Federal Register / Vol. 60, No. 187 / Wednesday, September 27, 1995 / Notices 49929

brackets or parentheses . . .’’ to

accommodate the editing style of the

various States.

Comments have not been received

from the remaining groups.

Status of Licensee Metrication Efforts

Reactors

Although there are no power reactor

licensees operating in the metric system,

some of the advanced reactors have

vendor-generated licensing documents

that use the metric system of

measurement. For example, both of

General Electric’s applications for the

ABWR and SBWR designs have their

Standard Safety Analysis Reports

(SSAR) in the SI system of

measurement. However, both the

Westinghouse AP600 and the ABB–CE

System 80+ have their SSARs in the

traditional inch-pound system. The

NRC’s completed Final Safety

Evaluation Reports (FSER) for the

System 80+ and the ABWR are in dual

units as prescribed by the Commission’s

policy statement. When the FSERs for

the AP600 and the SBWR are published,

they also will be in dual units.

Selected Examples of Metric Usage

There are varying degrees of use of the

metric system of measurement by the

non-power reactor nuclear industries.

Also, within a particular profession or

industry, there are varying degrees of

metric use. For example, in the field of

radiation oncology, the centigray (an SI

unit) has been the meter of therapy

doses, while the millicurie and curie

(traditional units) are used as the

measure expressing quantity or dosages.

Health Physics

It is also the case that most of the

operational health physics community

still uses the traditional system of

measurement because of the use of

instrumentation that is calibrated or

expressed in that system. Some newer

instrumentation that offers dual-unit

options will assist in metric conversion,

as the new instruments are being

integrated into existing stock.

Public Comment

The NRC staff, through this request, is

inviting comment from interested

individuals on the NRC’s metrication

efforts to learn if there is a need for the

Commission to revise its metrication

policy.

Electronic Access

Comments may be submitted

electronically, in either ASCII text or

Wordperfect format (version 5.1 or

later), by calling the NRC Electronic

Bulletin Board on FedWorld. The

bulletin board may be accessed using a

personal computer, a modem, and one

of the commonly available

communications software packages, or

directly via Internet.

If using a personal computer and

modem, the NRC subsystem on

FedWorld can be accessed directly by

dialing the toll free number: 1–800–

303–9672. Communication software

parameters should be set as follows:

Parity to none, data bits to 8, and stop

bits to 1 (N,8,1). Using ANSI or VT–100

terminal emulation, the NRC

rulemaking subsystems can then be

accessed by selecting the ‘‘Rules Menu’’

option from the ‘‘NRC Main Menu.’’ For

further information about options

available for NRC at FedWorld consult

the ‘‘Help/Information Center’’ from the

‘‘NRC Main Menu.’’ Users will find the

‘‘FedWorld Online User’s Guides’’

particularly helpful. Many NRC

subsystems and databases also have a

‘‘Help/Information Center’’ option that

is tailored to the particular subsystem.

The NRC subsystem on FedWorld can

also be accessed by a direct dial phone

number for the main FedWorld BBS:

703–321–8020; Telnet via Internet:

fedworld.gov (192.239.93.3); File

Transfer Protocol (FTP) via Internet:

ftp.fedworld.gov (192.239.92.205); and

World Wide Web using: http://

www.fedworld.gov (this is the Uniform

Resource Locator (URL)). If using a

method other than the toll free number

to contact FedWorld, then the NRC

subsystem will be accessed from the

main FedWorld menu by selecting the

‘‘F—Regulatory, Government

Administration and State Systems’’,

then selecting ‘‘A—Regulatory

Information Mall’’. At that point, a

menu will be displayed that has an

option ‘‘A—U.S. Nuclear Regulatory

Commission’’ that will take you to the

NRC Online main menu. You can also

go directly to the NRC Online area by

typing ‘‘/go nrc’’ at a FedWorld

command line. If you access NRC from

FedWorld’s main menu, then you may

return to FedWorld by selecting the

‘‘Return to FedWorld’’ option from the

NRC Online Main Menu. However, if

you access NRC at FedWorld by using

NRC’s toll-free number, you will have

full access to all NRC systems but you

will not have access to the main

FedWorld system. For more information

on NRC bulletin boards call Mr. Arthur

Davis, Systems Integration and

Development Branch, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555, telephone (301) 415–5780; email AXD3@nrc.gov.

Lastly, the Act has a reporting

requirement for Federal agencies to

include an annual metric report as part

of their annual budget submission to the

Congress. The reporting requirement

expires in the fiscal year after an agency

has fully implemented metric usage.

Unless the Commission receives

comment which would require it to

revise its policy, it will consider its

policy final and its conversion to the

metric system complete.

Dated at Rockville, Maryland this 14th day

of September 1995.

For the Nuclear Regulatory Commission.

James M. Taylor,

Executive Director for Operations.

[FR Doc. 95–23932 Filed 9–26–95; 8:45 am]

BILLING CODE 7590–01–P

Biweekly Notice

Applications and Amendments to

Facility Operating LicensesInvolving

No Significant Hazards Considerations

I. Background

Pursuant to Public Law 97-415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 97-415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from August 30,

1995, through September 15, 1995. The

last biweekly notice was published on

Wednesday, September 13, 1995 (60 FR

47613).

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
49930 Federal Register / Vol. 60, No. 187 / Wednesday, September 27, 1995 / Notices

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 October 27, 1995, the licensee may

file a request for a hearing with respect

to issuance of the amendment to the

subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

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
Federal Register / Vol. 60, No. 187 / Wednesday, September 27, 1995 / Notices 49931

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.

Carolina Power & Light Company, et

al., Docket No. 50-324, Brunswick

Steam Electric Plant, Unit 2, Brunswick

County, North Carolina

Date of amendment request: August 4,

1995

Description of amendment request:

The proposed amendment will allow

the loading and use of GE13 fuel

assemblies in the Brunswick Steam

Electric Plant (BSEP), Unit 2, during

Cycle 12 operation. The use of GE13

fuel assemblies requires that the safety

limit value for minimum critical power

ratio be revised. This safety limit is

established to maintain fuel cladding

integrity. Use of GE13 fuel also requires

an increase in the concentration of

sodium pentaborate solution required

by the Technical Specifications (TS) for

the standby liquid control system. This

change provides the additional

shutdown reactivity necessary to permit

use of this fuel.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Proposed Change 1:

The proposed amendment will allow the

loading and use of GE13 fuel assemblies in

the Brunswick Unit 2 reactor core. The use

of GE13 fuel assemblies requires that the

safety limit minimum critical power ratio

value also be revised. The safety limit

minimum critical power ratio is established

to maintain fuel cladding integrity. The GE13

fuel assembly design has been analyzed using

methods that have been previously approved

by the Nuclear Regulatory Commission and

documented in General Electric Nuclear

Energy’s reload licensing methodology

Topical Report (NEDE-24011-P-A-10,

‘‘General Electric Standard Application for

Reactor Fuel (GESTAR II)’’ dated February

1991).

The proposed revision of the safety limit

minimum critical power ratio does not alter

any plant safety-related equipment, safety

function, or plant operations that could

change the probability of an accident. The

change does not affect the design, materials,

or construction standards applicable to the

fuel bundles in a manner that could change

the probability of an accident.

A methodology that has been previously

reviewed and accepted by the Nuclear

Regulatory Commission was used to derive

the both existing and updated safety limit

minimum critical power ratio value. The

same methodology criteria have been applied

to derive the existing safety limit minimum

critical power ratio of 1.07 as that used to

derive the updated safety limit minimum

critical power ratio value of 1.09. The

updated safety limit minimum critical power

ratio assures that fuel cladding protection

equivalent to that provided with the existing

safety limit minimum critical power ratio

value is maintained. This ensures that the

consequences of previously evaluated

accidents are not significantly increased.

Proposed Change 2:

The standby liquid control system provides

a means of reactivity control that is

independent of the normal reactivity control

system. The standby liquid control system

must be capable of assuring that the reactor

core can be placed in a subcritical condition

at any time during reactor core life. Technical

Specification Figure 3.1.5-1 specifies the

acceptable range of concentrations and

volumes for sodium pentaborate solution

used as a neutron absorber (i.e., for reactivity

control). The portion of the sodium

pentaborate concentration range shown in

Technical Specification Figure 3.1.5-1

applicable to the lower range of tank volumes

is being revised to increase the required

concentration of sodium pentaborate

solution. This change is needed to account

for the additional shutdown reactivity

needed based on the planned use of GE13

fuel assemblies as reload fuel for the Unit 2

reactor core. Since the standby liquid control

system is independent from the normal

means of controlling reactor core reactivity

and not used to control core reactivity during

normal plant operations, the proposed

revision to the sodium pentaborate

concentration curve for the standby liquid

control system does not alter any plant

safety-related equipment, safety function, or

plant operations that could change the

probability of an accident.

The current volume-concentration range of

sodium pentaborate used in the standby

liquid control system will achieve a

sufficient concentration of boron in the

reactor vessel to ensure reactor shutdown.

Based on the increased reactivity of the new

GE13 reload fuel assemblies, the required

sodium pentaborate volume-concentration

range is being revised to ensure sufficient

neutron absorbing solution is available to

achieve reactor shutdown; therefore, the

consequences of an accident previously

evaluated are not significantly increased.

2. The proposed amendment would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

Proposed Change 1:

The GE13 fuel assembly has been designed

and complies with the acceptance criteria

contained in General Electric Nuclear

Energy’s standard application for reactor fuel

(GESTAR-II), which provides the latest

acceptance criteria for new General Electric

fuel designs. The GE13 fuel assembly

complies with GESTAR-II acceptance criteria

that have been previously reviewed and

accepted by the Nuclear Regulatory

Commission. The similarity of the GE13 fuel

design to the previously accepted GE11 fuel

design, in conjunction with the increased

critical power capability of the GE13 fuel

design, ensure that no new mode or

condition of plant operation is being

authorized by the loading and use of the

GE13 fuel type. The proposed revision of the

safety limit minimum critical power ratio

from 1.07 to 1.09 does not modify any plant

controls or equipment that will change the

plant’s responses to any accident or transient

as given in any current analysis. Therefore,

the proposed change to allow the loading and

use of the GE13 fuel type and the revision of

the safety limit minimum critical power ratio

value from 1.07 to 1.09 will not create the

possibility for a new or different kind of

accident from any accident previously

evaluated.

Proposed Change 2:

As discussed above, the standby liquid

control system provides a means of reactivity

control that is independent of the normal

reactivity control system and is capable of

assuring that the reactor core can be placed

in a subcritical condition at any time during

reactor core life. The proposed revision to the

sodium pentaborate concentration range does

not modify the standby liquid control system

or its controls, does not modify other plant

systems and equipment, and does not permit

a new or different mode of plant operation.

As such, the proposed revision to the

minimum pentaborate concentration value

does not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

3. The proposed license amendment does

not involve a significant reduction in a

margin of safety.

Proposed Change 1:

As previously discussed, the GE13 fuel

assembly design has been analyzed using

methods that have been previously approved

by the Nuclear Regulatory Commission and

documented in General Electric Nuclear

Energy’s reload licensing methodology

Topical Report (NEDE-24011-P-A-10,

‘‘General Electric Standard Application for

Reactor Fuel (GESTAR II)’’ dated February

1991). The safety limit minimum critical

power ratio value is selected to maintain the

fuel cladding integrity safety limit (i.e., that

99.9 percent of all fuel rods in the core be

expected to avoid boiling transition).
49932 Federal Register / Vol. 60, No. 187 / Wednesday, September 27, 1995 / Notices

Appropriate operating limit minimum

critical power ratio values are established,

based on the safety limit minimum critical

power ratio value, to ensure that the fuel

cladding fuel integrity safety limit is

maintained. The operating limit minimum

critical power ratio values are incorporated

in the Core Operating limits Report as

required by Technical Specification 6.9.3.1.

The new GE13 safety limit minimum critical

power ratio value of 1.09 is based on the

same fuel cladding integrity safety limit

criteria at that for the GE11 safety limit

minimum critical power ratio value of 1.07

(i.e., that 99.9 percent of all fuel rods in the

core be expected to avoid boiling transition);

therefore, the proposed change does not

result in a significant reduction in the margin

of safety.

Proposed Change 2:

As previously stated, the purpose of the

standby liquid control is to inject a neutron

absorbing solution into the reactor in the

event that a sufficient number of control rods

cannot be manually inserted to maintain

subcriticality. Sufficient solution is to be

injected such that the reactor will be brought

from maximum rated power conditions to

subcritical over the entire reactor

temperature range from maximum operating

to cold shutdown conditions. General

Electric reactor fuel methodology establishes

a fuel type dependent standby liquid control

system shutdown margin to account for

calculational uncertainties. General Electric

calculations show that an in-vessel

concentration of 660 ppm will provide an

estimated standby liquid control system

minimum shutdown margin of 4.1% delta k.

To achieve an in-vessel concentration of 660

ppm, the acceptable range of standby liquid

control system tank concentrations is being

revised for the lower range of tank volumes.

Thus, proposed revision of the standby liquid

control system sodium pentaborate volumeconcentration range ensures that there will

not be a significant reduction in the amount

of available shutdown margin and, therefore,

not 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: University of North Carolina at

Wilmington, William Madison Randall

Library, 601 S. College Road,

Wilmington, North Carolina 28403-

3297.

Attorney for licensee: R. E. Jones,

General Counsel, Carolina Power &

Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602

NRC Project Director: David B.

Matthews

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: April 10,

1995

Description of amendment request:

The requested amendment would

modify Technical Specification 4.6.4.3

to allow a reduction in the number of

hydrogen mitigation system igniters that

must be maintained Operable. This

would allow removal of the hydrogen

igniters in the incore instrument tunnel.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Criterion 1

The requested amendments will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated. No impact upon

accident probabilities will be created, since

the EHM System is not an accident initiating

system. In addition, it has been demonstrated

that based on the results of computer

analysis, and the review of results of an

external study performed for a similar type

containment, that hydrogen concentrations in

the cavity during degraded core accidents

will remain within acceptable limits. No

impact on the plant response to any accident

will be created (either design basis or

beyond-design basis).

Criterion 2

The requested amendments will not create

the possibility of a new or different kind of

accident from any accident previously

evaluated. As stated previously, the EHM

System is not an accident initiating system.

No new accident causal mechanisms will be

created as a result of deleting the affected

igniters. Plant operation will not be affected

by the proposed amendments and no new

failure modes will be created.

Criterion 3

The requested amendments will not

involve a significant reduction in a margin of

safety. No adverse impact upon any plant

safety margins will be created. As shown

previously, applicable computer analysis has

successfully demonstrated that the affected

igniters could be removed with no adverse

consequences. No fission product barriers are

being degraded. No change to the manner in

which the units are operated is being made.

Based upon the preceding analyses, Duke

Power Company concludes that the requested

amendments do not involve a significant

hazards consideration.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: 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

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 1, 1995

Description of amendment request:

Generic Letter 88-16 provided guidance

on removing cycle-specific parameters

which are calculated using NRC

approved methodologies from Technical

Specifications (TS). The parameters are

replaced in TS with a reference to a

named report which contains the

parameters, and a requirement that the

parameters remain within the limits

specified in the report. The proposed

changes incorporate NRC approved

methodologies, approved revisions to

previously approved methodologies, or

republished versions of previously

approved methodologies into Section

6.9 of the Catawba TS. For Catawba, the

limits to which these methodologies are

applied are explicitly listed in the TS.

Since the proposed changes only

incorporate NRC approved

methodologies into the TS the licensee

proposed that the changes are

administrative in nature and can be

assumed to have no impact, or potential

impact, on the health and safety of the

public.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The proposed changes will not create a

significant hazards consideration, as defined

by 10 CRF 50.92, because:

1) The proposed changes will not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes are administrative

in nature, and do not affect any system,

procedure, or manipulation of any equipment

which could affect the probability or

consequences of any accident.

2) The proposed changes will not create

the possibility of any new or different kind

of accident from any accident previously

evaluated.

The proposed changes are administrative

in nature, and cannot introduce any new

failure mode or transient which could create

any accident.
Federal Register / Vol. 60, No. 187 / Wednesday, September 27, 1995 / Notices 49933

3) The proposed changes will not involve

a significant reduction in a margin of safety.

The proposed changes are administrative

in nature, and will not affect any operating

parameters or limits which could result in a

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: 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

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 13, 1995

Description of amendment request:

The proposed amendments modify the

notation for the overpower deltatemperature (OPDT) reactor trip heatup

setpoint penalty coefficient to be

consistent with NUREG-0452, Revision

4, ‘‘Standard Technical Specifications

for Westinghouse Pressurized Water

Reactors’’ (STS). This change is

necessary in order to allow

implementation of the modification to

reduce the reactor coolant system hot

leg temperature as planned during the

Unit 2 end-of-cycle 7 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:

As required by 10CFR50.91, this analysis is

provided concerning whether the requested

amendments involve significant hazards

considerations, as defined by 10CFR50.92.

Standards for determination that an

amendment request involves no significant

hazards considerations are if operation of the

facility in accordance with the requested

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.

Criterion 1

The proposed amendments will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated. The amendments will

have no impact whatsoever upon the

probability of any accident being initiated,

since the reactor trip system is an accident

mitigating system. The amendments will

have no adverse impact upon any accident

consequences or upon the function of the

OPDT setpoint. The reactor trip heatup

setpoint penalty will continue to be applied

anytime T-avg is greater than T [double

prime] and will not be applied when T-avg

is less than or equal to T [double prime]. This

is consistent with the intent of this function.

Criterion 2

The proposed amendments will not create

the possibility of a new or different kind of

accident from any accident previously

evaluated. The function of the OPDT setpoint

will not be altered by the proposed changes.

As stated previously, the reactor trip system

is an accident mitigating system, so no new

failure modes can be created. No change to

any aspect of plant operation will result from

NRC approval of the proposed amendments.

Criterion 3

The proposed amendments will not

involve a significant reduction in a margin of

safety. The changes are necessary to allow

full implementation of the T-hot reduction

modification on Catawba Unit 2. The

proposed changes are consistent with the

terminology of both NUREG-0452, Revision 4

and NUREG-1431, Revision 1. OPDT setpoint

behavior will not be adversely impacted by

the proposed changes; therefore, no impact

upon any plant safety margins will result.

Based upon the preceding analyses, Duke

Power Company concludes that the requested

amendments do not involve a significant

hazards consideration.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: 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

Duke Power Company, Docket Nos. 50-

369 and 50-370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg

County, North Carolina

Date of amendment request: March

29, 1995

Description of amendment request:

The amendments would revise the

Technical Specification 3.4.9.3

requirements for the Low Temperature

Overpressure Protection (LTOP) system

and update the heatup and cooldown

curves. The intent of the proposed

amendments is to enhance overpressure

protection during low temperature

operations. These enhancements can be

fully implemented, improving startup

and shutdown operation of McGuire

Units 1 and 2.

Specifically, these changes are

categorized into five groups identified

as follows:

1) Revisions to the LCO requirements,

the Action Statements and the SR for

the Reactor Coolant System

Overpressure Protection System during

low temperature conditions,

2) A reduction in the Reactor Coolant

System (RCS) vent requirement from 4.5

square inches to 2.75 square inches,

3) The use of the Residual Heat

Removal suction relief valve (1ND3 and

2ND3) for overpressure protection under

restricted conditions. (RCS greater than

107°F and cooldown rate less than 20°F/

hr; or RCS greater than 167°F),

4) Revisions of the Pressure/

Temperature curves to 16 EFPY,

including the incorporation of the latest

radiation surveillance capsule results

and removal of instrumentation margins

from the Technical Specification

figures, and

5) Changes to format and consistency.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration for each of the five groups

listed above.

FIRST STANDARD

(Amendment would not) involve a

significant increase in the probablility or

consequences of an accident previously

evaluated.

1) Revised LCO [limiting conditions for

operation] and SR [surveillance

requirements] for LTOP:

The reduced maximum setpoint will

prevent the violation of the 10 CFR 50

Appendix G pressure/temperature curves (as

modified by the provisions of ASME Code

Case N-514) during overpressure transients at

low temperatures. Since the maximum

setpoint is reduced, the peak pressure for

LTOP [low-temperature overpressure

protection] events will be reduced as well.

Accordingly, the consequences of an LTOP

event would not change as result of the

proposed changes.

The analysis performed to determine the

setpoint is, in accordance with the methods

used in previous evaluations, found

acceptable by the NRC. The three possible

transients evaluated are; 1) a mass input from

an operable safety injection pump; 2) a mass

input from an operable centrifugal charging

pump; and 3) a heat input from a 50°F

temperature difference between the steam

generators and the NC system. The LTOP

setpoint of the PORV [power-operated relief

valve] proposed by this technical

specification change is not considered to be

an initiator of any of these three transients.

As such, the probability of an accident
49934 Federal Register / Vol. 60, No. 187 / Wednesday, September 27, 1995 / Notices

previously evaluated would not be increased

as a result of the proposed changes.

Two additional conditions for operability

of the LTOP system are defined (accumulator

isolation and only one NV or NI pump

operable) and new surveillance requirements

are specified as well. They provide

additional limitations, requirements and

restrictions that currently do not exist within

the technical specifications for McGuire. The

incorporation of these proposed changes are

consistent with what is specified within

NUREG-1341. Therefore, these changes do

not increase the probability of consequences

of an accident previously evaluated.

2) Reduction in NC vent opening:

The bases for the size of the vent to be

established per the technical specifications is

to ensure that the 10 CFR 50, Appendix G

pressure/temperature limits are not exceeded

during an LTOP event. The determination of

the size of the opening continues to preserve

the above design basis. The evaluation

performed demonstrated that a 2.75 square

inch opening would provide adequate

overpressure protection for the combined

capacity of a centrifugal charging pump and

a safety injection pump.

The only time that the vent path is to be

established is when the PORVs may not be

available. Defining the size of the vent is not

considered to be an initiator of any LTOP

events that have been previously evaluated.

As such, this change in the size of the vent

opening does not increase the probability of

an overpressure event during low

temperature conditions. The analysis

performed verifies that the size opening

specified is sufficient to mitigate the

consequences of an LTOP event.

Accordingly, the change in the size of the

opening for the vent will not impact the

consequences of LTOP events.

3) Use of RHR [residual heat removal]

suction relief valves:

By letter dated September 11, 1990, the

NRC authorized the deletion of the RHR

autoclosure interlock circuitry. A

modification which removed the RHR system

suction isolation valve autoclosure interlocks

has been completed. As such, the RHR

suction relief valve can be exposed to NC

system pressure and would be available to

mitigate LTOP events.

