss-of-voltage conditions is enhanced by the

design changes to ensure the timely

sequencing of loads onto the D/G and the

initiation of AFW pump starts for accident

mitigation. Because there are no reductions

in safety functions resulting from the design

logic, setpoint, and time-delay changes to the

loss-of-power instrumentation and offsite

dose levels for postulated accidents will not

be increased, the consequences of an

accident are not increased.

The applicable mode addition, TS 3.0.4

exclusion deletion, and response time

measurement clarification incorporated in

the proposed change do not affect plant

functions. These changes reflect the

requirements that SQN has been maintaining

and serve to clarify the requirements to

provide consistency of application and easier

understanding. The AFW footnote addition

and bases revision only clarify operability

conditions that are consistent with the plant

design for the AFW pump and loss-of-power

instrumentation. Because there are no

changes to plant functions or operations,

these revisions have no impact on accident

probabilities or consequences.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

As described above, the loss-of-power

instrumentation ensures adequate voltage to

safety-related loads by initiating D/G starts

and load shedding and provides for AFW

pump starting, but is not considered to be the

source of an accident. Although the design

logic, setpoint, and time-delay actuation

criteria have changed, the output functions to

various plant systems that actuate for load

shedding and D/G starts remain the same.

Therefore, actuation criteria have been

affected, but not safety functions, and the

TVA evaluation has confirmed that the new

design enhances the ability to maintain

adequate voltage to support safety functions.

Since safety functions have not changed and

the new loss-of-power instrumentation

design continues to support operability of

safety-related equipment, no new or different

accident is created.

The applicable mode addition, TS 3.0.4

exclusion deletion, and response time

measurement clarification, as well as the

AFW operability clarifications, do not affect

plant functions and will not create a new

accident.

3. Involve a significant reduction in a

margin of safety.

The proposed loss-of-power TS changes

support design logic, setpoint, and timedelay requirements that have been verified by
27344 Federal Register / Vol. 60, No. 99 / Tuesday, May 23, 1995 / Notices

TVA analysis to provide acceptable voltage

levels for safety-related components. In

determining the acceptability of these voltage

levels, the minimum voltage for operation as

well as detrimental component heating

resulting from sustained degraded-voltage

conditions were considered. This design

ensures that safety-related loads will be

available and operable for normal and

accident plant conditions. The applicable

mode addition, TS 3.0.4 exclusion deletion,

response time measurement clarification, and

AFW operability clarifications provide

enhancements to TS requirements and do not

affect plant functions. Therefore, no safety

functions are reduced by these changes and

there is no reduction in the margin of safety.

The NRC has reviewed the licensee’s

analysis and, based on this review, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402.

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H,

Knoxville, Tennessee 37902.

NRC Project Director: Frederick J.

Hebdon.

The Cleveland Electric Illuminating

Company, Centerior Service Company,

Duquesne Light Company, Ohio Edison

Company, Pennsylvania Power

Company, Toledo Edison Company,

Docket No. 50–440, Perry Nuclear

Power Plant, Unit No. 1, Lake County,

Ohio

Date of amendment request: April 28,

1995.

Description of amendment request:

The proposed amendment would extend

for one more operating cycle an

exception to Limiting Condition for

Operation (LCO) 3.0.4 as it applies to

the Technical Specification for the main

steam isolation valve leakage control

system. The existing LCO 3.0.4

exception was issued by Amendment 63

to the Operating License, and will

expire upon completion of the fifty

cycle of plant operation. The extension

is proposed for the duration of the sixth

cycle of operation to permit completion

of activities necessary to implement the

most appropriate permanent resolution

for the issue of secondary containment

bypass leakage through the main steam

line drains.

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.

This License Amendment application

proposes an extension for one operating cycle

of the exception to Limiting Condition for

Operation for Operation (LCO) 3.0.4 as it

applies to the Technical Specification for the

MSIV [main steam isolation valve] Leakage

Control system. This extension is proposed

for the duration of the sixth cycle of PNPP

(Perry Nuclear Power Plant) operation, to

permit completion of activities necessary to

implement the most appropriate permanent

resolution for the issue of secondary

containment bypass leakage through the

Main Steam Line drains. During the sixth

cycle, the drains will remain in their current

configuration, which seals off the bypass

leakage path. The sealed drain path results in

a temporary inoperability of the Inboard

MSIV Leakage control system (MSIV–LCS)

subsystem when the plant is operated below

50% power, due to condensate build-up in

the bottom of the steam lines between the

MSIVs. The requested 3.0.4 exception is

necessary to permit plant startups with this

temporary inoperability, for the duration of

the sixth operating cycle.

The probability of occurrence of a

previously evaluated accident is not affected

by the proposed extension of the LCO 3.0.4

exception since no change to the plant or to

the manner in which the plant is operated is

involved. The existing plant configuration

will be maintained for another operating

cycle, and possible concerns resulting from

that configuration have been analyzed. The

extra weight of the water pooled between the

MSIVs was analyzed with respect to piping

supports and seismic considerations and was

found to be acceptable, and any condensate

that is carried past the outboard MSIVs will

be drained to the condenser by drain

connections downstream of the outboard

MSIVs before it can reach the turbine. The

temporary inoperability of the Inboard

MSIV–LCS when below 50% power has no

impact on accident initiation probability,

since LCS does not serve to prevent

accidents, but is only used in mitigating the

consequences of Loss of Coolant Accidents

that have already occurred.

The consequences of an accident are not

significantly increased in that the Outboard

MSIV–LCS will be available to perform the

MSIV–LCS function by mitigating the

consequences of a Loss of Coolant Accident

(LOCA) during the temporary period in

which the Inboard MSIV–LCS is unavailable.

Any condensate that is carried past the

outboard MSIVs will be drained to the

condenser by drain connections downstream

of the outboard MSIVs; therefore no

impairment of the Outboard MSIV–LCS will

result from condensed water.

The Action statement for one inoperable

LCS subsystem remains the same, and the

limits plant operation to the previously

established 30-day Allowable Outage Time.

