Federal Register / Vol. 60, No. 50 / Wednesday, March 15, 1995 / Notices 14015

These time limits may be adjusted

depending on the number of

presentations and comment. The

workshop will be transcribed, and the

transcript will be available at the NRC

Public Document Room.

To foster meaningful discussions

during this session and to aid

participants in preparing their

presentations and comments,

participants should consider the

following set of questions:

• What impact will the CBLA

Administrative Letter have on those

organizations that the NRC regulates?

• Should the NRC develop a CBLA

database that could be made available to

the public?

• What are the reasons that the CBLA

program has not been used more widely

by licensees?

• What are the savings that can result

from conversion to the improved

Standard Technical Specifications?

Dated In Rockville, Maryland, this 9th day

of March , 1995.

For the Nuclear Regulatory Commission.

Eugene V. Imbro,

Director, RRG/CBLA Programs, Office of

Nuclear Reactor Regulation.

[FR Doc. 95–6341 Filed 3–14–95; 8:45 am]

BILLING CODE 7590–01–M

Biweekly Notice

Applications and Amendments to

Facility Operating Licenses Involving

No Significant Hazards Considerations

I. Background

Pursuant to Public Law 97-415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 97-415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from February 16,

1995, through March 3, 1995. The last

biweekly notice was published on

March 1, 1995.

Notice Of Consideration Of Issuance Of

Amendments To Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

And Opportunity For A Hearing

The Commission has made a

proposed determination that the

following amendment requests involve

no significant hazards consideration.

Under the Commission’s regulations in

10 CFR 50.92, this means that operation

of the facility in accordance with the

proposed amendment would not (1)

involve a significant increase in the

probability or consequences of an

accident previously evaluated; or (2)

create the possibility of a new or

different kind of accident from any

accident previously evaluated; or (3)

involve a significant reduction in a

margin of safety. The basis for this

proposed determination for each

amendment request is shown below.

The Commission is seeking public

comments on this proposed

determination. Any comments received

within 30 days after the date of

publication of this notice will be

considered in making any final

determination.

Normally, the Commission will not

issue the amendment until the

expiration of the 30-day notice period.

However, should circumstances change

during the notice period such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received before

action is taken. Should the Commission

take this action, it will publish in the

Federal Register a notice of issuance

and provide for opportunity for a

hearing after issuance. The Commission

expects that the need to take this action

will occur very infrequently.

Written comments may be submitted

by mail to the Rules Review and

Directives Branch, Division of Freedom

of Information and Publications

Services, Office of Administration, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555, and should cite

the publication date and page number of

this Federal Register notice. Written

comments may also be delivered to

Room 6D22, Two White Flint North,

11545 Rockville Pike, Rockville,

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

Federal workdays. Copies of written

comments received may be examined at

the NRC Public Document Room, the

Gelman Building, 2120 L Street, NW.,

Washington, DC. The filing of requests

for a hearing and petitions for leave to

intervene is discussed below.

By April 14, 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.
14016 Federal Register / Vol. 60, No. 50 / Wednesday, March 15, 1995 / Notices

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

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.

Baltimore Gas and Electric Company,

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

Cliffs Nuclear Power Plant, Unit Nos. 1

and 2, Calvert County, Maryland

Date of amendments request: January

31, 1995

Description of amendments request:

The proposed amendments would

revise the Technical Specifications

(TSs) for Calvert Cliffs, Unit Nos. 1 and

2, to increase the amount of Trisodium

Phosphate Dodecahydrate (TSP) located

in the containment sump baskets

required to be verified by TS

surveillance. The requested change is

the result of an reanalysis of the amount

of TSP necessary to maintain the

appropriate pH in the containment

sump water subsequent to a Loss of

Coolant Accident. Specifically, the

request would change the TS value of

TS 4.5.2.e.3 from the existing amount of

100 ft3 to 289 ft3. TS 4.5.2.e.4 would

also be changed by moving the amounts

of TSP and refueling water storage tank

water to be used in the required tests to

the TS Bases Section 3/4.5.2 and 3/

4.5.3. These Bases sections would also

be changed by modifying the test

methods.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Would not involve a significant increase

in the probability orconsequences of an

accident previously evaluated.

Trisodium Phosphate Dodecahydrate (TSP)

is stored in the containment lower level to

raise the pH of the sump and spray water

following a Loss of Coolant Accident (LOCA).

As the pH of the water increases, more

radioactive iodine is kept in solution and the

possibility of airborne radioactivity leakage is

decreased. An additional advantage of a

higher pH is the beneficial reduction in

chloride stress corrosion cracking of metal

components in the containment following an

accident.

This chemical is an accident mitigator, not

an accident initiator in that it is not used

until after an accident has occurred. At the

time it goes into solution, the accident has

occurred, containment spray has been

activated and water has collected in the

containment sump. Therefore, increasing the

Technical Specification minimum amount

verified to be in each containment will not

involve a significant increase in the

probability of an accident previously

evaluated.

Updated Final Safety Analysis Report,

Chapter 14.24, ‘‘Maximum Hypothetical

Accident’’, uses an assumption of a pre-RAS

minimum containment spray pH of 5.0 for

the iodine removal calculation and a postRAS sump pH of 7.0 for iodine retention.

Raising the pH to 7.0 does not increase the

consequences of an accident previously

evaluated.

The proposed change to Technical

Specification 4.5.2.e.4 would remove the

amounts of chemical and water used in the

test to the Bases. This relocation will not

alter the test method or acceptance criteria,

but will allow adjustments to the ratio of TSP

and borated water under the controls of 10

CFR 50.59 to reflect changes in plant

conditions. In the Bases, the amount of TSP

used in the test is changed to reflect the ratio

of TSP to water that would be found in the

containment following a LOCA. The

specified concentration of boron in the test

reflects the highest concentration that could

be found in the containment following a

LOCA. The test temperature is changed to

120°F which is well below the temperature

expected to be found in the containment

sump following a LOCA. The decanting of

the solution does not change the intent of the

test method since the dissolving period will

still be conducted without agitation.

Therefore, this change does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. Would not create the possibility of a new

or different type of accident from any

accident previously evaluated.
Federal Register / Vol. 60, No. 50 / Wednesday, March 15, 1995 / Notices 14017

The addition of more TSP does not

represent a significant change in the

configuration or operation of the plant.

Trisodium Phosphate Dodecahydrate is

currently present in the containment lower

level. There are no physical changes which

result from the increase in volume. The

proposed change to Technical Specification

4.5.2.e.4 would move the amounts of

chemical and water used in the test to the

Bases. This relocation will not alter the test

method or acceptance criteria, but will allow

adjustments to the ratio of TSP and borated

water under the controls of 10 CFR 50.59 to

reflect changes in plant conditions. In the

Bases, the amount of TSP used in the test is

changed to reflect the ratio of TSP to water

that would be found in the containment

following a LOCA. The specified

concentration of boron in the test reflects the

highest concentration that could be found in

the containment following a LOCA. The test

temperature is changed to 120°F which is

well below the temperature expected to be

found in the containment sump following a

LOCA. The decanting of the solution does

not change the intent of the test method since

the dissolving period will still be conducted

without agitation.

Therefore, this change would not create the

possibility of a new or different type of

accident from any accident previously

evaluated.

3. Would not involve a significant

reduction in a margin of safety.

Trisodium Phosphate Dodecahydrate is

stored in the containment lower level to raise

the pH of the sump and spray water

following a LOCA. As the pH of the water

increases, more radioactive iodine is kept in

solution and the possibility of airborne

radioactivity leakage is decreased.

Additionally, a higher pH has a beneficial

effect on chloride stress corrosion cracking of

metal components in the containment.

Technical Specification 4.5.2.e.3 requires

verification that a minimum volume of TSP

is contained in the storage baskets in each

containment. This change proposes to

increase that volume. The increased volume

will ensure the containment sump, when

filled with water, will have an acceptable pH

following a LOCA.

The proposed change to Technical

Specification 4.5.2.e.4 would move the

amounts of chemical and water used in the

test to the Bases. This relocation will not

alter the test method or acceptance criteria,

but will allow adjustments to the ratio of TSP

and borated water under the controls of 10

CFR 50.59 to reflect changes in plant

conditions. In the Bases, the amount of TSP

used in the test is changed to reflect the ratio

of TSP to water that would be found in the

containment following a LOCA. The

specified concentration of boron in the test

reflects the highest concentration that could

be found in the containment following a

LOCA. The test temperature is changed to

120°F which is well below the temperature

expected to be found in the containment

sump following a LOCA. The decanting of

the solution does not change the intent of the

test method since the dissolving period will

still be conducted without agitation.

Therefore, this change 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 amendments request

involves no significant hazards

consideration.

Local Public Document Room

location: Calvert County Library, Prince

Frederick, Maryland 20678.

Attorney for licensee: Jay E. Silbert,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Ledyard B.

Marsh

Boston Edison Company, Docket No.

50-293, Pilgrim Nuclear Power Station,

Plymouth County, Massachusetts

Date of amendment request: February

9, 1995

Description of amendment request:

The proposed amendment would

increase the Reactor High Water Level

Trip Level Setting for the Group 1

isolation. The change will allow an

increase to the main steam isolation

valve (MSIV) high water level isolation

setpoint.

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:

In accordance with 10 CFR 50.91, Boston

Edison submits the following analysis

addressing the no significant hazards

consideration. The proposed changes do not:

(1) Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Operation of the station in accordance with

the proposed Trip Level Setting will not

significantly increase the probability or

consequences of an accident previously

evaluated. The MSIV high water level

isolation signal is provided to protect against

rapid depressurization due to a pressure

regulator malfunction during plant startup.

The high water level isolation signal is not

functional when the mode switch is in the

RUN position. A high water level in the

reactor vessel indicates that fuel is covered.

Increasing the Trip Level Setting will have

minimal effect on moisture carryover in the

event of a pressure regulator failure at low

reactor power. MSIV closure (Group 1) is

initiated by low reactor pressure (810 psig)

approximately 30 seconds into the event. The

resulting reactor water level swell is not

sufficient to reach the bottom elevation of the

main steam lines.

The proposed Technical Specification

allowable value for the Reactor Low Level

Trip Level Setting and the Reactor Low Low

Water Level Trip Level setting does not

involve significant increase in the probability

or consequence of an accident.

(2) Create the possibility of a new or

different kind of accident from any

previously analyzed.

The proposed change does not affect the

Group 1 isolation safety function. The change

does not involve any plant hardware changes

that could introduce any new failure modes

or effects; thus, the change can not create the

possibility of a new or different kind of

accident from any previously analyzed.

(3) Involve a significant reduction in a

margin of safety.

The proposed change does not affect the

Group 1 isolation safety function. The

proposed change is consistent with the FSAR

[Final Safety Analysis Report] and Technical

Specification basis associated with reactor

vessel inventory control and main steam line

flooding.

The proposed change to the instrument

calibration range does not affect the margin

of safety for systems or components affected

by the change. Operating Pilgrim in

accordance with the proposed Trip Level

Setting 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 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: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts

02360.

Attorney for licensee: W. S. Stowe,

Esquire, Boston Edison Company, 800

Boylston Street, 36th Floor, Boston,

Massachusetts 02199.

NRC Project Director: Walter R. Butler

Carolina Power & Light Company, et

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

Nuclear Power Plant, Unit 1, Wake and

Chatham Counties, North Carolina

Date of amendment request: February

6, 1995

Description of amendment request:

The change proposes to relocate the

cycle specific core operating limits of

Figure 3.1-1, Shutdown Margin Versus

Boron Concentration, from Technical

Specification (TS) 3.1.1.2, Shutdown

Margins - Modes 3, 4, and 5, to the Core

Operating Limits Report (COLR).

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change of relocating TS

Figure 3.1-1, Shutdown Margin Versus Boron

Concentration to the COLR has no influence
14018 Federal Register / Vol. 60, No. 50 / Wednesday, March 15, 1995 / Notices

or impact to the probability or consequences

of an accident. The revised TS will continue

to implement the shutdown margin limits

through reference to the Shutdown Margin

Curve in the COLR. In addition, the COLR is

subject to the existing controls of TS 6.9.1.6.

Given that this change is an administrative

relocation of the Shutdown Margin Curve to

another TS controlled document, there

would be no increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

No safety-related equipment, safety

function, or plant operation will be altered as

a result of this proposed change. The TS will

continue to require operation within the

required core operating limits. Therefore, the

proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The proposed amendment does not

involve a significant reduction in the margin

of safety.

Relocation of the Shutdown Margin Curve

to the TS controlled COLR has no effect on

the core operating limits currently in force in

TS 3.1.1.2. Future revisions to the Shutdown

Margin Curve are governed by TS 6.9.1.6

which stipulates the specific TS that

reference the COLR limits and the

methodologies utilized in developing those

limits. Given that the change is an

administrative relocation of the Shutdown

Margin Curve to another TS controlled

document, the proposed changes do not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh,

North Carolina 27605

Attorney for licensee: R. E. Jones,

General Counsel, Carolina Power &

Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602

NRC Project Director: William H.

Bateman

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: January

12, 1995

Description of amendment request:

The proposed amendments would

revise and clarify portions of Technical

Specification (TS) Section 6.0,

‘‘Administrative Controls,’’ for the

McGuire, Catawba, and Oconee nuclear

stations. The licensee submitted a

combined amendment request covering

the three Duke Power nuclear stations.

The proposed changes are described

below.

1. Remove the specific assignment of

responsibilities for the review,

distribution, and approval activities

contained in the Technical Review and

Control Section of each station’s TS.

The proposed specifications state that

these activities will be performed by a

knowledgeable individual/organization.

Approval of the affected documents is to

be at the appropriate manager/

superintendent level as specified in

Duke administrative controls.

2. Move the requirement for the

review of proposed changes in the

stations’ TS and Operating Licenses by

the Duke Nuclear Safety Review Board

(NSRB) to Duke administrative

procedures (Selected Licensee

Commitments documents) and change

the wording of the requirements

covering NSRB meeting frequency. The

Oconee TS covering the NSRB are being

rewritten to be consistent with McGuire

and Catawba.

3. Add Technical Review and Control

Program implementation and Plant

Operations Review Committee (PORC)

implementation to the list of required

procedures and programs for each

nuclear station.

4. Change or clarify certain TS

administrative requirements covering

technical review and control activities

or records retention requirements.

5. For Oconee only, under ‘‘Station

Operating Procedures,’’ revise the TS

requirements covering the review and

approval of station procedures and

temporary procedure changes such that

these are now consistent with the

corresponding requirements for

McGuire and Catawba.

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:

(It should be noted that the licensee

submitted a combined analysis that covers

McGuire, Catawba, and Oconee nuclear

stations.)

Standard ι1. The proposed amendments

will not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The provisions of these proposed

amendments concern administrative changes

in the stations’ Technical Specifications

involving the Technical Review and Control,

Procedures and Programs/Station Operating

Procedures, and Records Retention/Station

Operating Records portions of the

Administrative Controls Section. The

requested changes primarily affect review

and control activities, but also include other

administrative changes affecting the approval

of station procedures (Oconee only), records

retention, and definition of the term ODCM

[offsite dose calculation manual] (McGuire

and [Catawba]). The provisions of the

proposed amendment primarily involve the

relocation of existing Technical

Specifications review, distribution, or

approval requirements to internal Duke

administrative controls. However,

implementation of the proposed amendment

does involve changes to several review/

distribution activities. Theses review/

distribution activities are primarily for: 1)

Proposed changes to the stations’ Technical

Specifications, 2) Proposed tests and

experiments which affect nuclear safety and

are not addressed in the stations’ FSAR

[Final Safety Analysis Report] or Technical

Specifications, 3) Environmental radiological

procedures, 4) Reportable events

documentation and reports of violations of

Technical Specifications, 5) Reports of

special reviews and investigations, and 6)

Reports of unplanned onsite releases of

radiological material to the environs. Planned

implementation of the proposed Technical

Specifications amendments utilizing Selected

Licensee Commitments will result in the

above items being reviewed/received by a

different organizational unit in the future.