The proposed amendments specify the

necessary requirements and controls to

ensure proper ND system alignments and

conditions will exist to protect the pressure/

temperature limits. This added relieving

capacity will enhance the current LTOP

system at McGuire in mitigating overpressure

events at low temperatures. As such, the

mitigation of previously evaluated LTOP

events would be improved by the proposed

technical specification changes. Further, the

proposed changes would not esult in the

initiation of an LTOP event or cause an

overpressure transient. Accordingly, the

proposed amendment would not involve an

increase in the consequences or the

probability of an accident previously

evaluated.

4) Revised pressure/temperature curves to

16 EFPY [effective full-power year]:

The proposed pressure/temperature curves,

provided by this amendment request, satisfy

all regulatory required material

embrittlement considerations including:

ASME Section XI Appendix G, 10 CFR 50

Appendix G, and Regulatory Guide 1.99,

Revision 2. In addition, the margins for

instrument error have been removed from the

curves. Instrument error will be

administratively handled by incorporating

them into the LTOP system setpoint selection

calculations and into appropriate controlling

procedures for unit operations.

The proposed changes to the pressure/

temperature curves are not considered to be

an initiator of LTOP events. The changes to

the curves proposed by this amendment

request will not cause an LTOP event. The

curves define the new limits that have been

defined in accordance with regulatory

requirements by which both units are to be

operated within. Accordingly, the proposed

amendment will not increase the probability

or the consequences of previously evaluated

accidents.

5) Format and consistency:

The changes associated within this group

are considered to be administrative in nature.

They do not affect station operability or

require any modifications to the facility.

Accordingly, the proposed amendment

request does not increase the probability or

consequences of any previously evaluated

accident.

SECOND STANDARD

(Amendment would not) create the

possibility of a new or different kind of

accident from any kind of accident

previously evaluated.

1) Revised LCO and SR for LTOP:

The only potential impact to plant systems,

structures and components, as a result of the

proposed changes associated with this group,

would be the setting of the PORV low

pressure setpoint. No other changes to plant

systems, structures or components would

occur. The proposed amendments, also,

would not impact the plant operation.

Although the value for the PORV pressure

setting specified within the technical

specification would be reduced per the

proposed amendment, the actual settings of

the PORV are now currently set low enough

to comply with the proposed lower setpoint

value. As such, the proposed lower setpoint

would not require any changes to the plant

nor how the plant is operated.

The additional requirements for LTOP

operability will not require any modifications

to the plant nor how the plant is operated.

Currently, when entering LTOP conditions,

the accumulators are isolated and only one

NV or NI pump is capable of injecting into

the reactor vessel. these actions are currently

controlled and are specified within the

operating procedures for heatup and

cooldown of the respective units. The

proposed changes will now specify these

current operating requirements within the

technical specifications as well.

Accordingly, the proposed revisions will

not create a new or different kind of accident

than what has already been previously

evaluated.

2) Reduction in NC vent opening:

The proposed changes to the technical

specifications associated with this group

involves the size of the vent opening. The

proposed amendment reduces the size of the

vent opening from 4.5 square inches to 2.75

square inches. The analysis that was

performed has determined that the proposed

size for the vent opening is adequate for

overpressure events. Therefore, this proposed

revision to the technical specifications will

not result in a new or different kind of

accident from any kind of accident

previously evaluated.

3) Use of RHR suction relief valves;

The proposed amendment associated with

this group will specify the necessary

requirements and controls to ensure the

appropriate use of the RHR suction relief

valve for overpressure protection. This added

relieving capacity will enhance the current

LTOP system in mitigating overpressure

events during low temperature conditions.

The analysis that has been performed

demonstrates the adequacy of the RHR

suction relief valve, in conjunction with a

PORV, in mitigating overpressure events at

low temperatures, assuming a worst case

single failure as well. As such, the use of the

RHR suction relief valve in the manner

prescribed by the proposed technical

specification amendment will not create a

new or different kind of accident from those

accidents that have been previously

evaluated.

4) Revised pressure/temperature curves to

16 EFPY:

The changes associated with this group,

provide new heatup and cooldown curves for

both Units 1 and 2, which will extend the

service period from 10 EFPY to 16 EFPY and

will remove the instrument error as well. The

proposed [heatup] and cooldown curves were

developed in accordance with all regulatory

required material embrittlement criteria.

Thus, operation of the units in accordance

with the proposed new pressure/temperature

curves will not create the possibility of a new

or different kind of accident from those

accident[s] that have been previously

evaluated.

5) Format and consistency:

The changes associated within this group

are considered to be administrative in nature.

They do not affect station operability or

require any modifications to the facility.

Accordingly, the proposed amendment will

create the possibility of a new or different

kind of accident from that previously

evaluated.

THIRD STANDARD

(Amendment would not) involve a

significant reduction in a margin of safety.

1) Revised LCO and SR for LTOP:

This proposed change will reduce the

maximum PORV setpoint such that, for LTOP

events, the maximum pressure in the vessel

would not exceed 110% of the pressure/

temperature limits that have been established

in accordance with ASME Appendix G. This

is congruous with the provisions of ASME

Code Case N-514. Currently, the maximum

PORV setpoint for LTOP events ensure that

the maximum pressure would not exceed

100% of the pressure/temperature curves. As

such, the proposed change appears to involve

a slight reduction in a margin of safety.

Although the proposed change may

involve a slight reduction in a margin of

safety, the proposed change will provide an
Federal Register / Vol. 60, No. 187 / Wednesday, September 27, 1995 / Notices 49935

equivalent margins of safety to the reactor

vessel during LTOP transients and will

satisfy the underlying purpose of 10 CFR

50.60 for fracture toughness requirements. By

letter dated June 28, 1994, an exemption

request and authorization to use ASME Code

Case N-514 at McGuire was submitted to the

NRC for review and approval. Approval for

the use of the code case was granted on

September 30, 1994. The proposed change to

reduce the maximum PORV setpoint,

coupled with the September 30, 1994 NRC

approval for the use of Code Case N-514

satisfies current regulatory acceptance

criteria. Therefore, the proposed change

would not involve a significant reduction in

a margin of safety.

This change group, also, defines two

additional conditions for the operability of

the LTOP system (accumulator isolation and

only one NV or NI pump operable) and

proposes new surveillance requirements and

restrictions that currently do not exist within

the technical specifications for McGuire. The

incorporation of these proposed changes are

consistent with what is specified within

NUREG-1341. Therefore, these changes do

not involve a significant reduction in a

margin of safety.

2) Reduction in NC vent opening:

The proposed changes to the technical

specifications associated with this group

involves the size of the vent opening. The

proposed amendment reduces the size of the

vent opening from 4.5 square inches to 2.75

square inches. The basis for the size of the

vent to be established per the technical

specifications is to ensure that the 10 CFR 50,

Appendix G pressure/temperature limits are

not exceeded during an LTOP event. The

determination of the size of the opening

continues to preserve the above design basis.

The evaluation performed demonstrated that

a 2.75 square inch opening would provide

adequate overpressure protection for the

combined capacity of a centrifugal charging

pump and a safety injection pump.

Accordingly, the proposed changes would

not involve a significant reduction in a

margin of safety.

3) Use of RHR suction relief valves:

The proposed amendment associated with

this group will specify the necessary

requirements and controls to ensure the

appropriate use of the RHR suction relief

valves for overpressure protection. This

added relieving capacity will enhance the

current LTOP system in mitigating

overpressure events during low temperature

conditions. The analysis that has been

performed demonstrates the adequacy of the

RHR suction relief valve, in conjunction with

a PORV, in mitigating overpressure events at

low temperatures.

Further, by letter dated September 11,

1990, the NRC approved amendments to

delete a portion of the surveillance

requirements regarding periodic verification

that the RHR suction isolation valves

automatically close on a RCS [reactor coolant

system] signal less than or equal to 560 psig.

This action, in effect, authorizes the removal

of the RHR autoclosure interlock circuitry. As

discussed within the NRC Safety evaluation

for the amendment, the Commission and

industry have recognized the safety benefits

of removing the ACI [automatic closure and

interlock] circuitry from the RHR system to

minimize, and thus reduce the risk

associated with loss of decay heat removal

events.

Therefore, the proposed amendments

associated with this change group will not

involve a significant reduction in a margin of

safety.

4) Revised pressure/temperature

curves to 16 EFPY:

The changes associated with this group

provide new heatup and cooldown curves for

both Units 1 and 2, which will extend the

service period from 10 EFPY to 16 EFPY and

will relocate the instrument error as well.

The proposed pressure/temperature curves

provided by this amendment request satisfy

all regulatory required material

embrittlement considerations including;

ASME Section XI Appendix G, 10 CFR 50

Appendix G, and Regulatory Guide 1.99,

Revision 2. The instrument error will be

administratively handled by incorporating

them into the LTOP system setpoint selection

calculations and into the controlling

procedures for unit operations.

The relocation of the instrument error to

licensee controlled documents is consistent

with the NRC actions proposed within

NUREG-1431, new standard technical

specifications for Westinghouse plants. As

prescribed within NUREG-1431, the

pressure/temperature limit curves are to be

relocated to a licensee controlled document

entitled ‘‘Pressure Temperature Limit Report

(PTLR)’’. Changes to the heatup and

cooldown curves would then be performed in

accordance with 10 CFR 50.59 criteria. For

the situation proposed by this amendment,

updates and revisions of the instrument error

associated with the pressure/temperature

limit curves will be processed in a similar

fashion. Thus, the proposed change to

relocate the instrument error to licensee

controlled documents is analogous with NRC

acceptable practices.

Accordingly, the proposed changes will

not reduce a margin of safety.

5) Format and consistency:

The changes associated within this group

are considered to be administrative in nature.

They do not affect station operability or

require any modifications to the facility.

Accordingly, there is no reduction in the

margin of safety of the LTOP system due to

the incorporation of these editorial/

administrative 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: Atkins Library, University of

North Carolina, Charlotte (UNCC

Station), North Carolina 28223

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, et

al., Docket Nos. 50-335 and 50-389, St.

Lucie Plant, Unit Nos. 1 and 2, St. Lucie

County, Florida

Date of amendment request: June 21,

1995

Description of amendment request:

The proposed amendments will revise

the action statements for a single

inoperable Emergency Diesel Generator

(EDG), TS 3.8.1.1.b, to extend the

allowed outage time (AOT) from 72

hours to 7 days, and permit a 10 day

AOT to be used once per refueling

cycle. This proposal is a result of a

cooperative study by participating

Combustion Engineering Owners Group

members which concluded that the

proposed AOT extension improves

plant operational flexibility while

adequately controlling overall plant

risk.

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 amendment would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed amendments for St. Lucie

Unit 1 and Unit 2 will extend the action

completion/allowed outage time (AOT) for a

single inoperable Emergency Diesel

Generator (EDG) from 72 hours to 7 days,

with provisions for a 10 day AOT once per

refueling cycle. The EDGs are designed as

backup AC power sources for essential safety

systems in the event of a loss of offsite

power. As such, the EDGs are not accident

initiators, and an extended AOT to restore

operability of an inoperable diesel generator

would not increase the probability of

occurrence of accidents previously analyzed.

The proposed technical specification

revisions involve the AOT for a single

inoperable EDG, and do not change the

conditions, operating configuration, or

minimum amount of operating equipment

assumed in the plant safety analyses for

accident mitigation. In addition, a Probability

Safety Assessment (PSA) was performed to

quantitatively assess the risk impact of the

proposed amendment. The impact on the

early radiological release probability for

design basis events was also evaluated. It was

concluded that the risk contribution from

this proposed AOT is very small, and that the

impact will be negligible.

Therefore, operation of either facility in

accordance with its proposed amendment

would not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

(2) Operation of the facility in accordance

with the proposed amendment would not
49936 Federal Register / Vol. 60, No. 187 / Wednesday, September 27, 1995 / Notices

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed amendments will not change

the physical plant or the modes of plant

operation defined in either Facility License.

The changes do not involve the addition or

modification of equipment, nor do they alter

the design of plant systems. Therefore,

operation of either facility in accordance

with its proposed amendment would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

(3) Operation of the facility in accordance

with the proposed amendment would not

involve a significant reduction in a margin of

safety.

The proposed amendments are designed to

improve EDG reliability by providing

flexibility in the scheduling and performance

of preventive and corrective maintenance

activities. The surveillance intervals or the

operability requirements are not changed by

the proposal; only the AOT for a single

inoperable EDG will be extended. The

proposed changes do not alter the basis for

any technical specification that is related to

the establishment of, or the maintenance of,

a nuclear safety margin. Moreover, an

integrated assessment of the risk impact of

extending the AOT for a single inoperable

EDG has determined that the risk

contribution is very small and can be offset

by improvements in EDG reliability.

Therefore, operation of either facility in

accordance with its proposed amendment

would not involve a significant reduction in

a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort

Pierce, Florida 34954-9003

Attorney for licensee: J. R. Newman,

Morgan, Lewis & Bockius, 1800 M

Street, NW., Washington, DC 20036

NRC Project Director: David B.

Matthews

Florida Power and Light Company, et

al., Docket Nos. 50-335 and 50-389, St.

Lucie Plant, Unit Nos. 1 and 2, St. Lucie

County, Florida

Date of amendment request: June 21,

1995

Description of amendment request:

The proposed amendments will revise

TS 3.5.2 to allow up to 7 days to restore

an inoperable Low Pressure Safety

Injection train to operable status. This

proposal is a result of a cooperative

study by participating Combustion

Engineering Owners Group members

which concluded that an extension of

the allowed outage time (AOT) from 72

hours to 7 days can improve plant

operational flexibility and is risk

beneficial.

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 amendment would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed amendments for St. Lucie

Unit 1 and Unit 2 will extend the action

completion/allowed outage time (AOT) for a

single inoperable Low Pressure Safety

Injection (LPSI) train from 72 hours to 7 days.

A LPSI train is designed as a part of each

Emergency Core Cooling System (ECCS)

subsystem to supplement Safety Injection

Tank (SIT) inventory during the early stages

of mitigating a Design Basis Accident. As

such, components of the LPSI system are not

accident initiators, and an extended AOT to

restore operability of an inoperable LPSI train

would not increase the probability of

occurrence of accidents previously analyzed.

The safety analyses for both St. Lucie Units

demonstrate that ECCS performance

acceptance criteria are satisfied with only

one of the two redundant ECCS subsystems

operating during the postulated Design Basis

Accident. The proposed technical

specification revisions involve the AOT for a

single inoperable LPSI train, and do not

change the conditions assumed for the

minimum amount of operating equipment

needed for accident mitigation. Therefore,

the consequences of an accident previously

evaluated will not be significantly increased.

In addition to the preceding evaluation, a

Probabilistic Safety Analysis (PSA) was

performed to quantitatively assess the risk

impact of the proposed amendments. It was

concluded from the results of that assessment

that the risk contribution of the AOT

extension is very small, and that the net

impact of the proposed amendment can be

risk beneficial.

Therefore, operation of either facility in

accordance with its proposed amendment

would not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

(2) Operation of the facility in accordance

with the proposed amendment would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed amendments will not change

the physical plant or the modes of plant

operation defined in either Facility License.

The changes do not involve the addition or

modification of equipment nor do they alter

the design of plant systems. Therefore,

operation of either facility in accordance

with its proposed amendment would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

(3) Operation of the facility in accordance

with the proposed amendment would not

involve a significant reduction in a margin of

safety.

The margin of safety associated with the

ECCS system is established by acceptance

criteria for system performance defined in 10

CFR 50.46. The proposed amendments will

not change this acceptance criteria nor the

operability requirements for equipment that

is used to achieve such performance as

demonstrated in the plant safety analyses.

Moreover, an integrated assessment of the

risk impact of extending the AOT for a single

inoperable LPSI train has concluded that the

risk contribution is very small, LPSI system

reliability can potentially be improved, and

the net impact of the proposed change can be

risk beneficial. Therefore, operation of either

facility in accordance with its proposed

amendment would not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort

Pierce, Florida 34954-9003

Attorney for licensee: J. R. Newman,

Morgan, Lewis & Bockius, 1800 M

Street, NW., Washington, DC 20036

NRC Project Director: David B.

Matthews

Florida Power and Light Company, et

al., Docket Nos. 50-335 and 50-389, St.

Lucie Plant, Unit Nos. 1 and 2, St. Lucie

County, Florida

Date of amendment request: June 21,

1995

Description of amendment request:

The proposed amendments will revise

the action statements and certain

surveillances of TS 3/4.5.1, Safety

Injection Tanks (SIT). This proposal is

based on the results of a cooperative

study performed by participating

Combustion Engineering Owners Group

members which investigated the impact

of a risk-based allowed outage time

(AOT) extension, and also included

recommendations for line-item TS

improvements from NUREG-1366 and

Generic Letter 93-05.

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 amendment would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The license amendments proposed for St.

Lucie Units 1 and 2 incorporate certain line-
Federal Register / Vol. 60, No. 187 / Wednesday, September 27, 1995 / Notices 49937

item Technical Specifications (TS)

improvements for the Safety Injection Tanks

(SIT), and include an extension of the

required action completion/allowed outage

time (AOT) from one hour to 72 hours to

restore an inoperable SIT (that is still able to

perform its safety function) to operable

status. In addition, an AOT of 24 hours,

based on risk assessment techniques, is

proposed for an SIT that may be unable to

perform its design function.

The SITs are passive components of the

Emergency Core Cooling System (ECCS). As

such, they are not accident initiators for any

transient evaluated in the plant safety

analyses, and an extension of the AOTs for

restoring an inoperable SIT to operable status

would not increase the probability of

occurrence of accidents previously analyzed.

The SITs, in combination with other ECCS

components, are used to mitigate the

consequences of a loss of coolant accident.

The TS revisions will provide a longer AOT

for a single inoperable SIT, but do not

involve a change to the ECCS configuration

or method of operation. The proposed

amendments will not change the conditions

assumed for the minimum amount of

operating equipment needed for accident

mitigation. Therefore, the consequences of an

accident previously evaluated will not be

significantly increased.

In addition to the preceding evaluation, a

Probability Safety Assessment (PSA) was

performed to quantitatively assess the risk

impact of the 24 hour AOT proposal. The

impact on the early radiological release

probability for design basis events was also

evaluated. It was concluded that the risk

contribution from this AOT is very small,

and that the impact is negligible.

Therefore, operation of either facility in

accordance with its proposed amendment

would not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

(2) Operation of the facility in accordance

with the proposed amendment would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed amendments will not change

the physical plant or the modes of plant

operation defined in either Facility License.

The changes do not involve the addition or

modification of equipment, nor do they alter

the design of plant systems. Therefore,

operation of either facility in accordance

with its proposed amendment would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

(3) Operation of the facility in accordance

with the proposed amendment would not

involve a significant reduction in a margin of

safety.

The margin of safety associated with the

ECCS system is established by acceptance

criteria for system performance defined in 10

CFR 50.46. The proposed amendments will

not change this criteria nor the operability

requirements for equipment that is used to

achieve such performance as demonstrated

by the plant safety analyses. Moreover, an

integrated assessment of the risk impact of

allowing 24 hours to restore an inoperable

SIT to operable status has concluded that this

impact is very small, and can be offset by

averting an unnecessary transition to the

shutdown modes. Therefore, operation of

either facility in accordance with its

proposed amendment would not involve a

significant reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort

Pierce, Florida 34954-9003

Attorney for licensee: J. R. Newman,

Morgan, Lewis & Bockius, 1800 M

Street, NW., Washington, DC 20036

NRC Project Director: David B.

Matthews

Florida Power and Light Company, et

al., Docket Nos. 50-335 and 50-389, St.

Lucie Plant, Unit Nos. 1 and 2, St. Lucie

County, Florida

Date of amendment request: August

16, 1995

Description of amendment request:

The revisions will modify Technical

Specification 3.6.6.1, Shield Building

Ventilation System (SBVS), to more

effectively address the design functions

performed by the SBVS for both the

Shield Building (secondary

containment) and the Fuel Handling

Building.

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 amendment would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed license amendment for St.

Lucie Unit 2 will clarify the Applicability

and the Actions required by Technical

Specification (TS) 3.6.6.1, and explicitly

account for the dual purpose of the Shield

Building Ventilation System (SBVS) to

perform design functions for both the Shield

Building (secondary containment) and the

Fuel Handling Building. The proposed

amendment is administrative in nature.

The SBVS only operates when actuated by

automatic control signals generated by

systems detecting postulated accident

conditions. The SBVS is not an accident

initiator, the proposed TS changes do not

involve any assumptions relative to accident

initiators used in the plant safety analyses,

and the amendment, therefore, will not

impact the probability of occurrence for

accidents previously analyzed. Relative to

accident consequences, at least one train of

the SBVS must operate to fulfill the design

function of evacuating filtered air from the

Shield Building during the postulated Loss of

Coolant Accident; and likewise assumed in

the analysis for the Fuel Handling Building

during a fuel handling accident. The

proposed changes simply remove elements of

ambiguity from TS 3.6.6.1; do not reduce the

existing operability requirements for the

system; and provide further assurance that

proper compensatory measures will be taken

in the event one or both SBVS trains become

inoperable.

Therefore, operation of the facility in

accordance with the proposed amendment

will not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

(2) Operation of the facility in accordance

with the proposed amendment would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed amendment is

administrative in nature and will not change

the physical plant or the modes of plant

operation defined in the facility license. The

changes do not involve the addition or

modification of equipment, nor do they alter

the design or methods of operation of plant

systems. Plant configurations that are

prohibited by TS will not be created by this

amendment. Therefore, operation of the

facility in accordance with the proposed

amendment would not create the possibility

of a new or different kind of accident from

any accident previously evaluated.

(3) Operation of the facility in accordance

with the proposed amendment would not

involve a significant reduction in a margin of

safety.

The proposed amendment will not change

the SBVS operability requirements nor

otherwise alter the basis for any technical

specification that is related to the

establishment of, or the maintenance of, a

nuclear safety margin. The proposed changes

are administrative in nature, and are

designed to provide assurance that the SBVS

capability to perform design functions

assumed available in the safety analyses will

remain available during the various plant

operating modes. Therefore, operation of the

facility in accordance with the proposed

amendment would not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort

Pierce, Florida 34954-9003

Attorney for licensee: J. R. Newman,

Morgan, Lewis & Bockius, 1800 M

Street, NW., Washington, DC 20036

NRC Project Director: David B.