The Action required if both the subsystems

of MSIV–LCS were to become inoperable also

remains the same. The MSIV function of

isolating the Main Stream Lines is also

unaffected by the existing plant

configuration, since MSIV performance will

not be affected by the existence of

accumulated water in the bottom of the steam

lines between the MSIVs during the plant

operation below 50% power. Therefore, if

necessary, the Main Steam Lines will be

isloated, and leakage past the MSIVs will be

routed for filtration as in the design-basis

radiological analyses, and the consequences

of previously evaluated accidents will remain

unaffected.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed change to permit

inoperability of the Inboard MSIV–LCS

during periods of startup and power

ascension to 50% RTP (rated thermal power)

and during shutdown below 50% RTP does

not create the possibility of a new or different

kind of accident from any previously

evaluated. The Inboard MSIV–LCS is only

credited during a Recirculation Line Break

LOCA wherein Reactor Coolant System

depressurization occurs. The temporary

unavailability of the Inboard MSIV–LCS. the

amendment to the Technical Specifications is

an administrative change that does not

involve any change to the current plant

design or methods of operation. No new

plant equipment failure modes or accident

initiators are introduced by the extension of

the LCO 3.0.4 exception.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The response to the Recirculation Line

Break LOCA will not be significantly affected

since the Outboard MSIV–LCS can be

assumed to be available. Allowing entry into

Operational Conditions 1, 2 and 3 while

utilizing the existing Action statement does

not significantly reduce the margin of safety

since the duration of time allowed for

remaining in that Action statement is not

increased. The proposed change will have no

adverse impact on the reactor coolant system

pressure boundary nor will any other system

protective boundary or safety limit be

affected.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room Location:

Perry Public Library, 3753 Main Street, Perry,

Ohio 44081.

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts & Trowbridge,

2300 N Street, NW., Washington, DC

20037.

NRC Project Director: Gail H. Marcus.
Federal Register / Vol. 60, No. 99 / Tuesday, May 23, 1995 / Notices 27345

The Cleveland Electric Illuminating

Company, Centerior Service Company,

Duquesne Light Company, Ohio Edison

Company, Pennsylvania Power

Company, Toledo Edison Company,

Docket No. 50–440, Perry Nuclear

Power Plant, Unit No. 1, Lake County,

Ohio

Date of amendment request: May 1,

1995.

Description of amendment request:

The proposed amendment would

eliminate selected response time testing

requirements, and incorporate guidance

provided by Generic Letter 93–08,

‘‘Relocation of Technical Specification

Tables of Instrument Response Time

Limits.’’

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 changes do not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

For those proposed changes dealing with

the elimination of selected response time test

requirements, the purpose of the proposed

Technical Specification change is to

eliminate response time testing requirements

for selected components in the Reactor

Protection System, Isolation system, and

Emergency Core Cooling System. The BWR

Owners’ Group has completed an evaluation

which demonstrates that the response time

testing is redundant to other Technical

Specification required testing. These other

tests, in conjunction with actions taken in

response to NRC Bulletin 90–01, ‘‘Loss of

Fill-Oil in Transmitters Manufactured by

Rosemount,’’ and Supplement 1, are

sufficient to identify failure modes or

degradations in instrument response time

and ensure operation of the associated

systems within acceptable limits. There are

no known failure modes that can be detected

by response time testing that cannot also be

detected by the other required Technical

Specification testing. This evaluation was

documented in NEDO–32291, ‘‘System

Analyses for Elimination of Selected

Response Time Testing Requirements,’’

January 1994, and the letter from T. Green to

P. Loeser dated April 15, 1994 which were

approved by an NRC Safety Evaluation dated

December 28, 1994. The applicability of this

evaluation to the Perry Nuclear Power Plant

(PNPP) has been confirmed. In addition,

PNPP will complete the additional actions

identified in the NRC staff’s Safety

Evaluation of NEDO–32291.

Because of the continued application of

other existing Technical Specification

required tests such as channel calibrations,

channel checks, channel functional tests, and

logic system functional tests, the response

times of these systems will be maintained

within the acceptance limits assumed in

plant safety analysis and required for

successful mitigation of an initiating event.

The proposed Technical Specification

changes do not affect the capability of the

associated systems to perform their intended

function within their required response time,

nor do the proposed changes themselves

affect the operation of any equipment. As a

result the proposed changes dealing with

elimination of selected response time tests do

not involve a significant increase in the

probability or the consequences of an

accident previously evaluated.

For those changes dealing with moving the

surveillance requirement for ECCS

RESPONSE TIME testing from the

instrumentation section to the system section

of the Technical Specifications, no change in

testing requirements (other than the

elimination of the instrument loops

implemented as part of the NEDO–32291

changes) has been introduced. The relaxation

in Applicability does not increase the

probability or the consequences of an

accident previously evaluated, since there are

no design basis events during

OPERATIONAL CONDITION 4 and 5 where

ECCS systems are relied upon.

For those changes dealing with relocation

of the response time limits from Technical

Specification Tables and into the Updated

Safety Analysis Report (USAR), the proposed

changes are administrative in nature in that

the test requirements and time limits are still

requirements, but the placement of the limits

have been relocated from the Technical

Specifications and into the USAR. Therefore

these changes do not involve a significant

increase in the probability or the

consequences of an accident previously

evaluated.

2. The changes do not create the possibility

of a new or different kind of accident from

any previously evaluated.

None of the proposed Technical

Specification changes affect the capability of

the associated systems to perform their

intended function within the acceptance

limits assumed in plant safety analyses and

required for successful mitigation of an

initiating event. The proposed changes also

do not change the manner in which any plant

equipment is operated. Therefore, the

proposed changes do not create the

possibility of a new or different kind of

accident from any previously evaluated.

3. The changes do not involve a significant

reduction in the margin of safety.

The current Technical Specification

response times are based on the maximum

allowable value assumed in the plant safety

analyses. These analyses conservatively

establish the margin of safety. As described

above, the proposed Technical Specification

changes do not affect the capability of the

associated systems to perform their intended

function within the allowed response time

used as the basis for the plant safety analyses.

Plant and system response to an initiating

event will remain in compliance within the

assumptions of the safety analyses, and

therefore the margin of safety is not affected.

Although not explicitly evaluated, the

proposed Technical Specification changes

dealing with response time testing

elimination will provide an improvement to

plant safety and operation by reducing the

time safety systems are unavailable, reducing

safety system actuation, reducing plant

shutdown risk, limiting radiation exposure to

plant personnel, and eliminating the

diversion of key personnel to conduct

unnecessary testing. Therefore, the proposed

changes do not result in a significant

reduction in a margin of safety, and may

result in an overall increase in the margin of

safety.

The changes dealing with relocation of the

time response limits from the Technical

Specifications to the USAR is an

administrative change that does not affect

either the requirements to perform response

time testing or the limits associated with the

response time tests. Future changes to the

limits will be controlled by 10 CFR 50.59.

Therefore, this portion of the change does not

result in a significant decrease 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: Perry Public Library, 3753

Main Street, Perry, Ohio 44081.

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts & Trowbridge,

2300 N Street, NW., Washington, DC

20037.