The organizational unit is to be either the

recently initiated Plant Operations Review

Committee (PORC) or the General Manager,

Environmental Services. Personnel serving

on the PORC, and the General Manager,

Environmental Services will be qualified

based upon education and experience to

review the operational and technical

considerations involved with the applicable

items listed above. No required reviews are

being eliminated by the requested

amendments, only the organizational units

responsible for performing the reviews will

be changed. Future reviews of theses items

under the auspices of the PORC or the

General Manager, Environmental Services

will maintain a quality level equivalent to

that being currently achieved by Duke’s

Qualified Reviewer Program, the Station

Managers, or the

Duke Nuclear Safety Review Board as

applicable. Consequently, merely changing

the organizational units performing future

reviews, or making the additional

administrative changes described above,

results in no increase in the probability or

consequences of an accident previously

evaluated because the review function will

continue to be conducted in an equivalent

manner.

The implementing SLC will also permit

proposed amendments to the stations’

Technical Specifications and Operating

Licenses to be approved for the Station

Manager by a designee. However, this

individual will occupy a position equivalent

to, or higher, in the Duke organization as the

Station manager.

Additionally, the proposed changes do not

directly impact the design or operation of any

plant systems or components any more so

than the review and approval processes

currently being conducted in accordance
Federal Register / Vol. 60, No. 50 / Wednesday, March 15, 1995 / Notices 14019

with existing approved Technical

Specifications.

Standard ι2. The proposed amendments

will not create the possibility of a new or

different kind of accident from any

previously evaluated.

The proposed changes are administrative

in nature and primarily cover the review,

distribution, and/or approval function

performed for items identified in existing

Technical Specifications. The quality level of

the future reviews will not decrease and the

ability of Duke to identify the possibility for

the concurrence of new or different kinds of

accidents prior to implementation will be

maintained. Of specific interest in the

consideration of Standard ι2 is the review of

proposed tests and experiments which affect

station nuclear safety and are not addressed

in the FSAR or Technical Specifications. The

Technical Specifications required reviews of

these tests and experiments are not being

proposed for removal by these requested

amendments. Only the organizational unit

conducting the review of proposed tests and

experiments is being changed by the

requested amendments. The PORC, instead of

the Station Manager, is being assigned the

responsibility for conducting the reviews of

proposed tests and experiments in the future.

It is believed that the combined expertise of

the PORC membership will enhance Duke’s

ability to identify potential situations which

could possibly involve a new, or different,

kind of accident.

Standard ι3. The proposed amendments

will not involve a significant reduction in

any margin of safety.

The changes contained in the requested

amendments are administrative in nature and

do not impact the design capabilities or

operation of any plant structures, systems, or

components. There will be no reduction in

margin of safety as a result of implementing

these requested amendments. Impact upon

margin of safety is a consideration primarily

included in the 10 CFR 50.59 evaluation

process conducted for station procedures,

procedure changes, and nuclear station

modifications. The 10 CFR 50.59 evaluation

process in conducted under the auspices of

the Duke Qualified Reviewer Program and is

not affected by these requested amendments.

The impact on margin of safety for future

Technical Specifications and Operating

License changes will be reviewed by the

PORC, but these reviews will be equivalent

in quality to the reviews presently conducted

by the Qualified Reviewers.

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

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: January

13, 1995

Description of amendment request:

The proposed amendments would

increase the surveillance test intervals

and allowed outage times for Reactor

Trip System (RTS) and Engineered

Safety Features Actuation System

(ESFAS) equipment based upon

analyses by Westinghouse for the

Westinghouse Owners Group and

approved by the NRC. The proposed

changes to the RTS and ESFAS

instrumentation are based upon WCAP10271, its supplements, and the NRC’s

safety evaluation reports.

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 - Operation of McGuire in

accordance with the proposed license

amendment[s] [do] not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

The determination that the results of the

proposed changes are within all acceptable

criteria was established in the SERs prepared

for WCAP-10271, WCAP-10271 Supplement

1, WCAP-10271 Supplement 2, and WCAP10271 Supplement 2, Revision 1 issued by

letters dated February 21, 1985, February 22,

1989, and April 30, 1990. Implementation of

the proposed changes is expected to result in

an acceptable increase in total RTS yearly

unavailability. This increase, which is

primarily due to less frequent surveillance,

results in an increase of similar magnitude in

the probability of an Anticipated Transient

Without Scram (ATWS) and in the

probability of core melt resulting from an

ATWS and also results in a small increase in

core damage frequency (CDF) due to ESFAS

unavailability.

Implementation of the proposed changes is

expected to result in a significant reduction

in the probability of core melt from

inadvertent reactor trips. This is a result of

a reduction in the number of inadvertent

reactor trips (0.5 fewer inadvertent reactor

trips per unit per year) occurring during

testing of RTS instrumentation. This

reduction is primarily attributable to testing

in bypass and less frequent surveillance.

The reduction in core melt frequency from

inadvertent reactor trips is sufficiently large

to counter the increase in ATWS core melt

probability resulting in an overall reduction

in total core melt probability.

The values determined by the WOG and

presented in the WCAP for the increase in

CDF were verified by Brookhaven National

Laboratory (BNL) as part of an audit and

sensitivity analysis for the NRC staff. Based

on the small value of the increase compared

to the range of uncertainty in the CDF, the

increase is considered acceptable.

Changes to surveillance test frequencies for

the RTS [reactor trip system] interlocks do

not represent a significant reduction in

testing. The currently specified test interval

for interlock channels allows the surveillance

requirement to be satisfied by verifying that

the permissive logic is in its required state

using the permissive annunciator window.

The surveillance as currently required only

verifies the status of the permissive logic and

does not address verification of channel

setpoint or operability. The setpoint

verification and channel operability are

verified after a refueling shutdown. The

definition of the channel check includes

comparison of the channel status with other

channels for the same parameter. The

requirement to routinely verify permissive

status is a different consideration than the

availability of trip or actuation channels

which are required to change state on the

occurrence of an event and for which the

function availability is more dependent on

the surveillance interval. The change in

surveillance requirement to at least once

every refueling does not therefore represent

a significant change in channel surveillance

and does not involve a significant increase in

unavailability of the RTS.The proposed

changes do not result in an increase in the

severity or consequences of an accident

previously evaluated. Implementation of the

proposed changes affects the probability of

failure of the RTS but does not alter the

manner in which protection is afforded nor

the manner in which limiting criteria are

established.

Criterion 2 - The proposed license

amendment[s] [do] not create the possibility

of a new or different kind of accident from

any accident previously evaluated.

The proposed changes do not result in a

change in the manner in which the RTS

provides plant protection. No change is being

made which alters the functioning of the RTS

(other than in a test mode). Rather, the

likelihood or probability of the RTS

functioning properly is affected as described

above. Therefore, the proposed changes do

not create the possibility of a new or different

kind of accident.

The proposed changes do not involve

hardware changes except those necessary to

implement testing in bypass. Some existing

instrumentation is designed to be tested in

bypass and current Technical Specifications

allow testing in bypass. Testing in bypass is

also recognized by IEEE standards. Therefore,

testing in bypass has been previously

approved and implementation of the

proposed changes for testing in bypass does

not create the possibility of a new or different

kind of accident from any previously

evaluated. Furthermore, since the other

proposed changes do not alter the

functioning of the RTS, the possibility of a

new or different kind of accident from any

previously evaluated has not been created.

Criterion 3 - The proposed license

amendment[s] [do] not involve a significant

reduction in a margin of safety.

The proposed changes do not alter the

manner in which safety limits, limiting safety
14020 Federal Register / Vol. 60, No. 50 / Wednesday, March 15, 1995 / Notices

system setpoints, or limiting conditions for

operation are determined. The impact of

reduced testing other than as addressed

above is to allow a longer time interval over

which instrument uncertainties (e.g., drift)

may act. Experience has shown that the

initial uncertainty assumptions are valid for

reduced testing.

Implementation of the proposed changes is

expected to result in an overall improvement

in safety by:

1) Less frequent testing will result in fewer

inadvertent reactor trips and actuation of

Engineered Safety Features Actuation System

components.

2) Higher quality repairs leading to

improved equipment reliability due to longer

allowable repair times.

3) Improvements in the effectiveness of the

operating staff in monitoring and controlling

plant operation. This is due to less frequent

distraction of the operator and shift

supervisor to attend to instrumentation

testing.

The foregoing analysis demonstrates that

the proposed amendment[s] to McGuire’s

Technical Specifications [do] not involve a

significant increase in the probability or

consequences of a previously evaluated

accident, [do] not create the possibility of a

new or different kind of accident, and [do]

not involve a significant reduction in a

margin of safety.

Based upon the preceding analysis, Duke

Power Company concludes that the proposed

amendment[s] [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: 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

Duke Power Company, Docket Nos. 50-

269, 50-270 and 50-287, Oconee

Nuclear Station, Units 1, 2 and 3,

Oconee County, South Carolina

Date of amendment request: January

12, 1995

Description of amendment request:

The proposed amendments would

revise and clarify portions of Technical

Specification (TS) Section 6.0,

‘‘Administrative Controls,’’ for the

McGuire, Catawba, and Oconee nuclear

stations. The licensee submitted a

combined amendment request covering

the three Duke Power nuclear stations.

The proposed changes are described

below.

1. Remove the specific assignment of

responsibilities for the review,

distribution, and approval activities

contained in the Technical Review and

Control Section of each station’s TS.

The proposed specifications state that

these activities will be performed by a

knowledgeable individual/organization.

Approval of the affected documents is to

be at the appropriate manager/

superintendent level as specified in

Duke administrative controls.

2. Move the requirement for the

review of proposed changes in the

stations’ TS and Operating Licenses by

the Duke Nuclear Safety Review Board

(NSRB) to Duke administrative

procedures (Selected Licensee

Commitments documents) and change

the wording of the requirements

covering NSRB meeting frequency. The

Oconee TS covering the NSRB are being

rewritten to be consistent with McGuire

and Catawba.

3. Add Technical Review and Control

Program implementation and Plant

Operations Review Committee (PORC)

implementation to the list of required

procedures and programs for each

nuclear station.

4. Change or clarify certain TS

administrative requirements covering

technical review and control activities

or records retention requirements.

5. For Oconee only, under ‘‘Station

Operating Procedures,’’ revise the TS

requirements covering the review and

approval of station procedures and

temporary procedure changes such that

these are now consistent with the

corresponding requirements for

McGuire and Catawba.

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:

(It should be noted that the licensee

submitted a combined analysis that covers

McGuire, Catawba, and Oconee nuclear

stations.)

Standard ι1. The proposed amendments

will not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The provisions of these proposed

amendments concern administrative changes

in the stations’ Technical Specifications

involving the Technical Review and Control,

Procedures and Programs/Station Operating

Procedures, and Records Retention/Station

Operating Records portions of the

Administrative Controls Section. The

requested changes primarily affect review

and control activities, but also include other

administrative changes affecting the approval

of station procedures (Oconee only), records

retention, and definition of the term ODCM

[offsite dose calculation manual] (McGuire

and [Catawba]). The provisions of the

proposed amendment primarily involve the

relocation of existing Technical

Specifications review, distribution, or

approval requirements to internal Duke

administrative controls. However,

implementation of the proposed amendment

does involve changes to several review/

distribution activities. These review/

distribution activities are primarily for: 1)

Proposed changes to the stations’ Technical

Specifications, 2) Proposed tests and

experiments which affect nuclear safety and

are not addressed in the stations’ FSAR

[Final Safety Analysis Report] or Technical

Specifications, 3) Environmental radiological

procedures, 4) Reportable events

documentation and reports of violations of

Technical Specifications, 5) Reports of

special reviews and investigations, and 6)

Reports of unplanned onsite releases of

radiological material to the environs. Planned

implementation of the proposed Technical

Specifications amendments utilizing Selected

Licensee Commitments will result in the

above items being reviewed/received by a

different organizational unit in the future.

The organizational unit is to be either the

recently initiated Plant Operations Review

Committee (PORC) or the General Manager,

Environmental Services. Personnel serving

on the PORC, and the General Manager,

Environmental Services will be qualified

based upon education and experience to

review the operational and technical

considerations involved with the applicable

items listed above. No required reviews are

being eliminated by the requested

amendments, only the organizational units

responsible for performing the reviews will

be changed. Future reviews of these items

under the auspices of the PORC or the

General Manager, Environmental Services

will maintain a quality level equivalent to

that being currently achieved by Duke’s

Qualified Reviewer Program, the Station

Managers, or the Duke Nuclear Safety Review

Board as applicable. Consequently, merely

changing the organizational units performing

future reviews, or making the additional

administrative changes described above,

results in no increase in the probability or

consequences of an accident previously

evaluated because the review function will

continue to be conducted in an equivalent

manner.

The implementing SLC will also permit

proposed amendments to the stations’

Technical Specifications and Operating

Licenses to be approved for the Station

Manager by a designee. However, this

individual will occupy a position equivalent

to, or higher, in the Duke organization as the

Station Manager.

Additionally, the proposed changes do not

directly impact the design or operation of any

plant systems or components any more so

than the review and approval processes

currently being conducted in accordance

with existing approved Technical

Specifications.

Standard ι2. The proposed amendments

will not create the possibility of a new or

different kind of accident from any

previously evaluated.

The proposed changes are administrative

in nature and primarily cover the review,
Federal Register / Vol. 60, No. 50 / Wednesday, March 15, 1995 / Notices 14021

distribution, and/or approval function

performed for items identified in existing

Technical Specifications. The quality level of

the future reviews will not decrease and the

ability of Duke to identify the possibility for

the occurrence of new or different kinds of

accidents prior to implementation will be

maintained. Of specific interest in the

consideration of Standard ι2 is the review of

proposed tests and experiments which affect

station nuclear safety and are not addressed

in the FSAR or Technical Specifications. The

Technical Specifications required reviews of

these tests and experiments are not being

proposed for removal by these requested

amendments. Only the organizational unit

conducting the review of proposed tests and

experiments is being changed by the

requested amendments. The PORC, instead of

the Station Manager, is being assigned the

responsibility for conducting the reviews of

proposed tests and experiments in the future.

It is believed that the combined expertise of

the PORC membership will enhance Duke’s

ability to identify potential situations which

could possibly involve a new, or different,

kind of accident.

Standard ι3. The proposed amendments

will not involve a significant reduction in

any margin of safety.

The changes contained in the requested

amendments are administrative in nature and

do not impact the design capabilities or

operation of any plant structures, systems, or

components. There will be no reduction in

margin of safety as a result of implementing

these requested amendments. Impact upon

margin of safety is a consideration primarily

included in the 10 CFR 50.59 evaluation

process conducted for station procedures,

procedure changes, and nuclear station

modifications. The 10 CFR 50.59 evaluation

process is conducted under the auspices of

the Duke Qualified Reviewer Program and is

not affected by these requested amendments.

The impact on margin of safety for future

Technical Specifications and Operating

License changes will be reviewed by the

PORC, but these reviews will be equivalent

in quality to the reviews presently conducted

by the Qualified Reviewers.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Oconee County Library, 501

West South Broad Street, Walhalla,

South Carolina 29691

Attorney for licensee: J. Michael

McGarry, III, Winston and Strawn, 1200

17th Street, NW., Washington, DC 20036

NRC Project Director: Herbert N.