Matthews
49938 Federal Register / Vol. 60, No. 187 / Wednesday, September 27, 1995 / Notices

Florida Power and Light Company, et

al., Docket Nos. 50-335 and 50-389, St.

Lucie Plant, Unit Nos. 1 and 2, St. Lucie

County, Florida

Date of amendment request: August

16, 1995

Description of amendment request:

The proposed amendments revise St.

Lucie Units 1 and 2 Technical

Specifications to relocate selected

Technical Specification Monitoring

Instrumentation utilizing the Final

Policy Statement on Technical

Specification Improvement for Nuclear

Power Reactors, 58 FR 39132, July 22,

1993. The proposed amendments also

include relocation of Technical

Specifications related to the Emergency

and Security Plan review process

utilizing the guidance contained in NRC

Generic Letter 93-07, ‘‘Modification of

the Technical Specification

Administrative Requirements for

Emergency and Security Plans.’’

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 changes to the Selected

Technical Specification Requirements

Related to Instrumentation are administrative

in nature in that the specifications for

operation and surveillance of the selected

Technical Specification instrumentation will

be relocated from Appendix A of the facility

operating license to the Updated Final Safety

Analysis Report (UFSAR) for each unit. Once

relocated, future changes will be controlled

by 10 CFR 50.59 and the UFSARs updated

pursuant to 10 CFR 50.71(e). Relocation of

these requirements to the UFSAR is

consistent with the NRC ‘‘Final Policy

Statement on Technical Specifications

Improvements for Nuclear Power Reactors’’

published in the Federal Register (58 FR

39132) dated July 22, 1993.

The selected Technical Specification

instruments are not accident initiators nor a

part of the success path(s) which function to

mitigate accidents evaluated in the plant

safety analyses. The proposed Technical

Specification change does not involve any

change to the configuration or method of

operation of any plant equipment that is used

to mitigate the consequences of an accident,

nor do the changes alter any assumptions or

conditions in any of the plant accident

analyses. Therefore, operation of the facility

in accordance with the proposed amendment

would not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

The Technical Specifications changes

associated with Emergency Plan and Security

Plan requirements are proposed in

accordance with Generic Letter 93 07. The

changes being proposed are administrative in

nature and do not affect assumptions

contained in plant safety analyses, the

physical design and/or operation of the plant,

nor do they affect Technical Specifications

that preserve safety analysis assumptions.

Therefore, operation of the facility in

accordance with the proposed amendments

would not affect the probability or

consequences of an accident previously

analyzed.

(2) Use of the modified specification would

not create the possibility of a new or different

kind of accident from any previously

evaluated.

The proposed amendment to relocate the

existing Technical Specification

requirements for selected Technical

Specification instrumentation to the UFSAR

will not change the physical plant or the

modes of plant operation defined in the

Facility License. The change does not involve

the addition or modification of equipment

nor does it alter the design or operation of

plant systems. Therefore, operation of the

facility in accordance with the proposed

amendment would not create the possibility

of a new or different kind of accident from

any accident previously evaluated.

The proposed amendments, in accordance

with Generic Letter 93-07, change the

Technical Specifications to remove the audit

of the emergency and security plans and

implementing procedures from the list of

responsibilities of the Facility Review Group.

The changes being proposed are

administrative in nature and will not change

the physical plant or the modes of operation

defined in the Facility License. The change

does not involve the addition or modification

of equipment nor does it alter the design or

operation of plant systems. 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) Use of the modified specification would

not involve a significant reduction in a

margin of safety.

The proposed changes are administrative

in nature in that operating and surveillance

requirements for the selected Technical

Specification instrumentation will be

relocated from Appendix A of the facility

license to the appropriate Updated Final

Safety Analysis Report for each unit. These

selected instruments are not used to actuate

safety-related equipment, provide interlocks,

or otherwise perform plant control functions.

Conditions evaluated in plant accident and

transient analyses do not involve these

selected instruments. The proposed changes

do not alter the basis for any technical

specification that is related to the

establishment of, or the maintenance of, a

nuclear safety margin. Therefore, operation of

the facility in accordance with the proposed

amendment would not involve a significant

reduction in a margin of safety.

The proposed amendments, in accordance

with Generic Letter 93-07, change the

Technical Specifications to remove the audit

of the emergency and security plans and

implementing procedures from the list of

responsibilities of the Facility Review Group.

The changes being proposed are

administrative in nature and do not alter the

bases for assurance that safety-related

activities are performed correctly or the basis

for any Technical Specification that is related

to the establishment of or maintenance of a

safety margin. Therefore, operation of the

facility in accordance with the proposed

amendments would not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort

Pierce, Florida 34954-9003

Attorney for licensee: J. R. Newman,

Morgan, Lewis & Bockius, 1800 M

Street, N.W., Washington, DC 20036

NRC Project Director: David B.

Matthews

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

Duane Arnold Energy Center, Linn

County, Iowa

Date of amendment request: July 21,

1995

Description of amendment request:

The proposed amendment would make

administrative changes to various

sections of the Duane Arnold Energy

Center (DAEC) Technical Specifications

(TS). These changes replace a

conditional surveillance if one

emergency service water (ESW) pump or

loop is determined to be inoperable (TS

4.8.E.2); credit successful emergency

diesel generator (EDG) tests performed

in the previous 24 hours (TS 4.8.E.2);

clarify the requirements governing spent

and new fuel storage in Section 5.5 of

the DAEC TS; and eliminate the

Operations Committee reviews of

procedures in support of the DAEC

Emergency Plan and Security Plan, as

specified in Sections 6.5 and 6.8 of the

TS. DAEC TS Section 4.8.E.2 states the

surveillance requirement applicable

when one ESW pump or loop is

determined to be inoperable. This

amendment request deletes the

surveillance requirement to physically

test the opposite train’s EDG and

replaces it with a requirement to verify

OPERABILITY of the opposite train low

pressure core and containment cooling

systems and EDG.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the
Federal Register / Vol. 60, No. 187 / Wednesday, September 27, 1995 / Notices 49939

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed revision does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated. The changes are administrative in

nature and are consistent with previouslypublished NRC guidance. The proposed

revision does not change any accident

analysis, plant safety analysis or calculations;

degrade existing plant programs; or modify

any functions of safety related systems or

accident mitigation functions for which the

DAEC has previously been credited. The

proposed revision to the Surveillance

Requirements will continue to assure

OPERABILITY as required, but eliminate

unnecessary operation of an EDG.

2. The proposed revision does not create

the possibility of a new or different kind of

accident from any accident previously

evaluated. The proposed revision does not

alter any plant parameters, revise any safety

limit setpoint, or provide any new release

pathways. In addition, the proposed revision

does not modify the operation or function of

any safety-related equipment, nor introduce

any new modes of operation, failure modes,

or physical changes to the plant.

3. The proposed revision does not involve

a significant reduction in a margin of safety.

The proposed revision does not alter any

plant parameters, revise any safety limit

setpoint, or provide any new release

pathways. In addition, the proposed revision

does not modify the operation or function of

any safety-related equipment, nor introduce

any new modes of operation, failure modes,

or physical changes to the plant.

The NRC staff has reviewed the

licensee’s analysis and, based on

thisreview, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Cedar Rapids Public Library,

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

Iowa 52401

Attorney for licensee: Jack Newman,

Kathleen H. Shea, Morgan, Lewis &

Bockius, 1800 M Street, NW.,

Washington, DC 20036-5869

NRC Project Director: Gail H. Marcus

Nebraska Public Power District, Docket

No. 50-298, Cooper Nuclear Station,

Nemaha County, Nebraska

Date of amendment request: May 5,

1995, as revised by letter dated July 14,

1995

Description of amendment request:

The proposed changes would amend the

Cooper Nuclear Station (CNS) Technical

Specifications (TS) sections 3/4.5.F.1,

3.5.F.2, 3.9.B.1, 3.9.B.2, 4.9.A.2, and the

associated bases. These changes would

revise the TS to: 1) verify that the

redundant diesel generator is operable

upon the loss of one diesel generator,

and implement provisions to verify that

the operable diesel generator does not

have a common cause failure; 2)

incorporate provisions to allow a

modified start for the diesel generators;

and 3) remove the requirement that the

reactor power level be reduced to 25%

of rated power upon loss of both diesel

generator units or both incoming power

sources (start-up and emergency

transformers). In addition, the period of

time allowed for continued reactor

operation with both diesels inoperable

would be reduced from 24 to two hours.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

10 CFR 50.91(a)(1), requires that licensee

requests for operating license amendments be

accompanied by an evaluation of significant

hazards posed by the issuance of the

amendment. NPPD has reviewed the

proposed changes in accordance with

10CFR50.92 and concludes that the changes

do not involve a significant hazards

consideration (SHC). The basis for this

conclusion is that the three criteria of

10CFR50.92(c) are not compromised. The

proposed changes do not involve a SHC

because the changes would not:

1. Involve a significant increase in the

probability or consequences of an accident

previously analyzed.

Proposed Revision 1:

This proposed revision serves to ensure

that an emergency diesel generator is always

available to perform on demand and that

lowering the number of demands to

demonstrate operability reduces the

probability of equipment failure. The

required action no longer requires the

redundant emergency diesel generator to be

demonstrated operable immediately.

Therefore, this requirement has been deleted

from TS 4.5.F.1.

The proposed change includes provisions

to determine if the redundant diesel

generator has been made inoperable by a

common cause failure or perform a

demonstration test. The redundant

emergency diesel generator will remain in

service during the entire period of

inoperability of the out of service emergency

diesel generator. If a common cause failure

cannot be ruled out, the redundant diesel

generator will be tested in accordance with

the surveillance requirements of TS section

4.9.A.2.a.1 to assure operability.

Since this proposed revision does not

affect the design or negatively affect the

performance of the diesel generators, the

change will not result in an increase in the

consequences or probability of an accident

previously analyzed. This proposed revision

will increase diesel generator reliability and

availability, thereby increasing overall plant

safety.

Proposed Revision 2:

This proposed revision only affects

emergency diesel generator periodic testing.

The diesel generators are not accident

initiators and the method of testing the diesel

generators cannot initiate an accident and

therefore will not increase the probability of

an accident. This change to the diesel

generator testing method does not impact any

Updated Safety Analysis Report (USAR)

safety analysis. The proposed surveillances

will still provide assurance that the diesel

generators are available to mitigate the

consequences of accidents previously

evaluated. Thus the consequences of an

accident previously evaluated are not

increased.

The revised periodic testing will still

demonstrate that the emergency diesel

generators are ready to perform their safety

function. An overall improvement in diesel

engine reliability and availability can be

gained by performing diesel generator starts

for surveillance testing using engine

prelubes, warmups and other manufacturer

recommended practices to reduce engine

stress and wear. Since this proposed revision

does not affect the design or negatively affect

the performance of the diesel generators, the

change will not result in an increase in the

consequences or probability of an accident

previously analyzed. This proposed revision

will increase diesel generator reliability,

thereby increasing overall plant safety.

Proposed Revision 3:

This proposed revision does not affect the

operation of the emergency diesel generators

or the incoming power sources (start-up and

emergency transformers). Both the diesel

generators and the incoming power sources

function to mitigate the consequences of

postulated accidents. As such, removing the

requirement to reduce power level upon the

loss of both redundant components in either

of these systems does not create an increase

in the probability of an accident. By

eliminating this requirement, the potential

for plant transients during power reduction

to 25% are also eliminated. Eliminating this

requirement will not increase the

consequences of a postulated accident

because the redundant components will

remain available. Additionally, the loss of

both offsite power sources condition becomes

more restrictive by requiring a plant

shutdown instead of notification within 24

hours.

The proposed changes do not alter the

conditions or assumptions in any of the

Updated Safety Analysis Report (USAR)

accident analyses. Since the USAR accident

analyses remains bounding, the radiological

consequences previously evaluated are not

adversely affected by the proposed changes.

Therefore, no significant increase in the

probability or consequences of an accident

previously analyzed would occur.

The proposed rearrangement of

information, and rewording of some the TS

requirements are included to enhance

usability and alleviate any possible

confusion. These changes are strictly

editorial have no impact, and do not alter

technical content or meaning of the

specifications. These editorial changes do not

involve a significant increase in the

probability or consequences of an accident

previously analyzed.
49940 Federal Register / Vol. 60, No. 187 / Wednesday, September 27, 1995 / Notices

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

Proposed Revision 1:

Accidents involving loss of off-site power

and single failure have been previously

evaluated, and this proposed change does not

impact any of those assumptions. This

proposed revision does not introduce any

new mode of plant operation or new accident

precursors, involve any physical alterations

to plant configurations, or make changes to

system setpoints which could initiate a new

or different kind of accident. Operation of the

facility in accordance with the proposed

revised changes does not create the

possibility of a new or different kind of

accident from any previously evaluated.

Proposed Revision 2:

This proposed revision only affects

emergency diesel generator periodic testing.

The diesel generators are not accident

initiators and the method of testing the diesel

generators cannot initiate an accident. This

revision does not relieve the operation of the

diesel generator from existing requirements

and the diesel generators remain bounded by

the assumptions in the USAR accident

analysis. The method of testing provides

assurance that the diesel generators are

available when needed. The proposed

revision does not involve any changes in

setpoints, plant equipment, plant operation,

protective functions, or the design basis of

the plant. Therefore, a change in the method

of starting the diesel generators during

periodic testing would not create a different

kind of accident than previously evaluated.

Proposed Revision 3:

This proposed revision does not add or

change any equipment or logic, nor do the

changes associated with this revision alter

any system operability requirements. The

proposed changes for this revision do not

introduce any new failure modes for any

plant system or component important to

safety nor has any new limiting failure been

identified as a result of the proposed

revision. Since there are no changes to the

function, or operation of any system,

equipment, or component, the possibility of

a new or different kind of accident is not

created.

The proposed rearrangement of

information, and rewording of some [of] the

TS requirements are included to enhance

usability and alleviate any possible

confusion. These changes are strictly

editorial have no impact, and do not alter

technical content or meaning of the

specifications. These editorial changes do not

create the possibility of a new or different

kind of accident from any previously

analyzed.

3. Involve a significant reduction in the

margin of safety.

Proposed Revision 1:

This proposed revision does not result in

an overall reduction in the margin of safety.

The reduction in margin going from

‘‘immediately’’ testing an operable diesel

generator to 24 hours to determine no

common cause, is offset by the increase in

margin resulting from increased diesel

generator reliability and availability

associated with implementing the vendor

recommendations for testing and not

exposing the diesel generator to potential

grid disturbances when a diesel generator is

found to be inoperable. No physical

modification to the plant or change in the

procedurally prescribed operator actions

result from the proposed changes associated

with this revision. Operation of the facility in

accordance with the proposed revision does

not involve a significant reduction in a

margin of safety.

Proposed Revision 2:

This proposed revision is made to increase

the reliability and availability of the

emergency diesel generators thus enhancing

the safety of the plant. Changing the way

periodic testing of the diesel generators is

conducted does not involve a reduction in

safety. The test still demonstrates the ability

of the diesel generator to start within the time

required, and reach rated voltage and

frequency as required in the accident

analysis. The test also demonstrates the

ability of the diesel generator to start reliably,

carry the required load, and ensures the

capabilities of the cooling system and other

support systems are operable. Therefore,

assurance that the diesel generators operate

within the limits determined to be acceptable

continues to be provided. Implementing

manufacturer’s recommendations to

minimize stress and wear of the diesel engine

does not involve a significant reduction in

the margin of safety, but rather enhances

safety.

Proposed Revision 3:

This proposed revision deletes the

requirement to reduce reactor power level to

25% of rated power upon the loss of either

both diesel generators or both incoming

power sources. The elimination of this

requirement will allow the plant to maintain

the existing power level rather than subject

the plant to an unnecessary transient.

Maintaining the plant at the existing power

level provides a more stable operating

environment. The equipment and

components of the diesel generators or the

incoming power sources are not impacted in

any way as a result of the proposed revisions.

The margin of safety for the diesel generators

and the incoming power sources are not

significantly reduced since these systems are

not altered in any way, and will continue to

be surveillance tested as required. Assurance

of operability is provided by the normal,

scheduled surveillances which have been

established at a sufficient interval to provide

reasonable assurance of operability.

Therefore, the proposed changes do not

involve a significant reduction in the margin

of safety.

The proposed rearrangement of

information, and rewording of some [of] the

TS requirements are included to enhance

usability and alleviate any possible

confusion. These changes are strictly

editorial have no impact, and do not alter

technical content or meaning of the

specifications. These editorial 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. The licensee’s July 14, 1995,

letter revised the proposed changes in

their letter of May 5, 1995, to further

limit the period of time that continued

reactor operation would be allowed

with both emergency diesel generators

inoperable from 24 to two hours. This

revision to the proposed changes is

more restrictive and does not impact the

licensee’s analysis of the criteria of 10

CFR 50.92(c). Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Auburn Public Library, 118

15th Street, Auburn, Nebraska 68305

Attorney for licensee: Mr. John R.

McPhail, Nebraska Public Power

District, Post Office Box 499, Columbus,

Nebraska 68602-0499

NRC Project Director: William D.

Beckner

Northeast Nuclear Energy Company

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

Nuclear Power Station, Unit 1, New

London County, Connecticut

Date of amendment request: August

31, 1995

Description of amendment request:

The proposed amendment modifies the

definition of HOT SHUTDOWN and

COLD SHUTDOWN to specify that the

definitions are not applicable during the

performance of an inservice hydrostatic

and leak test (IHLT). Technical

Specification Section 3.6.B and 4.6.B

would be modified by adding Section

3.6.B.1.b and 4.6.B.1.b to identify the

requirements that must be satisfied to

consider the reactor in COLD

SHUTDOWN during the performance of

an IHLT. In addition, the proposed

amendment will change temperature

specific requirements on several pages

to mode or condition specific

requirements; make several editorial

changes; and change the associated

Bases.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has rovided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

NNECO has reviewed the proposed

changes in accordance with 10CFR50.92 and

concluded that the changes do not involve a

significant hazards consideration (SHC). The

bases for this conclusion is that the three

criteria of 10CFR50.92(c) are not

compromised. The proposed changes do not

involve an SHC because the changes would

not:
Federal Register / Vol. 60, No. 187 / Wednesday, September 27, 1995 / Notices 49941

1. Involve a significant increase in the

probability or consequences of an accident

previously analyzed.

The proposed changes will allow the

reactor to be considered in COLD

SHUTDOWN during an IHLT with the

average reactor coolant temperature greater

than 212°F but less than 280°F. The change

to allow the reactor to be in COLD

SHUTDOWN during the performance of

IHLT will not increase the probability or

consequences of an accident. The probability

of a leak in the reactor pressure boundary

during this testing is not increased by

considering the reactor to be in COLD

SHUTDOWN. The IHLT is performed near

water solid, all control rods inserted, and

with an appropriate availability of

engineering safety features. The stored energy

in the reactor core will be very low and the

potential for failed fuel and a subsequent

increase in coolant activity are minimal. In

addition, secondary containment will be

operable and capable of handling airborne

radioactivity from leaks that could occur

during the performance of an IHLT.

Requiring secondary containment to be

operable will further ensure that potential

airborne radiation from leaks will be filtered

by one or both trains of SBGT [standby gas

treatment], thereby limiting releases to the

environment. Therefore, the changes will not

significantly increase the consequences of an

accident.

In the unlikely event of a large pressure

boundary leak, the reactor vessel would

rapidly depressurize, allowing one or both of

the operable core spray systems to operate.

Small system leaks would be detected by

leakage inspections before significant

inventory loss occurred, since leakage

inspections are an integral part of the IHLT

program.

Based upon the above, the proposed

changes do not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

The IHLT conditions remain unchanged.

The potential for a system leak remains

unchanged since the reactor coolant system

is designed for temperatures exceeding 500°F

with similar pressures. The change in

operable engineered safety features available

to mitigate a postulated accident does not

reduce the ability to

safely mitigate a postulated accident.

Adequate ECCS [emergency core cooling

system] equipment will be available to

mitigate a LOCA [loss of coolant accident]

with an assumed single failure. Therefore,

this will not create the possibility of a new

or different kind of accident from any

previously evaluated.

3. Will not involve a significant reduction

in a margin of safety.

The proposed changes will not have any

significant impact on any design basis

accident or safety limit. The various

engineered safety features which are required

by the proposed change will ensure

appropriate mitigation of postulated events.

Since the test is performed at a near water

solid condition and at low decay heat values,

no fuel damage is expected in case of an

accident such as a LOCA. Nevertheless,

secondary containment and the SBGT system

will be maintained operable to process airborne radioactivity from a steam leak that

could occur during the performance of the

IHLT. Therefore, the proposed change does

not constitute 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: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360.

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

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

NRC Project Director: Phillip F.

McKee

Northeast Nuclear Energy Company, et

al., Docket No. 50-336, Millstone

Nuclear Power Station, Unit No. 2, New

London, Connecticut

Date of amendment request: August

31, 1995

Description of amendment request:

The proposed change to the Millstone 2

Technical Specifications would remove

the phrase ‘‘other than Millstone Unit

No. 2’’ from Section 6.3.1 on page 6-2.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

NNECO has reviewed the proposed change

in accordance with 10CFR50.92 and

concluded that the change does not involve

a significant hazards consideration (SHC).

The basis for this conclusion is that the three

criteria of 10CFR50.92(c) are not

compromised. The proposed change does not

involve an SHC because the change would

not:

1. Involve a significant increase in the

probability or consequences of an accident

previously analyzed.

The proposed change does not affect any

system or equipment of Millstone Unit No. 2.

The proposed change does not affect the

qualification of any of the licensed

individuals involved in the day-to-day

operation of Millstone Unit No. 2. The

proposed change corrects a statement which

could be interpreted such that an individual

who once held a Millstone Unit No. 2 SRO

[Senior Reactor Operator] license would not

be eligible to be Operations Manager. Since

this change does not affect any equipment or

operating procedures, does not affect the

level of expertise and

training required for on-shift personnel,

and does not reduce the level of expertise

required of operations management, this

change does not involve a significant

increase in the probability or consequences

of an accident previously analyzed.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

This change does not affect any equipment

or operating procedures, does not affect the

level of expertise and training required for

on-shift personnel, and does not reduce the

level of expertise required of operations

management. Therefore, this change does not

create the possibility of a new or different

kind of accident.

3. Involve a significant reduction in the

margin of safety.