NRC Project Director: Gail H. Marcus.

Union Electric Company, Docket No.

50–483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of amendment request: April 26,

1995.

Description of amendment request:

the proposed amendment would revise

Technical Specification (TS)

Surveillance Requirements 3/4.7.6 and

associated Bases to reduce the upper

limit on the control room filtration

subsystem flow rate. It would also adopt

ASTM D–3803–1989 as the laboratory

testing standard for control room

filtration and control building

pressurization charcoal absorber.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The proposed revision does not involve a

significant hazards consideration because

operation of Callaway Plant with this change

would not:

(1) Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Overall protection system performance will

remain within the bounds of the accident

analysis documented in FSAR Chapter 15

* * * since no hardware changes are

proposed.
27346 Federal Register / Vol. 60, No. 99 / Tuesday, May 23, 1995 / Notices

The Control Room Emergency Ventilation

System (CREVS) will continue to function in

a manner consistent with the above analysis

assumptions and the plant design basis.

There will be no degradation in the

performance of or an increase in the number

of challenges to equipment assumed to

function during an accident situation.

These Technical Specification revisions do

not involve any hardware changes nor do

they affect the probability of any event

initiators. The change to the control room

filtration flow rate is consistent with the

original licensing basis and will ensure an

average atmosphere residence time of greater

than or equal to 0.25 sec. There will be no

change to ESF (engineered safety feature)

actuation setpoints or accident mitigation

capabilities. The laboratory testing will

demonstrate the required absorber

performance after a design basis LOCA (lossof-coolant accident).

The control room dose analyses assume a

total flow rate through the control room

filtration units that is less than the proposed

upper limit. As such, there will be no

changes required to the control room dose

analyses.

Based on the above, these Technical

Specification changes will not increase the

probability or consequences of an accident or

malfunction.

(2) Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

As discussed above, there are no hardware

changes associated with these Technical

Specification revisions nor are there any

changes in the method by which any safetyrelated plant system performs its safety

function.

Revisions to the Surveillance Requirements

for the CREVS will ensure that the control

room does analysis assumptions made in

support of OL (operating license)

Amendment No. 96 are valid. Changes to the

control room filtration unit flow rate are

more limiting than that currently specified

and have already been implemented by

resetting the open limit switches on the

respective units’ outlet dampers. This flow

rate is consistent with the design basis for the

filtration units as originally licensed.

No new accident scenarios, transient

precursors, failure mechanisms, or limiting

single failures are introduced as a result of

these changes. There will be no adverse effect

or challenges imposed on any safety-related

system as a result of these changes.

Therefore, the possibility of a new or

different kind of accident is not created.

(3) Involve a significant reduction in a

margin of safety.

There will be no margin reduction since

these changes are in the conservative

direction and have already been approved by

NRC via the approval of OL Amendment No.

96. The reduced upper bound flow rate for

the control room filtration units is consistent

with their design basis and will maintain an

average atmosphere residence time greater

than or equal to 0.25 sec under both clean

and dirty filter conditions. The new charcoal

absorber sample laboratory testing protocol is

more stringent than the current testing

practice and more accurately demonstrates

the required performance after a design basis

LOCA.

There will be no effect on the manner in

which safety limits or limiting safety system

settings are determined nor will there be any

effect on those plant systems, necessary to

assure the accomplishment of protection

functions. There will no impact on the

overpower limit, DNBR (departure from

nucleate boiling ratio) limits, FQ, F[delta]H,

LOCA PCT (peak cladding temperature), peak

local power density, or any other margin of

safety. These changes will ensure that the

criteria of GDC 19 are met.

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: Callaway County Public

Library, 710 Court Street, Fulton,

Missouri 65251.

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW,

Washington, DC 20037.

NRC Project Director: Gail H. Marcus.

Wisconsin Electric Power Company,

Docket Nos. 50–266 and 50–301, Point

Beach Nuclear Power Plant, Unit Nos.

1 and 2, Town of Two Creeks.

Manitowoc County, Wisconsin

Date of amendment request: April 17,

1995.

Description of amendment request:

The proposed amendment would

modify Technical Specification (TS)

Section 15.6.2, ‘‘Organization,’’ and TS

Section 15.6.3, ‘‘Facility Staff

Qualifications.’’ The training

requirements for the Operations

Manager and other staff would be

changed to provide staffing flexibility.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration which is presented below:

The proposed change does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated; create the possibility of a new or

different kind of accident from any

previously evaluated; or create the possibility

of a new or different kind of accident from

any previously evaluated.

1. The proposed change affects only an

administrative control, which was based on

industry guidance in ANSI N18.1–1971, that

recommended the Operations Manager hold

an SRO (senior reactor operator) license. This

administrative control is being updated to

meet the current guidance in ANSI/ANS 3.1–

1987.

2. The proposed qualification requirements

for the Operations Manager ensures the

individual filling the position meets

knowledge levels equivalent to the present

requirements. It also ensures that individuals

responsible for directing the activities of

licensed operators continue to hold SRO

licenses as required by 10 CFR 50.54(l).

3. Since the proposed specifications ensure

regulatory requirements are met and ensures

knowledge levels equivalent to existing

license requirements for operations

management, the proposed changes are

considered administrative. The design of

plant systems and equipment is not being

altered. Plant operations will continue to be

directed and performed by qualified

personnel. Therefore, the probability or

consequences of accidents previously

evaluated are not affected, a new or different

type of accident is not created, nor is a

margin of safety reduced.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin

54241.

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts, and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Gail H. Marcus.

Wisconsin Electric Power Company,

Docket Nos. 50–266 and 50–301, Point

Beach Nuclear Power Plant, Unit Nos.

1 and 2, Town of Two Creeks,

Manitowoc County, Wisconsin

Date of amendment request: April 27,

1995.

Description of amendment request:

The proposed amendment would

modify Technical Specification (TS)

Table 15.3.5–1, ‘‘Engineered Safety

Features Initiation Instrument Setting

Limits,’’ and TS Table 15.35–3,

‘‘Engineered Safety Features.’’ Setting

limits would be modified and references

would be changed. The bases for TS

Section 15.3.5, ‘‘Instrumentation

System,’’ would also be changed to be

consistent with the TS changes.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration which is presented below:

1. Operation of this facility under the

proposed Technical Specifications will not

create a significant increase in the probability

or consequences of an accident previously

evaluated.