Berkow

Entergy Operations Inc., Docket No. 50-

382, Waterford Steam ElectricStation,

Unit 3, St. Charles Parish, Louisiana

Date of amendment request: January

27, 1995

Description of amendment request:

The requested change would modify

Section 5.3.1, Fuel Assemblies, of the

Waterford 3 technical specifications.

The requested change increases the

maximum enrichment for the spent fuel

pool and containment temporary storage

rack from 4.1 to 4.9 weight percent U235 when fuel assemblies contain fixed

poisons. Waterford 3 plans to use higher

enriched fuel in the next fuel cycle

(Cycle 8) to meet the energy plans and

maintain a reload batch size similar to

that used in Cycles 6 and 7.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The proposed change will increase the fuel

enrichment limit in order to meetthe cycle

energy requirements while maintaining fuel

batch sizes consistent with previous cycle

designs. The calculated k-effective, including

uncertainties, demonstrate substantial margin

to criticality in the storage racks for both

normal and accident conditions. No changes

to the facility are required. No new modes of

operating the fuel storage or transfer systems

are required, except a restriction to limit the

use of the new fuel vault to fuel with a

maximum enrichment of 4.1 weight percent

U-235. This restriction will be implemented

by administrative controls. Since the plant

equipment and operation are essentially the

same, there is no significant increase in the

probability of a criticality accident. Since a

criticality event is demonstrated to be

unfeasible, there are no increased adverse

consequences for such a postulated event.

As previously discussed, the proposed

change will not result in a physical change

to the facility nor will it result in a significant

change to the operation of the facility;

therefore, it does not create the possibility of

a new or different kind of accident from any

accident previously evaluated.

The proposed change has been analyzed to

establish a k-effective, including

uncertainties, at or below the NRC criticality

acceptance criteria of k-effective below 0.95

including uncertainties at the 95/95

probability/confidence level; therefore, there

is no reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, Louisiana 70122

Attorney for licensee: N.S. Reynolds,

Esq., Winston & Strawn 1400 L Street

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

NRC Project Director: William D.

Beckner

GPU Nuclear Corporation, et al.,

Docket No. 50-289, Three Mile Island

Nuclear Station, Unit No. 1, Dauphin

County, Pennsylvania

Date of amendment request: January

16, 1995

Description of amendment request:

The proposed amendment would revise

the TMI-1 Technical Specifications (TS)

to incorporate certain improvements

from the Revised Standard Technical

Specifications (TS) for Babcock &

Wilcox nuclear power plants (NUREG1430). The amendment would also

change the bases incorporating the

results of analyses to support allowance

for drift of the pressurizer code safety

valve setpoint. One of the proposed STS

improvements involves a change to

Chapter 6, Administrative Controls,

affecting both TMI-1 and TMI-2 TSs. A

separate notice of consideration of

issuance of amendment to facility

operating license is being issued for the

proposed TMI-2 TSs Change.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Operation of the facility in accordance

with the proposed amendment would not

involve a significant increase in the

probability of occurrence or the

consequences of an accident previously

evaluated.

The proposed amendments involve a) an

administrative change to both the TMI-1 and

TMI-2 Technical Specifications which is

consistent with the B&W Standard Technical

Specifications (STS), NUREG-1430, and b)

changes to the TMI-1 Technical

Specifications which are consistent with the

STS. This change does not involve any

change to system or equipment configuration.

The proposed amendment revises certain

surveillance requirements, extends certain

surveillance intervals as evaluated above, or

involves changes that are purely

administrative. The reliability of systems

and components relied upon to prevent or

mitigate the consequences of accidents

previously evaluated is not degraded by the

proposed changes. Assurance of system and

equipment availability is maintained.

Therefore, this change does not increase the

probability of occurrence or the

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 changes only

involve changes to surveillance requirements

that are consistent with STS and with the

ASME Code. No new failure modes are

created and thus the changes are bounded by

accidents previously evaluated.

3. Operation of the facility in accordance

with the proposed amendment would not
14022 Federal Register / Vol. 60, No. 50 / Wednesday, March 15, 1995 / Notices

involve a significant reduction in a margin of

safety. Each of these changes is compatible

with the STS and has been evaluated to

preserve the level of safety assured by the

current TS.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Government Publications

Section, State Library of Pennsylvania,

Walnut Street and Commonwealth

Avenue, Box 1601, Harrisburg, PA

17105.

Attorney for licensee: Ernest L. Blake,

Jr., Esquire, Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Phillip F.

McKee

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: January

20, 1995

Description of amendment request:

The proposed amendment would revise

the fire hazards analysis for the River

Bend Station (RBS) by allowing a

deviation from 10 CFR 50, Appendix R,

Section III.G.3 with respect to the

requirement for a fixed fire suppression

system in fire area C-17. This area

houses the control building heating,

ventilation and air conditioning (HVAC)

systems and the loss due to a fire could

cause the loss of main control room

habitability. C-17 does not have a fixed

fire suppression system but depends

upon the use of the existing remote

shutdown system as described in the

updated safety analysis report (USAR).

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 request does not involve a

significant increase in the probability or

consequences of accident previously

evaluated.

The event of concern is a fire in fire area

C-17. The low fire loading and sparse

concentration of exposed combustible

material in fire area C-17 would limit fire

spread. However, for this scenario all

equipment in fire area C-17 will be assumed

lost. Fire area C-17 contains the air handling

units for the main control room envelope.

The loss of both air handling units would

cause the control building chillers to stop

running due to a logic tie requiring air flow

through the air handling equipment for the

chilled water system to operate during

normal operation. The loss of the HVAC

system in the control building would cause

the main control room and the equipment

rooms to begin heating up if exposed to

design summer conditions. Operator actions

can be accomplished to minimize the heat up

rates for the rooms prior to the areas reaching

equipment temperature limits. This would

allow the operators to begin the shutdown

process from the main control room. If the

main control room continued to heat up, the

operators could accomplish the shutdown

using the remote shutdown system. HVAC

for the remote shutdown panel is located in

fire area C-4 and would not be damaged by

a fire in fire area C-17. Operation of the

control building HVAC system from the

remote shutdown panel bypasses the logic

between the chilled water system and the air

handling system. This would allow restart of

the HVAC system for all areas except the

main control room. The scenario would

conclude in a manner similar to that

described in RBS USAR Appendix 15A,

Event 52, ‘‘Reactor Shutdown From Outside

Main Control Room.’’

In summary, the probability of a fire

occurring in fire area C-17 is not increased.

However, if a fire were to occur in fire area

C-17 which caused the loss of main control

room HVAC, the remote shutdown system

would provide an acceptable method of

shutdown. The low fire loading and sparse

concentration of exposed combustible

material in fire area C-17 would limit fire

spread. Therefore, this request does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2) The request does not create the

possibility of occurrence of a new or different

kind of accident from any accident

previously evaluated.

The event of concern is a fire in fire area

C-17. Fire area C-17 does not have a fixed

suppression system as required by 10 CFR

50, Appendix R, Section III.G.3. Fire

suppression systems are generally used to

limit fire spread, once the heat of the fire

opens thermally sensitive sprinklers. The low

fire loading and sparse concentration of

exposed combustible material in fire area C17 would limit fire spread. However, for the

purpose of event analysis, all equipment in

fire area C-17 is assumed lost. Thus a fire in

fire area C-17 is bounded by the same

analysis with or

without a fixed suppression system in

terms of equipment availability.

The proposed method of shutdown for a

fire in fire area C-17 will be changed in that

the remote shutdown system will be credited.

Use of the remote shutdown system is

bounded by RBS USAR Appendix 15A, Event

52, ‘‘Reactor Shutdown From Outside Main

Control Room.’’ The HVAC for the remote

shutdown panel is located in fire area C-4

and would be undamaged by a fire in fire

area C-17. Operation of the control building

HVAC system from the remote shutdown

panel bypasses the logic between the chilled

water system and the air handling system.

This would allow restart of the HVAC system

for all areas except the main control room.

In summary, if a fire were to occur in fire

area C-17 which caused the loss of main

control room HVAC, the remote shutdown

system would provide an acceptable method

of shutdown. Since, for the purpose of event

analysis, all equipment in fire area C-17 is

assumed lost, a fire in fire area C-17 is

bounded by the same analysis with or

without a fixed suppression system in terms

of equipment availability. Therefore, this

request does not create the possibility of

occurrence of a new or different kind of

accident from any accident previously

evaluated.

3) The request does not involve a

significant reduction in a margin of safety.

In this case, the margin of safety is implicit

rather than being explicitly expressed as a

numerical value. An implicit margin of safety

involves conditions for NRC acceptance.

Since the RBS Technical Specification Bases

do not specifically address a margin of safety

for fire protection, the SAR, the NRC’s Safety

Evaluation Report (SER), and appropriate

other licensing basis documents were

reviewed to determine if the proposed

change would result in a reduction in a

margin of safety. As stated, in part, in

Attachment 4 to NPF-47:

EOI shall implement and maintain in effect

all provisions of the approved fire protection

program as described in the Final Safety

Analysis Report for the facility through

Amendment 22 and as approved in the SER

dated May 1984 and Supplement 3 dated

August 1985 subject to provisions 2 and 3....

As discussed in the Reason for Request,

SSER 3 dated August 1985 states, in part:

On the basis of its evaluation the staff finds

that the applicant’s fire protection program

with approved deviations is in conformance

with the guidelines of BTP CMEB 9.5-1,

sections III.G, III.J, and III.O of Appendix R

to 10CFR50, and GDC 3, and is, therefore,

acceptable.

Thus, the margin of safety in this case can

be defined as conformance with the specified

fire protection guidelines. 10 CFR 50,

Appendix R, Section III.G.3, requires, in part,

that alternative shutdown capability be

provided for areas where adequate separation

of redundant safe shutdown components

cannot be provided. In addition, fire

detection and a fixed fire suppression system

must be installed in the area, room, or zone

under consideration. Since fire area C-17

does not have a fixed suppression system,

use of the remote shutdown system for a fire

in this fire area would deviate from the

requirements of 10 CFR 50, Appendix R,

Section III.G.3. However, as discussed

previously, the low fire loading and sparse

amount of exposed combustibles compensate

for the lack of a fixed fire suppression

system. There is no adverse impact on the

ability to achieve and maintain safe

shutdown. Therefore, this request does not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the
Federal Register / Vol. 60, No. 50 / Wednesday, March 15, 1995 / Notices 14023

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Government Documents

Department, Louisiana State University,

Baton Rouge, Louisiana 70803

Attorney for licensee: Mark

Wetterhahn, Esq., Winston & Strawn,

1400 L Street, N.W., Washington, D.C.

20005

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: July 28,

1994

Description of amendment request:

The proposed amendment would add a

footnote to Technical Specifcaiton 3.5.C.

The footnote would state that the

operability of the feedwater coolant

injection (FWCI) system be independent

of its seismic capability.

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.

Any postulated failure in the non-seismic

portion of the FWCI subsystem may result in

a loss of feedwater flow transient. However

comparing the probability of occurrence of a

seismic event, any increase in the probability

of occurrence of a loss of feedwater event

would be small. The proposed change would

have no impact on the probability of

occurrence of any other accident, including

LOCAs [loss of coolant accidents].

The FWCI subsystem will continue to be

maintained as QA Category 1 (except for the

seismic attribute). Therefore, it will remain

available for accident mitigation for most

scenarios. Nevertheless, LOCA analyses have

been reevaluated to demonstrate that FWCI is

not necessary to show compliance with

10CFR50.46. Potentially limiting LOCA

scenarios have been analyzed without the

FWCI subsystem using an approved LOCA

methodology. An active single failure was

postulated in addition to not taking credit for

the FWCI subsystem. Based on the results of

these analyses, the current design basis large

and small break LOCAs remain bounding.

Moreover, FWCI is not credited in mitigating

any of the non-LOCA transients/accidents.

Safe shutdown following a seismic event

can be achieved using the LPCI [low pressure

coolant injection] and ESW [emergency

service water] systems, and the SRVs [safety

relief valves], which are all seismically

qualified. Therefore, the FWCI system is not

required to mitigate a seismic event.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

Seismic reclassification of portions of

FWCI does not create the possibility of a new

kind of an accident. The portion of the piping

up to the second isolation valve (from the

RPV [reactor pressure vessel]), is seismically

qualified and will remain classified as

seismic. This ensures that a postulated

failure in the non-seismic portion of piping

or components does not degrade containment

integrity or result in a blowdown of the RPV.

Consequential and environmental effects of a

FW [feedwater] piping failure have been

analyzed in the HELB [high energy line

break] program and have been found to be

acceptable.

3. Involve a significant reduction in the

margin of safety.

All accidents, including LOCAs, can be

mitigated without using FWCI. FWCI is also

not necessary for safe shutdown following a

seismic event. The intended function of the

FWCI subsystem is to reduce the likelihood

of core uncovery during the lifetime of the

plant. The CS [core spray] and LPCI

subsystems provide redundant and diverse

means of injecting water to the RPV. The

FWCI subsystem provides an additional

diverse means to inject water. Since FWCI

will be maintained QA Category 1 (except for

the seismic attribute), it will continue to

provide the additional diversity to the

injection systems. Considering the intended

function of the subsystem and the credit

taken in the accident analysis, reclassifying

FWCI to be non-seismic does not

significantly reduce 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 Resource Center,

Three Rivers Community-Technical

College, Thames Valley Campus, 574

New London Turnpike, Norwich, CT

06360.

Attorney for licensee: Ms. L. M.

Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company,

Post Office Box 270, Hartford, CT

06141-0270.

NRC Project Director: Phillip F.

McKee

Northern States Power Company,

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

Island Nuclear Generating Plant, Unit

Nos. 1 and 2, Goodhue County,

Minnesota

Date of amendment requests: January

9, 1995, as supplemented February 7,

1995

Description of amendment requests:

The proposed amendments would

revise Prairie Island Nuclear Generating

Plant Technical Specification (TS) 4.12,

‘‘Steam Generator Tube Surveillance,’’

to incorporate revised acceptance

criteria for steam generator tubes with

degradation in the tubesheet roll

expansion region. These criteria for

steam generator tube acceptance were

developed by Westinghouse Electric

Corporation and are known as F* (≥FStar’’) and L* (≥L-Star’’). These criteria

would be utilized to avoid unnecessary

plugging and sleeving of steam

generator tubes.

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[s] will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

F* Steam Generator Tube Repair Criteria

The supporting technical and safety

evaluations of the subject criterion

demonstrate that the presence of the

tubesheet will enhance the tube integrity in

the region of the hardroll by precluding tube

deformation beyond its initial expanded

outside diameter. The resistance to both tube

rupture and tube collapse is strengthened by

the presence of the tubesheet in that region.

The results of hardrolling of the tube into the

tubesheet is an interference fit between the

tube and the tubesheet. Tube rupture cannot

occur because the contact between the tube

and tubesheet does not permit sufficient

movement of tube material. The radial

preload developed by the rolling process will

secure a postulated separated tube end

within the tubesheet during all plant

conditions. In a similar manner, the

tubesheet does not permit sufficient

movement of tube material to permit

buckling collapse of the tube during

postulated LOCA [loss-of-coolant accident]

loadings.