This change eliminates a phrase which

could be interpreted to prevent an individual

who had possessed a Millstone Unit No. 2

SRO license from becoming the Operations

Manager. The training and experience

necessary to possess a Millstone Unit No. 2

SRO license is equivalent to that of other

PWRs. Therefore, this proposed change does

not involve a significant reduction in the

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360.

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

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

NRC Project Director: Phillip F.

McKee

Philadelphia Electric Company, Docket

Nos. 50-352 and 50-353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of amendment request: July 28,

1995

Description of amendment request:

The proposed amendment would

eliminate the Technical Specifications

requirements to perform 10 CFR 50,

Appendix J, Type C hydrostatic testing

on certain valves that are within closed

systems and are assured a water seal

following a Design Basis Accident.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed TS changes do not involve

a significant increase in the probability or
49942 Federal Register / Vol. 60, No. 187 / Wednesday, September 27, 1995 / Notices

consequences of an accident previously

evaluated.

The primary containment (drywell and

suppression pool) and the affected closed

systems are accident mitigators not accident

initiators. The proposed change to the scope

of Appendix J, Type C testing does not affect

the probability of the DBA [Design Basis

Accident]. The valves will continue to be

maintained in an operable state, and in their

current design configuration. There is no

correlation between the scope of Appendix J,

Type C testing and accident probability.

There are no physical or operational changes

to the containment structure, system or

components being made as a result of the

proposed changes. Therefore, the

consequences of a malfunction of equipment

important to safety is not increased from

those previously evaluated.

The consequences of loss-of-coolant

accidents (LOCAs) under the proposed

change were considered where a single active

failure of a containment isolation valve (CIV)

or a passive failure of the closed system were

reviewed, within the limits of the existing

licensing basis. Under the existing licensing

basis, a pipe rupture of the seismically

qualified ECCS piping does not have to be

assumed concurrent with the LOCA, except

if it is a consequence of the LOCA.

Consideration of consequential failures can

be eliminated, since a LOCA inside

containment is separated from the affected

piping by the containment structure.

Consideration of consequential failures of the

ECCS piping from LOCAs outside

containment are outside the Appendix J

design considerations. A single active failure

of the CIV, under the LOCA condition, can

be accommodated since the closed and water

sealed system piping remains as the leakage

barrier. The ECCS passive failure criterion

does require consideration of system leaks,

but not pipe breaks, beyond the initiating

LOCA. The capability to make-up water

inventory to the suppression pool is adequate

to ensure that postulated seat leakage and

pipe leakage does not result in a condition

that jeopardizes pool level. Make-up

capability exists for the suppression pool via

the Condensate Storage Tank and Ultimate

Heat Sink Spray Pond. Operator actions to

make-up the suppression pool are delineated

in existing Operating Procedures.

The subject valves are single isolation

valves associated with lines that penetrate

the primary containment, but are not

connected directly to the primary

containment atmosphere or the reactor

coolant pressure boundary. This

configuration is described in the LGS

UFSAR, Section 6.2.4.3.1.3.1, which states

‘‘the systems which the lines from the

suppression pool connect to outside

containment are closed systems meeting the

appropriate requirements of closed systems.’’

The integrity of these closed systems are also

monitored and controlled in accordance with

TS Section 6.8.4.a. Any leakage that may

escape the confines of the closed system will

be contained within the Reactor Building,

treated by standby gas and radwaste systems,

and, therefore, are within the existing LGS

licensing bases.

Finally, the affected penetrations will

continue to be subjected to the periodic 10

CFR 50, Appendix J, Type A test (Integrated

Containment Leakage Rate Test).

The suppression pool level is designed and

operated so that water level is maintained in

accordance with current TS, and the

associated bases. The supply of water in the

suppression pool is assured for 30 days

during all DBA, post-accident modes of

operation. The lowest water level which the

suppression pool will reach was analyzed,

and it was determined that the affected lines

will remain below this minimum level,

thereby assuring a water seal. The valves will

continue to be tested and maintained to

ensure their operability, and the closed

systems’ integrity will continue to be

monitored and controlled in accordance with

TS 6.8.4.a and the performance of the

periodic 10 CFR 50, Appendix J, Type A test.

Therefore, the proposed changes will not

increase the probability or consequences of

an accident previously evaluated.

2. The proposed TS changes do not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes do not change the

plant response to accident scenarios, and do

not introduce new or different scenarios. The

primary containment (drywell and

suppression pool) and the affected closed

systems are accident mitigators not accident

initiators. The proposed change to the scope

of Appendix J, Type C hydrostatic testing

maintains the existing barriers to primary

containment bypass leakage by the assurance

that a water seal will be maintained for 30

days during all DBA, post-accident modes of

operation. The valves will continue to be

tested and maintained to ensure their

operability, and the closed systems’ integrity

will continue to be monitored and controlled

in accordance with TS 6.8.4.a. Therefore, the

proposed changes cannot cause an accident,

and the plant response to the design basis

events is unchanged, whereby the change

does not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

3. The proposed TS changes do not involve

a significant reduction in a margin of safety.

The water seal provided by the assurance

of a minimum suppression pool level will

prevent post-accident containment bypass

leakage. Appendix J does not require air leak

testing of the valves since the 30 day postaccident supply of water is maintained. In

addition, the closed systems’ integrity is

monitored and controlled in accordance with

TS 6.8.4.a. Any leakage that may escape the

confines of the closed system will be

contained within the Reactor Building, and is

within the existing LGS licensing bases.

Therefore, the proposed TS changes do not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania

19464.

Attorney for licensee: J. W. Durham,

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

Counsel, Philadelphia Electric

Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101

NRC Project Director: John F. Stolz

Philadelphia Electric Company, Docket

Nos. 50-352 and 50-353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of amendment request: July 28,

1995

Description of amendment request:

The proposed amendments, which are

consistent with the Improved Standard

Technical Specifications (NUREG-1433),

delete the operability and surveillance

requirements involving secondary

containment differential pressure

instrumentation.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed Technical Specifications

(TS) changes do not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

Deleting the operability and surveillance

requirements for the secondary containment

differential pressure instrumentation does

not involve any changes to the design,

function, or operation of any plant

components or safety-related systems. There

are no changes to the separation,

redundancy, qualification, quality assurance

or fire protection requirements for the

associated components and systems, nor are

there any new failure modes created. This

activity only removes operability and

surveillance requirements from the Technical

Specifications for selected plant components

associated with the secondary containment

differential pressure trip functions. No credit

for operation of these trip functions is taken

in any design basis accidents valuated in the

SAR [Safety Analysis Report].

Thesecomponents will be maintained in

accordance with the plant preventive

maintenance program. The failure of any of

these components does not result in the

occurrence of an accident. Consequently,

there is no increase in the probability of

occurrence of an accident previously

evaluated in the SAR.

The Outside Atmosphere to Reactor

Enclosure Delta Pressure-Low and Outside

Atmosphere To Refueling Area Delta

Pressure-Low trip functions are not

symptomatic of a design basis accident. No

credit for operation of the trip functions is

taken in any design basis accidents evaluated

in the SAR. Neither failure of the differential

pressure components nor failure to generate

the associated trip functions affects the

consequences of an accident previously

evaluated in the SAR. The appropriate
Federal Register / Vol. 60, No. 187 / Wednesday, September 27, 1995 / Notices 49943

accident prevention and mitigation actions

are generated from other plant parameters

symptomatic of an accident. Sufficient plant

parameters symptomatic of a design basis

accident are monitored to initiate the

appropriate actions as evaluated in the SAR.

Furthermore, all safety-related systems will

still be able to perform all of their design

basis safety-related functions. Consequently,

there is no increase in the consequences of

an accident previously evaluated in the SAR.

Therefore, the proposed TS changes do not

involve an increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed TS changes do not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The failure of the differential pressure

automatic isolation instrumentation

components does not result in the occurrence

of an accident. The failure to generate the

associated trip functions does not result in

the occurrence of an accident. This activity

does not involve any changes to the design,

function, or operation of any plant

components or safety-related systems. There

are no changes to the separation,

redundancy, qualification, quality assurance

or fire protection requirements for the

associated components and systems. These

components will be maintained in

accordance with the plant preventative

maintenance program. Consequently, there is

no possibility of an accident of a different

type than previously evaluated in the SAR.

Therefore, the proposed TS changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed TS changes do not involve

a significant reduction in a margin of safety.

The ability of secondary containment to

minimize any ground level release of

radioactive material which may result from

any accident is not affected. Surveillance and

operability requirements for secondary

containment SGTS [Standby Gas Treatment

System] and RERS [Reactor Enclosure

Recirculation System] are not changed by

this activity. Draw down time, leakage

factors, secondary containment system

ratings, and secondary containment system

response to a LOCA [Loss of Coolant

Accident] or refueling accident are not

affected by this activity. SGTS and RERS

initiation will continue to occur when plant

parameters symptomatic of a LOCA or

refueling accident exceed predetermined

values. There are no changes to the inputs for

the post-LOCA offsite dose analysis.

Therefore, the proposed TS changes do not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania

19464.

Attorney for licensee: J. W. Durham,

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

Counsel, Philadelphia Electric

Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101

NRC Project Director: John F. Stolz

Philadelphia Electric Company, Docket

Nos. 50-352 and 50-353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of amendment request: July 28,

1995

Description of amendment request:

The proposed amendment would

modify Technical Specifications (TS)

Surveillance Requirements 4.9.1.1,

4.9.1.2, 4.9.3, 4.9.5, and 4.9.8 to delete

specific requirements to perform

surveillances just prior to beginning or

resuming core alterations or control rod

withdrawal associated with refueling

activities. This proposed TS change

would delete the phrase ‘‘incore

instrumentation’’ from the footnote in

TS Section 3/4.9.5, ‘‘Communications.’’

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed Technical Specifications

(TS) changes do not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

The proposed TS changes do not involve

any physical changes to plant systems or

equipment. The proposed TS changes only

delete those Surveillance Requirements (SRs)

pertaining to the performance of tests just

prior to beginning or resuming core

alterations or control rod withdrawal, and

revises a footnote description to be consistent

with the current TS definition of ‘‘Core

Alteration.’’ The proposed TS changes do not

revise any of the other applicable periodic

SRs, or modify any procedural controls

currently in place governing fuel handling

operations. The periodic surveillance test

frequencies provide adequate assurance that

the equipment will remain in an operable

condition. The normal periodic surveillance

intervals bound those surveillance intervals

for the tests that are being altered by this

proposed TS change. In the event that one of

the periodic surveillances has not been

performed within the specified time interval,

entry into the specified condition (i.e.,

performance of core alterations, control rod

withdrawal, or handling of fuel or control

rods) is not permitted as required by TS 4.0.4

until the surveillance has been satisfactorily

completed.

The consequences of an accident are not

increased by the proposed TS changes, since

the changes only involve revising the

frequency of conducting surveillance tests.

The method of operation or performance of

plant structures, systems, or plant

components are not affected by the proposed

TS changes. The proposed TS changes will

not impact the operation of any fuel handling

equipment, and therefore, the potential for a

Fuel Handling Accident as described in

Section 15.7.4 of the LGS [Limerick

Generating Station] Updated Final Safety

Analysis Report (UFSAR) is not increased.

In addition, any unexpected reduction of

water level in the reactor cavity or fuel pool

at the start of fuel handling or control rod

handling will be immediately apparent to

operators by direct observation. Plant

procedures utilized by the refueling

personnel require the suspension of core

component transfers in the event of loss of

water inventory.

Therefore, the proposed TS change does

not involve an increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed TS changes do not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed TS changes only involve

changes to the frequency in which the

specified surveillance tests are performed.

The proposed TS changes do not revise any

of the other applicable periodic SRs, or

modify any procedural controls currently in

place governing fuel handling operations.

The periodic surveillance test frequencies

provide adequate assurance that the

equipment will remain in operable condition.

The periodic surveillance intervals bound

those surveillance intervals for the tests that

are being altered by this proposed TS change.

The refueling interlock system combined

with strict procedural controls provide

multiple barriers to preclude an inadvertent

criticality.

The proposed TS changes do not involve

any physical changes to plant systems or

equipment. The proposed TS changes do not

alter the configuration of the plant or the way

that the plant is operated. The associated

plant equipment will continue to function as

designed. This equipment is not designed to

perform any other function than it is

presently capable of, and therefore, will not

affect the operation of any other plant

equipment.

Therefore, the proposed TS change does

not create the possibility of a new or different

kind of accident from any previously

evaluated.

3. The proposed TS changes do not involve

a significant reduction in a margin of safety.

The proposed TS changes do not involve

any physical changes to plant systems or

equipment. The reactor will continue to be

maintained subcritical during refueling

operations and reactor water level will be

maintained at the required level (i.e., above

the vessel flange). The proposed TS changes

do not affect the operation of other plant

systems and equipment essential in

maintaining reactor water temperature during

refueling operations, or the capability in

responding to a postulated Fuel Handling

Accident.

The proposed changes do not adversely

affect reliability of the refueling interlocks or

refuel platform communications equipment.
49944 Federal Register / Vol. 60, No. 187 / Wednesday, September 27, 1995 / Notices

Since the proposed changes only impact the

frequency in which certain surveillance tests

are performed, and do not change the plant

configuration or setpoints, there is

substantial assurance that the reactor will be

maintained subcritical during refueling.

Therefore, the proposed TS changes do not

involve a reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania

19464.

Attorney for licensee: J. W. Durham,

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

Counsel, Philadelphia Electric

Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101

NRC Project Director: John F. Stolz

Philadelphia Electric Company, Docket

Nos. 50-352 and 50-353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of amendment request: July 28,

1995

Description of amendment request:

The proposed amendment would revise

Technical Specifications Table 4.3.1.1-1,

‘‘Reactor Protection System

Instrumentation Surveillance

Requirements’’, to reflect changes to the

surveillance test frequency requirements

for various Reactor Protection System

[RPS] instrumentation.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed Technical Specifications

(TS) changes do not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

In all of the applicable SAR [Safety

Analysis Report] evaluated events, the IRM

[Intermediate Range Monitor] and APRM

[Average Range Power Monitor]

instrumentation is credited for performing a

mitigating function (i.e., initiating a scram),

to terminate the transient prior to a safety

limit being exceeded. The proposed TS

changes do not alter the RPS configuration,

or RPS instrumentation setpoints, nor do

they change the manner in which the IRM

and APRM instrumentation carry out the

scram functions. Therefore the consequences

of any potential malfunction of equipment

important to safety will remain unchanged.

In each case where a startup surveillance

test requirement is proposed to be deleted,

(i.e., IRM and APRM), the normal

surveillance test frequency specified for the

required Operational Condition remains

unchanged (except for the APRM Upscale

Setdown functional test). The startup

surveillance requirement is conservatively

bounded by the normal surveillance test

interval which is greater than or equal to any

interval associated with the startup

surveillance requirement and ensures that

the IRM and APRM instrumentation

reliability is unchanged. This is in

accordance with the Improved Standard

Technical Specifications, NUREG-1433,

issued September 28, 1992.

The reliability of the APRM Upscale

Setdown scram function will not be

decreased due to changing the functional test

frequency from Weekly (W), to Quarterly (Q),

in Operational Conditions 2, 3, and 5

(Startup, Hot Shutdown and Refueling,

respectively). Plant operational data taken

from each of the APRM calibration/

functional tests performed since August 1992

until present at LGS Units 1 and 2, shows

that setpoint reliability will be maintained if

the functional test frequency is increased to

quarterly, as proposed. Presently, each time

an APRM calibration/functional test is

performed, both the Upscale Setdown and

the Flow Reference scram circuits are tested.

The results of the quarterly tests confirm that

the APRM Upscale Setdown function already

has over 2.5 years of performance without

failure in Operational Condition 1, thus being

extremely reliable.

Therefore, the proposed TS changes do not

involve an increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed TS changes do not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed TS changes affect only the

required surveillance test intervals, not the

RPS configuration or RPS instrumentation

setpoints. The proposed TS changes do not

introduce a new failure mode for the IRM or

APRM instrumentation. Plant operating

experience data confirms that at LGS Units

1 and 2, the IRM and APRM instrumentation

will continue to perform their safety function

as currently designed, with the same degree

of reliability.

The proposed TS changes do not alter the

configuration of the plant, nor the way the

plant is operated.

Therefore, the proposed TS changes do not

create the possibility of a new or different

kind of accident, from any accident

previously evaluated.

3. The proposed TS changes do not involve

a significant reduction in a margin of safety.

The following TS Bases were reviewed for

potential reduction in the margin of safety:

B 2.2.1 Reactor Protection System

Instrumentation Setpoints

B 3/4.1.4 Control Rod Program Controls

B 3/4.2 Power Distribution Limits

B 3/4.3.1 Reactor Protection System

Instrumentation

B 3/4.3.6 Control Rod Block

Instrumentation

The surveillance test frequency changes

proposed for the RPS instrumentation section

of TS do not adversely affect the IRM or

APRM instrumentation, which will continue

to perform the RPS functions required to

maintain the present margin of safety.

Changes to the IRM instrumentation startup

surveillance intervals are already bounded by

the existing surveillance requirements, and

are in accordance with the Improved

Standard Technical Specifications, NUREG1433, issued September 28, 1992. The same

statement applies to the APRM

instrumentation, with respect to deletion of

the startup surveillance requirement. The

change of the APRM Upscale Setdown

Channel functional test surveillance interval

from Weekly to Quarterly was evaluated to

ensure that the APRM instrumentation would

perform that function, with the same degree

of reliability as presently experienced. A

review of the plant operating experience data

at LGS Units 1 and 2 shows that APRM

instrumentation is extremely reliable for a

quarterly surveillance test interval. The

proposed TS changes do not modify plant

configuration, RPS instrumentation setpoints,

or RPS operation. The margin of safety

remains unchanged.

Therefore, the proposed TS changes do not

involve a reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania

19464.

Attorney for licensee: J. W. Durham,

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

Counsel, Philadelphia Electric

Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101

NRC Project Director: John F. Stolz

Philadelphia Electric Company, Docket

Nos. 50-352 and 50-353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of amendment request: August 1,

1995

Description of amendment request:

The proposed amendment would

modify Technical Specifications Section

3/4.9.1, ‘‘Reactor Mode Switch,’’ in

order to provide alternate actions to

allow the continuation of core

alterations in the event certain Reactor

Manual Control System (RMCS) and

refueling interlocks are inoperable,

while preserving the intended function

of the inoperable interlocks.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed Technical Specifications

(TS) changes do not involve a significant
Federal Register / Vol. 60, No. 187 / Wednesday, September 27, 1995 / Notices 49945

increase in the probability or consequences

of an accident previously evaluated.

The refueling and one-rod-out interlocks

impose barriers to preclude an inadvertent

criticality during refueling operations.

Section 7.7.2.15.1 of the LGS Updated Final

Safety Analysis Report (UFSAR) clearly

delineates the functions of the interlocks and

the criteria used in assessing correct refueling

and one-rod-out interlock operation in the

following statement.

In all cases, correct operation of the

refueling interlock prevents either the

operation of loaded refueling equipment over

the core when any control rod is withdrawn,

or the withdrawal of any control rod when

fuel-loaded refueling equipment is operating

over the core. In addition, when the reactor

mode switch is in REFUEL position, only one

rod can be withdrawn, and selection of a

second control rod initiates a rod block.

The proposed TS changes provide

operational flexibility while strictly

conforming to, and preserving, the intended

function of the refueling and one-rod-out

interlocks. The proposed TS changes that

could affect interlock capabilities are

identified below, along with the appropriate

justification to substantiate that the proposed

TS changes will not result in an increase in

the probability or consequences of an

accident previously evaluated.

a.TS Section 3.9.1, ACTION Statement b.

The proposed change to this existing TS

ACTION will add a verification that all

control rods are fully inserted, and then

disabled from being withdrawn as a suitable

alternative to placing the reactor mode

switch in the SHUTDOWN position when the

one-rod-out interlock is not operable. In

addition, the proposed change to this TS

section includes a caveat of non-applicability

for those control rods already removed in

accordance with requirements stipulated in

TS Sections 3.9.10.1 and 3.9.10.2. As

indicated in LGS UFSAR which is described

in the statement above, it is expected that the

refuel and one-rod-out interlocks will permit

the withdrawal of only one (1) control rod at

a time with the reactor mode switch in the

REFUEL position, and no control rods can be

moved when fuel-loaded refueling

equipment is operating over the core. By

verifying all control rods are inserted, then

disabling withdraw capabilities of all rods, as

requested, the most limiting requirements for

control rod motion will be met. The potential

for having more than one (1) control rod out

at a time, or having any control rod not fully

inserted while fuel-loaded refueling

equipment is operating over the core, does

not exist when applying the alternative.

Therefore, the intended functions of the

refuel and one-rod-out interlocks are

operationally preserved. Since TS Sections

3.9.10.1 and 3.9.10.2 have specific

requirements for removing surrounding fuel

prior to control rod blade removal, the

control rods already removed are no longer

required to carry out a safety function in the

defueled cell, and as a result would not apply

for this specific proposed TS change. From

a control rod withdrawal perspective, there is

no functional difference between the

proposed TS change and the existing, and

still remaining, TS ACTION of locking the

reactor mode switch in SHUTDOWN

position.

b. TS Section 3.9.1, ACTION Statement c.

This existing TS ACTION requires that core

alterations be suspended in the event that a

refueling interlock is not operable. The

proposed TS change to this TS ACTION

leaves this requirement in place, but makes

this ACTION specifically applicable to the

refueling platform, and adds three (3) new

additional ACTION alternatives. The

wording for changes to this TS section are

such that implementation of any one of the

three (3) new alternatives can be substituted

for suspending core alterations. The

proposed wording for these three (3) new

alternatives and justification is provided

below.

1) Verify all control rods are fully inserted

and disable withdraw capabilities of all

control rods***.

Since this alternative ensures all control

rods are, and will remain fully inserted, all

required conditions of the associated

refueling and one-rod-out interlocks are met.