The probabilities of accidents previously

evaluated are based on the probability of

initiating events for these accidents.
Federal Register / Vol. 60, No. 99 / Tuesday, May 23, 1995 / Notices 27347

Initiating events for accidents previously

evaluated for Point Beach include: control

rod withdrawal and drops, CVCS (chemical

and volume control system) malfunction

(Boron Dilution), startup of an inactive

reactor coolant loop, reduction in feedwater

enthalpy, excessive load increase, losses of

reactor coolant flow, loss of external

electrical load, loss of normal feedwater, loss

of all AC power to the auxiliaries, turbine

overspeed, fuel handling accidents,

accidental releases of water liquid or gas,

steam generator tube rupture, steam pipe

rupture, control rod ejection, and primary

coolant system ruptures.

This license amendment request proposes

to correct some minor errors, include

appropriate operability requirements for the

modification to include the safety injection

signal in the time delay for the 4.16KV

degraded voltage protection logic, slightly

lower the degraded voltage setting limit,

change the format of the 4.16 KV degraded

voltage and loss of voltage setting limits, and

change the time delays associated with the

4.16 KV degraded voltage, 4.16 KV loss of

voltage and 480 V loss of voltage protection

functions.

These proposed changes do not cause an

increase in the probabilities of any accidents

previously evaluated because these changes

will not cause an increase in the probability

of any initiating events for accidents

previously evaluated. In particular, these

proposed changes affect time delay and

format of the setting limits associated with

the 4.16 KV degraded voltage, 4.16 KV loss

of voltage, and 480 V loss of voltage

protection functions. These are protection

functions and do not cause accidents.

The consequences of the accidents

previously evaluated in the PBNP FSAR

(Final Safety Analysis Report) are determined

by the results of analyses that are based on

initial conditions of the plant, the type of

accident, transient response of the plant, and

the operation and failure of equipment and

systems. The changes proposed in this

license amendment request provide

appropriate limiting conditions for operation,

action settlements, allowable outage times,

setting limits, and time delays for the Point

Beach Nuclear Plant Technical Specifications

for the 4.16 KV degraded voltage, 4.16 KV

loss of voltage, and 480 V loss voltage

protection functions.

The proposed changes affect functions that

are required to ensure the proper operation

of engineered features equipment. The

proposed changes do not increase the

probability of failure of this equipment or its

ability to operate as required for the

accidents previously evaluated in the PBNP

FSAR.

The modifications to reduce the time delay

limit associated with the 4.16 KV degraded

voltage protection function when the

degraded voltage condition is coincident

with a safety injection signal, have been

designed and installed in accordance with

the requirements for PBNP. The probability

of occurrence of degraded voltage conditions

at PBNP has not been increased. The

modifications and proposed Technical

Specifications will ensure the proper

operation of ESF (engineered safety feature)

equipment. These changes do not increase

the possibility of failure of this equipment.

Therefore, this proposed license

amendment does not affect the consequences

of any accident previously evaluated in the

Point Beach Nuclear Plant FSAR, because the

factors that are used to determine the

consequences of accidents are not being

changed.

2. Operation of this facility under the

proposed Technical Specifications change

will not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

New or different kinds of accidents can

only be created by new or different accident

initiators or sequences. New and different

types of accidents (different from those that

were originally analyzed for Point Beach)

have been evaluated and incorporated into

the licensing basis for Point Beach Nuclear

Plant. Examples of different accidents that

have been incorporated into the Point Beach

Licensing basis include anticipated transients

without scram and station blackout.

The changes proposed by this license

amendment request do not create any new or

different accident initiators or sequences

because these changes to the 4.16 KV

degraded voltage, 4.16 KV loss of voltage,

and 480 V loss of voltage protection

functions will not cause failures of

equipment or accident sequences different

than the accidents previously evaluated.

Therefore, these modifications and proposed

Technical Specification changes do not

create the possibility of an accident of a

different type than any previously evaluated

in the Point Beach FSAR.

3. Operation of this facility under the

proposed Technical Specifications change

will not create a significant reduction in a

margin of safety.

The margins of safety for Point Beach are

based on the design and operation of the

reactor and containment and the safety

systems that provide their protection.

The changes proposed by this license

amendment request provide the appropriate

setting limits and time delays for the 4.16 KV

degraded voltage, 4.16 KV loss of voltage,

and 480 V loss of voltage protection

functions. This ensures that the safety

systems that protect the reactor and

containment will operate as required. The

design and operation of the reactor and

containment are not affected by these

proposed changes. Therefore, the margins of

safety for Point Beach are not being reduced

because the design and operation of the

reactor and containment are not being

changed and the safety systems that provide

their protection that are being changed are

being modified in accordance with the

applicable design and installation

requirements for Point Beach Nuclear Plant.

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: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin

54241.

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts, and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Gail H. Marcus.

Previously Published Notices of

Consideration of Issuance of

Amendments to Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination

and Opportunity for a Hearing

The following notices were previously

published as separate individual

notices. The notice content was the

same as above. They were published as

individual notices either because time

did not allow the Commission to wait

for this biweekly notice or because the

action involved exigent circumstances.

They are repeated here because the

biweekly notice lists all amendments

issued or proposed to be issued no

significant hazards consideration.

For details, see the individual notice

in the Federal Register on the day and

page cited. This notice does not extend

the notice period of the original notice.

Northeast Nuclear Energy Company, et

al., Docket No. 50–336, Millstone

Nuclear Power Station, Unit No. 2, New

London County, Connecticut

Date of amendment request: April 21,

1995.

Description of amendment request:

The proposed amendment would revise

the Technical Specifications (TS)

3.1.2.4, ‘‘Charging Pumps-Operating,’’

by adding a note that indicates that the

provisions of TS 3.0.4 and 4.0.4 are not

applicable for entry into MODE 4 from

MODE 5.

Date of publication individual notice

in Federal Register: May 2, 1995 (60 FR

21558).

Expiration date of individual notice:

June 1, 1995.

Local Public Document Room

location: Learning Resource Center,

Three Rivers Community-Technical

College, Thames Valley Campus, 574

New London Turnpike, Norwich,

Connecticut 06360.

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
27348 Federal Register / Vol. 60, No. 99 / Tuesday, May 23, 1995 / Notices

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 Ch. 1, 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.

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:

August 19, 1994, as supplemented

November 3, 1994.

Brief description of amendment: The

amendment requests a line-item

improvement to the Radiological

Effluent Technical Specifications

pursuant to the guidance of Generic

Letter 89–01 and incorporates the

requirements of revised 10 CFR part 20

and 10 CFR 50.36a.

Date of issuance: May 1, 1995.

Effective date: May 1, 1994.

Amendment No.: 58.