The F* length of roll expansion is

sufficient to preclude tube pullout from tube

degradation located below the F* distance,

regardless of the extent of the tube

degradation. The existing Technical

Specification leakage rate requirements and

accident analysis assumptions remain

unchanged in the unlikely event that

significant leakage from this region does

occur. As noted above, tube rupture and

pullout is not expected for tubes using the F*

criterion. Any leakage out of the tube from
14024 Federal Register / Vol. 60, No. 50 / Wednesday, March 15, 1995 / Notices

within the tubesheet at any elevation in the

tubesheet is fully bounded by the existing

steam generator tube rupture analysis

included in the Prairie Island Plant USAR

[Updated Safety Analysis Report]. For plants

with partial depth roll expansion like Prairie

Island, a postulated tube separation within

the tube near the top of the roll expansion

(with subsequent limited tube axial

displacement) would not be expected to

result in coolant release rates equal to those

assumed in the USAR for a steam generator

tube rupture event due to the limited gap

between the tube and tubesheet. The

proposed plugging criterion does not

adversely impact any other previously

evaluated design basis accident.

Leakage testing of roll expanded tubes

indicates that for roll lengths approximately

equal to the F* distance, any postulated

faulted condition primary to secondary

leakage from F* tubes would be insignificant.

L* Steam Generator Tube Repair Criteria

The presence of the tubesheet enhances

steam generator tube integrity in the region

of the hardroll by precluding tube

deformation beyond its initial expanded

outside diameter. The resistance to both tube

rupture and tube collapse is strengthened by

the presence of the tubesheet in that region.

The result of the hardroll of the tube into the

tubesheet is an interference fit between the

tube and the tubesheet. Tube rupture cannot

occur because the contact between the tube

and tubesheet does not permit sufficient

movement of tube materials. In a similar

manner, the tubesheet does not permit

sufficient movement of tube material to

permit buckling collapse of the tube during

postulated LOCA loadings.

The type of degradation for which the L*

criteria has been developed (cracking with an

axial or near axial orientation) has been

found not to significantly reduce the axial

strength of a tube. An evaluation including

analysis and testing has been done to

determine the strength reduction for the axial

loads with simulated axial and near axial

cracks. This evaluation provided the basis for

the acceptance criteria for tube degradation

subject to the L* criteria.

The length of roll expansion above L* is

sufficient to preclude significant leakage

from tube degradation located below the L*

distance. The existing Technical

Specification leakage rate requirements and

accident analysis assumptions remain

unchanged in the unlikely event that

significant leakage from this region does

occur. As noted above, tube rupture and

pullout is not expected for tubes using the

alternate plugging criteria.

Any leakage out of the tube from within

the tubesheet at any elevation in the

tubesheet is fully bounded by the existing

steam generator tube rupture analysis

included in the Prairie Island Updated Safety

Analysis Report. The proposed alternate

plugging criteria do not adversely impact any

other previously evaluated design basis

accident.

2. The proposed amendment[s] will not

create the possibility of a new or different

kind of accident from any accident

previously analyzed.

F*

Implementation of the proposed F*

criterion does not introduce any significant

changes to the plant design basis. Use of the

criterion does not provide a mechanism to

initiate an accident outside of the region of

the expanded portion of the tube. Any

hypothetical accident as a result of any tube

degradation in the expanded portion of the

tube would be bounded by the existing tube

rupture accident analysis. Tube bundle

structural integrity will be maintained. Tube

bundle leaktightness will be maintained such

that any postulated accident leakage from F*

tubes will be negligible with regards to offsite

doses.

L*

Implementation of the proposed alternate

tubesheet tube plugging criteria does not

introduce changes to the plant design basis.

Use of the criteria does not provide a

mechanism to result in an accident outside

of the region of the tubesheet expansion. Any

hypothetical accident as a result of any tube

degradation in the expanded portion of the

tube would be bounded by the existing tube

rupture accident analysis.

3. The proposed amendment[s] will not

involve a significant reduction in the margin

of safety.

F*

The use of the F* criterion has been

demonstrated to maintain the integrity of the

tube bundle commensurate with the

requirements of Reg Guide 1.121 [≥Bases for

Plugging Degraded PWR Steam Generator

Tubes≥] (intended for indications in the free

span of tubes) and the primary to secondary

pressure boundary under normal and

postulated accident conditions. Acceptable

tube degradation for the F* criterion is any

degradation indication in the tubesheet

region, more than the F* distance below the

bottom of the transition between the roll

expansion and the unexpanded tube. The

safety factors used in the verification of the

strength of the degraded tube are consistent

with the safety factors in the ASME Boiler

and Pressure Vessel Code used in steam

generator design. The F* distance has been

verified by testing to be greater than the

length of roll expansion required to preclude

both tube pullout and significant leakage

during normal and postulated accident

conditions. Resistance to tube pullout is

based upon the primary to secondary

pressure differential as it acts on the surface

area of the tube, which includes the tube wall

cross-section, in addition to the inner

diameter based area of the tube. The leak

testing acceptance criteria are based on the

primary to secondary leakage limit in the

Technical Specifications and the leakage

assumptions used in the USAR accident

analysis.

Implementation of the tubesheet plugging

criterion will decrease the number of tubes

which must be taken out of service with tube

plugs or repaired with sleeves. Both plugs

and sleeves reduce the RCS (reactor coolant

system) flow margin; thus, implementation of

the F* criterion will maintain the margin of

flow that would otherwise be reduced in the

event of increased plugging or sleeving.

Based on the above, it is concluded that the

proposed change does not result in a

significant reduction in margin with respect

to plant safety as defined in the USAR or the

Technical Specification Bases.

L*

The use of the alternate tubesheet plugging

criteria has been demonstrated to maintain

the integrity of the tube bundle

commensurate with the requirements of Reg.

Guide 1.121 for indications in the free span

of tubes and the primary to secondary

pressure boundary under normal and

postulated accident conditions. Acceptable

tube degradation for the L* criteria is any

degradation indication with axial or nearly

axial cracking in the tubesheet region, more

than the L* distance below the bottom of the

transition between the roll expansion and the

unexpended tube. For tubes with axial or

nearly axial cracks the strength of the tube

relative to an axial load would not be

reduced below the strength required to resist

potential axial loads. The safety factors used

in the verification of the strength of the

degraded tube are consistent with the safety

factors in the ASME Boiler and Pressure

Vessel Code used in steam generator design.

The L* distance has been verified by testing

to be greater than the length of roll expansion

required to preclude significant leakage

during normal and postulated accident

conditions. The leak testing acceptance

criteria are based on the primary to

secondary leakage limit in the Technical

Specifications and the leakage assumptions

used in the USAR accident analyses.

Implementation of the proposed tubesheet

plugging criteria will decrease the number of

tubes which must be taken out of service

with tube plugs or repaired with sleeves.

Both plugs and sleeves reduce the RCS flow

margin, thus implementation of the alternate

plugging criteria will maintain the margin of

flow that would otherwise be reduced in the

event of increased plugging or sleeving.

Based on the above, it is concluded that the

proposed change does not result in a

significant reduction in margin with respect

to plant safety as defined in the Updated

Safety Analysis Report or the bases of the

Technical Specifications.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment requests involve no

significant hazards consideration. This

notice supersedes the staff’s previous

notice which was published in the

Federal Register February 1, 1995 (60

FR 6307).

Local Public Document Room

location: Minneapolis Public Library,

Technology and Science Department,

300 Nicollet Mall, Minneapolis,

Minnesota 55401

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts, and Trowbridge,

2300 N Street, NW, Washington, DC

20037

NRC Project Director: Cynthia

Carpenter, Acting
Federal Register / Vol. 60, No. 50 / Wednesday, March 15, 1995 / Notices 14025

Northern States Power Company,

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

Island Nuclear Generating Plant, Unit

Nos. 1 and 2, Goodhue County,

Minnesota

Date of amendment requests:

February 23, 1995

Description of amendment requests:

The proposed amendments would

revise the wording in the Prairie Island

technical specifications to allow

implementation of exemptions to the

schedule requirements of 10 CFR Part

50, Appendix J. A related exemption

request would grant temporary relief

from the requirements of 10 CFR Part

50, Appendix J, Section III.D.1.(a) which

requires Prairie Island Unit 2 to perform

a Type A test in the May 1995 refueling

outage.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed amendment[s] will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed amendment is an

administrative change which allows

implementation of approved exemptions to

the regulations and by itself does not change

any retest schedules.

Therefore, the probability or consequences

of an accident previously evaluated are not

affected by the proposed amendment.

2. The proposed amendment[s] will not

create the possibility of a new or different

kind of accident from any accident

previously analyzed.

The proposed amendment is an

administrative change which allows

implementation of approved exemptions to

the regulations and by itself does not change

any retest schedules.

Therefore, the possibility of a new or

different kind of accident from any accident

previously evaluated would not be created by

the proposed amendment.

3. The proposed amendment[s] will not

involve a significant reduction in the margin

of safety

The proposed amendment is an

administrative change which allows

implementation of approved exemptions to

the regulations and by itself does not change

any retest schedules.

Therefore, a significant reduction in the

margin of safety would not be involved with

the proposed amendment.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment requests involve no

significant hazards consideration.

Local Public Document Room

location: Minneapolis Public Library,

Technology and Science Department,

300 Nicollet Mall, Minneapolis,

Minnesota 55401

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts, and Trowbridge,

2300 N Street, NW, Washington, DC

20037

NRC Project Director: Cynthia

Carpenter, Acting

Omaha Public Power District, Docket

No. 50-285, Fort Calhoun Station,Unit

No. 1, Washington County, Nebraska

Date of amendment request: February

10, 1995

Description of amendment request:

The proposed amendment to the

technical specifications (TSs) would

relocate the requirements for the incore

instrumentation (ICI) system from the

TS to the Updated Safety Analysis

Report (USAR).

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) The proposed changes do not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The Incore Instrumentation (ICI) System is

used to measure core power distribution for

the purpose of Limiting Conditions for

Operation (LCO) monitoring of Technical

Specification (TS) limits on linear heat rate,

unrodded planer radial peaking factor,

unrodded integrated radial peaking factor,

and azimuthal power tilt. The ICI System has

no safety purpose itself; it measures

parameters which have safety significance.

No change to the monitored parameters is

proposed. The proposed changes will

relocate requirements on the number and

distribution of incore detectors used by the

ICI System when measuring these parameters

from the TS to the Updated Safety Analysis

Report (USAR). Changes to the requirements

can be made without NRC approval when the

changes meet the criteria of 10 CFR 50.59.

Changes to the ICI System requirements that

do not meet the criteria of 10 CFR 50.59 must

be approved by the NRC by license

amendment.

Relocation of the requirements on the ICI

System from the TS to the USAR does not

increase the probability or consequences of

any accident previously analyzed because the

ICI System is neither a precursor nor a

mitigator for any analyzed accident. The ICI

System is used to ensure that operation

within the LCOs for linear heat rate,

unrodded planer radial peaking factor,

unrodded integrated radial peaking factor,

and azimuthal power tilt is maintained.

However, its operation serves no mitigation

function associated with any USAR Section

14 accident analysis. The parameters

measured by the ICI System are important

parameters in many accident analyses;

however, this proposed change does not

remove or revise the limits on these

parameters.

Additionally, it is proposed to revise TS

2.10.4(1)(b) to clarify its requirements.

Currently TS 2.10.4(1) part (b) applies while

operating under the provisions of part (a) if

the plant computer incore detector alarms

become inoperable. This is incorrect in that

part (a) applies when the linear heat rate is

being monitored by the ICI System and the

linear heat rate is exceeding its limits as

indicated by valid detector alarms. Part (b) of

this specification should apply only if the

linear heat rate is being monitored by the ICI

System, is within its limits, and the plant

computer incore detector alarms are

inoperable.

Administrative changes are also proposed

which correct grammar and renumber/

relocate portions of the TS and bases to other

TS, to correspond to the proposed change to

relocate ICI System requirements.

Therefore, the proposed change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

(2) The proposed changes do not create the

possibility of a new or different kind of

accident from any previously evaluated.

The ICI System will continue to be used to

monitor TS limits on core power distribution.

There will be no physical alterations to the

plant configuration, changes to setpoint

values, or changes to the implementation of

setpoints or limits as a result of this proposed

change.

The proposed change to TS 2.10.4(1)(b)

only clarifies its requirements. The proposed

change is more restrictive in that TS

2.10.4(1)(b), as currently written, could be

interpreted to allow continued operation for

up to seven days with the linear heat rate

exceeding its limits. The proposed change

clarifies this specification to ensure that TS

2.10.4(1)(a) is applied if the linear heat rate

is exceeded while being monitored by the ICI

System. TS 2.10.4(1)(a) requires that the

linear heat rate be restored within one hour

or a plant shutdown initiated.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any previously

evaluated.

(3) The proposed changes do not involve

a significant reduction in a

margin of safety.

The ICI System is used to measure core

power distribution parameters which are a

direct measure of the margin of safety. The

limits on these parameters are not changed.

Therefore, the proposed change (i.e.,

relocation of the ICI System operability

requirements to the USAR and/or plant

procedures) does not involve a significant

reduction in a margin of safety.

The proposed change to TS 2.10.4(1)(b)

helps ensure that the margin of safety is

maintained by clarifying when the TS is

applicable. This clarification ensures that the

more restrictive actions of TS 2.10.4(1)(a) are

taken if the linear heat rate is exceeded while

being monitored by the ICI System.

Therefore, the proposed change does not

involve a significant reduction in a margin of

safety.
14026 Federal Register / Vol. 60, No. 50 / Wednesday, March 15, 1995 / Notices

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska

68102

Attorney for licensee: LeBoeuf, Lamb,

Leiby, and MacRae, 1875 Connecticut

Avenue, NW., Washington, DC 20009-

5728

NRC Project Director: Theodore R.

Quay

Pacific Gas and Electric Company,

Docket Nos. 50-275 and 50-323,

DiabloCanyon Nuclear Power Plant,

Unit Nos. 1 and 2, San Luis Obispo

County, California

Date of amendment requests:

December 30, 1994 (Reference LAR 94-

12)

Description of amendment requests:

The proposed amendments would

revise the combined Technical

Specifications (TS) for the Diablo

Canyon Nuclear Power Plant, Unit Nos.

1 and 2, to revise TS 2.2, 3/4.3.1, 3/

4.3.2, 3/4.3.3, 3/4.4.4, 3/4.4.9, 3/4.5.2, 3/

4.8.1, 3/4.8.2, 3/4.9.2, 3/4.9.9, and 3/

4.10.3. The specific TS changes

proposed are as follows:

(1) The TS issued in License

Amendments (LAs) 84/83 would be

changed to (a) revise the value of the

overpower Delta-temperature (OPDT)

constant K6 in TS 2.2.1, Table 2.2-1,

Note 3; (b) revise the reactor coolant

system (RCS) loop Delta-T function; and

(c) make editorial corrections for

clarification and consistency to TS 2.2.1

(and TS 2.2.1 Bases), TS 3/4.3.1, and TS

3/4.3.2.

In revising the RCS loop Delta-T

function, the licensee would (a)

incorporate the 0.99 multiplying factor

listed in TS 2.2.1, Table 2.2-1, Note 5,

and TS 3/4.3.2, Table 3.3-4, Note 2, into

constants B1 through B4; (b) change

‘‘Steam Generator (SG) Water Level

Low-Low’’ in TS 3/4.3.2, Table 3.3-3

and Table 4.3-2, Functional Unit 6.c,

‘‘Auxiliary Feedwater’’ (AFW), by

deleting the Mode 3 applicability of the

RCS loop Delta-T function and by

adding a footnote to the Mode 3

applicability of the SG water level lowlow function requiring that the trip time

delay (TTD) associated with the SG

water level low-low channel be less

than or equal to 464.1 seconds; (c)

change TS 3/4.3.1, Table 3.3-1, Action

27, and TS 3/4.3.2, Table 3.3-3, Action

29, by allowing up to four RCS loop

Delta-T channels to be inoperable with

the TTD threshold power level for zero

seconds time adjusted to 0-percent rated

thermal power (RTP) and by allowing

the affected SG water level low-low

channels to be placed in the tripped

condition, with one inoperable RCS

loop Delta-T channel; and (d) change

the Table 3.3-1 and Table 3.3-3

‘‘Channels to Trip’’ and ‘‘Minimum

Channels Operable’’ columns to not

applicable (N.A.).