The refueling interlock is satisfied since a

fuel-loaded refueling platform operating over

the core would be assured that all control

rods are fully inserted and prevented from

being withdrawn. The one-rod-out interlock

is satisfied since control rod withdrawal is

disabled for all control rods, which is an

even more conservative requirement than the

one-rod-out interlock itself. While operating

in this configuration, there will be no

associated travel or hoist restrictions for the

refueling platform over the core, which is

normal for the current refuel interlock

design. The potential for having any control

rod not fully inserted while a fuel-loaded

refueling platform is operating over the core,

does not exist when applying this proposed

alternative. Therefore, the intended function

of the refueling platform refuel interlocks are

operationally preserved with the

implementation of this proposed alternative,

and there will be no increase in the

probability of occurrence of an accident. This

proposed alternative also maintains an

exclusion (via a reference to the proposed

*** footnote) for control rods removed in

accordance [with] TS Sections 3.9.10.1 and

3.9.10.2. This exclusion does not apply to

inadvertent criticality concerns, as

previously discussed in Item 1.a above.

2) Verify Refuel Platform is not over core

(limit switches not reached) and disable

refuel platform travel over core.

As previously stated above, LGS UFSAR

Section 7.7.2.15.1 stipulates that the

refueling platform position interlocks initiate

a control rod block whenever a fuel-loaded

refueling platform is over the core, and stop

a fuel-loaded refueling platform from moving

over the core if a control rod is already

withdrawn. This specific proposed TS

change satisfies both these requirements by

precluding the possibility of the platform

from being over the core. If a control rod is

being withdrawn, the platform will not be

over the core, and the withdrawal will be in

accordance with the current design. If a

control rod is already withdrawn, disabling

platform travel over the core, before reaching

the over-core limit switches, is performing

the same function as the existing refueling

platform reverse and forward motion blocks.

Therefore, the potential for having any

control rod not fully inserted while a fuelloaded refueling platform is operating over

the core, does not exist when applying this

proposed alternative. The intended refueling

interlock functions are operationally

preserved with the implementation of this

proposed alternative.

3) Verify that no Refuel Platform hoist is

loaded and disable all Refuel Platform hoists

from picking up (grappling) a load.

As previously stated above, UFSAR

Section 7.7.2.15.1 stipulates that blocking

control rod withdrawal with a refueling

platform over the core, and restricting

refueling platform travel from going over the

core with a control rod already withdrawn,

are based on the refueling platform hoist

being fuel-loaded. An unloaded platform

without grappling capabilities poses no

threat to erroneous fuel bundle or control rod

removal, and eliminates the potential for

having any control rod not fully inserted

while a fuel-loaded refueling platform is

operating over the core. Therefore,

implementing this proposed alternative

operationally preserves the intended

interlock functions.

c. TS Section 3.9.1, ACTION Statement d.

The proposed TS change adds this new TS

ACTION section to specify the refueling

interlock requirements for the service

platform, since the applicability of ACTION

Statement c above is being revised to

specifically address refueling interlocks

associated with the refueling platform. The

proposed TS changes for new this TS section

retain the existing requirement to suspend

core alterations if the service platform

associated refueling interlock is not operable,

unless the service platform is not installed

over vessel. The specific proposed TS

changes add two (2) new additional ACTION

alternatives. The proposed wording for these

two (2) new ACTION statements are such

that implementation of any one of the two (2)

new alternatives can be substituted for

suspending core alterations. Not enforcing

operability requirements on the service

platform refueling interlocks when the

service platform is not over the vessel does

not pose an inadvertent criticality concern

since there is no associated hoist to

manipulate fuel bundles or control rods.

These two (2) new alternatives are not

applicable unless the service platform is

installed over the vessel, and are described

below.

1) Verify all control rods are fully inserted

and disable withdraw capabilities of all

control rods***.

This alternative ensures that all control

rods are, and will remain, fully inserted

which meets the required conditions for

proper refueling and one-rod-out interlock

operation. The refueling interlock is satisfied

since a fuel-loaded service platform hoist

operating over-core is assured that all control

rods are fully inserted and prevented from

being withdrawn. The one-rod-out interlock

is satisfied since all control rods are disabled,

an even more conservative requirement than

the one-rod-out interlock itself. While

operating in this configuration, there will be

no associated hoist restrictions for the service
49946 Federal Register / Vol. 60, No. 187 / Wednesday, September 27, 1995 / Notices

platform, which is normal for the current

refuel interlock design. The potential for

having any control rod not fully inserted

while a service platform hoist is fuel-loaded

over the core, does not exist when utilizing

this proposed alternative. Therefore, the

intended function of the service platform

refuel interlocks are operationally preserved

with the implementation of this proposed

alternative. This proposed alternative also

maintains an exclusion (via a reference to the

proposed *** footnote) for control rods

removed in accordance with the

requirements of TS Sections 3.9.10.1 and

3.9.10.2. This exclusion is not applicable to

inadvertent criticality concerns as discussed

in Item 1.a above.

2) Verify Service Platform hoist is not

loaded and disable Service Platform hoist

from picking up (grappling) a load.

As previously described above, UFSAR

Section 7.7.2.15.1 stipulates that blocking

control rod withdrawal with the service

platform over the core is based on the service

platform hoist being fuel-loaded. An

unloaded hoist without grappling capabilities

poses no threat to erroneous fuel bundle or

control rod removal and eliminates the

potential for having any control rod not fully

inserted while a fuel-loaded service platform

is operating over the core. Therefore,

implementing this proposed alternative

operationally preserves the intended

refueling interlock functions.

As discussed in the LGS UFSAR, the use

of the refueling and one-rod-out interlocks

are evaluated from a prevention, not a

mitigation, perspective. A Rod Withdrawal

Error (RWE) transient event during refueling

is concerned with an inadvertent criticality,

and assumes the reactor vessel head is off,

and the plant is shutdown (i.e, Operating

State A). As described in the LGS UFSAR

under Nuclear Safety Operational Analysis

(NSOA) Event 16, it is assumed that the

Reactor Protection System (RPS) terminates

the event should the reactor actually reach

Operating State B (i.e., head off and not shut

down), which is conditional on the reactor

mode switch being in the STARTUP position.

The proposed TS changes only pertain to the

refueling and one-rod-out interlocks. Since

these interlocks act only in a preventive

mode, the consequences of an inadvertent

criticality accident during refueling remain

unchanged.

Since the proposed TS changes are limited

to the one-rod-out and refueling interlocks,

they do not affect the reliability of the

associated equipment. The proposed TS

changes specify alternative actions that can

be taken in the event that an interlock is

inoperable. These alternative actions serve to

ensure the failed interlock function is

preserved, and do not affect the probability

of malfunction of the interlocks.

The one-rod-out and refueling interlocks,

as evaluated in the LGS UFSAR, are designed

to preclude an inadvertent criticality during

refueling operations by placing strict controls

on fuel bundle and control rod

manipulations, using the following methods.

a. Preventing operation of a fuel-loaded

refueling platform or service platform hoist

while over the core if a control rod is already

withdrawn.

b. Preventing a fuel-loaded refueling

platform from traveling over the core if a

control rod is already withdrawn.

c. Preventing any control rod from being

withdrawn if a fuel-loaded refueling platform

or service platform is already operating over

the core.

d. Preventing the withdrawal of more than

one control rod at a time with the reactor

mode switch in the REFUEL position.

The LGS UFSAR indicates that a single

component failure does not cause an

interlock failure and that a single interlock

failure does not cause an accident. The

proposed TS changes provide alternative

actions that can be taken in the event of an

associated component or interlock

malfunction. Implementing the proposed TS

changes will continue to ensure that the

intended interlock functions are maintained

and operationally preserved, as described in

the LGS UFSAR.

Therefore, the proposed TS changes do not

involve an increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed TS changes do not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed TS changes only pertain to

the refueling and one-rod-out interlocks. The

refueling and one-rod-out interlocks impose

barriers to preclude an inadvertent criticality

during refueling operations. The proposed TS

changes provide operational flexibility, while

strictly conforming to, and preserving, the

intended function of the refueling and onerod-out interlocks. There is no other potential

failure mode for these interlocks than has

already been evaluated and described in the

LGS UFSAR. Implementation of these

proposed changes will maintain and

operationally preserve the intended interlock

functions. Therefore, the malfunction of any

associated component or interlock will not

adversely impact the plant and any other

equipment important to safety, directly or

indirectly.

Therefore, the proposed TS changes do not

create the possibility of a new or different

kind of accident from any previously

evaluated.

3. The proposed TS changes do not involve

a significant reduction in a margin of safety.

The proposed TS changes only affect the

TS associated with the one-rod-out and

refueling interlocks. The associated TS Bases

Section 3/4.9, ‘‘Refueling Operations,’’ states

that the one-rod-out and refueling interlocks

maintain conditions during refueling

activities that reinforce refueling procedures

and reduce the potential for the probability

of occurrence of each of the following

conditions:

a. Inadvertent criticality,

b. Damage to reactor internals or fuel

assemblies, and

c. Exposure of personnel to excessive

radioactivity.

The proposed TS changes do not adversely

affect the one-rod-out or refueling interlocks.

The associated interlocks will continue to

perform the refueling functions required to

maintain the present margin of safety. The

proposed TS changes only contain alternative

actions that can be taken in the event an

interlock is inoperable. These proposed

alternative actions ensure that the intent of

the interlocks is preserved, and that there is

no reduction in the ability of the interlocks

to maintain adequate refueling conditions.

The proposed TS changes will preserve the

intended interlock functions, and maintain

the existing level of protection against

refueling errors that could lead to an

inadvertent criticality, damage to reactor

internals or fuel assemblies, or excessive

personnel radiation exposure. The one-rodout and refueling interlocks will continue to

function with their present degree of

reliability. The proposed TS changes will

continue to maintain strict controls on fuel

bundle and control rod manipulations to

avoid inadvertent criticality. The proposed

TS changes provide the same level of

assurance regar[d]ing the manipulation of

control rods during refueling operations as

that currently described in the LGS UFSAR,

and as discussed below.

a. Preventing operation of a fuel-loaded

refueling platform or service platform hoist

while over the core if a control rod is already

withdrawn.

b. Preventing a fuel-loaded refueling

platform from traveling over the core if a

control rod is already withdrawn.

c. Preventing any control rod from being

withdrawn if fuel-loaded refueling platform

or service platform is already operating over

the core.

d. Preventing the withdrawal of more than

one control rod at a time with the reactor

mode switch in the REFUEL position.

Therefore, the proposed TS changes do not

involve a reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

eview, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania

19464.

Attorney for licensee: J. W. Durham,

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

Counsel, Philadelphia Electric

Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101

NRC Project Director: John F. Stolz

South Carolina Electric & Gas Company

(SCE&G), South Carolina Public Service

Authority, Docket No. 50-395, Virgil C.

Summer Nuclear Station, Unit No. 1,

Fairfield County, South Carolina

Date of amendment request: February

21, 1995, as revised on August 31, 1995

Description of amendment request:

The proposed amendment would revise

the Technical Specifications (TS) to

reflect changes to 10 CFR Part 20

(including Appendix B, Table 2

concentrations) and provide additional

administrative corrections.
Federal Register / Vol. 60, No. 187 / Wednesday, September 27, 1995 / Notices 49947

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The probability or consequences of an

accident previously evaluated does not

involve a significant increase.

The proposed TS changes showing the

relocation of the old 10 CFR 20.106

requirements to the new 10 CFR 20.1302, the

old 10 CFR 20.203(c)(2) requirements to the

new 10 CFR 20.1601(a), and the old 10 CFR

20.407 requirements to the new 10 CFR

20.2206(b) will not involve a significant

increase in the probability or consequences

of an accident previously evaluated because

there will be no change in the types and

amounts of effluents that will be released,

nor will there be an increase in individual or

cumulative occupational radiation exposures.

The proposed revision to the liquid and

gaseous release rate limits will not involve a

significant increase in the probability or

consequences of an accident previously

evaluated because there will be no change in

the types and amounts of effluents that will

be released, nor will there be an increase in

individual or cumulative occupational

radiation exposures. This is only a change to

the method of (algorithm) determining

release rate limits and will not change net

limits or change the more restrictive 10 CFR

50 Appendix I dose limits.

The proposed revision to the radioactive

material quantity in the settling pond and its

associated TS Bases will not involve a

significant increase in the probability or

consequences of an accident previously

evaluated because there will be no change in

the types of effluents that will be released,

nor will there be an increase in individual or

cumulative occupational radiation exposures.

This is only a change to the quantity of

radioactive material in the settling pond and

will conservatively lower net limits.

The proposed revision to the TS bases for

the liquid holdup tank activity limit will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated because there will be no

change in the types and amounts of effluents

that will be released, nor will there be an

increase in individual or cumulative

occupational radiation exposures. The curie

limit is not affected, therefore, the change

does not represent a decrease in the level of

control previously evaluated.

The proposed revision to the distance at

which dose rates are measured from the

radiation source or surface will not involve

a significant increase in the probability or

consequences of an accident previously

evaluated because there will be no increase

in the individual or cumulative occupational

radiation exposures. The change in distance

is conservative in its effect on worker

protection and is in conformance with 10

CFR 20.1601 requirements.

2. The possibility of a new or different kind

of accident from any previously evaluated is

not created.

The proposed TS changes showing the

relocation of the old 10 CFR 20.106

requirements to the new 10 CFR 20.1302,

relocation of the old 10 CFR 20.203(c)(2)

requirements to the new 10 CFR 20.1601(a),

and relocation of the old 10 CFR 20.407

requirements to the new 10 CFR 20.2206(b)

will not create the possibility of a new or

different kind of accident from any

previously evaluated because the revisions

are administrative and will not change the

types and amounts of effluents that will be

released.

The proposed revision to the liquid and

gaseous release rate limits will not create the

possibility of a new or different kind of

accident from any previously evaluated

because the revision is administrative and

will not change the types and amounts of

effluents that will be released.

The proposed revision to the quantity of

radioactive material in the settling pond and

its associated TS Bases will not create the

possibility of a new or different kind of

accident from any previously evaluated

because the revision will not change the

types of effluents that will be released. This

is only a change to the quantity of radioactive

material in the settling pond and will

conservatively lower net limits.

The proposed revision to the TS bases for

the liquid holdup tank activity limit will not

create the possibility of a new or different

kind of accident from any previously

evaluated because the revision is

administrative and will not change the types

and amounts of effluents that will be

released.

Implementation of the more conservative

distance at which dose rates are measured

will not create the possibility of a new or

different kind of accident from any

previously evaluated.

3. A significant reduction in a margin of

safety is not involved.

The proposed revisions due to the location

of requirements will not reduce a margin of

safety because they are administrative in

nature. No equipment or procedural changes

are postulated. There is no impact on any

margin of safety.

The proposed revision to liquid and

gaseous release rate limits will not reduce a

margin of safety because it is administrative

in nature. These revisions preserve the

existing level of effluent control. No changes

to the more restrictive 10 CFR 50 Appendix

I dose limits are made. There are no

equipment or operational procedure changes,

therefore, no accidents of any kind will be

created by this change.

The proposed revision to the quantity of

radioactive material in the settling pond and

its associated TS Bases will not reduce a

margin of safety because it is conservative in

nature and preserves the existing level of

effluent control. There are no equipment or

operational procedure changes required,

therefore, no accidents of any kind will be

created by this change.

The proposed revision to the TS bases for

the liquid holdup tank activity limit will not

reduce a margin of safety because it is

administrative in nature and preserve[s] the

existing level of effluent control. No

equipment or procedural changes are

postulated. There is no impact on any margin

of safety.

The change in distance for a High

Radiation Area classification from 18 in.(45

cm) to (30 cm)12 in. from the radiation

source or surface will not reduce the margin

of safety because this change will reduce the

worker’s stay time in the area and therefore

minimize exposure.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Fairfield County Library, 300

Washington Street, Winnsboro, SC

29180

Attorney for licensee: Randolph R.

Mahan, South Carolina Electric & Gas

Company, Post Office Box 764,

Columbia, South Carolina 29218

NRC Project Director: Frederick J.

Hebdon

Southern California Edison Company,

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

San Onofre Nuclear Generating Station,

Unit Nos. 2 and 3, San Diego County,

California

Date of amendment requests: July 19,

1995

Description of amendment requests:

The licensee proposes to revise

technical specifications (TSs) to (1)

support modifications to the

containment area radiation monitors, to

either upgrade or replace existing

equipment with state-of-the-art

equipment, (2) relocate the setpoint and

allowable values for the control room

airborne radiation monitors to be

consistent with the containment

airborne radiation monitors TS, and (3)

make minor editorial changes to the TS

pertaining to the control room airborne

radiation monitors and the containment

airborne radiation monitors. The

proposed changes affect TS Tables 3.3-

3, 3.3-4, 3.3-5, 3.3-6, 4.3-2, and 4.3-3.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

Control Room Airborne Radiation Monitors

The proposed change would permit

relocation of the setpoint and allowable

values for the monitors from the Technical

Specifications (TSs) to the administrative

control procedures. This change is consistent

with the existing Containment Airborne

Radiation Monitor TSs. This change will not

prevent the radiation monitors from
49948 Federal Register / Vol. 60, No. 187 / Wednesday, September 27, 1995 / Notices

performing their intended function following

a design basis accident. Therefore, operation

of the facility in accordance with this change

will not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Containment Area Radiation Monitors

The proposed change deletes the existing

Containment Area Radiation Monitors RE7856-1 and RE-7857-2 and their Engineered

Safety Feature Actuation System (ESFAS)

function to initiate containment purge

isolation on high radiation in containment.

The deletion of this ESFAS function does not

create a precursor to any analyzed accident

since these monitors are for accident

mitigation only.

Currently, no release of radioactivity is

assumed during a Fuel Handling Accident in

containment since the Containment Area

Radiation Monitors detect and isolate

containment purge prior to release. The

proposed deletion will cause some release

prior to detection and isolation of purge by

the remaining noble gas containment

monitors. The consequences of a Fuel

Handling Accident inside containment were

previously re-evaluated, assuming no

containment purge isolation, to resolve

inconsistencies in the original analysis

assumptions and methodology. The results of

the calculation indicated off-site doses well

within the limits of 10 CFR 100 and Control

Room doses that met the limits of 10 CFR 50

Appendix A General Design Criterion 19.

Containment purge isolation on high gaseous

activity during a Fuel Handling Accident will

still be available with this proposed change

but is not required for the dose consequences

to remain within the dose criteria. Therefore,

the proposed change will not significantly

increase the consequences of a Fuel Handling

Accident inside containment.

The Loss of Coolant Accident (LOCA)

function of the Containment Purge Isolation

System (CPIS) signal will be essentially

unaffected by this proposed change.

Currently, containment purge isolation

(containment minipurge) on high radiation

signals is a diverse signal with Safety

Injection Actuation System (SIAS) and

Containment Isolation Actuation System

(CIAS). In a LOCA event, containment purge

isolation is expected to occur on either SIAS

or CIAS prior to a CPIS signal on high

radiation in containment. While this

proposed change reduces the diversity of

radiation monitoring inputs, the diversity of

parameters measured (pressure and

radiation) is still preserved. Therefore, the

proposed change will not increase the

consequences of a LOCA.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

Control Room Airborne Radiation Monitors

Relocating the monitor setpoint and

allowable values from the TSs to the

administrative procedures would not alter

the design and operational interface between

the Control Room Isolation System

instrumentation and existing plant

equipment. As such, the monitors would

continue to operate and perform their

intended safety function to isolate the control

room following a design basis accident as

before. Therefore, operation of the facility in

accordance with this proposed change will

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

Containment Area Radiation Monitors

The deletion of the Containment Area

Radiation Monitors will not alter the

operation of CPIS. The remaining interface

between CPIS and existing plant equipment

will continue to perform their intended

safety function to isolate containment purge

by closing the containment purge valves.

This function will continue to be performed

by Containment Airborne Radiation Monitors

2(3) RT-7804-1 and 2(3) RT-7807-2.

Therefore, operation of the facility in

accordance with this proposed change will

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.

Control Room Airborne Radiation Monitors

Relocating the monitor setpoint and

allowable values to the administrative

procedures would not alter the existing

margin of safety. The relocation would only

relinquish control of the setpoint and

allowable values from the TSs to qualityaffecting (changes will require a 10 CFR

50.59 evaluation) procedures. Therefore,

operation of the facility will not involve a

significant reduction in a margin of safety.

Containment Area Radiation Monitors

The proposed change does not affect the

margin of safety in Modes 1 through 4 since

the diversity of the parameters measured is

maintained for minipurge isolation. Either

SIAS, CIAS, CPIS, or manual operation will

close the containment mini purge valves. The

main purge is sealed closed during Modes 1

through 4 with the purge valves closed and

deactivated.

The diversity of the parameters measured

is not maintained for the containment main

purge isolation. The main purge is only

applicable during Modes 5 and 6 and main

purge isolation is initiated only by either

CPIS or manual operation. This proposed

change along with the previously submitted

PCN-299 reduces the diversity of radiation

sensing in containment for CPIS generation

from four types (gaseous, iodine, particulate,

and gamma) to one type (gaseous activity).

Since the consequences of a Fuel Handling

Accident inside containment without purge

isolation have been calculated to be well

within 10 CFR 100 dose limits, the loss of

diversity for this accident does not result in

a significant reduction in a margin of safety.

Therefore, this proposed change will not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 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: Main Library, University of

California, P. O. Box 19557, Irvine,

California 92713

Attorney for licensee: T. E. Oubre,

Esquire, Southern California Edison

Company, P. O. Box 800, Rosemead,

California 91770

NRC Project Director: William H.

Bateman

Tennessee Valley Authority, Docket No.

50-328, Sequoyah Nuclear Plant, Unit 2,

Hamilton County, Tennessee

Date of amendment request: May 19,

1995; revised September 11, 1995 (TS

95-13)

Description of amendment request:

The proposed change would revise

License Condition 2.C.(17) to extend the

required surveillance interval to May

18, 1996, for Surveillance Requirement

4.3.2.1.3. The proposed change would

extend the Engineered Safety Features

Response Time instrument tests

required at 36-month intervals shown in

Table 3.3-3 associated with safety

injection, feedwater isolation,

containment isolation Phase A,

auxiliary feedwater pump, essential raw

cooling water system, emergency gas

treatment system, containment spray,

containment isolation Phase B, turbine

trip, 6.9-kilovolt shutdown boarddegraded voltage or loss of voltage, and

automatic switchover to containment

sump actuations. The proposed

extension will limit the interval past the

allowable extension provided by TS

4.0.2 to 5 months.