Facility Operating License No. NPF–

63: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: October 12, 1994 (60 FR

51617) The Commission’s related

evaluation of the amendment, and

NRC’s response to the public comments

received, are contained in a Safety

Evaluation dated May 1, 1995.

No significant hazards consideration

comments received: Yes.

Local Public Document Room

location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh,

North Carolina 27605.

Duke Power Company, Docket Nos. 50–

269, 50–270, and 50–287, Oconee

Nuclear Station, Units 1, 2, and 3,

Oconee County, South Carolina

Date of application of amendments:

November 22, 1994, as supplemented by

letters dated January 30, March 2, March

13, and May 2, 1995.

Brief description of amendments: The

amendments revise Technical

Specification 3.8 to establish restricted

loading patterns and associated burnup

criteria for placing fuel in the Oconee

spent fule pools. In addition, the Design

Features sections associated with the

reactor and fuel storage are also revised.

Date of issuance: May 3, 1995.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment Nos.: 209, 209, and 206.

Facility Operating License Nos. DPR–

38, DPR–47, and DPR–55: The

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 15, 1995 (60 FR

8746); Re-Noticed March 29, 1995 (60

FR 16185).

The May 2, 1995, letter did not

change the scope of the November 22,

1994, application and the initial

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated May 3, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Oconee County Library, 501

West South Broad Street, Walhalla,

South Carolina 29691.

Duquesne Light Company, et al., Docket

No. 50–412, Beaver Valley Power

Station, Unit 2, Shippingport,

Pennsylvania

Date of application for amendment:

April 10, 1995, as supplemented April

12, 1995, and April 20, 1995.

Brief description of amendment: This

amendment revises Technical

Specification 4.6.2.2.d to delete the

reference to the specific test acceptance

criteria for the Containment

Recirculation Spray Pumps and replace

the specific test acceptance criteria with

reference to the developed head

required by the plant’s safety analysis.

In addition, the 18-month test frequency

would be replaced with the test

frequency requirements specified in the

IST Program. The current footnote (1)

pertaining to the performance of

recirculation spray pump 2RSS*P21A

would be deleted.

Date of issuance: May 3, 1995.

Effective date: May 3, 1995.

Amendment No.: 68.

Facility Operating License No. NPF–

73: Amendment revised the Technical

Specifications.

Public comments requested as to

proposed no significant hazards

consideration: Yes (60 FR 19417, April

18, 1995) That notice provided an

opportunity to submit comments on the

Commission’s proposed no significant

hazards consideration determination.

No comments have been received. The

notice also provided for an opportunity

to request a hearing by May 18, 1995,

but indicated that if the Commission

makes a final no significant hazards

consideration any such hearing would

take place after issuance of the

amendment.

The Commission’s related evaluation

of the amendment, finding of exigent

circumstances, and final determination

of no significant hazards consideration

are contained in a Safety Evaluation

dated May 3, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa,

Pennsylvania 15001.

Entergy Operations, Inc., Docket Nos.

50–313 and 50–368, Arkansas Nuclear

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

Arkansas

Date of amendment request: August

30, 1994 as supplemented January 19,

1995.

Brief description of amendments: The

amendments changed requirements

related to the site perimeter security

system.

Date of issuance: April 28, 1995.

Effective date: April 28, 1995.

Amendment Nos.: Unit 1—

Amendment No. 180; Unit 2—

Amendment No. 161

Facility Operating License Nos. DPR–

51 and NPF–6: Amendments revised the

licenses.

Date of initial notice in Federal

Register: April 12, 1995 (60 FR 18625).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated April 28, 1995.

No significant hazards consideration

comments received: No.
Federal Register / Vol. 60, No. 99 / Tuesday, May 23, 1995 / Notices 27349

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801.

Entergy Operations, Inc., Docket No.

50–382, Waterford Steam Electric

Station, Unit 3, St. Charles Parish,

Louisiana

Date of amendment request:

December 14, 1993, as supplemented by

letter dated March 3, 1995.

Brief description of amendment: The

amendment changed the Appendix A

Technical Specifications by removing

the reactor vessel material specimen

withdrawal schedule and by updating

the reactor coolant system pressuretemperature (P–T) curves.

Date of issuance: May 8, 1995.

Effective date: May 8, 1995.

Amendment No.: 106.

Facility Operating License No. NPF–

38.: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 19, 1994 (59 FR 2867).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 8, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, Louisiana 70122.

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:

October 20, 1994.

Brief description of amendments:

These amendments change the

definition of ‘‘core alteration’’ to

exclude the movement of items not

associated with reactivity. The second

change involves allowing the personnel

airlock (PAL) doors to remain open

during fuel movement and core

alterations under certain conditions.

Date of issuance: May 11, 1995.

Effective date: May 11, 1995.

Amendment Nos.: 173 and 167.

Facility Operating License No. DPR–

31 and DPR–41: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: November 9, 1994 (59 FR

55869).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated May 11, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Florida International

University, University Park, Miami,

Florida 33199.

GPU Nuclear Corporation, et al.,

Docket No. 50–219, Oyster Creek

Nuclear Generating Station, Ocean

County, New Jersey

Date of application for amendment:

February 28, 1995.

Brief description of amendment: The

amendment revises Technical

Specification (TS) Section 6.5.1.12 to

delete the requirement to render

determinations in writing with regard to

whether or not activities listed in TS

Sections 6.5.1.2 and 6.5.1.5 constitute

an unreviewed safety question. These

activities are changes to Appendix A

Technical Specifications (6.5.1.2) and

investigations of all violations of the

TSs (6.5.1.5). This change is consistent

with NUREG–1433 Standard Technical

Specifications General Electric Plants,

BWR/4 Revision 0, dated September 28,

1992.

Date of issuance: May 1, 1995.

Effective date: May 1, 1995.

Amendment No.: 180.

Facility Operating License No. DPR–

16.: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 29, 1995 (60 FR 16188).

The Commission’s related evaluation

of this amendment is contained in a

Safety Evaluation dated May 1, 1995.

No significant hazards consideration

comments received: Yes.

By letter dated April 5, 1995, Mr. Kent

W. Tosch, of the State of New Jersey

Department of Environmental Protection

commented that they concur with GPU

Nuclear’s rationale that these

unreviewed safety question reviews

serve no value since these activities

specifically require NRC review and

approval. The State official had no other

comments.

Local Public Document Room

location: Ocean County Library,

Reference Department, 101 Washington

Street, Toms River, NJ 08753.

Houston Lighting & Power Company,

City Public Service Board of San

Antonio, Central Power and Light

Company, City of Austin, Texas, Docket

Dos. 50–498 and 50–499, South Texas

Projects, Units 1 and 2, Matagorda

County, Texas

Date of amendment request: February

15, 1995.