(2) The TS issued in LAs 70/69 would

be changed to (a) delete references to the

plant vent noble gas activity monitors

(RM-14A and RM-14B) and footnote

references to applicability of the

containment ventilation exhaust

radiation monitors (RM-44A and RM44B) in TS Tables 3.3-3, 3.3-4, 3.3-5, 3.3-

6, 4.3-2, and 4.3-3 and TS 4.9.9; and (b)

revise the ‘‘Trip Setpoint and Allowable

Values’’ column in TS Table 3.3-4,

Functional Unit 3.c.4), to reference the

offsite dose calculation procedure

(ODCP).

(3) Cycle-specific information in TS

4.3.2.1, TS 3.3.3.6, TS 4.4.4.1, TS 4.5.2,

TS 3.8.1.1, TS 3.8.2.1, and TS 3.8.2.2

that is no longer necessary would be

deleted.

(4) The word ‘‘analog’’ would be

deleted from TS 4.4.9.3.1, TS 4.9.2, and

TS 4.10.3.2.

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:

a. Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

The proposed change to the OPDT constant

K6 is conservative and will not cause any

design or analysis acceptance criteria to be

exceeded. There is no effect on the structural

and functional integrity of any plant system.

The OPDT function is part of the accident

mitigation response and is not itself an

initiator for any transient. This change does

not affect the integrity of the fission product

barriers for mitigation of radiological dose

consequences as a result of an accident.

The proposed change to incorporate the

0.99 multiplier into the TTD constants is an

administrative change and has no effect on

plant operation. The proposed change to

delete Mode 3 applicability of the RCS Loop

Delta-T function does not affect any design

or analysis results. Allowing up to 4 RCS

Loop Delta-T channels to be inoperable with

the TTD threshold power level for zero

seconds time delay adjusted to 0% RTP is

conservative with respect to ESFs

[engineered safety features] and reactor trip

actuation time. Allowing the SG [steam

generator] water level low-low channels

affected by the inoperable RCS Loop DeltaT channels to be placed in the tripped

condition is also conservative with respect to

reactor trip and AFW pumps start. The

change to the Channels to Trip and Minimum

Channels Operable columns is a clarifying

change to reflect the proposed changes to the

action statements and identifies that the RCS

Loop Delta-T does not provide a reactor trip

function. Therefore, the proposed changes to

the RCS Loop Delta-T function do not affect

any of the accident analysis results.

The proposed changes to revise Table 3.3-

4, Functional Unit 3.c.4), and to delete cyclespecific TS, TS references to RM-14A and

RM-14B, and the word ‘‘analog’’ from the

analog channel operation test are

administrative and have no effect on plant

operation.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

b. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

The proposed change to the OPDT constant

K6 does not affect the assumed accident

initiation sequences. No new operating

configuration is being imposed by the change

to K6 that would create a new failure

scenario. No new failure modes are being

created for any plant equipment.

The proposed changes to the RCS Loop

Delta-T function do not involve any physical

modification to any plant system or change

the methodology by which any safety-related

system performs its function.

1The proposed changes to revise Table 3.3-

4, Functional Unit 3.c.4), and to delete cyclespecific TS, TS references to RM-14A and

RM-14B, and the word ‘‘analog’’ from the

analog channel operation test are

administrative, would not result in any

physical alteration to any plant system, and

would not be a change in the method by

which any safety-related system performs its

function.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

c. Does the change involve a significant

reduction in a margin of safety?

The proposed change to the OPDT constant

K6 will not affect any accident analysis

assumptions, initial conditions, or results.

The proposed changes to the RCS Loop

Delta-T function do not affect any accident

analysis assumptions, initial conditions, or

results.

The proposed changes to revise Table 3.3-

4, Functional Unit 3.c.4), and to delete cyclespecific TS, TS references to RM-14A and

RM-14B, and the word ‘‘analog’’ from the

analog channel operation test are

administrative and clarify the TS. These

proposed changes have no effect on current

operating methodologies or actions that

govern plant performance.

Therefore, the proposed changes do not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment requests
Federal Register / Vol. 60, No. 50 / Wednesday, March 15, 1995 / Notices 14027

involve no significant hazards

consideration.

Local Public Document Room

location: California Polytechnic State

University, Robert E. Kennedy Library,

Government Documents and Maps

Department, San Luis Obispo, California

93407

Attorney for licensee: Christopher J.

Warner, Esq., Pacific Gas and Electric

Company, P.O. Box 7442, San

Francisco, California 94120

NRC Project Director: Theodore R.

Quay

Philadelphia Electric Company, Public

Service Electric and Gas

Company,Delmarva Power and Light

Company, and Atlantic City Electric

Company,Dockets Nos. 50-277 and 50-

278, Peach Bottom Atomic Power

Station,Units Nos. 2 and 3, York

County, Pennsylvania

Date of application for amendments:

September 26, 1994

Description of amendment request:

The proposed TS changes extend

surveillance test intervals and allowable

out-of-service times for the testing and/

or repair of instrumentation that actuate

the Reactor Protection System, Primary

Containment Isolation, Core and

Containment Cooling systems, Control

Rod Blocks, Radiation Monitoring

systems, and Alternate Rod Insertion/

Recirculation Pump Trip.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed TS changes increase the STIs

and AOTs for actuation instrumentation

based on analyses described and justified in

Licensing Topical Reports (References 2

through 8) [see licensee’s September 26, 1994

application for reference information] which

have been evaluated in associated Safety

Evaluation Reports. These changes were

incorporated into PBAPS Technical

Specifications consistent with NUREG-1433.

TS requirements that govern Operability or

routine testing of plant instruments are not

assumed to be initiators of any analyzed

event because these instruments are intended

to prevent, detect or mitigate accidents.

Therefore, these changes will not involve an

increase in the probability of occurrence of

an accident previously evaluated.

Additionally, these changes will not increase

the consequences of an accident previously

evaluated because the proposed change will

not involve any physical changes to plant

systems, structures, or components (SSC), or

the manner in which these SSC are operated,

maintained, modified, or inspected. The

changes will not alter the operation of

equipment assumed to be available for the

mitigation of accidents or transients by the

plant safety analysis or licensing basis. As

justified in References 1 through 8, the

proposed changes establish or maintain

adequate assurance that components are

operable when necessary for the prevention

or mitigation of accidents or transients and

that plant variables are maintained within

limits necessary to satisfy the assumptions

for initial conditions in the safety analyses.

These changes establish or modify time

limits allowed for operation with inoperable

instrument channels based on the analyses in

References 1 through 8 and will not allow

continuous plant operation with plant

conditions such that a single failure will

result in a loss of any safety function.

Therefore, these changes will not increase the

consequences of an accident previously

evaluated.

2) The proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

These proposed changes will not involve

any physical changes to SSC, or the manner

in which these SSC are operated, maintained,

modified, tested, or inspected. Therefore,

these changes will not create the possibility

of a new or different kind of accident from

any accident previously evaluated. The

changes in methods governing normal plant

operation are consistent with the current

safety analysis assumptions. Therefore, these

changes will not create the possibility of a

new or different kind of accident from any

accident previously evaluated.

3) The proposed changes do not involve a

significant reduction in a margin of safety.

The proposed TS changes increase the STIs

and AOTs for actuation instrumentation

based on analyses described and justified in

Licensing Topical Reports (References 2

through 8) which have been evaluated in

associated Safety Evaluation Reports. These

changes were incorporated into PBAPS

Technical Specifications consistent with

NUREG-1433. These changes can be

classified into one of the following three

categories:

a. Changes to the minimum STIs and AOTs

for the testing and/or repair of

instrumentation based on the results of

generic analyses in References 1 through 8;

b. Changes to conditions, required actions,

and completion times needed to make

PBAPS TS requirements consistent with the

assumptions used in the analyses in

References 1 through 8; and,

c. Changes that reformat, renumber, and/or

reword existing requirements to incorporate

the changes above.

All of the proposed changes will be

incorporated into the PBAPS custom

Technical Specifications using the same

approach and specific requirements used in

Reference 12.

There is no significant reduction in the

margin of safety resulting from changes to the

STIs and AOTs for the testing and/or repair

of instrumentation based on the results of the

analyses in References 1 through 8. These

analyses determined that there is no

significant change in the availability and/or

reliability of instrumentation as a result of

this change in STIs and AOTs. PECO Energy

performed reviews that confirmed these

analyses are applicable to PBAPS and that

there would be no effect on the identification

of excessive instrument setpoint drift as a

result of increasing from monthly to quarterly

the minimum interval between instrument

functional tests. The proposed required

actions ensure that actions to mitigate loss of

single failure tolerance is initiated within 24

hours (12 hours for RPS) in accordance with

the results of the analyses in References 1

through 8 and action to mitigate a loss of

instrument function is initiated within 1

hour.

The proposed changes which replace the

shutdown actions associated with inoperable

instrumentation with actions to declare the

supported system inoperable does not

involve a reduction in a margin of safety. The

proposed changes ensure that appropriate

compensatory measures are taken

commensurate with approved TS Actions for

the affected systems and the safety analyses.

In addition, the proposed changes provide

the benefit of avoiding an unnecessary

shutdown transient when appropriate

measures are available to compensate for the

inoperable instrumentation.

There is no significant reduction in the

margin of safety resulting from changes that

reformat, renumber, and/or reword existing

requirements to incorporate the changes

above.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Education

Building, Walnut Street and

Commonwealth Avenue, Box 1601,

Harrisburg, Pennsylvania 17105.

Attorney for licensee: J. W. Durham,

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

Counsel, Philadelphia Electric

Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101

NRC Project Director: John F. Stolz

Philadelphia Electric Company, Public

Service Electric and Gas

Company,Delmarva Power and Light

Company, and Atlantic City Electric

Company,Dockets Nos. 50-277 and 50-

278, Peach Bottom Atomic Power

Station,Units Nos. 2 and 3, York

County, Pennsylvania

Date of application for amendments:

November 17, 1994

Description of amendment request:

The proposed changes to the Technical

Specifications (TS) are being requested

to support modifications 5384 and 5386

which upgrade the Main Stack and Vent

Stack Radiation Monitoring Systems.
14028 Federal Register / Vol. 60, No. 50 / Wednesday, March 15, 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 changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

Neither the Main Stack nor the Vent Stack

Radiation Monitoring Systems serve as an

initiator or contributor to any accidents

previously evaluated. The systems provide

indication and detection of radioactivity and

effluent release in the main and vent stacks.

The new systems perform the same function

as the old, and have equal or better

performance characteristics. Installation and

operation of the new radiation monitoring

systems do not degrade any active or passive

equipment that responds to an accident.

The proposed increase in the surveillance

test interval of the subject radiation

monitoring systems from 12 to 18 months is

consistent with vendor recommendations,

and is based on operating experience with

instrumentation of a similar design.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed changes do not create the

possibility of a new or different kind of

accident from any previously evaluated.

Both modifications replace obsolete

radiation monitoring equipment and have the

same failure modes as the existing

equipment. The upgraded systems are

considered enhancements to the existing

systems and are considered neither a

contributor nor initiator of any accidents

previously evaluated.

Based on the above, the proposed changes

do not create the possibility of a new or

different kind of accident from any

previously evaluated.

3. The proposed changes do not involve a

significant reduction in a margin of safety.

Neither the accuracy nor the

responsiveness of the existing radiation

monitoring equipment will be degraded as a

result of the installation of modifications

5384 and 5386. Revisions to the calibration

and surveillance frequencies are based on

vendor information and experience with

instrumentation of similar design. The

changes associated with setpoints and the

lower limit of detection are in the

conservative direction. The upgraded main

stack system continues to provide a nonsafety related trip signal to Group III isolation

valves during purging of the containment

through the SBGTS [standby gas treatment

system]. The revisions to parameter

descriptions and instrument designation are

considered administrative.

Therefore, based on the above, the

proposed changes do not involve a

significant reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Education

Building, Walnut Street and

Commonwealth Avenue, Box 1601,

Harrisburg, Pennsylvania 17105.

Attorney for licensee: J. W. Durham,

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

Counsel, Philadelphia Electric

Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101

NRC Project Director: John F. Stolz

Public Service Electric & Gas Company,

Docket No. 50-311, Salem Nuclear

Generating Station, Unit No. 2, Salem

County, New Jersey

Date of amendment request: February

3, 1994, supplemented September 19,

1994, and November 23, 1994

Description of amendment request:

The proposed amendment revises the

Technical Specifications to reflect a

reduction in the Reactor Coolant System

flow.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Involve a significant increase in the

probability or consequences of an accident

previously analyzed.

No component modification, system

realignment, or change in operations will

occur which could affect the probability of

any accident or transient. The proposed

reduction in RCS loop and total flow rates

will not change the probability of a challenge

to any Engineered Safeguard Feature or other

device. The consequences of previously

analyzed accidents have been found to

remain within acceptable licensing basis

limits when the reduced flow rates are

assumed. The system transient response is

not affected by the initial RCS flow

assumption, unless the initial assumption is

so low as to impair the steady-state core

cooling capability or steam generator heat

transfer capability. This is clearly not the

case with a 1% reduction in RCS flow. The

proposed change to the wording of the

parameter title on Table 3.2-1 is editorial for

clarity. Therefore, the proposed changes do

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.

No component modification, system

realignment, or change in operating

procedure will occur which could create the

possibility of a new event not previously

considered. The proposed reduction in RCS

loop and total flow rates will not initiate any

new events. Therefore, the proposed changes

would not create the possibility of a different

or new kind of accident.

3. Involve a significant reduction in a

margin of safety.

The proposed decrease in RCS loop and

total flow rates has been analyzed and found

to have an insignificant effect on the

applicable transient analyses found in the

FSAR. The proposed change to the wording

of the parameter title on Table 3.2-1 is

editorial for clarity. Therefore, the proposed

changes would not involve a significant

reduction in any margin of safety.

Therefore, based on the information

presented above, PSE&G has concluded there

is no 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: Salem Free Public library, 112

West Broadway, Salem, New Jersey

08079

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW, Washington,

DC 20005-3502

NRC Project Director: John F. Stolz

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 amendment request: January

30, 1995

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) 4.6.1.2.a

and associated Bases for 3/4.6.1.2 to

state that Type A tests for overall

integrated containment leakage rate

shall be conducted in accordance with

the requirements specified in Appendix

J of 10 CFR 50, as modified by NRCapproved exemptions. Additionally, TS

4.6.1.2.b would be revised to eliminate

the reference to the schedule contained

in TS 4.6.1.2.a.

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:

Toledo Edison has reviewed the proposed

change and determined that a significant

hazards consideration does not exist because

operation of the Davis-Besse Nuclear Power

Station, Unit No. 1, in accordance with these

changes would:

1a. Not involve a significant increase in the

probability of an accident previously

evaluated because no accident initiators,
Federal Register / Vol. 60, No. 50 / Wednesday, March 15, 1995 / Notices 14029

conditions or assumptions are significantly

affected by the proposed changes.

The proposed change would revise

Technical Specification (TS) Surveillance

Requirement (SR) 4.6.1.2.a to allow overall

integrated containment leakage rate (Type A)

testing to be scheduled in accordance with 10

CFR 50 Appendix J, as modified by approved

exemptions, and would make associated

administrative changes to TS SR 4.6.1.2.b

and to TS Bases 3/4.6.1.2. As stated above,

none of these proposed changes involve

accident initiators, conditions, or

assumptions.