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:

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 change is temporary and

allows a one-time extension of Surveillance

Requirement 4.3.2.1.3 for Cycle 7 to allow

surveillance testing to coincide with the

seventh refueling outage. The proposed

surveillance interval extension will not cause

a significant reduction in system reliability

nor affect the ability of the systems to

perform their design function. Current

monitoring of plant conditions and

continuation of the surveillance testing

required during normal plant operation will

continue to be performed to ensure

conformance with TS operability

requirements. Therefore, this change does not

involve a significant increase in the
Federal Register / Vol. 60, No. 187 / Wednesday, September 27, 1995 / Notices 49949

probability or consequences of an accident

previously evaluated.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

Extending the surveillance interval for the

performance of specific testing will not create

the possibility of a new or different kind of

accidents. 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 previously

evaluated.

3. Involve a significant reduction in a

margin of safety.

Surveillance interval extensions will not

impact any plant safety analyses since the

assumptions used will remain unchanged.

The safety limits assumed in the accident

analyses and the design function of the

equipment required to mitigate the

consequences of any postulated accidents

will not be changed since only the

surveillance test interval is being extended.

Historical performance generally indicates a

high degree of reliability, and surveillance

testing performed during normal plant

operation will continue to be performed to

verify proper performance. Therefore, the

plant will be maintained within the analyzed

limits, and the proposed extension will not

significantly reduce the margin of saety.

The NRC has reviewed the licensee’s

analysis and, based on thisreview, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: 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

Virginia Electric and Power Company,

Docket Nos. 50-338 and 50-339, North

Anna Power Station, Units No. 1 and

No. 2, Louisa County, Virginia

Date of amendment request:

September 1, 1995

Description of amendment request:

The proposed changes to the Technical

Specifications (TS) for the North Anna

Power Station, Units 1&2 (NA-1&2)

would allow a single outage of up to 14

days for each emergency diesel

generator (EDG) once every 18 months.

The purpose of the outage is the

performance of a preventive

maintenance inspection, appropriate for

diesels used for this class of standby

service, which requires disassembly of

the EDG. Currently this maintenance

inspection is performed during refueling

outages. The proposed changes would

permit this maintenance inspection to

be performed during Modes 1 to 4 in

addition to the current allowance during

Modes 5 or 6.

A probabilistic safety analysis (PSA)

has been performed which demonstrates

that a fourteen (14) day maintenance

inspection outage, once every eighteen

(18) months for each EDG, results in no

significant change in core damage

frequency assuming adequate

compensatory measures are in place.

The compensatory measures include

requirements that the other EDGs, offsite power supply, and the alternate

A.C. diesel (AAC DG) be operable

during the preventive maintenance

inspection outage.

The effect of the proposed change has

been calculated to be an increase in core

damage frequency of approximately 1E6 per year, which is not considered to

be a significant change (i.e., an

acceptable change in risk, or a non-risk

significant change) from the baseline

core damage frequency of 4.1E-5.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Specifically, operation of North Anna

Power Station in accordance with the

[proposed] Technical Specifications changes

will not:

a. involve a significant increase in the

probability or consequences of an accident

previously evaluated. The probabilistic safety

analysis (PSA) demonstrates that the increase

in core damage frequency due to performing

the EDG maintenance inspection over a

fourteen day period once every 18 months is

not significant as long as the AAC DG is

operable to act as a source of emergency

power to replace the EDG. The period of time

during which the EDG is unavailable is short

enough to limit the impact of using the

manually operated AAC DG as a replacement

for the automatically operated EDG.

b. create the possibility of a new or

different kind of accident from any accident

previously evaluated. The proposed

Technical Specifications changes only

modify the operability of an EDG for a

limited and defined period of time. The

UFSAR [Updated Final Safety Analysis

Report] accidents are analyzed assuming that

the EDG is the worst single failure. This

assumption is more severe than the proposed

Technical Specifications changes which

replaces the EDG with the AAC DG.

Similarly, the PSA performed to evaluate the

proposed Technical Specifications changes

considered all of the initiating events defined

for the PSA performed for the Individual

Plant Examination. No new initiators were

defined as a result of a review of the PSA

model. Therefore, it is concluded that no new

or different kind of accident from any

previously evaluated has been created.

c. The proposed Technical Specifications

changes do not result in a reduction in

margin of safety as defined in the basis for

any Technical Specifications. The PSA was

performed to evaluate the concept of a one

time outage. The results of the analyses show

no significant change in the core damage

frequency. As described above the proposed

Technical Specifications changes only

modify the operability of an EDG for a

limited and defined period of time. Thus,

operation with slightly increased EDG

unavailability due to maintenance, and the

AAC DG operable is acceptable.

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: The Alderman Library, Special

Collections Department, University of

Virginia, Charlottesville, Virginia 22903-

2498.

Attorney for licensee: Michael W.

Maupin, Esq., Hunton and Williams,

Riverfront Plaza, East Tower, 951 E.

Byrd Street, Richmond, Virginia 23219.

NRC Project Director: David B.

Matthews

Wolf Creek Nuclear Operating

Corporation, Docket No. 50-482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request: March

24, 1995, as supplemented by letter

dated August 16, 1995.

Description of amendment request:

This request proposes to revise

Technical Specification 1.7,

‘‘Containment Integrity,’’ Technical

Specification 3/4.6.1, ‘‘Containment

Integrity,’’ Technical Specification 3/

4.6.3, ‘‘Containment Isolation Valves,’’

and their associated Bases. These

proposed changes will remove

Technical Specification Table 3.6-1

‘‘Containment Isolation Valves,’’ to Wolf

Creek Generating Station (WCGS)

procedures. This proposed change is in

accordance with the guidance provided

in Generic Letter 91-08, ‘‘Removal of

Component Lists from Technical

Specifications,’’ dated May 6, 1991. In

addition, this request proposes to add a

footnote to Technical Specification 3.6.3

extending the allowed outage time for

the component cooling water (CCW)

system reactor coolant pump seal water

supply and return valves. This

determination supersedes the staff’s

proposed no significant hazards

consideration determination evaluation

for the requested changes that was

published on April 26, 1995 (60 FR

20532).
49950 Federal Register / Vol. 60, No. 187 / Wednesday, September 27, 1995 / Notices

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes simplify the

technical specifications, meet the regulatory

requirements for control of containment

isolation, and are consistent with the

guidelines of GL 91-08. The procedural

details of Technical Specification Table 3.6-

1 have not been changed, but only relocated

to a different controlling document. The

proposed changes are administrative in

nature, should result in improved

administrative practices, and do not affect

plant operations. The addition of the footnote

to allow up to 12 hours for valve testing the

CCW MOVs [motor-operated valves] does not

affect the severity of any accident previously

evaluated. This footnote does not impact

plant safety since the second isolation device

in the affected penetrations would still be

available to provide isolation between the

RCS and the outside atmosphere.

The probability of occurrence of a

previously evaluated accident is not

increased because this change does not

introduce any new potential accident

initiating conditions. The consequences of an

accident previously evaluated is not

increased because the ability of containment

to restrict the release of any fission product

radioactivity to the environment will not be

degraded by this change.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes are administrative

in nature, do not result in physical

alterations or changes to the operation of the

plant, and cause no change in the method by

which any safety-related system performs its

function. The addition of the footnote to

allow up to 12 hours for valve testing the

CCW MOVs does not affect the severity of

any accident previously evaluated. The

additional time provides assurance that the

inoperable valve is in proper working order

prior to returning it to OPERABLE status.

Therefore, this proposed change will not

create the possibility of a new or different

kind of accident from any previously

evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The administrative change to relocate

Technical Specification Table 3.6-1 to

appropriate plant procedures does not alter

the basic regulatory requirements for

containment isolation and will not adversely

affect containment isolation capability for

credible accident scenarios. Adequate control

of the content of the table is assured by

existing plant procedures. The additional

footnote to extend the allowed outage time to

12 hours for the CCW MOVs does not affect

containment isolation capability since the

second isolation device in the affected

penetrations would still be available to

provide isolation between the RCS and the

outside atmosphere, and to ensure that a

release of radioactive material to the

environment following an accident will not

exceed the assumptions used in the LOCA

Analyses.

The proposed relocation of the Technical

Specification Table 3.6-1 does not alter the

requirements for containment isolation valve

operability currently in the technical

specifications. The LCO and Surveillance

Requirements would be retained in the

revised technical specifications. Therefore,

the proposed change will not affect the

meaning, application, and function of the

current technical specification requirements

for the valves in Table 3.6-1.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

locations: Emporia State University,

William Allen White Library, 1200

Commercial Street, Emporia, Kansas

66801 and Washburn University School

of Law Library, Topeka, Kansas 66621

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts and Trowbridge,

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

20037

NRC Project Director: William H.

Bateman

Wolf Creek Nuclear Operating

Corporation, Docket No. 50-482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request: August

22, 1995

Description of amendment request:

The proposed license amendment

request would relocate Technical

Specification Tables 3.3-2, ‘‘Reactor

Trip System Instrumentation Response

Times,’’ and 3.3-5, ‘‘Engineered Safety

Features Response Times,’’ and

applicable Bases discussions, to

Updated Safety Analysis Report (USAR)

Chapter 16. The NRC has already

implemented this line-item technical

specification improvement in the new

Standard Technical Specifications

(NUREG-1431 for Westinghouse plants).

This amendment request follows the

guidance provided by the NRC in

Generic Letter 93-08, ‘‘Relocation of

Technical Specification Tables of

Instrument Response Time Limits,’’ for

relocating instrument response time

tables.

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 request does not

change any Reactor Trip System (RTS) or

Engineered Safety Features Actuation System

(ESFAS) instrument response times or

surveillance intervals currently prescribed in

Technical Specification Tables 3.3-2 and 3.3-

5. The RTS and ESFAS will continue to

function in a manner consistent with the

assumptions in the Updated Safety Analysis

Report Chapter 15 accident analyses and the

plant design basis. Therefore, overall

protection system performance will remain

within the bounds of the accident analyses

documented in USAR Chapter 15. As such,

there will be no degradation in system

performance, nor will there be an increase in

the number of challenges to equipment

assumed to function during an accident

situation.

The proposed technical specification

revision does not involve any hardware

changes or changes to any instrumentation

setpoints, system operating parameters, or

system accident mitigation capabilities, nor

do the changes affect the probability of any

event initiators. Thus, the proposed change

will not result in an increase in the

consequences of or the probability of

occurrence of any accident or safety-related

equipment malfunction.

2. The proposed change does not 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 this proposed

amendment request, nor are there any

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

function. The normal manner of plant

operation is not affected by this proposed

change.

No new accident scenarios, transient

precursors, failure mechanisms, or limiting

single failures are introduced as a result of

the proposed 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 by

the proposed changes.

3. The proposed change does not involve

a significant reduction in a margin of safety.

No response times will be changed by this

amendment request. The proposed request

only changes the document where the

response times will be listed. This proposed

amendment request will not affect the

manner in which safety limits or limiting

safety system settings are determined, nor

will there by [be] any effect on plant systems

necessary to assure the accomplishment of

protection functions. The proposed change

will not impact any margin of safety defined

in the basis for any Technical Specification.

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. 187 / Wednesday, September 27, 1995 / Notices 49951

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

locations: Emporia State University,

William Allen White Library, 1200

Commercial Street, Emporia, Kansas

66801 and Washburn University School

of Law Library, Topeka, Kansas 66621

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts and Trowbridge,

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

20037

NRC Project Director: William H.

Bateman

Previously Published Notices Of

Nonsideration Of Issuance Of

Amendments To Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

and opportunity for a hearing

The following notices were previously

published as separate individual

notices. The notice content was the

same as above. They were published as

individual notices either because time

did not allow the Commission to wait

for this biweekly notice or because the

action involved exigent circumstances.

They are repeated here because the

biweekly notice lists all amendments

issued or proposed to be issued

involving no significant hazards

consideration.

For details, see the individual notice

in the Federal Register on the day and

page cited. This notice does not extend

the notice period of the original notice.

Indiana Michigan Power Company,

Docket No. 50-315, Donald C. Cook,

Nuclear Plant, Unit No. 1, Berrien

County, Michigan

Date of application for amendment:

August 4, 1995 (AEP:NRC:1129E)

Description of amendment request:

The proposed amendment would

modify Technical Specification 4.4.5.4

and 4.4.5.5, on steam generators, to

allow for repair of hybrid expansion

joint sleeves under redefined repair

boundary limits.

Date of publication of individual

notice in the Federal Register: August

14, 1995 (60 FR 41904)

Expiration date of individual notice:

For comments: August 29, 1995; hearing

requests: September 13, 1995

Local Public Document Room

location: Maud Preston Palenske

Memorial Library, 500 Market Street, St.

Joseph, Michigan 49085

Tennessee Valley Authority, Docket

Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of application for amendments:

August 21, 1995Description of

amendments request: Amend technical

specification 3.7.5.c to allow an increase

in the average essential raw cooling

water supply header temperature from

84.5°F to 87°F until September 30, 1995.

Date of publication of individual

notice in the Federal Register: August

28, 1995 (60 FR 44517)

Expiration date of individual notice:

September 12, 1995

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402

Notice Of Issuance Of Amendments To

Facility Operating Licenses

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application

complies with the standards and

requirements of the Atomic Energy Act

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

Commission’s rules and regulations.

The Commission has made appropriate

findings as required by the Act and the

Commission’s rules and regulations in

10 CFR Chapter I, which are set forth in

the license amendment.

Notice of Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for A Hearing in

connection with these actions was

published in the Federal Register as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the applications for

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

the Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, the Gelman Building, 2120 L

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

local public document rooms for the

particular facilities involved.

Arizona Public Service Company, et al.,

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

and STN 50-530, Palo Verde Nuclear

Generating Station, Units 1, 2, and 3,

Maricopa County, Arizona

Date of application for amendments:

June 17, 1994

Brief description of amendments:

These amendments revise the

surveillance requirement and Bases

section of TS 4.7.1.6 to increase the

minimum nitrogen accumulator

pressure for the atmospheric dump

valves (ADVs).

Date of issuance: September 6, 1995

Effective date: September 6, 1995

Amendment Nos.: Unit 1 -

Amendment No. 99; Unit 2 -

Amendment No. 87; Unit 3 -

Amendment No. 70

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

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 17, 1994 (59 FR 42333)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated September 6, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona

85004

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:

March 31, 1995

Brief description of amendments: The

amendments clarify the shutdown

margin definition, change the shutdown

margin applicability and surveillance

requirements to comply with the safety

analysis assumptions for subcritical

inadvertent control element assembly

withdrawal (UFSAR Section 15.4, and

expand the applicability for core

protection calculator (CPC) operability.

In addition, the amendments add a

reference to the Core Operating Limits

Report for the MODE 6 refueling boron

concentration limit. The amendments

also change the power calibration

requirements for the linear power level,

the CPC delta T power, and CPC nuclear

power signals to allow more

conservative settings than previously

required.

Date of issuance: September 1, 1995
49952 Federal Register / Vol. 60, No. 187 / Wednesday, September 27, 1995 / Notices

Effective date: September 1, 1995

Amendment Nos.: Unit 1 -

Amendment No. 98; Unit 2 -

Amendment No. 86; Unit - Amendment

No. 69

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

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 6, 1995 (60 FR 29871)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated September 1, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Phoenix Public Library, 1221

N. Central, Phoenix, Arizona 85004

Carolina Power & Light Company,

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

SteamElectric Plant, Unit No. 2,

Darlington County, South Carolina

Date of application for amendment:

June 3, 1995, as supplemented on

August 7, 1995. The supplemental

submittal did not expand the scope of

the original Federal Register notice or

change the no significant hazards

determination.

Brief description of amendment: The

amendment clarifies the definition of

operability of the charging pumps by

adding a footnote to TS Section 3.2.2.a

that states that the connectibility of the

emergency power sources is not

required for charging pump operability.

The bases statement for TS 3.2.2 is also

changed for clarification.

Date of issuance: September 5, 1995

Effective date: September 5, 1995

Amendment No.: 166

Facility Operating License No. DPR23. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: July 5, 1995 (60 FR 35063) The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated September 5, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Hartsville Memorial Library,

147 West College Avenue, Hartsville,

South Carolina 29550

Commonwealth Edison Company,

Docket Nos. STN 50-454 and STN 50-

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

Ogle County, IllinoisDocket Nos. STN

50-456 and STN 50-457, Braidwood

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

Illinois

Date of application for amendments:

February 21, 1995

Brief description of amendments: The

amendments revise the technical

specifications to permit replacement of

the reactor coolant resistance

temperature detector (RTD) bypass

manifold system with fast response

RTDs mounted in thermowells welded

directly into the reactor coolant system

piping.

Date of issuance: September 5, 1995

Effective date: September 5, 1995

Amendment Nos.: 74 and 66

Facility Operating License Nos. NPF37, NPF-66, NPF-72 and NPF-77: The

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 5, 1995 (60 FR 35063) The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated September 5, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: For Byron, the Byron Public

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

Box 434, Byron, Illinois 61010; for

Braidwood, the Wilmington Public

Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481

Commonwealth Edison Company,

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

Cities Nuclear Power Station, Units 1

and 2, Rock Island County, Illinois

Date of application for amendments:

February 23, 1995

Brief description of amendments: The

amendment revises the Quad Cities

Nuclear Power Station, Units 1 and 2,

operating licenses to reflect the transfer

of the Iowa-Illinois Gas and Electric

Company’s 25 percent undivided

ownership to MidAmerican Energy

Company.

Date of issuance: September 11, 1995

Effective date: As of the

consummation of the merger between

Iowa-Illinois Gas and Electric Company,

Midwest Power Systems, Inc.,

MidAmerican Energy Company, and

Midwest Resources, Inc.

Amendment Nos.: 159 and 155

Facility Operating License Nos. DPR29 and DPR-30: The amendments

revised the operating licenses.

Date of initial notice in Federal

Register: July 5, 1995 (60 FR 35054) The

Commission’s related evaluation of the

amendments is contained in an

Environmental Assessment and Finding

of No Significant Impact dated March

21, 1995, and in a Safety Evaluation

dated September 11, 1995.No significant

hazards consideration comments

received: No

Local Public Document Room

location: Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois 61021

Commonwealth Edison Company,

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

Nuclear Power Station Units 1 and 2,

Lake County, Illinois

Date of application for amendments:

May 31, 1995

Brief description of amendments: The

amendments authorize an alternative

repair criteria for defects found in the

tube expansion region within the

tubesheet.

Date of issuance: September 11, 1995

Effective date: September 11, 1995

Amendment Nos.: 168 and 155

Facility Operating License Nos. DPR39 and DPR-48: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: July 5, 1995 (60 FR 35067) The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated August 11, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois

60085

Duke Power Company, et al., Docket

Nos. 50-413 and 50-414, Catawba

Nuclear Station, Units 1 and 2, York

County, South Carolina

Date of application for amendments:

September 19, 1994, as supplemented

April 26 and June 19, 1995

Brief description of amendments:

These changes to the Technical

Specifications (TS) 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 TS 3/4.9.12 and 3/4.9.13 are

added to explain the basis for each LCO,

Action Statement and Surveillance
Federal Register / Vol. 60, No. 187 / Wednesday, September 27, 1995 / Notices 49953

Requirement covered by the subject TS;

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

reflect limits for criticality analysis for

fuel storage; and (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.

Date of issuance: August 31, 1995

Effective date: As of the date of

issuance to be implemented within 30

days

Amendment Nos.: 134 and 128

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

the Technical Specifications.

Date of initial notice in Federal

Register: May 23, 1995 (60 FR 27338)

The June 19, 1995, letter provided

clarifying information that did not

change the scope of the September 19,

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 August 31, 1995, and

Environmental Assessment dated

August 15, 1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: York County Library, 138 East

Black Street, Rock Hill, South Carolina

29730

Duquesne Light Company, et al., Docket

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

Power Station, Unit Nos. 1 and 2,

Shippingport, Pennsylvania

Date of application for amendments:

August 31, 1994, as supplemented May

18, 1995.

Brief description of amendments:

These amendments delete Beaver Valley

Power Station, Unit 2, License

Conditions 2.C.(3), 2.C.(5), 2.C.(7),

2.C.(8), 2.C.(9) and 2.C.(10) to reflect

completion of activities required by

these license conditions and make the

following revisions to the Beaver Valley

Power Station, Units 1 and 2, TSs:

1. Eliminate references to specific

frequencies for each of the TS required

audits (TS 6.2.2.8).

2. Eliminate references to reviews and

audits of the Emergency plan and

Security Plant (TSs 6.5.2.8 and 6.8.1).

3. Include Offsite Dose Calculation

Manual and Process Control Program

and associated implementing

procedures into the list of required

audits (TS 6.5.2.8).

4. Editorial changes which were

necessitated by a reorganization (TS

6.2.1, 6.2.3.1, 6.2.3.4, 6.5.2.2, 6.5.2.8,

6.5.2.9, and 6.5.2.10).

5. Eliminate reference to Appendix A

of 10 CFR Part 55 (TS 6.4.1).

6. Separate the Inservice Inspection

(ISI) and Inservice Testing (IST)

Programs surveillance requirements and

remove the requirement that relief

requests be granted before they are

implemented for both IST and ISI (TS

4.0.5).

The May 18, 1995, letter requested

withdrawal of the proposed changes to

TS 6.5.2.8 dealing with audits of the

Beaver Valley Power Station, Units 1

and 2, fire protection program and

withdrawal of a proposed 25-percent

grace period for all audit frequencies

(Item 6 in August 31, 1994 application).

Date of issuance: August 31, 1995

Effective date: Units 1 and 2, as of the

date of issuance and shall be

implemented within 60 days.

Amendment Nos.: 191 and 74

Facility Operating License Nos. DPR66 and NPF-73: Amendments revised

the Units 1 and 2 Technical

Specifications, and the Unit 2 License.

Date of initial notice in Federal

Register: (59 FR 65812) December 21,

1994. The May 18, 1995, letter did not

change the original no significant

hazards consideration determination or

expand the scope of the December 21,

1994, Federal Register notice. The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated August 31, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa,

Pennsylvania 15001

Duquesne Light Company, et al., Docket

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

Power Station, Unit Nos. 1 and 2,

Shippingport, Pennsylvania

Date of application for amendments:

October 11, 1994, as supplemented June

23, 1995, and August 24, 1995

Brief description of amendments:

These amendments revise Beaver Valley

Power Station Technical Specifications

(TSs) 1.18, ‘‘Quadrant Power Tilt Ratio,’’

3/4.2.4, ‘‘Quadrant Power Tilt Ratio,’’

the table Notation of TS Table 3.3-1,

‘‘Reactor Trip System Instrumentation,’’

and associated Bases to incorporate the

guidance provided in the NRC’s

Improved Standard Technical

Specifications (NUREG-1431, Revision

1) to these TSs. The amendments clarify

the requirements of the subject TSs with

regard to the use of excore power range

neutron flux detectors to monitor

quadrant power tilt ratio when an

excore power range neutron flux

instrument is inoperable. The changes

also make several minor editorial

changes in the subject TSs.