Brief description of amendment: The

amendment modified Technical

Specification 4.6.2.3.a.2 (and associated

Bases) to reflect the reactor containment

fan cooler flow rate assumed in the

accident analysis and to specify that this

flow is provided by the component

cooling water system.

Date of issuance: May 2, 1995.

Effective date: May 2, 1995, to be

implemented within 30 days.

Amendment Nos.: Unit 1—

Amendment No. 74; Unit 2—

Amendment No. 63.

Facility Operating License Nos. NPF–

76 AND NPF–80. The amendment

revised the Technical Specifications.

Date of initial notice in Federal

Register: March 29, 1995 (60 FR 16189)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated May 2, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Wharton County Junior

College, J.M. Hodge Learning Center,

911 Boling Highway, Wharton, TX

77488.

Illinois Power Company and Soyland

Power Cooperative, Inc., Docket No.

50–461, Clinton Power Station, Unit No.

1. DeWitt County, Illinois

Date of application for amendment:

February 10, 1995.

Brief description of amendment: The

amendment changes Technical

Specification 3.3.2.1, ‘‘Control Rod

Block Instrumentation,’’ to revise two

surveillance requirements and their

associated notes for the Rod Withdrawal

Limiter mode of the Rod Pattern Control

System. The changes are consistent with

the Clinton Power Station Technical

Specifications prior to implementation

of the improved Technical

Specifications (Amendment No. 95) and

eliminates the potential for unnecessary

power reductions.

Date of issuance: May 2, 1995.

Effective date: May 2, 1995.

Amendment No.: 100.

Facility Operating License No. NPF–

62. The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: March 29, 1995. (60 FR 16190)

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 2, 1995.

No significant hazard consideration

comments received: No.

Local Public Document Room

location: The Vespasian Warner Public

Library, 120 West Johnson Street,

Clinton, Illinois 61727.

Niagara Mohawk Power Corporation,

Docket No. 50–410, Nine Mile Point

Nuclear Station, Unit 2, Oswego

County, New York

Date of application for amendment:

July 22, 1993, as supplemented

February 4, August 23, September 16,

October 6, and December 2, 1994, and

January 3, January 9, March 8, and April

10, 1995.
27350 Federal Register / Vol. 60, No. 99 / Tuesday, May 23, 1995 / Notices

Brief description of amendment: The

amendment modified Facility Operating

License No. NPF–69 and the NMP–2

TSs to authorize an increase in the

maximum power level of NMP–2 from

3323 megawatts thermal (MWt) to 3467

MWt. The amendment also approves

changes to the TSs to implement

uprated power operation.

Date of issuance: April 28, 1995.

Effective date: As of the date of

issuance to be implemented prior to

restart from refueling outage number 4.

Amendment No.: 66.

Facility Operating License No. NPF–

69: Amendment revises the Technical

Specifications and modifies Facility

Operating License No. NPF–69.

Date of initial notice in Federal

Register: March 16, 1994 (59 FR 12360).

The letters dated February 4, August

23, September 16, October 6, and

December 2, 1994, and January 3,

January 9, March 8, and April 10, 1995,

provided clarifying information that

did not change the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 28, 1995.

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:

October 18, 1994, a supplemented

February 21, 1995.

Brief description of amendment: The

amendment changes Surveillance

Requirement 4.6.1.2.a (Overall

Integrated Containment Leakage Rate

Tests) by revising the surveillance

interval for Type A tests from 40 plus

or minus 10 months to approximately

equal intervals during each 10-year

inservice period. The amendment also

removes a note that expired upon

completion of Cycle II refueling outage.

Date of issuance: May 3, 1995.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 187.

Facility Operating License No. DPR–

65. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 29, 1995 (60 FR 16191).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated may 3, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Learning Resource Center,

Three Rivers Community-Technical

College, Thames Valley Campus, 574

New London turnpike, Norwich, CT

06360.

Northeast Nuclear Energy Company, et

al., Docket No. 50-423, Millstone

Nuclear Power Station, Unit no. 3, New

London County, Connecticut

Date of application for amendment:

December 23, 1994.

Brief description of amendment: The

amendment changes the acceptance

criteria for the peak transient generator

voltage from 4784 volts to 5000 volts

during full load rejection tests of the

diesel generator (DG), and also deletes

the 10-year surveillance requirement to

perform a 110% pressure test of the DG

fuel oil system.

Date of issuance: May 1, 1995.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 110.

Facility Operating License No. NPF–

49. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 15, 1995 (60 FR

8751).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 1, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, Thames Valley Campus, 574

New London Turnpike, Norwich, CT

06360.

Northeast Nuclear Energy Company, et

al., Docket No. 50–423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of application for amendment:

September 28, 1994.

Brief description of amendment: The

amendment revises Surveillance

Requirement 4.6.1.2.a of the Technical

Specification to eliminate the

requirement to perform Type A tests on

an interval of 40 plus or minus 10

months while reiterating the Appendix

J requirement that the Type A tests be

performed three times, at approximately

equal intervals, during each 10 year

service period. In addition, a footnote is

added which states that the third Type

A test will be performed during the

sixth refueling outage. This reflects an

exemption to Appendix J which

separates the third Type A test from the

10 year inservice inspection.

Date of issuance: May 8, 1995.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 111.

Facility Operating License No. NPF–

49. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: November 23, 1994 (59 FR

60384)

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 8, 1995.

No significant hazards consideration

comments received: NO.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, Thames Valley Campus, 574

New London Turnpike, Norwich, CT

06360.

Public Service Electric & Gas Company,

Docket Nos. 50–272 and 50–311, Salem

Nuclear Generating Station, Unit Nos. 1

and 2, Salem County, New Jersey

Date of application for amendment:

August 19, 1994, as supplemented

March 15, 1995.

Brief description of amendment: The

amendments add a new action

statement to Technical Specification

3.1.3.2.1., ‘‘Position Indication

Systems—Operating’’.

Date of issuance: May 3, 1995.

Effective date: May 3, 1995.

Amendment No.: 166 and 148.

Facility Operating License Nos. DPR–

70 and DPR–75. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: October 12, 1994 (59 FR

51626) The March 15, 1995 supplement

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 may 3, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Salem Free Public Library, 112

West Broadway, Salem, New Jersey

08079.

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of application for amendments:

March 19, 1993; superseded May 16,

1994; superseded February 10, 1995;

supplemented February 17, 1995 (TS

93–04).