1b. Not involve a significant increase in the

consequences of an accident previously

evaluated because no accident conditions or

assumptions are affected by the proposed

changes.

The results of the previous Type A testing

demonstrate a high degree of containment

integrity. The Type B and C testing

performed since the last Type A test provides

confidence that the high degree of

containment integrity will be maintained

during the interval to the next Type A test.

Therefore, the proposed changes do not alter

the source term, containment isolation, or

allowable releases, and will not increase the

radiological consequences of a previously

evaluated accident.

2. Not create the possibility of a new or

different kind of accident from any accident

previously evaluated because no new or

different accident initiators or assumptions

are introduced by the proposed changes. The

proposed changes do not affect the design or

operation of any plant system, structure, or

component. The proposed changes do not

affect any accident initiators and are not

initiators themselves. The proposed changes

do not alter any accident scenarios.

3. Not involve a significant reduction in a

margin of safety. The initial conditions and

methodologies used in the accident analyses

remain unchanged. As described above, the

proposed changes do not significantly reduce

or adversely affect the confidence that the

present high degree of containment integrity

will be maintained.

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 Toledo Library,

Documents Department, 2801 Bancroft

Avenue, Toledo, Ohio 43606.

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, N.W.,

Washington, DC 20037.

NRC Project Director: Leif J. Norrholm

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 amendment request: January

30, 1995

Description of amendment request:

The proposed amendment would

provide new Reactor Coolant Pressure

Boundary (RCPB) pressure-temperature

limit curves that are applicable up to 21

effective full power years (EFPY).

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:

Toledo Edison had reviewed the proposed

change and determined that a significant

hazards consideration does not exist because

operation of Davis-Besse Nuclear Power

Station, Unit 1, in accordance with this

change would:

1a. Not involve a significant increase in the

probability of an accident previously

evaluated because: (1) revision of the

pressure-temperature curves and the

extended applicability of the pressurizer

level/RCS pressure limit curves for periods

when relief valve DH4849 is inoperable will

continue to provide the same level of

protection of the RCPB as was previously

evaluated, and (2) the revision to License

Condition 2.C(3)(d) is administrative to

reflect the validity of the present analyses to

21 EFPY and (3) the revision to the Technical

Specification Bases

to reflect the extension to 21 EFPY is

administrative and does not affect any

previously analyzed accidents.

1b. Not involve a significant increase in the

consequences of an accident previously

evaluated because: (1) revision of the

pressure-temperature curves and the

extended applicability of the pressurizer

level/RCS pressure limit curves for periods

when relief valve DH4849 is inoperable will

continue to provide the same level of

protection of the RCPB as was previously

evaluated, and (2) the revision to License

Condition 2.C(3)(d) is administrative to

reflect the validity of the present analyses to

21 EFPY and (3) the revision to the Technical

Specification Bases to reflect the extension to

21 EFPY is administrative and does not affect

any previously analyzed accidents.

2. Not create the possibility of a new or

different kind of accident from any accident

previously evaluated because: (1) revision of

the pressure-temperature curves and the

extended applicability of the pressurizer

level/RCS pressure limit curves will continue

to provide protection against reactor vessel

failure due to brittle fracture concerns under

all postulated circumstances, and (2) the

revision to License Condition 2.C(3)(d) is

administrative to reflect the validity of the

present analyses to 21 EFPY and (3) the

revision to the Technical Specification Bases

to reflect the extension to 21 EFPY is an

administrative change and does not affect

any activities or equipment in plant

operation.

3. Not involve a significant reduction in a

margin of safety because: (1) revision of the

pressure-temperature curves and the

extended applicability of the pressurizer

level/RCS pressure limit curves maintains

the present margin of safety from reactor

vessel brittle fracture as required by 10 CFR

50, Appendix G, and (2) the revision to

License Condition 2.C(3)(d) and the Bases

revision are administrative and do not affect

any analyses which provide the basis for the

Technical Specifications.

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 Toledo Library,

Documents Department, 2801 Bancroft

Avenue, Toledo, Ohio 43606.

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, N.W.,

Washington, DC 20037.

NRC Project Director: Leif J. Norrholm

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 amendment request: February

14, 1995

Description of amendment request:

The proposed change revises Technical

Specification 4.4.D to reference the

testing requirements of 10 CFR Part 50,

Appendix J, and to state that the

Nuclear Regulatory Commissionapproved exemptions to the applicable

regulatory requirements are permitted.

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:

Virginia Electric and Power Company has

performed an evaluation of ... the proposed

administrative Technical Specification

change, in accordance with 10 CFR

50.91(a)(1) regarding no significant hazards

considerations using the standards in 10 CFR

50.92(c). A discussion of these standards as

they relate to this ... amendment request

follows.

Criterion 1 - Does Not Involve a Significant

Increase in the Probability or Consequences

of an Accident Previously Evaluated.

The proposed change ... revises Technical

Specification 4.4.D to reference the testing

frequency requirements of 10 CFR 50

Appendix J and to state that NRC approved

exemptions to the applicable regulatory
14030 Federal Register / Vol. 60, No. 50 / Wednesday, March 15, 1995 / Notices

requirements are permitted. The current

Technical Specification requires retests in

accordance with Section III.D.1(a) of

Appendix J. The proposed administrative

change simply includes the statement ‘‘as

modified by NRC approved exemptions.’’ No

new requirements are added, nor are any

existing requirements deleted. Any specific

changes to the requirements of Section

III.D.1(a) will require a submittal from

Virginia Electric and Power Company under

10 CFR 50.12 and subsequent review and

approval by the NRC prior to

implementation. The proposed change is

stated generically to avoid the need for

further Technical Specification changes if

different exemptions are approved in the

future.

The proposed change, in itself, does not

affect reactor operations or accident analyses

and has no radiological consequences. The

change provides clarification so that future

Technical Specifications changes will not be

necessary to correspond to applicable NRC

approved exemptions from the requirements

of Appendix J.

Therefore, this proposed change does not

involve a significant increase in the

probability or consequences of any accident

previously evaluated.

Criterion 2 - Does Not Create the Possibility

of a New or Different Kind of Accident from

any Previously Evaluated.

The proposed Technical Specification

amendment provides clarification to a

specification that paraphrases a codified

requirement.

Since the proposed change would not

change the design, configuration or method

of operation of the plant, it would not create

the possibility of a new or different kind of

accident from any previously evaluated.

Criterion 3 - Does Not Involve a Significant

Reduction in the Margin of Safety.

The proposed Technical Specification

change is administrative and clarifies the

relationship between the requirements of TS

4.4.D, Appendix J, and any approved

exemptions to Appendix J. It does not, in

itself, change a safety limit or [a] Limiting

Condition for Operation. The NRC will

directly approve any proposed change or

exemption to III.D.1(a) of Appendix J prior to

implementation.

Therefore, this change does not involve a

significant reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 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: Swem Library, College of

William and Mary, Williamsburg,

Virginia 23185.

Attorney for licensee: Michael W.

Maupin, Esq., Hunton and Williams,

Riverfront Plaza, 951 E. Byrd Street,

Richmond, Virginia 23219.

NRC Project Director: David B.

Matthews

Previously Published Notices Of

Consideration Of Issuance Of

Amendments To Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

And Opportunity For A Hearing

The following notices were previously

published as separate individual

notices. The notice content was the

same as above. They were published as

individual notices either because time

did not allow the Commission to wait

for this biweekly notice or because the

action involved exigent circumstances.

They are repeated here because the

biweekly notice lists all amendments

issued or proposed to be issued

involving no significant hazards

consideration.

For details, see the individual notice

in the Federal Register on the day and

page cited. This notice does not extend

the notice period of the original notice.

Illinois Power Company and Soyland

Power Cooperative, Inc., Docket No. 50-

461, Clinton Power Station, Unit No. 1,

DeWitt County, Illinois

Date of amendment request: February

14, 1995

Brief description of amendment

request: The amendment request

proposes changes to Technical

Specification 3.8.2, ‘‘AC SourcesShutdown;’’ 3.8.5, ‘‘DC SourcesShutdown;’’ and 3.8.8, ‘‘InvertersShutdown.’’ The proposed changes

would revise the operability

requirements for the Division 3 diesel

generator and the Division 3 and 4

batteries, battery chargers, and inverters

to apply only when the high pressure

core spray system is required to be

operable.Date of publication of

individual notice in Federal Register:

February 17, 1995 (60 FR 9412).

Expiration date of individual notice:

March 20, 1995

Local Public Document Room

location: Vespasian Warner Public

Library, 120 West Johnson Street,

Clinton, Illinois 61727.

Union Electric Company, Docket No.

50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of amendment request:

September 8, 1994

Brief description of amendment

request: The amendment request

proposes changes to Technical

Specification Section 3/4.9.1 to

establish administrative controls to

address a possible boron dilution event

directly from the reactor makeup water

system.

Date of publication of individual

notice in Federal Register: March 1,

1995 (60 FR 11151).

Expiration date of individual notice:

March 31, 1995

Local Public Document Room

location: Callaway County Public

Library, 710 Court Street, Fulton,

Missouri 65251.

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 No. STN 50-529, Palo Verde

Nuclear Generating Station, Unit 2,

Maricopa County, Arizona

Date of application for amendment:

November 30, 1994, as supplemented by

letter dated January 27, 1995

Brief description of amendment: The

amendment changed the pressurizer

code safety valve lift setting from 2500
Federal Register / Vol. 60, No. 50 / Wednesday, March 15, 1995 / Notices 14031

psia to 2475 psia. The lift setting is

being changed to permit Unit 2 to

operate with up to 1500 plugged tubes

in each steam generator.

Date of issuance: March 1, 1995

Effective date: March 1, 1995

Amendment No.: 78

Facility Operating License No. NPF74: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: January 4, 1995 (60 FR 496)

The additional information contained in

the January 27, 1995, supplemental

letter was clarifying in nature and thus

within the scope of the initial notice

and did not affect the NRC staff’s

proposed no significant hazards

consideration determination. The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated March 1, 1995.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona

85004

Boston Edison Company, Docket No.

50-293, Pilgrim Nuclear Power

Station,Plymouth County,

Massachusetts

Date of application for amendment:

September 6, 1994

Brief description of amendment: The

proposed amendment relocates the

alarms for the drywell to suppression

chamber vacuum breaker to a different

annunicator panel.

Date of issuance: February 16, 1995

Effective date: To be implemented prior

to startup from refueling outage ι10.

Amendment No.: 158

Facility Operating License No. DPR35: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 26, 1994 (59 FR

53839) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

No significant hazards consideration

comments received: No

Local Public Document Room

location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts

02360.

Commonwealth Edison Company,

Docket Nos. STN 50-456 and STN 50-

457, Braidwood Station, Unit Nos. 1

and 2, Will County, Illinois

Date of application for amendments:

January 5, 1994, as supplemented by

letters dated April 26, 1994, September

30, 1994, and January 12, 1995.

Brief description of amendments: The

amendments change the Braidwood

Technical Specifications to remove the

requirement to verify, every 18 months,

that the control room ventilation can be

manually isolated.

Date of issuance: February 28, 1995

Effective date: February 28, 1995

Amendment Nos.: 60 and 60

Facility Operating License Nos. NPF72 and NPF-77: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: January 25, 1995 (60 FR 4930).

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated February 28, 1995. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Wilmington Township Public

Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481.

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:

August 31, 1993, as supplemented July

19, 1994.

Brief description of amendments: The

amendments revise the technical

specifications by increasing the allowed

outage time for an inoperable chiller

only in MODES 1 through 4, adding an

optional ACTION statement in MODES

5 and 6, and adding a surveillance

requirement for the control room

ventilation system.

Date of issuance: March 2, 1995

Effective date: March 2, 1995

Amendment Nos.: 70, 70, 61 and 61

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: January 25, 1995 (60 FR 4932).

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated March 2, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: For Byron, the Byron Public

Library, 109 N. Franklin, P.O. Box 434,

Byron, Illinois 61010; for Braidwood,

the Wilmington Township Public

Library, 201 S. Kankakee Street,

Wilmington, Illinois

60481.Commonwealth Edison

Company, Docket Nos. 50-237 and 50-

249, Dresden Nuclear Power Station,

Units 2 and 3, Grundy County, Illinois;

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

Cities Nuclear Power Station, Units 1

and 2, Rock Island County, IllinoisDate

of application for amendments: July 29,

1992, as supplemented January 14,

1993, and February 16, 1993

Brief description of amendments:

Dresden and Quad Cities Technical

Specification Upgrade Program. Date of

issuance: February 16, 1995Effective

date: Immediately, to be implemented

by December 31, 1995.

Amendment Nos.: 131, 125, 152, and

148

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

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 23, 1993 (58 FR 34071)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated February 16, 1995. No

significant hazards consideration

comments received: No

Local Public Document Room

location: For Dresden, The Morris

Public Library, 604 Liberty Street,

Morris, Illinois 60450; For Quad Cities,

The Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois

61021.

Commonwealth Edison Company,

Docket Nos. 50-237 and 50-249,

Dresden Nuclear Power Station, Units 2

and 3, Grundy County, Illinois

Date of application for amendments:

July 29, 1992, as supplemented January

14, 1993, February 16, 1993 and January

27, 1995

Brief description of amendments: The

July 29, 1992, application, is one of

twelve applications which have been

submitted by Commonwealth Edison

Company (ComEd) in an effort to

upgrade the existing custom Technical

Specifications (TS) to the Boiling Water

Reactor (BWR) Standard Technical

Specifications (STS). Dresden has

recently rescheduled the Unit 2

refueling outage from March 4, 1995,

until June 1995. Currently, the

surveillance frequency for certain

Inservice Testing (IST) requirements

expires on February 21, 1995. The

current TSs do not make provisions for

a grace period for surveillance

frequencies of the IST program. In

accordance with BWR STS guidance,

the TSs regarding IST proposed in the

July 29, 1992, application, allow the

flexibility to perform these tests

appropriately during refueling outages

(where applicable) by providing a 25

percent extension to IST surveillance

intervals. The January 27, 1995,

supplement requested the staff to review

and approve just that portion of the July

29, 1992, application dealing with the

implementation of the IST program in

Section 3.0/4.0 of the proposed TS.
14032 Federal Register / Vol. 60, No. 50 / Wednesday, March 15, 1995 / Notices

Date of issuance: February 22,

1995Effective date: February 22, 1995

Amendment Nos.: 132 and 126

Facility Operating License Nos. DPR19 and DPR-25: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: June 23, 1993 (58 FR 34071)

The January 27, 1995, letter did not

change the initial proposed no

significant hazards consideration

determination. The Commission’s

related evaluation of the amendments is

contained in a Safety Evaluation

datedFebruary 22, 1995.No significant

hazards consideration comments

received: No

Local Public Document Room

location: Morris Public Library, 604

Liberty Street, Morris, Illinois 60450.

Connecticut Yankee Atomic Power

Company and Northeast Nuclear

Energy Company, Docket Nos. 50-213

and 50-245, Haddam Neck Plant and

Millstone Nuclear Power Station, Unit

1, Middlesex County and New London

County, Connecticut

Date of application for amendments:

October 31, 1994, as supplemented

February 14, 1995.

Brief description of amendments: The

amendments renew the existing license

conditions for both plants to implement

and maintain Integrated Implementation

Schedule Program Plans (the Program

Plans). The Program Plans provide a

methodology to be followed for

scheduling plant modifications and

engineering evaluations.