Date of issuance: September 15, 1995

Effective date: As of date of issuance,

to be implemented within 60 days.

Amendment Nos.: 192 and 75

Facility Operating License Nos. DPR66 and NPF-73: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: August 2, 1995 (60 FR 39436)

The August 24, 1995, letter provided

typed final TS pages, with minor

editorial changes, for issuance of these

amendments. The August 24, 1995,

letter did not change the initial

proposed no significant hazards

consideration determination or expand

the scope of the August 2, 1995, Federal

Register notice. The Commission’s

related evaluation of the amendments is

contained in a Safety Evaluation dated

September 15, 1995. No significant

hazards consideration comments

received: No

Local Public Document Room

location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa,

Pennsylvania 15001

Entergy Operations, Inc., Docket No.

50-382, Waterford Steam

ElectricStation, Unit 3, St. Charles

Parish, Louisiana

Date of amendment request: June 22,

1994

Brief description of amendment: The

amendment changes the Appendix A

Technical Specifications by removing

the seismic and meteorological

monitoring instrumentation

requirements. These requirements are to

be relocated in the Updated Final Safety

Analysis Report.

Date of issuance: September 5, 1995

Effective date: September 5, 1995

Amendment No.: 112

Facility Operating License No. NPF38. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 3, 1994 (59 FR 39585)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 5, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: University of New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, LA 70122

Entergy Operations, Inc., Docket No.

50-382, Waterford Steam

ElectricStation, Unit 3, St. Charles

Parish, Louisiana

Date of amendment request: June 22,

1994, and December 9, 1994

Brief description of amendment: The

amendment changes the Appendix A

TSs by revising the plant protection

system trip setpoints and allowable
49954 Federal Register / Vol. 60, No. 187 / Wednesday, September 27, 1995 / Notices

values such that they will be consistent

with the current setpoint/uncertainty

methodology being implemented at

Waterford 3.

Date of issuance: September 5, 1995

Effective date: September 5, 1995

Amendment No.: 113

Facility Operating License No. NPF38. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 3, 1994 (59 FR 39586)

and February 1, 1995 (60 FR 6300)The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated September 5, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: University of New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, LA 70122

Entergy Operations, Inc., Docket No.

50-382, Waterford Steam

ElectricStation, Unit 3, St. Charles

Parish, Louisiana

Date of amendment request:

December 9, 1994, as supplemented by

letter dated July 25, 1995

Brief description of amendment: The

requested changes revised the allowable

opening tolerances on the pressurizer

safety valves (PSVs) and the main steam

line safety valves (MSSVs) from plus or

minus 1 percent to plus or minus 3

percent. However, following testing, the

as-left lift setting of the PSVs and

MSSVs will be within plus or minus 1

percent of the pressure specified in the

Technical Specifications.

Date of issuance: September 11, 1995

Effective date: September 11, 1995

Amendment No.: 111

Facility Operating License No. NPF38. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 1, 1995 (60 FR 6300)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 11,

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 No. 50-335, St. Lucie Plant, Unit

No. 1, St. Lucie County, Florida

Date of application for amendment:

May 23, 1994

Brief description of amendment: The

amendment revises Technical

Specification 3.5.2 for Emergency Core

Cooling Systems (ECCS) by removing

the option that allows High Pressure

Safety Injection (HPSI) Pump 1C to be

used as an alternative to the preferred

pump for subsystem operability. HPSI

pump 1C is an installed spare which is

not required to be maintained in an

operable status, and this change

upgrades the ECCS operability

requirements consistent with actual

plant operating needs.

Date of issuance: September 11, 1995

Effective date: September 11, 1995

Amendment No.: 139

Facility Operating License No. DPR67: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 6, 1994 (59 FR 34663) The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated September 11,

1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort

Pierce, Florida 34954-9003

Florida Power and Light Company, et

al., Docket No. 50-389, St. Lucie Plant,

Unit No. 2, St. Lucie County, Florida

Date of application for amendment:

February 27, 1995

Brief description of amendment: This

amendment will change Table 3.3-3 and

3.3-4 to accommodate an improved

coincidence logic and relay replacement

for the 4.16 kV Loss of Voltage Relays.

Actions required for certain trip units

with the number of operable channels

one less than the total number of

channels will also be changed. In

addition, the format used to state the

time delay for the 4.16 kV Degraded

Voltage trip unit will be revised.

Date of issuance: September 1, 1995

Effective date: September 1, 1995

Amendment No.: 79

Facility Operating License No. NPF16: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 29, 1995 (60 FR 16187)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 1, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort

Pierce, Florida 34954-9003

Georgia Power Company, Oglethorpe

Power Corporation, Municipal Electric

Authority of Georgia, City of Dalton,

Georgia, Docket Nos. 50-321 and 50-

366, Edwin I. Hatch Nuclear Plant,

Units 1 and 2, Appling County, Georgia

Date of application for amendments:

January 13, 1995, as supplemented by

letters dated April 5 and June 20, 1995.

Brief description of amendments: The

amendments modify

Facility Operating License Nos. DRP57 and NPF-5 and the corresponding TS

for Hatch Units 1 and 2, respectively, to

authorize an increase in the maximum

power level from 2436 megawatts

thermal (MWt) to 2558 MWt. The

amendments also approve changes to

the Technical Specification to

implement uprated power operation.

Date of issuance: August 31, 1995

Effective date: As of the date of

issuance to be implemented prior to

startup in Cycle 17 for Unit 1; and prior

to startup in Cycle 13 for Unit 2

Amendment Nos.: 197 and 138

Facility Operating License Nos. DPR57 and NPF-5. Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: July 5, 1995 (60 FR 35072) The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated August 31, 1995 and

an Environmental Assessment dated

July 21, 1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: Appling County Public

Library, 301 City Hall Drive, Baxley,

Georgia 31513

Georgia Power Company, Oglethorpe

Power Corporation, Municipal Electric

Authority of Georgia, City of Dalton,

Georgia, Docket Nos. 50-321 and 50-

366, Edwin I. Hatch Nuclear Plant,

Units 1 and 2, Appling County, Georgia

Date of application for amendments:

June 6, 1995, as supplemented August 9,

1995.

Brief description of amendments: The

amendments revise Technical

Specification Surveillance

Requirements (SR) 3.6.4.1.3 and

3.6.4.1.4 for the secondary containment

drawdown. The revision reduces the SR

acceptance criteria to greater than or

equal to 0.20 inch water gauge (wg)

negative pressure from greater than or

equal to O.25 inch wg negative pressure.

The appropriate TS Bases pages are also

changed to reflect the TS revision.

Date of issuance: September 11, 1995

Effective date: As of the date of

issuance to be implemented within 60

days

Amendment Nos.: 198 and 139
Federal Register / Vol. 60, No. 187 / Wednesday, September 27, 1995 / Notices 49955

Facility Operating License Nos. DPR57 and NPF-5. Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: June 21, 1995 (60 FR 32364)

The August 9, 1995, letter provided

clarifying information that did not

change the scope of the June 6, 1995,

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

September 11, 1995.No significant

hazards consideration comments

received: No

Local Public Document Room

location: Appling County Public

Library, 301 City Hall Drive, Baxley,

Georgia 31513

GPU Nuclear Corporation, et al.,

Docket No. 50-219, Oyster Creek

Nuclear Generating Station, Ocean

County, New Jersey

Date of application for amendment:

June 26, 1995

Brief description of amendment: The

amendment revises the snubber visual

inspection intervals to match the

schedule developed by the NRC staff for

use with a 24-month refueling interval.

This schedule was documented in

Generic Letter 90-09. The amendment

also revises the bases for the snubber

visual inspection interval to be

consistent with the bases described in

Generic Letter 90-09.

Date of issuance: September 6, 1995

Effective date: September 6, 1995

Amendment No.: 182

Facility Operating License No. DPR16. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 2, 1995 (60 FR 39440).

The Commission’s related evaluation of

this amendment is contained in a Safety

Evaluation dated September 6, 1995. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Ocean County Library,

Reference Department, 101 Washington

Street, Toms River, NJ 08753

Gulf States Utilities Company, Cajun

Electric Power Cooperative, and

Entergy Operations, Inc., Docket No.

50-458, River Bend Station, Unit 1,

West Feliciana Parish, Louisiana

Date of amendment request: May 25,

1995

Brief description of amendment: The

amendment revises the Physical

Security Plan vital island requirements.

Date of issuance: September 12, 1995

Effective date: September 12, 1995

Amendment No.: 83

Facility Operating License No. NPF47. The amendment revised the

operating license.

Date of initial notice in Federal

Register: July 19, 1995 (60 FR 37091)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 12,

1995.No significant hazards

consideration comments received. No

Local Public Document Room

location: Government Documents

Department, Louisiana State University,

Baton Rouge, LA 70803

Houston Lighting & Power Company,

City Public Service Board of San

Antonio, Central Power and Light

Company, City of Austin, Texas, Docket

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

Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request: May 25,

1995, as supplemented by letter dated

August 3, 1995.

Brief description of amendments: The

amendments revised the technical

specifications (TSs) on containment

leakage, making the action statement

consistent with the need to perform

Type C testing at power, and replacing

the surveillance requirements with a

single requirement to apply the

requirements of Appendix J as modified

by approved exemptions. The

amendments also revised the TSs on

containment integrity, containment

leakage, and containment air locks, to

eliminate the numerical value of

calculated peak containment internal

pressure related to the design basis

accident.

Date of issuance: September 7, 1995

Effective date: September 7, 1995

Amendment Nos.: Unit 1 -

Amendment No. 80; Unit 2 -

Amendment No. 69

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

revised the Technical Specifications.

Date of initial notice in Federal

Register: July 19, 1995 (60 FR 37092)

The August 3, 1995, supplement

provided clarifying information and did

not change the original no significant

hazards consideration determination.

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated September 7, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Wharton County Junior

College, J. M. Hodges Learning Center,

911 Boling Highway, Wharton, Texas

77488

Houston Lighting & Power Company,

City Public Service Board of San

Antonio, Central Power and Light

Company, City of Austin, Texas, Docket

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

Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request: May 31,

1995, as supplemented by letter dated

August 2, 1995

Brief description of amendments: The

amendments modified (by relocation to

the Technical Requirements Manual) TS

3/4.1.2.1, Boration Systems/Flow Paths

- Shutdown, TS 3/4.1.2.2, Boration

Systems/Flow Paths - Operating, TS 3/

4.1.2.3, Charging Pumps - Shutdown, TS

3/4.1.2.4, Charging Pumps - Operating,

TS 3/4.1.2.5, Borated Water Sources -

Shutdown, TS 3/4.1.2.6, Borated Water

Sources - Operating, TS 3/4.4.2.1, Safety

Valves - Shutdown, and the associated

Bases.

Date of issuance: September 5, 1995

Effective date: September 5, 1995

Amendment Nos.: Unit 1 -

Amendment No. 79; Unit 2 -

Amendment No. 68

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

revised the Technical Specifications.

Date of initial notice in Federal

Register: August 2, 1995 (60 FR 39441)

The additional information contained in

the supplemental letter dated August 2,

1995, was clarifying in nature and thus,

within the scope of the initial notice

and did not affect the staff’s proposed

no significant hazards consideration

determination.The Commission’s

related evaluation of the amendments is

contained in a Safety Evaluation dated

September 5, 1995.No significant

hazards consideration comments

received: No

Local Public Document Room

location: Wharton County Junior

College, J. M. Hodges Learning Center,

911 Boling Highway, Wharton, TX

77488

Indiana Michigan Power Company,

Docket No. 50-315, Donald C. Cook,

Nuclear Plant, Unit No. 1, Berrien

County, Michigan

Date of application for amendment:

February 3, 1995, as supplemented

April 25, 1995.

Brief description of amendment: The

amendment modifies the technical

specifications to extend the interim

steam generator tube plugging criteria

used in Cycle 14 to the next operating

cycle (Cycle 15).

Date of issuance: September 13, 1995

Effective date: September 13, 1995

Amendment No.: 200
49956 Federal Register / Vol. 60, No. 187 / Wednesday, September 27, 1995 / Notices

Facility Operating License No. DPR58. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: July 19, 1995 (60 FR 37093)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 13,

1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: Maud Preston Palenske

Memorial Library, 500 Market Street, St.

Joseph, Michigan 49085

Indiana Michigan Power Company,

Docket No. 50-316, Donald C. Cook,

Nuclear Plant, Unit No. 2, Berrien

County, Michigan

Date of application for amendment:

February 15, 1994, as supplemented

June 29, 1995

Brief description of amendment: The

amendment deletes Technical

Specification section 3/4.3.4, associated

bases, and associated index listings for

the Unit 2 turbine overspeed protection

system.

Date of issuance: September 1, 1995

Effective date: September 1, 1995

Amendment No.: 185

Facility Operating License No. DPR74. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: March 30, 1994 (59 FR 14890)

The licensee’s submittal of June 29,

1995, did not change the basis for the

proposed no significant hazards

consideration determination.The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated September 1, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Maud Preston Palenske

Memorial Library, 500 Market Street, St.

Joseph, Michigan 49085

Nebraska Public Power District, Docket

No. 50-298, Cooper Nuclear Station,

Nemaha County, Nebraska

Date of amendment request: June 15,

1995

Brief description of amendment: The

amendment changes the Technical

Specifications to revise the definition

for logic system functional test and

revises the surveillance interval for

emergency core cooling system logic

system functional testing from 6 months

to 18 months.

Date of issuance: September 7, 1995

Effective date: September 7, 1995

Amendment No.: 171

Facility Operating License No. DPR46. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 19, 1995 (60 FR 37096)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 7, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Auburn Public Library, 118

15th Street, Auburn, NE 68305

Niagara Mohawk Power Corporation,

Docket No. 50-410, Nine Mile

PointNuclear Station, Unit 2, Oswego

County, New York

Date of application for amendment:

January 6, 1995

Brief description of amendment: The

amendment incorporates Limiting

Condition for Operation 3.3.3.1 from

Standard Technical Specifications into

Technical Specification (TS) 3/4.3.7.5,

Accident Monitoring Instrumentation

and make associated changes in TS 3/

4.4.2, Safety Relief Valves.

Date of issuance: September 11, 1995

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 69

Facility Operating License No. NPF69: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: February 15, 1995 (60 FR

8748)The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

September 11, 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

Niagara Mohawk Power Corporation,

Docket No. 50-410, Nine Mile

PointNuclear Station, Unit 2, Oswego

County, New York

Date of application for amendment:

January 6, 1995, as supplemented April

18, 1995

Brief description of amendment: The

amendment revises Technical

Specifications (TSs) Sections 3.8.1.1 and

3.8.1.2; TS Surveillance Requirements

Section 4.8.1.1.2; TS Bases Section 3/

4.8.1.3; and TS Administrative Controls

Section 6.8.4. The changes include:

updating the minimum day tank and

storage tank oil inventory, specific

actions required if oil level fall below

minimum required, revising and

relocating the fuel oil sampling and

testing criteria to the associated Bases,

and specific action to be taken if the fuel

oil properties do not meet the specified

limits. In addition, a requirement was

added for a diesel fuel oil testing

program. These changes are consistent

with guidance provided in NUREG1434.

Date of issuance: September 15, 1995

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 70

Facility Operating License No. NPF69: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: February 15, 1995 (60 FR

8747) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

September 15, 1995.No significant

hazards consideration comments

received: No

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126

North Atlantic Energy Service

Corporation, Docket No. 50-443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request: April 16,

1995.

Description of amendment request:

The amendment revises the Appendix A

Technical Specifications (TS) relating to

containment building penetrations.

Specifically, the amendment modifies

Limiting Conditions for Operation 3.9.4

to permit both doors of one personnel

airlock to be open during core

alterations or irradiated fuel movement

if certain conditions are met and to add

equivalent and alternate penetration

closure methodologies. Surveillance

Requirement 4.9.4 is changed to reflect

that the penetrations are to be verified

to be in the condition required. Bases

Section 3/4 9.4 also is revised to reflect

the changes described above.

Date of issuance: August 31, 1995

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment No.: 40

Facility Operating License No. NPF86. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 21, 1995 (60 FR 32369)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated August 31, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Exeter Public Library,

Founders Park, Exeter, NH 03833
Federal Register / Vol. 60, No. 187 / Wednesday, September 27, 1995 / Notices 49957

North Atlantic Energy Service

Corporation, Docket No. 50-443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request: May 30,

1995.

Description of amendment request:

The amendment revises the Appendix A

Technical Specifications (TS) relating to

Moderator Temperature Coefficient. The

amendment changes the upper limit for

the moderator temperature coefficient

(MTC) for certain operating conditions.

Additionally, a reference for the

analytical method used to determine the

cycle-specific MTC upper limit is added

to TS 6.8.1.6.b.

Date of issuance: September 14, 1995

Effective date: As of the date of

issuance to be implemented within 60

days.

Amendment No.: 41

Facility Operating License No. NPF86. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 5, 1995 (60 FR 35082).

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 14, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Exeter Public Library,

Founders Park, Exeter, NH 03833

North Atlantic Energy Service

Corporation, Docket No. 50-443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request: June 16,

1995

Description of amendment request:

The amendment revises the Appendix A

Technical Specifications (TS) relating to

core reactivity control available from

borated water sources. The amendment

changes the minimum boron

concentration specified for the refueling

water storage tank (RWST) in Limiting

Condition for Operation (LCO) in TS

3.1.2.5 and replaces the minimum

specified concentration for boron with

an acceptable range of boron

concentration for the RWST and the

accumulators in the LCOs for TS 3.1.2.6,

3.5.1.1, and 3.5.4.

Date of issuance: September 14, 1995

Effective date: As of the date of

issuance, to be implemented prior to

entering MODE 4 following the fourth

refueling outage.

Amendment No.: 42

Facility Operating License No. NPF86. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 2, 1995 (60 FR 39442).

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 14, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Exeter Public Library,

Founders Park, Exeter, NH 03833

Northeast Nuclear Energy Company,

Docket No. 50-245, MillstoneNuclear

Power Station, Unit 1, New London

County, Connecticut

Date of application for amendment:

July 11, 1995

Brief description of amendment: The

amendment modifies Technical

Specification 3.5.F.7 to also allow the

use of pull-to-lock switches to defeat the

automatic initiation of the emergency

core cooling system while in the refuel

condition. The amendment also makes

editorial corrections and makes changes

to the associated Bases section.

Date of issuance: September 13, 1995

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 88

Facility Operating License No. DPR21. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 2, 1995 (60 FR 39442).

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 13,

1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360

Northeast Nuclear Energy Company,

Docket No. 50-245, MillstoneNuclear

Power Station, Unit 1, New London

County, Connecticut

Date of application for amendment:

July 18, 1995

Brief description of amendment: The

amendment adds operability and

surveillance requirements for reactor

pressure vessel overfill protection

instrumentation. The amendment also

adds the associated Bases.

Date of issuance: September 13, 1995

Effective date: As of the date of

issuance to be implemented within 60

days.

Amendment No.: 87

Facility Operating License No. DPR21. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 2, 1995 (60 FR 39443)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 13,

1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360

Omaha Public Power District, Docket

No. 50-285, Fort Calhoun Station,Unit

No. 1, Washington County, Nebraska

Date of amendment request: April 7,

1995

Brief description of amendment: The

amendment revises the technical

specifications (TS) to relocate the axial

power distribution limits to the Core

Operating Limits Report (COLR).

Date of issuance: September 1, 1995

Effective date: September 1, 1995

Amendment No.: 170

Facility Operating License No. DPR40. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: May 23, 1995 (60 FR 27339)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 1, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska

68102

Omaha Public Power District, Docket

No. 50-285, Fort Calhoun Station,Unit

No. 1, Washington County, Nebraska

Date of amendment request: May 8,

1995, as supplemented by letter dated

July 11, 1995.

Brief description of amendment: The

amendment changes Sections 2.3, 3.1,

3.2, 3.3, and 3.6 of the Technical

Specifications in accordance with the

guidance of Generic Letter (GL) 93-05,

‘‘Line Item Technical Specifications

Improvements to Reduce Surveillance

Requirements for Testing During Power

Operation,’’ dated September 27, 1993.

The changes are consistent with Station

operating experience and NUREG-1366,

‘‘Improvements to Technical

Specifications Surveillance

Requirements,’’ dated December 1992.

In addition, a change was made to TS

Section 3.1 in accordance with the

Commission’s Final Policy Statement on

Technical Specifications Improvements

for Nuclear Power Reactors. Also,

changes were made to the TS sections

identified above for clarity and to

correct administrative errors.

Date of issuance: September 7, 1995

Effective date: September 7, 1995

Amendment No.: 171
49958 Federal Register / Vol. 60, No. 187 / Wednesday, September 27, 1995 / Notices

Facility Operating License No. DPR40. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 6, 1995 (60 FR 29883)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 7, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska

68102

Pennsylvania Power and Light

Company, Docket No. 50-387,

Susquehanna Steam Electric Station,

Unit 1, Luzerne County, Pennsylvania

Date of application for amendment:

April 11, 1995

Brief description of amendment: This

amendment extends on a one-time basis

the allowed outage time from 3 to 7 days

for one offsite circuit being out of

service.

Date of issuance: August 31, 1995

Effective date: As of the date of

issuance and is to be implemented

within 30 days.

Amendment No.: 153

Facility Operating License No. NPF14: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: June 6, 1995 (60 FR 29886).

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated August 31, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Osterhout Free Library,

Reference Department, 71 South

Franklin Street, Wilkes-Barre,

Pennsylvania 18701

Pennsylvania Power and Light

Company, Docket Nos. 50-387 and 50-

388 Susquehanna Steam Electric

Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of application for amendments:

February 2, 1995

Brief description of amendments:

These amendments change the

Technical Specifications for the two

Susquehanna units to increase the

licensed discharge fuel assembly for

SPC 9X9-2 fuel from 40 to 45 GWD/

MTU. This change is consistent with the

Commissions approval of Topical

Report PL-NF-94-005-P, ‘‘Technical

Basis for SPC 9X9-2 Extended Fuel

Exposure at Susquehanna SES,’’

documented in a letter to PP&L dated

December 15, 1994.