Brief description of amendment: The

amendments clarify the Limiting
Federal Register / Vol. 60, No. 99 / Tuesday, May 23, 1995 / Notices 27351

Conditions for Operation applicable to

the dual function of the containment

vacuum relief isolation lines by

specifying the actions that would be

required should one or more of the

vacuum relief isolation lines by

specifying the actions that would be

required should one or more of the

vacuum relief lines be incapable of

performing the containment isolation

function or incapable of performing the

vacuum relief function.

Date of issuance: April 28, 1995.

Effective date: April 28, 1995.

Amendment No.: 197 and 188.

Facility Operating License Nos. DPR–

77 and DPR–79: Amendments revise the

technical specifications.

Date of initial notice in Federal

Register: May 12, 1994 (58 FR 28060);

renoticed June 22, 1994 (59 FR 32237),

and March 29, 1995 (60 FR 16202).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 28, 1995.

No significant hazards consideration

comments received: None.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402.

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of application for amendment:

November 15, 1994; superseded March

7, 1995 (TS 94–12).

Brief description of amendments: The

amendments remove the frequencies

specified in the Technical

Specifications for performing audits and

delete the requirement to perform the

Radiological Emergency Plan, Physical

Security Plan, and Safeguard

Contingency Plan reviews.

Date of issuance: May 10, 1995.

Effective date: May 10, 1995.

Amendment No.: 198 and 189.

Facility Operating License Nos. DPR–

77 and DPR–79: Amendments revise the

technical specifications.

Date of initial notice in Federal

Register: December 21, 1994 (59 FR

65823); renoticed March 29, 1995 (60 FR

16203)

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 10, 1995.

No significant hazards consideration

comments received: None.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402.

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50–346, Davis-Besse Nuclear Power

Station, Unit No. 1, Ottawa County,

Ohio

Date of application for amendment:

January 30, 1995.

Brief description of amendment: This

amendment revises Technical

Specification (TS) 4.6.1.2.a,

‘‘Containment Systems, Containment

Leakage, Surveillance Requirements

(SR)’’ and Bases 3/4.6, ‘‘Containment

Systems,’’ to state that Type A tests for

overall integrated containment leakage

rate testing shall be conducted in

accordance with the requirements

specified in appendix J of 10 CFR part

50, as modified by NRC-approved

exemptions. Additionally, TS SR

4.6.1.2.a.

Date of issuance: May 3, 1995.

Effective date: May 3, 1995.

Amendment No.: 198.

Facility Operating License No. NPF–3.

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 15, 1995 (60 FR 14028).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 3, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of Toledo Library,

Documents Department, 2801 Bancroft

Avenue, Toledo, Ohio 43606.

Virginia Electric and Power Company,

et al., Docket Nos. 50–338 and 50–339,

North Anna Power Station, Units No. 1

and No. 2, Louisa County, Virginia

Date of application for amendments:

July 8, 1993, as supplemented by letters

dated July 12, 1994, and March 7, 1995.

Brief description of amendments: The

amendments revise the NA–1&2

Technical Specifications by deleting the

requirements to periodically review

certain administrative and technical

procedures.

Date of issuance: May 1, 1995.

Effective date: May 1, 1995.

Amendment Nos.: 190 and 171.

Facility Operating License Nos. NPF–

4 and NPF–7: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: August 4, 1993 (58 FR 41518).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated May 1, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: The Alderman Library, Special

Collections Department, University of

Virginia, Charlottesville, Virginia

22903–2498.

Virginia Electric and Power Company,

et al., Docket Nos. 50–338 and 50–339;

North Anna Power Station, Units No. 1

and No. 2, Louisa County, Virginia

Date of application for amendments:

December 27, 1993, as supplemented

September 6, 1994, and March 7, 1995.

Brief description of amendments: The

amendments revise the NA–1&2

Technical Specifications regarding the

review responsibilities of the Station

Nuclear Safety and Operating

Committee and the Management Safety

Review Committee.

Date of issuance: May 2, 1995.

Effective date: May 2, 1995.

Amendment Nos.: 191 and 172.

Facility Operating License Nos. NPF–

4 and NPF–7: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: February 16, 1994 (59 FR

7700).

The September 6, 1994, and March 7,

1995 submittals provided additional

information only, and did not change

the staff’s initial proposed

determination of no significant hazards

consideration.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated May 2, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: The Alderman Library, Special

Collections Department, University of

Virginia, Charlottesville, Virginia

22903–2498.

Virginia Electric and Power Company,

Docket Nos. 50–280 and 50–281, Surry

Power Station, Unit Nos. 1 and 2, Surry

County, Virginia

Date of application for amendments:

June 28, 1991.

Brief description of amendments:

These amendments incorporate

operability and surveillance

requirements for power-operated relief

valves to conform with Generic Letter

90–06.

Date of issuance: May 2, 1995.

Effective date: May 2, 1995.

Amendment Nos.: 198 and 198.

Facility Operating License Nos. DPR–

32 and DPR–37: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: October 2, 1991 (56 FR 49929).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 2, 1995.

No significant hazards consideration

comments received: No.
27352 Federal Register / Vol. 60, No. 99 / Tuesday, May 23, 1995 / Notices

Local Public Document Room

location: Swem Library, College of

William and Mary, Williamsburg,

Virginia 23185.

Notice of Issuance of Amendments to

Facility Operating Licenses and Final

Determination of No Significant

Hazards Consideration and

Opportunity for a Hearing (Exigent

Public Announcement or Emergency

Circumstances)

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application for the

amendment complies with the

standards and requirements of the

Atomic Energy Act of 1954, as amended

(the Act), and the Commission’s rules

and regulations. The Commission has

made appropriate findings as required

by the Act and the Commission’s rules

and regulations in 10 CFR Ch. I, which

are set forth in the license amendment.

Because of exigent or emergency

circumstances associated with the date

the amendment was needed, there was

not time for the Commission to publish,

for public comment before issuance, its

usual 30-day Notice of Consideration of

Issuance of Amendment, Proposed No

Significant Hazards Consideration

Determination, and Opportunity for a

Hearing.

For exigent circumstances, the

Commission has either issued a Federal

Register notice providing opportunity

for public comment or has used local

media to provide notice to the public in

the area surrounding a licensee’s facility

of the licensee’s application and of the

Commission’s proposed determination

of no significant hazards consideration.

The Commission has provided a

reasonable opportunity for the public to

comment, using its best efforts to make

available to the public means of

communication for the public to

respond quickly, and in the case of

telephone comments, the comments

have been recorded or transcribed as

appropriate and the licensee has been

informed of the public comments.