Date of issuance: February 23, 1995

Effective date: February 23, 1995

Amendment Nos.: 183 for Haddam

Neck, 80 for Millstone 1

Facility Operating License Nos. DPR61 and DPR-21. Amendments revise the

Licenses.

Date of initial notice in Federal

Register: December 7, 1994 (59 FR

63117)The February 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 amendments is

contained in a Safety Evaluation dated

February 23, 1995.No significant

hazards consideration comments

received: No.

Local Public Document Room

locations: Russell Library, 123 Broad

Street, Middletown, CT 06457, for the

Haddam Neck Plant, and the Learning

Resource Center, Three Rivers

Community-Technical College, Thames

Valley Campus, 574 New London

Turnpike, Norwich, CT 06360, for

Millstone Unit 1.

Consumers Power Company, Docket

No. 50-255, Palisades Plant, Van Buren

County, Michigan

Date of application for amendment:

October 5, 1994, as supplemented

February 10, 20, and 22, 1995.

Brief description of amendment: The

amendment revises primary coolant

system (PCS) pressure-temperature

limits, power-operated relief valve

setting limits, and primary coolant

pump starting limits to accommodate

reactor vessel fluence for an additional

4 effective full power years. The

amendment also revises the emergency

core cooling system technical

specifications to render two highpressure safety injection pumps

incapable of injecting into the PCS

when the PCS is below 300°F rather

than rendering both inoperable below

260°F. In addition, it revises the

pressurizer heatup to achieve

consistency between design

assumptions and technical

specifications limits.

Date of issuance: March 2, 1995

Effective date: March 2, 1995

Amendment No.: 163

Facility Operating License No. DPR20. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 4, 1995 (60 FR 501)

The February 10, 20, and 22, 1995,

submittals provided

clarifyinginformation which was within

the scope of the initial application and

did not affect the staff’s initial proposed

no significant hazards consideration

findings. The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

March 2, 1995.No significant hazards

consideration comments received: No.

Local Public Document Room

location: Van Wylen Library, Hope

College, Holland, Michigan 49423.

Duke Power Company, et al., Docket

Nos. 50-413 and 50-414, Catawba

Nuclear Station, Units 1 and 2, York

County, South Carolina

Date of application for amendments:

January 10, 1994, as supplemented

March 21 and September 15, 1994, and

January 5, 1995

Brief description of amendments: The

amendments revised Technical

Specification Table 2.2-1 and TS 4.2.5 to

allow a change in the method for

measuring reactor coolant system (RCS)

flow rate from the calorimetric heat

balance method to a method based on a

one-time calibration of the RCS cold leg

elbow differential pressure taps.

Date of issuance: February 17, 1995

Effective date: To be implemented

within 30 days from the date of issuance

Amendment Nos.: 128 and 122

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

the Technical Specifications.

Date of initial notice in Federal

Register: January 26, 1994 (59 FR 3743)

for Unit 1; and March 1, 1994 (59 FR

9785) for Unit 2

The March 21 and September 15,

1994, and January 5, 1995, letters

provided additional information that

did not change the initial scope of the

January 10, 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 February 17,

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

Entergy Operations, Inc., Docket No.

50-313, Arkansas Nuclear One,Unit No.

1, Pope County, Arkansas

Date of amendment request: August

30, 1994

Brief description of amendment: The

amendment revised the Technical

Specifications to address the installation

of two battery chargers on each 125 vdc

power train in lieu of the ‘‘swing’’

battery charger that is currently used.

Date of issuance: February 17, 1995

Effective date: February 17, 1995

Amendment No.: 176

Facility Operating License No. DPR51. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 17, 1995 (60 FR 3439)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated February 17, 1995.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas

72801

Entergy Operations, Inc., Docket No.

50-313, Arkansas Nuclear One,Unit No.

1, Pope County, Arkansas

Date of amendment request: June 22,

1994.

Brief description of amendment: The

amendment extends the allowable

outage time for one inoperable train of

emergency feedwater from 36 hours to

72 hours, clarifies the specifications and

their associated bases, and relocates

information within the specifications.

Date of issuance: March 1, 1995

Effective date: 30 days following the

date of issuance.
Federal Register / Vol. 60, No. 50 / Wednesday, March 15, 1995 / Notices 14033

Amendment No.: 177

Facility Operating License No. DPR51. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 17, 1994, (59 FR

42339) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

March 1, 1995.No significant hazards

consideration comments received: No.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas

72801

Entergy Operations, Inc., Docket No.

50-382, Waterford Steam

ElectricStation, Unit 3, St. Charles

Parish, Louisiana

Date of amendment request: January

19, 1995

Brief description of amendment: The

amendment changed the Appendix A

technical specifications (TSs) by adding

TS 3.0.5 and its associated Bases. This

new specification will allow equipment

removed from service or declared

inoperable to comply with ACTIONS to

be returned to service under

administrative controls soley to perform

testing required to demonstrate its

OPERABILITY or the OPERABILITY of

other equipment.

Date of issuance: March 1, 1995

Effective date: March 1, 1995

Amendment No.: 101

Facility Operating License No. NPF38. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 27, 1995 (60 FR 5441)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated March 1, 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.

Entergy Operations, Inc., Docket No.

50-382, Waterford Steam

ElectricStation, Unit 3, St. Charles

Parish, Louisiana

Date of amendment request: August

11, 1994, as supplemented by letter

dated December 2, 1994.

Brief description of amendment: The

amendment revised the Technical

Specifications for the Waterford Steam

Electric Station, Unit 3, by modifying

the specifications having cycle-specific

parameter limits by replacing the values

of those limits with a reference to a core

operating limits report for the values of

those limits. These changes are in

accordance with the requirements of

Generic Letter 88-16.

Date of issuance: March 1, 1995

Effective date: March 1, 1995

Amendment No.: 102

Facility Operating License No. NPF38. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: December 21, 1994 (59 FR

65812) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

March 1, 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.

Entergy Operations, Inc., Docket No.

50-382, Waterford Steam

ElectricStation, Unit 3, St. Charles

Parish, Louisiana

Date of amendment request: August

19, 1994, as supplemented by letter

dated October 14, 1994.

Brief description of amendment: The

amendment changed the Appendix A

technical specification (TSs) by

removing the Limiting Condition For

Operation (LCO) 3/4.3.4, the associated

surveillance requirements, and Bases

information from the TSs. This

information and requirements will be

incorporated into the Waterford 3

Updated Final Safety Analysis Report

(UFSAR) and maintained under the

provisions of 10 CFR 50.59.

Date of issuance: March 2, 1995

Effective date: March 2, 1995

Amendment No.: 103

Facility Operating License No. NPF38. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 31, 1994 (59 FR 45023)

The additional information contained in

the supplemental letter dated October

14, 1994, was clarifying in nature and

thus, within the scope of the initial

notice and did not affect the staff’s

proposed no significant hazards

consideration determination. The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated March 2, 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.

Entergy Operations, Inc., System

Energy Resources, Inc., South

Mississippi Electric Power Association,

and Mississippi Power & Light

Company,Docket No. 50-416, Grand

Gulf Nuclear Station, Unit 1, Claiborne

County, Mississippi

Date of application for amendment:

April 21, 1993

Brief description of amendment: The

amendment revised the requirement for

control rod testing to increase the

‘‘notch testing’’ surveillance interval for

partially withdrawn control rods from

once per 7 days to once per 31 days. The

change is consistent with the format and

content of the Improved Standard

Technical Specifications (NUREG-1434,

Revision 0).

Date of issuance: February 16, 1995

Effective date: February 16, 1995

Amendment No: 115

Facility Operating License No. NPF29. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: May 12, 1993 (58 FR 28055)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated February 16, 1995. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Judge George W. Armstrong

Library, 220 S. Commerce Street,

Natchez, Mississippi 39120.

Entergy Operations, Inc., System

Energy Resources, Inc., South

Mississippi Electric Power Association,

and Mississippi Power & Light

Company,Docket No. 50-416, Grand

Gulf Nuclear Station, Unit 1, Claiborne

County, Mississippi

Date of application for amendment:

July 14, 1993

Brief description of amendment: The

amendment revised technical

specification requirements for the

hydrogen ignition system (HIS). The

amendment also removed several tables

related to the HIS in accordance with

guidance contained in Generic Letter

91-08, ‘‘Removal of Component Lists

From Technical Specifications.’’

Date of issuance: February 16, 1995

Effective date: February 16, 1995

Amendment No: 116

Facility Operating License No. NPF29. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: September 1, 1993 (58 FR

46232) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

February 16, 1995. No significant

hazards consideration comments

received: No
14034 Federal Register / Vol. 60, No. 50 / Wednesday, March 15, 1995 / Notices

Local Public Document Room

location: Judge George W. Armstrong

Library, 220 S. Commerce at

Washington, Natchez, Mississippi

39120.

Entergy Operations, Inc., System

Energy Resources, Inc., South

Mississippi Electric Power Association,

and Mississippi Power & Light

Company,Docket No. 50-416, Grand

Gulf Nuclear Station, Unit 1, Claiborne

County, Mississippi

Date of application for amendment:

August 11, 1993

Brief description of amendment: The

amendment deleted the requirements of

Limiting Condition for Operation (LCO)

3.3.3.9 and Surveillance Requirement

4.3.3.9 related to loose-part detection

instrumentation. The deleted

requirements will be relocated to

documents that are controlled by the

licensee under the provisions of 10 CFR

50.59. The change is consistent with the

format and content of the Improved

Standard Technical Specifications

(NUREG-1434, Revision 0).

Date of issuance: February 16, 1995

Effective date: February 16, 1995

Amendment No: 117

Facility Operating License No. NPF29. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: September 1, 1993 (58 FR

46232) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

February 16, 1995. No significant

hazards consideration comments

received: No

Local Public Document Room

location: Judge George W. Armstrong

Library, 220 S. Commerce Street,

Natchez, Mississippi 39120.

Entergy Operations, Inc., System

Energy Resources, Inc., South

Mississippi Electric Power Association,

and Mississippi Power & Light

Company,Docket No. 50-416, Grand

Gulf Nuclear Station, Unit 1, Claiborne

County, Mississippi

Date of application for amendment:

August 11, 1993

Brief description of amendment: The

amendment deleted certain accident

monitoring instruments from Technical

Specification Table 3.3.7.5-1 ‘‘Accident

Monitoring Instrumentation’’ and

deleted the corresponding Surveillance

Requirements from Table 4.3.7.5-1,

‘‘Accident Monitoring Instrumentation

Surveillance Requirements.’’ The

deleted requirements will be relocated

to documents that are controlled by the

licensee under the provisions of 10 CFR

50.59. The change is consistent with the

format and content of the Improved

Standard Technical Specifications

(NUREG-1434, Revision 0).

Date of issuance: February 16, 1995

Effective date: February 16, 1995

Amendment No: 118

Facility Operating License No. NPF29. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: September 1, 1993 (58 FR

46234) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

February 16, 1995.No significant

hazards consideration comments

received: No

Local Public Document Room

location: Judge George W. Armstrong

Library, 220 S. Commerce Street,

Natchez, Mississippi 39120.

Entergy Operations, Inc., System

Energy Resources, Inc., South

Mississippi Electric Power Association,

and Mississippi Power & Light

Company, Docket No. 50-416, Grand

Gulf Nuclear Station, Unit 1, Claiborne

County, Mississippi

Date of application for amendment:

October 22, 1993, as supplemented by

letters dated February 10, and 14, 1995.

Brief description of amendment: The

amendment modified the testing

frequencies for the drywell bypass test

and the airlock test, relocated certain

drywell airlock tests from the technical

specifications to administrative

procedures, and incorporates various

improvements from the Improved

Standard Technical Specifications

(NUREG-1434, Revision 0).

Date of issuance: February 16, 1995

Effective date: February 16, 1995

Amendment No: 119

Facility Operating License No. NPF29. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: December 8, 1993 (58 FR

64607) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

February 16, 1995. No significant

hazards consideration comments

received: No

Local Public Document Room

location: Judge George W. Armstrong

Library, 220 S. Commerce Street,

Natchez, Mississippi 39120.

GPU Nuclear Corporation, et al.,

Docket No. 50-289, Three Mile Island

Nuclear Station, Unit No. 1, Dauphin

County, Pennsylvania

Date of application for amendment:

May 17, 1993, as supplemented on

December 23, 1994

Brief description of amendment: The

amendment changes the action

statement for inoperable degraded grid

and loss of voltage relays and their

associated auxiliary relays and timers.

Date of issuance: January 31, 1995

Effective date: January 31, 1995

Amendment No.: 193

Facility Operating License No. DPR50. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: November 10, 1993 (58 FR

59750). The December 23, 1994, letter

provided additional information that

did not change the initial proposed no

significant hazards consideration

determination. The Commission’s

related evaluation of this amendment is

contained in a Safety Evaluation dated

January 31, 1995. No significant hazards

consideration comments received: No.

Local Public Document Room

location: Government Publications

Section, State Library of Pennsylvania,

Walnut Street and Commonwealth

Avenue, Box 1601, Harrisburg, PA

17105. The above Notice was to be

published in the Federal Register of

February 15, 1995. The notice that was

inadvertently published at 60 FR 8762

relates to a licensing action which has

not been completed.

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:

November 7, 1994, as supplemented by

letters dated December 20, 1994, and

January 23, 1995.

Brief description of amendments: The

amendments changed the number of

standby diesel generators (SDGs)

(emergency power source) required to

be operable during Mode 6 with greater

than or equal to 23 feet of water above

the reactor vessel flange, from two to

one. The amendment also allows

limited substitution of an alternate

onsite emergency power source for one

of the two required SDGs, in Mode 5,

and in Mode 6 with less than 23 feet of

water. In addition, certain system

specifications that are affected by the

changes for the emergency power source

were also changed.

Date of issuance: February 14, 1995

Effective date: February 14, 1995, to

be implemented within 31 days of

issuance.

Amendment Nos.: Unit 1 -

Amendment No. 34; Unit 2 -

Amendment No. 20

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

revised the Technical
Federal Register / Vol. 60, No. 50 / Wednesday, March 15, 1995 / Notices 14035

Specifications.Public comments

requested as to proposed no significant

hazards consideration: Yes (60 FR 5739,

dated January 30, 1995). The notice

provided an opportunity to submit

comments on the Commission’s

proposed no significant hazards

consideration determination. No

comments have been received. The

notice also provided for an opportunity

to request a hearing by March 1, 1995,

but stated that, if the Commission makes

a final no significant hazards

consideration determination, any such

hearing would take place after issuance

of the amendments.

The Commission’s related evaluation

of the amendments, finding of exigent

circumstances, and final determination

of no significant hazards consideration

is contained in a Safety Evaluation

dated February 14, 1995.

Local Public Document Room

location: Wharton County Junior

College, J. M. Hodges Learning Center,

911 Boling Highway, Wharton, Texas

77488

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

Duane Arnold Energy Center, Linn

County, Iowa

Date of application for amendment:

August 15, 1994, as supplemented on

December 21, 1994, and January 20,

1995. The licensee’s submittals of

December 21, 1994, and January 20,

1995, provided clarification and did not

change the original no significant

hazards consideration.

Brief description of amendment: The

proposed amendment would revise the

Technical Specifications by increasing

the allowable main steam isolation

valve (MSIV) leakage and deleting the

requirements applicable to the MSIV

leakage control system.

Date of issuance: February 22, 1995

Effective date: February 22, 1995 and

to be implemented within 90 days.

Amendment No.: 207

Facility Operating License No. DPR49. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: September 14, 1994 (59 FR

47169) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

February 22, 1995.No significant

hazards consideration comments

received: No.

Local Public Document Room

location: Cedar Rapids Public Library,

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

Iowa 52401.