Date of issuance: September 12, 1995

Effective date: September 12, 1995

Amendment Nos.: 154 and 124

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

revised the Technical Specifications.

Date of initial notice in Federal

Register: March 29, 1995 (60 FR 16194)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated September 12,

1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: Osterhout Free Library,

Reference Department, 71 South

Franklin Street, Wilkes-Barre,

Pennsylvania 18701

Pennsylvania Power and Light

Company, Docket No. 50-388,

Susquehanna Steam Electric Station,

Unit 2, Luzerne County, Pennsylvania

Date of application for amendment:

March 31, 1995

Brief description of amendment: This

amendment changes Technical

Specification Section 6.9.3.2 to allow

four GE demonstration assemblies to be

loaded into Susquehanna Unit 2, Cycle

8 core.

Date of issuance: September 13, 1995

Effective date: As of date of issuance

and shall be implemented within 30

days.

Amendment No.: 125

Facility Operating License No. NPF22. This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: April 26, 1995 (60 FR 20523)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 13,

1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: Osterhout Free Library,

Reference Department, 71 South

Franklin Street, Wilkes-Barre,

Pennsylvania 18701

Pennsylvania Power and Light

Company, Docket No. 50-388,

Susquehanna Steam Electric Station,

Unit 2, Luzerne County, Pennsylvania

Date of application for amendment:

May 5, 1995, and supplemented by

letter dated August 18, 1995

Brief description of amendment: This

amendment deletes from SSES

Technical Specification Table 3.6.3-1,

‘‘Primary Containment Isolation

Valves,’’ three relief valves in the

residual heat removal system. These

specific valves which were originally

intended to support the steam

condensing mode, were previously

eliminated from the plant design. The

valves are being replaced during the

September Unit 2 refueling outage and

will be replaced by blind flanges.

Date of issuance: September 11, 1995

Effective date: September 11, 1995

Amendment No.: 123

Facility Operating License No. NPF22. This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: July 5, 1995 (60 FR 35083 and

July 17, 1995 (60 FR 36449)The

supplemental letter provided clarifying

information that did not change the

initial proposed no significant hazards

consideration determination or expand

the scope of the original Federal

Register notice. The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

September 11, 1995.No significant

hazards consideration comments

received: No

Local Public Document Room

location: Osterhout Free Library,

Reference Department, 71 South

Franklin Street, Wilkes-Barre,

Pennsylvania 18701

Power Authority of The State of New

York, Docket No. 50-286, Indian Point

Nuclear Generating Unit No. 3,

Westchester County, New York

Date of application for amendment:

February 23, 1995, as supplemented

July 28, 1995

Brief description of amendment: The

amendment revised the minimum

emergency diesel generator (EDG) fuel

oil requirements, as indicated in

Technical Specification (TS) Section 3.7

(Auxiliary Electrical Systems), from

7056 to 6671 gallons. The actual

minimum fuel oil level had always been

6671 gallons; however, the previous TS

limit of 7056 gallons was based on a

level indicator that had an accuracy of

+/- 385 gallons. This revision clarified

the TS such that any level indicator can

now be used as long as an actual

minimum level of 6671 gallons is

assured.

Date of issuance: August 30, 1995

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 161

Facility Operating License No. DPR64: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 29, 1995 (60 FR 16196)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated August 30, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10610
Federal Register / Vol. 60, No. 187 / Wednesday, September 27, 1995 / Notices 49959

Power Authority of The State of New

York, Docket No. 50-286, Indian Point

Nuclear Generating Unit No. 3,

Westchester County, New York

Date of application for amendment:

March 2, 1995

Brief description of amendment: The

amendment revised the titles of several

management positions as described in

Technical Specifications Section 6.0

(Administrative Controls). Specifically,

the title of Executive Vice President and

Chief Nuclear Officer and the title of

Shift Supervisor were changed to Chief

Nuclear Officer and Shift Manager,

respectively. In addition, the position

titles of Senior Reactor Operator and

Reactor Operator were deleted and

replaced with qualification

requirements.

Date of issuance: August 31, 1995

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 162

Facility Operating License No. DPR64: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 29, 1995 (60 FR 16197)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated August 31, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10610

Power Authority of the State of New

York, Docket No. 50-333, James A.

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of application for amendment:

October 3, 1994

Brief description of amendment: The

amendment proposed changes to

FitzPatrick TSs which will extend the

instrumentation functional test interval

and allowable out-of-service times,

remove the average power range

monitor downscale scram function and

the instrument response time values,

and incorporate several editorial,

clarification, and correction changes.

Date of issuance: September 11, 1995

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 227

Facility Operating License No. DPR59: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: November 9, 1994 (59 FR

55887) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

September 11, 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

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of application for amendment:

September 29, 1994

Brief description of amendment: This

amendment revises Table 4.3.6-1,

‘‘Control Rod Block Instrumentation

Surveillance,’’ of the Hope Creek TS.

The channel calibration frequencies for

the Source Range Monitor (SRM) and

the Intermediate Range Monitor (IRM),

in TS Table 4.3.6-1, are changed for the

up-scale and the down-scale trip

functions on each instrument from

‘‘SA’’ (once-per-184 days) to ‘‘R’’ (onceper-refueling cycle).

Date of issuance: September 12, 1995

Effective date: As of the date of

issuance and shall be implemented

within 60 days.

Amendment No.: 78

Facility Operating License No. NPF57: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: June 18, 1995 (60 FR 3676).

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 12, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey

08070

Public Service Electric & Gas Company,

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

Nuclear Generating Station, Unit Nos. 1

and 2, Salem County, New Jersey

Date of application for amendments:

May 2, 1995

Brief description of amendments: The

amendments eliminate the monthly

manual initiation of auxiliary feedwater

from Technical Specification Tables 3.3-

3, 3.3.-4 and 4.3-2.

Date of issuance: September 6, 1995

Effective date: Units 1 and 2, as of the

date of issuance, to be implemented

within 60 days.

Amendment Nos. 175 and 156

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

revised the Technical Specifications.

Date of initial notice in Federal

Register: June 6, 1995 (60 FR 29887)The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated September 6, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Salem Free Public Library, 112

West Broadway, Salem, New Jersey

08079

South Carolina Electric & Gas

Company, South Carolina Public

ServiceAuthority, Docket No. 50-395,

Virgil C. Summer Nuclear Station, Unit

No. 1, Fairfield County, South Carolina

Date of application for amendment:

June 19, 1995

Brief description of amendment: The

amendment restructures the primary

containment and primary containment

leakage technical specifications to

reduce the repetition of those

requirements contained in NRC

regulations such as Appendix J to 10

CFR Part 50.

Date of issuance: September 5, 1995

Effective date: September 5, 1995

Amendment No.: 126

Facility Operating License No. NPF12: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: July 19, 1995 (60 FR

37099)The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

September 5, 1995.No significant

hazards consideration comments

received: No

Local Public Document Room

location: Fairfield County Library, 300

Washington Street, Winnsboro, SC

29180

Southern California Edison Company,

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

San Onofre Nuclear Generating Station,

Unit Nos. 2 and 3, San Diego County,

California

Date of application for amendments:

May 20, 1994

Brief description of amendments: The

amendments revise Technical

Specification 3/4.7.3, ‘‘Component

Cooling Water System,’’ and the

corresponding Bases to support the

addition of the component cooling

water surge tank backup nitrogen

supply (BNS) system.

Date of issuance: September 13, 1995

Effective date: September 13, 1995

Amendment Nos.: Unit 2 -

Amendment No. 125; Unit 3 -

Amendment No. 114

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

revised the Technical Specifications.

Date of initial notice in Federal

Register: August 31, 1994 (59 FR
49960 Federal Register / Vol. 60, No. 187 / Wednesday, September 27, 1995 / Notices

45034)The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

September 13, 1995.No significant

hazards consideration comments

received: No

Local Public Document Room

location: Main Library, University of

California, P.O. Box 19557, Irvine,

California 92713

Tennessee Valley Authority, Docket

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

Browns Ferry Nuclear Plant, Units 1, 2,

and 3, Limestone County, Alabama

Date of application for amendments:

March 31, 1995, supplemented July 14,

1995 (TS 349)

Brief description of amendment:

These amendments revise the Browns

Ferry Nuclear Plant (BFN) Units 1, 2,

and 3 reactor vessel pressuretemperature curves and bolt-up

temperatures.

Date of issuance: September 13, 1995

Effective date: September 13, 1995

Amendment Nos.: 224, 239, 198

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

revised the Technical Specifications.

Date of initial notice in Federal

Register: June 6, 1995 (60 FR 29888)The

July 14, 1995 letter 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 September 13,

1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: Athens Public library, South

Street, Athens, Alabama 35611

Tennessee Valley Authority, Docket

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

Browns Ferry Nuclear Plant, Units 1, 2,

and 3, Limestone County, Alabama

Date of application for amendments:

May 11, 1995, supplemented June 30,

1995 (TS 359)

Brief description of amendment: The

amendments provide for the addition of

a reactor trip on low scram pilot air

header pressure for BFN Unit 3, and

revise a note regarding instrumentation

requirements for all three BFN reactors.

Date of issuance: August 29, 1995

Effective date: August 29, 1995

Amendment Nos.: 223, 228 and 197

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

revised the Technical Specifications.

Date of initial notice in Federal

Register: June 6, 1995 (60 FR 29889)The

June 30, 1995 letter 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 August 29, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Athens Public library, South

Street, Athens, Alabama 35611

Tennessee Valley Authority, Docket

Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of application for amendments:

June 29, 1995 (TS 95-14)

Brief description of amendments: The

amendments revise Technical

Specification 3.9.4, Containment

Building Penetrations, to allow both sets

of containment personnel airlock doors

to be open during core alterations and

fuel movement provided one door is

capable of closure and one train of

auxiliary building gas treatment remains

operable.

Date of issuance: September 6, 1995

Effective date: September 6, 1995

Amendment Nos.: 209 and 199

Facility Operating License Nos. DPR77 and DPR-79: Amendments revise the

technical specifications.

Date of initial notice in Federal

Register: July 19, 1995 (60 FR

37100)The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

September 6, 1995. No significant

hazards consideration comments

received: No

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 amendments:

August 7, 1995 (TS 95-11)

Brief description of amendments: The

amendments revise the time constant

used in the overtemperature delta

temperature and the overpower delta

temperature trip equations of Technical

Specification Table 2.2-1.

Date of issuance: September 15, 1995

Effective date: September 15, 1995

Amendment Nos.: 211 and 201

Facility Operating License Nos. DPR77 and DPR-79: Amendments revise the

technical specifications.

Date of initial notice in Federal

Register: April 26, 1995 (60 FR 20527);

superseded August 15, 1995 (60 FR

42187) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

September 15, 1995.No significant

hazards consideration comments

received: No

Local Public Document Room

location: 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 amendments:

August 21, 1995 (TS 95-21)

Brief description of amendments: The

amendments change Technical

Specification 3.7.5.c to allow an

increase in the average essential raw

cooling water supply header

temperature from 84.5°F to 87°F

untilSeptember 30, 1995.

Date of issuance: September 13, 1995

Effective date: September 13, 1995

Amendment Nos.: 210 and 200

Facility Operating License Nos. DPR77 and DPR-79: Amendments revise the

technical specifications.Public

comments requested as to proposed no

significant hazards consideration: Yes

(August 28, 1995, 60 FR 44517). That

notice provided an opportunity to

submit comments on the Commission’s

proposed no significant hazards

determination. No comments have been

received. The notice also provided an

opportunity to request a hearing, by

September 12, 1995, but indicated that

if the Commission makes a final no

significant hazards consideration

determination before the expiration of

the notice period, any such hearing

would take place after issuance of the

amendments.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 September 13, 1995.

Local Public Document Room

location: Chattanooga-Hamilton County

Library,1101 Broad Street, Chattanooga,

Tennessee 37402

The Cleveland Electric Illuminating

Company, Centerior Service Company,

Duquesne Light Company, Ohio Edison

Company, Pennsylvania Power

Company, Toledo Edison Company,

Docket No. 50-440, Perry Nuclear

Power Plant, Unit No. 1, Lake County,

Ohio

Date of application for amendment:

April 28, 1995

Brief description of amendment: The

amendment extends for one additional

operating cycle the exception to

Limiting Condition for Operation 3.0.4

as it applies to the main steam isolation
Federal Register / Vol. 60, No. 187 / Wednesday, September 27, 1995 / Notices 49961

valve leakage control system Technical

Specification.

Date of issuance: September 8, 1995

Effective date: September 8, 1995

Amendment No.: 71

Facility Operating License No. NPF58: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: May 23, 1995 (60 FR

27344)The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

September 8, 1995. No significant

hazards consideration comments

received: No

Local Public Document Room

location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50-346, Davis-Besse Nuclear Power

Station, Unit No. 1, Ottawa County,

Ohio

Date of application for amendment:

April 10, 1995

Brief description of amendment: This

amendment changes auxiliary feedwater

system, motor driven feedwater pump,

and condensate system Technical

Specifications to increase clarity and

changes format to more closely follow

improved standard technical

specifications and increases content of

Bases discussions.

Date of issuance: September 5, 1995

Effective date: September 5, 1995

Amendment No.: 200

Facility Operating License No. NPF-3.

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 2, 1995 (60 FR 39453)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 5, 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,

Docket Nos. 50-280 and 50-281, Surry

Power Station, Unit Nos. 1 and 2, Surry

County, Virginia.

Date of application for amendments:

November 10, 1994

Brief description of amendments:

Clarify the surveillance requirementsfor

the Reactor Protection and Engineered

Safeguards Systems instrumentation

and actuation logic.

Date of issuance: September 14, 1995

Effective date: September 14, 1995

Amendment Nos.: 205 and 205

Facility Operating License Nos. DPR32 and DPR-37: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: April 12, 1995 (60 FR 18630)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 14,

1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: Swem Library, College of

William and Mary, Williamsburg,

Virginia 23185

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:

July 14, 1995

Brief description of amendments:

These amendments would provide a 2-

hour allowed outage time for one

residual heat removal loop to

accommodate plant safety and

emergency power systems surveillance

testing, permit depressurizing safety

injection accumulators in lieu of

accumulator isolation, and make

administrative changes.

Date of issuance: September 1, 1995

Effective date: September 1, 1995

Amendment Nos.: 204 and 204

Facility Operating License Nos. DPR32 and DPR-37: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: August 2, 1995 (60 FR 39455)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 1, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Swem Library, College of

William and Mary, Williamsburg,

Virginia 23185

Wisconsin Public Service Corporation,

Docket No. 50-305, Kewaunee

NuclearPower Plant, Kewaunee County,

Wisconsin

Date of application for amendment:

December 16, 1994.

Brief description of amendment: The

amendment revises the Kewaunee

Nuclear Power Plant (KNPP) Technical

Specifications Sections 3.4 and 4.1 by

removing the limiting conditions for

operation (LCO) and the surveillance

requirements for the turbine overspeed

protection system (TOPS). The TOPS

requirements will be relocated to the

Updated Safety Analysis Report

(USAR).

Date of issuance: August 31, 1995

Effective date: August 31, 1995

Amendment No.: 121

Facility Operating License No. DPR43. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 18, 1995 (60 FR 3676).

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated August 31, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: University of Wisconsin

Library Learning Center, 2420 Nicolet

Drive, Green Bay, Wisconsin 54301

Notice Of Issuance Of Amendments To

Facility Operating Licenses And Final

Determination Of No Significant

Hazards Consideration And

Opportunity For A Hearing (Exigent

Public Announcement Or Emergency

Circumstances)

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application for the

amendment complies with the

standards and requirements of the

Atomic Energy Act of 1954, as amended

(the Act), and the Commission’s rules

and regulations. The Commission has

made appropriate findings as required

by the Act and the Commission’s rules

and regulations in 10 CFR Chapter I,

which are set forth in the license

amendment.

Because of exigent or emergency

circumstances associated with the date

the amendment was needed, there was

not time for the Commission to publish,

for public comment before issuance, its

usual 30-day Notice of Consideration of

Issuance of Amendment, Proposed No

Significant Hazards Consideration

Determination, and Opportunity for a

Hearing.

For exigent circumstances, the

Commission has either issued a Federal

Register notice providing opportunity

for public comment or has used local

media to provide notice to the public in

the area surrounding a licensee’s facility

of the licensee’s application and of the

Commission’s proposed determination

of no significant hazards consideration.

The Commission has provided a

reasonable opportunity for the public to

comment, using its best efforts to make

available to the public means of

communication for the public to

respond quickly, and in the case of

telephone comments, the comments

have been recorded or transcribed as

appropriate and the licensee has been

informed of the public comments.

In circumstances where failure to act

in a timely way would have resulted, for
49962 Federal Register / Vol. 60, No. 187 / Wednesday, September 27, 1995 / Notices

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

October 27, 1995, the licensee may file

a request for a hearing with respect to

issuance of the amendment to the

subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses. Since the Commission has

made a final determination that the

amendment involves no significant

hazards consideration, if a hearing is

requested, it will not stay the

effectiveness of the amendment. Any

hearing held would take place while the

amendment is in effect.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

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

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

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

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

operator should be given Datagram

Identification Number N1023 and the

following message addressed to (Project

Director): petitioner’s name and

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear
Federal Register / Vol. 60, No. 187 / Wednesday, September 27, 1995 / Notices 49963

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).

Niagara Mohawk Power Corporation,

Docket No. 50-410, Nine Mile

PointNuclear Station, Unit 2, Oswego

County, New York

Date of application for amendment:

August 28, 1995

Brief description of amendment: The

amendment revises Primary

Containment Purge System Technical

Specification Section 3.6.1.7, Limiting

Condition for Operation. The revision

extends the amount of time the 12-inch

and 14-inch purge system supply and

exhaust lines may be used for venting or

purging from 90 to 135 hours per 365

days. In addition, expired footnotes

were deleted as an editorial change and

the associated Bases section was

revised.

Date of issuance: August 31, 1995

Effective date: As of the date of

issuance to be implemented upon

receipt.

Amendment No.: 68

Facility Operating License No. NPF69: Amendment revises the Technical

Specifications.Public comments

requested as to proposed no significant

hazards consideration: NoThe

Commission’s related evaluation of the

amendment, emergency circumstances

and consultation with the State, and

final determination of no significant

hazards consideration are contained in

a Safety Evaluation dated August 31,

1995.

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston & Strawn,

1400 L Street, NW., Washington, DC

20005-3502.

NRC Project Director: Ledyard B.

Marsh

For the Nuclear Regulatory Commission

John N. Hannon,

Acting Director, Division of Reactor Projects

- III/IV Office of Nuclear Reactor Regulation

[Doc. 95–23806 Filed 9–26–95; 8:45 am]

BILLING CODE 7590–01–F

[Docket Nos. STN 50–454, STN 50–455, STN

50–456 and STN 50–457]

Commonwealth Edison Company;

Notice of Consideration of Issuance of

Amendments to Facility Operating

Licenses, Proposed no Significant

Hazards Consideration Determination,

and Opportunity for A Hearing

The U.S. Nuclear Regulatory

Commission (the Commission) is

considering issuance of amendments to

Facility Operating License Nos. NPF–37,

NPF–66, NPF–72, and NPF–77, issued

to Commonwealth Edison Company for

operation of the Byron Station, Units 1

and 2, located in Ogle County, Illinois

and Braidwood Station, Units 1 and 2,

located in Will County, Illinois.

The proposed amendments would

revise the present voltage-based repair

criteria in the Byron 1 and Braidwood

1 Technical Specifications (TSs). These

proposed revisions would raise the

lower voltage limit from its present

value of 1.0 volt to 3.0 volts; there

would no longer be an upper voltage

limit.

The Braidwood 1 TSs were revised by

License Amendment No. 54, issued on

August 18, 1994, to add voltage-based

repair criteria to the existing steam

generator (SG) tube repair criteria. The

Byron 1 TSs were revised in a similar

manner by License Amendment No. 66,

issued on October 24, 1994.

The voltage-based repair criteria in

the subject TSs are applicable only to a

specific type of SG tube degradation

which is predominantly axially-oriented

outer diameter stress corrosion cracking

(ODSCC). This particular form of SG

tube degradation occurs entirely within

the intersections of the SG tubes with

the tube support plates (TSPs).

The present voltage values for the

ODSCC repair criteria are based on the

assumption of a ‘‘free span’’ exposure of

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

for any constraint against burst or

leakage, which may be provided by the

presence of the TSPs. This approach is,

in turn, based on the assumption that

under postulated accident conditions,

the TSPs may be displaced sufficiently

by blowdown hydrodynamic loads such

that a SG tube flaw which was fully

confined within the thickness of the

TSP prior to the accident would then be

fully exposed. This approach was first

advanced by the NRC staff in a draft

generic letter issued on August 12, 1994,

which was subsequently modified

slightly and issued as Generic letter (GL)

95–05, ‘‘Voltage-Based Repair Criteria

For Westinghouse Steam Generator

Tubes Affected by Outside Diameter

Stress Corrosion Cracking,’’ dated

August 3, 1995. The previous license

amendments related to the issue of

ODSCC were based to a large extent on

the draft generic letter cited above.

The fundamental difference between

the pending proposal to raise the lower

voltage repair limit to 3.0 volts and the

methodology contained in GL 95–05, is

that the licensee proposes to install

certain modifications to the SG internal

structures, thereby limiting to a small

value, the maximum displacement of

the TSPs under accident conditions.

The proposed structural modifications

consist of expanding a limited number

of SG tubes only on the hot leg side of

the TSP, at each of the intersections of

the tubes with the TSPs. The purpose of

this approach would be to greatly

reduce the probability of SG tube burst

under postulated accident conditions by

several orders of magnitude. There

would be a negligible impact on the

primary-to-secondary SG tube leakage

under accident conditions.

While the voltage-based repair criteria

for ODSCC flaws are applicable only to

Byron 1 and Braidwood 1, the pending

request for license amendments

involves all four units in that both

stations have a common set of TSs.

Before issuance of the proposed

license amendments, the Commission

will have made findings required by the

Atomic Energy Act of 1954, as amended

(the Act) and the Commission’s

regulations.

The Commission has made a

proposed determination that the

amendment request involves 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

amendments 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.

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 previously evaluated accidents of

interest are steam generator tube burst and

main steam line break [MSLB]. Their

potential impact on public health and safety

due to the change in SG tube plugging

criteria proposed in this amendment request

is very low as discussed below. Tube burst

related to the types of cracks under