In circumstances where failure to act

in a timely way would have resulted, for

example, in derating or shutdown of a

nuclear power plant or in prevention of

either resumption of operation or of

increase in power output up to the

plant’s licensed power level, the

Commission may not have had an

opportunity to provide for public

comment on its no significant hazards

consideration determination. In such

case, the license amendment has been

issued without opportunity for

comment. If there has been some time

for public comment but less than 30

days, the Commission may provide an

opportunity for public comment. If

comments have been requested, it is so

stated. In either event, the State has

been consulted by telephone whenever

possible.

Under its regulations, the Commission

may issue and make an amendment

immediately effective, notwithstanding

the pendency before it of a request for

a hearing from any person, in advance

of the holding and completion of any

required hearing, where it has

determined that no significant hazards

consideration is involved.

The Commission has applied the

standards of 10 CFR 50.92 and has made

a final determination that the

amendment involves no significant

hazards consideration. The basis for this

determination is contained in the

documents related to this action.

Accordingly, the amendments have

been issued and made effective as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the application for

amendment, (2) the amendment to

Facility Operating License, and (3) the

Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment, as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, the Gelman Building, 2120 L

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

local public document room for the

particular facility involved.

The Commission is also offering an

opportunity for a hearing with respect to

the issuance of the amendment. By June

23, 1995, the licensee may file a request

for a hearing with respect to issuance of

the amendment to the subject facility

operating license and any person whose

interest may be affected by this

proceeding and who wishes to

participate as a party in the proceeding

must file a written request for a hearing

and a petition for leave to intervene.

Requests for a hearing and a petition for

leave to intervene shall be filed in

accordance with the Commission’s

‘‘Rules of Practice for Domestic

Licensing Proceedings’’ in 10 CFR part

2. Interested persons should consult a

current copy of 10 CFR 2.714 which is

available at the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington, DC and

at the local public document room for

the particular facility involved. If a

request for a hearing or petition for

leave to intervene is filed by the above

date, the Commission or an Atomic

Safety and Licensing Board, designated

by the Commission or by the Chairman

of the Atomic Safety and Licensing

Board Panel, will rule on the request

and/or petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) The nature of the

petitioner’s right under the Act to be

made a party 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
Federal Register / Vol. 60, No. 99 / Tuesday, May 23, 1995 / Notices 27353

1 15 U.S.C. 78s(b)(1).

2 17 CFR 240.19b–4.

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 lest one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses. Since the Commission has

made a final determination that the

amendment involves no significant

hazards consideration, if a hearing is

requested, it will not stay the

effectiveness of the amendment. Any

hearing held would take place while the

amendment is in effect.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

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

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

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

(800) 342–6700). The Western Union

operator should be given Datagram

Identification Number N1023 and the

following message addressed to (Project

Director): petitioner’s name and

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555, and to the attorney for the

licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of the

factors specified in 10 CFR

2.714(a)(1)(i)–(v) and 2.714(d).

Commonwealth Edison Company,

Docket Nos. 50–295 and 50–304, Zion

Nuclear Power Station, Units 1 and 2

Date of application for amendments:

April 24, 1995.

Brief description of amendments: the

amendments change the Technical

Specifications by modifying the

surveillance testing periodicity

requirements of the automatic actuation

logic of engineered safeguards

equipment.

Date of issuance: May 5, 1995.

Effective date: May 5, 1995.

Amendment Nos.: 162 and 150.

Facility Operating Licenses Nos. DPR–

39 and DPR–48. The amendments

revised the Technical Specifications.

Public comments requested as to

proposed no significant hazards

consideration: No.

The Commission’s related evaluation

of the amendments, finding of

emergency circumstances, and final

determination of no significant hazards

consideration are contained in a Safety

Evaluation dated May 5, 1995.

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60690.

Local Public Document Room

location: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois

60085.

NRC Project Director: Robert A. Capra.

Baltimore Gas and Electric Company,

Docket No. 50–317, Calvert Cliffs

Nuclear Power Pant, Unit No. 1, Calvert

County, Maryland

Date of application for amendment:

April 28, 1995.

Brief description of amendment: The

amendment revises the control room

emergency ventilation system TS

3.7.6.1, Limiting Condition For

Operation. The revision extends the

one-time increase in the allowed outage

time for loss of emergency power only,

from the 30 days previously approved,

to 45 days. This extension is necessary

to allow time to repair the Number 21

emergency diesel generator which failed

its operability tests subsequent to

modifications which have been recently

completed.

Date of issuance: May 2, 1995.

Effective date: As of the date of

issuance to be implemented upon

receipt.

Amendment No.: 205.

Facility Operating License No. DPR–

53: Amendment revised the Technical

Specifications.

Public comments requested as to

proposed no significant hazards

consideration: No.

The Commission’s related evaluation

of the amendment, consultation with

the State, and final determination of no

significant hazards consideration are

continued in a Safety Evaluation dated

May 2, 1995.

Local Public Document Room

location: Calvert County Library, Prince

Frederick, Maryland 20678.

Attorney for licensee: Jay E. Silbert,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N. Street, NW.,

Washington, DC 20037.

NRC Project Director: Ledyard B.

Marsh.

Dated at Rockville, MD, this 17th day of

May, 1995.

For the Nuclear Regulatory Commission,

Elinor G. Adensam,

Acting Director, Division of Reactor Projects—

III/IV, Office of Nuclear Reactor Regulation.

[FR Doc. 95–12538 Filed 5–22–95; 8:45 am]

BILLING CODE 7590–01–M

SECURITIES AND EXCHANGE

COMMISSION

[Release No. 34–35723; File No. SR–Amex–

95–08]

Self-Regulatory Organizations;

American Stock Exchange, Inc.; Order

Granting Approval to Proposed Rule

Change Relating to Membership

Structure and Requirements and the

Exchange’s Gratuity Fund

May 16, 1995.

I. Introduction

On February 17, 1995, the American

Stock Exchange, Inc. (‘‘Amex’’ or

‘‘Exchange’’) filed with the Securities

and Exchange Commission (‘‘SEC’’ or

‘‘Commission’’), pursuant to Section

19(b)(1) of the Securities Exchange Act

of 1934 (‘‘Act’’) 1 and Rule 19b–4

thereunder,2 a proposed rule change to

amend its Construction, Rules and

Membership Lease Plan to allow

organizations, including certain pension

plans, to own memberships legally as

well as beneficially and to allow

individuals and organizations to own

multiple memberships. The Exchange

also is proposing to revise its Gratuity

Fund to reflect the above changes, to

increase the death benefit paid

thereunder, and to allow options

principal members to participate

therein.

The proposed rule change was

published for comment in Securities