Illinois Power Company and Soyland

Power Cooperative, Inc., Docket No. 50-

461, Clinton Power Station, Unit No. 1,

DeWitt County, Illinois

Date of application for amendment:

August 12, 1994, as supplemented on

October 14, 1994 and February 6, 1995.

Brief description of amendment: The

amendment modifies Clinton Power

Station Technical Specification 3.6.5.1,

‘‘Drywell,’’ to permit a one-time only

change to forego performance of the

drywell bypass leakage rate test during

the fifth refueling outage scheduled to

begin in March 1995.

Date of issuance: March 1, 1995

Effective date: March 1, 1995

Amendment No.: 96

Facility Operating License No. NPF62. The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: September 28, 1994 (59 FR

49427). The October 14, 1994, and

February 6, 1995, submittals consisted

of revisions and clarifications which did

not change the staff’s initial proposed

no significant hazards consideration

determination or expand the scope of

the original notice.The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

March 1, 1995. No significant hazards

consideration comments received: No

Local Public Document Room

location: The Vespasian Warner Public

Library, 120 West Johnson Street,

Clinton, Illinois 61727.

Indiana Michigan Power Company,

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

C. Cook Nuclear Plant, Unit Nos. 1 and

2, Berrien County, Michigan

Date of application for amendments:

November 18, 1994

Brief description of amendments: The

amendments revise Technical

Specification 4.0.5 to delete the wording

‘‘except where specific written relief has

been granted by the Commission

pursuant to 10 CFR 50, Section

50.55a(g)(6)(i).’’ This change allows the

licensee to implement certain 10 CFR

50.55a relief requests while the relief

requests are being reviewed by the NRC

at the beginning of an updated interval.

Date of issuance: February 23, 1995

Effective date: February 23, 1995

Amendment Nos.: 190/176

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

the Technical Specifications.

Date of initial notice in Federal

Register: December 21, 1994 (59 FR

65817) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

February 23, 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.

Northeast Nuclear Energy Company, et

al., Docket No. 50-423,

MillstoneNuclear Power Station, Unit

No. 3, New London County, Connecticut

Date of application for amendment:

May 18, 1994

Brief description of amendment: The

amendment modifies the operability

requirements for the fuel building

exhaust filter system. The amendment

will result in modifications to the

applicability, surveillance requirement,

and bases sections of Technical

Specification 3/4.9.12, ‘‘Fuel Building

Exhaust Filter System.’’

Date of issuance: February 22, 1995

Effective date: As of the date of

issuance to be implemented within30

days.

Amendment No.: 105

Facility Operating License No. NPF49. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 22, 1994 (59 FR 32234)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated February 22, 1995.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, Thames Valley Campus, 574

New London Turnpike, Norwich, CT

06360.

Pacific Gas and Electric Company,

Docket Nos. 50-275 and 50-323,

DiabloCanyon Nuclear Power Plant,

Unit Nos. 1 and 2, San Luis Obispo

County,California

Date of application for amendments:

July 9, 1992

Brief description of amendments: The

amendments extend the operating

licenses for the Diablo Canyon Nuclear

Power Plant, Units 1 and 2 to recover or

recapture the construction period of the

reactors. Specifically, the amendments

extend the expiration date of the Unit 1

license from April 23, 2008, to

September 22, 2021, and the expiration

date of the Unit 2 license from

December 9, 2010, to April 26, 2025.

Date of issuance: March 1, 1995

Effective date: March 1, 1995

Amendment Nos.: 97 and 96

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

revised the license.

Date of initial notice in Federal

Register: July 22, 1992 (57 FR 32575)
14036 Federal Register / Vol. 60, No. 50 / Wednesday, March 15, 1995 / Notices

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated March 1, 1995.No

significant hazards consideration

comments received: Yes. Comments

from the San Luis Obispo Mothers for

Peace (MFP) and their contentions were

admitted into this proceeding. These

contentions concern the adequacy of the

licensee’s maintenance and surveillance

program and interim corrective actions

in lieu of Thermo-Lag. The Atomic

Safety and Licensing Board, in its initial

decision dated November 4, 1994 (LBP94-35), authorized the staff to extend the

DCPP operating license expiration dates.

Because a hearing was held prior to

license issuance, the staff does not need

to make a final no significant hazards

consideration determination.

Local Public Document Room

location: California Polytechnic State

University, Robert E. Kennedy Library,

Government Documents and Maps

Department, San Luis Obispo, California

93407

Pennsylvania Power and Light

Company, Docket No. 50-387,

Susquehanna Steam Electric Station,

Unit 1, Luzerne County, Pennsylvania

Date of application for amendment:

July 27, 1994, as supplemented October

27, 1994 and February 3, 1995

Brief description of amendment: The

amendment raises the authorized Power

Level from 3293 MWt to a new limit of

3441 MWt.

Date of issuance: February 22, 1995

Effective date: As of date of issuance

and is to be implemented prior to

startup in Cycle 9, currently scheduled

to occur in May 1995.

Amendment No.: 143

Facility Operating License No. NPF14: This amendment revised the

Technical Specifications and license.

Date of initial notice in Federal

Register: September 14, 1994 (59 FR

47171) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

February 22, 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:

June 23, 1994

Brief description of amendments: The

amendment revises Technical

Specification 4.0.5, which provides the

requirements for inservice inspection

and testing of ASME Code components,

to conform to Standard Technical

Specifications (NUREG-1433).

Date of issuance: February 28, 1995

Effective date: February 28, 1995

Amendment Nos.: 144 and 113

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

revised the Technical Specifications.

Date of initial notice in Federal

Register: August 3, 1994 (59 FR 39595)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated February 28, 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:

October 28, 1994, and supplemented by

letter dated December 29, 1994

Brief description of amendments:

These amendments change the

Technical Specifications (TS) for the

two units by adding reference ι20 (Unit

1) and reference ι18 (Unit 2) to Section

6.9.3.2 as ‘‘PL-NF-90-001, Supplement

1, ’Application of Reactor Analysis

Methods for BWR Design and Analysis:

Loss of Feedwater Heating Changes and

Use of RETRAN MOD 5.1,’ September

1994.’’ These additions reflect changes

to the methodology that the licensee is

using to perform its nuclear fuel reload

analysis for the two units.

Date of issuance: February 28, 1995

Effective date: February 28, 1995

Amendment Nos.: 145 and 114

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

revised the Technical Specifications.

Date of initial notice in Federal

Register: December 21, 1994 (59 FR

65819) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

February 28, 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.

Philadelphia Electric Company, Docket

Nos. 50-352 and 50-353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of application for amendments:

August 31, 1994

Brief description of amendments:

These amendments address Section 5,

‘‘Remove Temperature Requirement for

Operational Condition 5 (TSCR 94-44-

0), by revising TS Table 1.2 and TS

Bases 3/4.9.11 to remove the average

reactor coolant temperature requirement

in Operational Condition (OPCON) 5,

Refueling.

Date of issuance: January 27, 1995

Effective date: January 27,

1995Amendment Nos. 88 and 50

Facility Operating License Nos. NPF39 and NPF-85. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: November 9, 1994 (59 FR

55884) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

January 27, 1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania

19464.

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:

November 16, 1994

Brief description of amendment: The

amendment revises Technical

Specifications Section 3.10.8 and the

associated Bases, to reduce the

maximum allowable control rod drop

time from 2.4 to 1.8 seconds.

Date of issuance: February 21, 1995

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 160

Facility Operating License No. DPR64: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 20, 1995 (60 FR 4203)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated February 21, 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. 50 / Wednesday, March 15, 1995 / Notices 14037

Saxton Nuclear Experimental

Corporation, Docket No. 50-146, Saxton

Nuclear Reactor Facility

Date of application for amendment:

August 8, 1994, as supplemented on

October 28, 1994, and January 12, 1995.

Brief description of amendment: The

amendment adds characterization as an

authorized activity at Saxton and

improves the wording of the technical

specifications.

Date of issuance: February 22, 1995

Effective date: February 22, 1995

Amendment No.: 12Amended Facility

License No. DPR-4: Amendment

changed the Technical Specifications

Date of initial notice in Federal

Register: November 9, 1994. The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated February 22, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Saxton Community Library,

911 Church Street, Saxton,

Pennsylvania 16678

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:

December 30, 1993, as supplemented by

letters dated June 3, 1994, August 25,

1994, and January 3, 19, and 30, 1995.

Brief description of amendments:

These amendments revise TS 3.9.4,

‘‘Containment Building Penetrations,’’

and the associated bases to allow both

doors of the containment personnel

airlock to be open at the same time

during refueling operations provided

certain conditions are met.

Date of issuance: February 28, 1995

Effective date: February 28, 1995

Amendment Nos.: 117 and 106

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

revised the Technical Specifications.

Date of initial notice in Federal

Register: September 28, 1994 (59 FR

49434). The additional information

contained in the January 3, 19, and 30,

1995, letters were clarifying in nature,

within the scope of the initial notice

and did not affect the NRC staff’s

proposed no significant hazards

consideration determination. The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated February 28, 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

Southern Nuclear Operating Company,

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

Joseph M. Farley Nuclear Plant, Units

1 and 2, Houston County, Alabama.

Date of amendments request:

December 19, 1994

Brief description of amendments: The

amendments to Technical Specifications

include: (1) a revision in Table 3.7-3 to

the main steam safety valve (MSSV)

setpoint tolerance from plus or minus 1

percent to plus or minus 3 percent, (2)

modification of the bases to 3/4.7.1.1 to

increase the relieving capacity of the

MSSVs to at least 12,984,660 pounds

per hour which corresponds to

approximately 112 percent of total

secondary steam flow at 100 percent

rated thermal power, (3) modifications

to Table 3.7-1 to reduce the allowable

power range neutron flux high setpoints

for multiple inoperable steam generator

safety valves, and (4) an editorial

correction to Bases 3/4.7.1.2 to indicate

required auxiliary feedwater flow at

‘‘1133 psia’’ rather than ‘‘1133 psig.’’

Date of issuance March 1, 1995

Effective date: March 1, 1995

Amendment Nos.: 112 and 103

Facility Operating License Nos. NPF2 and NPF-8. Amendments revise the

Technical Specifications.

Date of initial notice in Federal

Register: January 4, 1995 (60 FR 505)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated March 1, 1995No

significant hazards consideration

comments received: No

Local Public Document Room

location: Houston-Love Memorial

Library, 212 W. Burdeshaw Street, Post

Office Box 1369, Dothan, Alabama

36302The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

March 1, 1995No significant hazards

consideration comments received: No

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:

September 29, 1993

Brief description of amendment: The

proposed changes increase the amount

of boron required in the standby liquid

control system.

Date of issuance: February 28, 1995

Effective date: February 28, 1995

Amendment Nos.: 217, 233 and 191

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

revised the Technical Specifications.

Date of initial notice in Federal

Register: June 8, 1994 (59 FR 29635)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated February 28, 1995.No

significant hazards consideration

comments received: None

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:

September 30, 1993 (TS 336)

Brief description of amendment: The

proposed changes revise and clarify the

spent fuel pool water level, temperature,

sampling, and analysis surveillance

requirements.

Date of issuance: March 2, 1995

Effective date: March 2, 1995

Amendment Nos.: 218, 334 and 192

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

revised the Technical Specifications.

Date of initial notice in Federal

Register: December 22, 1993 (58 FR

67862) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

March 2, 1995.No significant hazards

consideration comments received: None

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:

March 31, 1994

Brief description of amendment: For

Browns Ferry Units 1 and 3, the

proposed changes provide for operation

in the extended load line limit region

and revised rod block monitor

operability requirements. For all three

Browns Ferry units, the changes delete

a obsolete value for rated loop

recirculation flow rate, relocate cyclespecific equations to the Core Operating

Limits report, and provide other

miscellaneous changes.

Date of issuance: February 24, 1995

Effective date: February 24, 1995

Amendment Nos.: 216, 232, 190

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

revised the Technical Specifications.

Date of initial notice in Federal

Register: September 28, 1994 (59 FR

49437) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

February 24, 1995.No significant

hazards consideration comments

received: None
14038 Federal Register / Vol. 60, No. 50 / Wednesday, March 15, 1995 / Notices

Local Public Document Room

location: Athens Public library, South

Street, Athens, Alabama 35611

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:

October 7, 1994

Brief description of amendment:

Eliminates redundancy in system

leakage test requirements by revising TS

3/4.5.2 and its associated basis for the

Emergency Core Cooling System and TS

3/4.6.2 and its associated basis for the

Containment Spray System.

Date of issuance: February 27, 1995

Effective date: February 27, 1995 and

to be implemented within 90 days.

Amendment No. 195

Facility Operating License No. NPF-3.

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: November 9, 1994 (59 FR

55893) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

February 27, 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.

TU Electric Company, Docket Nos. 50-

445 and 50-446, Comanche Peak Steam

Electric Station, Unit Nos. 1 and 2,

Somervell County, Texas

Date of amendment request:

November 18, 1994 (published in

Federal Register as November 11, 1994)

Brief description of amendments: The

proposed amendments would provide

for cycle-specific allowances to account

for increases in the Heat Flux Hot

Channel Factor between monthly

surveillances.

Date of issuance: March 1, 1995

Effective date: March 1, 1995, to be

implemented within 30 days of

issuance.

Amendment Nos.: Unit 1 -

Amendment No. 34; Unit 2 -

Amendment No. 20

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

revised the Technical Specifications.

Date of initial notice in Federal

Register: December 7, 1994 (59 FR

63127) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

March 1, 1995.No significant hazards

consideration comments received: No.

Local Public Document Room

location: University of Texas at

Arlington Library, Government

Publications/Maps, 702 College, P.O.

Box 19497, Arlington, Texas 76019.

Washington Public Power Supply

System, Docket No. 50-397, Nuclear

Project No. 2, Benton County,

Washington

Date of application for amendment:

October 31, 1994

Brief description of amendment: The

amendment modifies the Techical

Specifications (TS) to (1) add two action

statements that would provide allowed

outage times for either one or both of the

scram discharge volume (SDV) vent or

drain valves less stringent than the

current requirements of TS 3.0.3., and

(2) change the surveillance requirements

for the SDV vent and drain valves to

conduct the testing during shutdown

conditions rather than at power as

currently required.

Date of issuance: February 27, 1995

Effective date: February 27, 1995

Amendment No.: 134

Facility Operating License No. NPF21: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: December 21, 1994 (59 FR

65828) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

February 27, 1995.No significant

hazards consideration comments

received: No

Local Public Document Room

location: Richland Public Library, 955

Northgate Street, Richland, Washington

99352

Wisconsin Public Service Corporation,

Docket No. 50-305, Kewaunee

NuclearPower Plant, Kewaunee County,

Wisconsin

Date of application for amendment:

February 23, 1994

Brief description of amendment: The

amendment revises Kewaunee Nuclear

Power Plant (KNPP) Technical

Specification (TS) 6.8.c by removing the

requirement to conduct a biennial

review of plant procedures in

accordance with American National

Standards Institute (ANSI) N18.7-1976,

Section 5.2.15. Alternate programs that

are described in the KNPP Operational

Quality Assurance Program Description

(OQAPD) will be used to ensure that

procedures are reviewed and

maintained current.

Date of issuance: February 23, 1995

Effective date: February 23, 1995 and

to be implemented within 30 days.

Amendment No.: 115

Facility Operating License No. DPR43. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 30, 1994 (59 FR 14903)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated February 23, 1995.No

significant hazards consideration

comments received: None.

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
Federal Register / Vol. 60, No. 50 / Wednesday, March 15, 1995 / Notices 14039

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

April 14, 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
14040 Federal Register / Vol. 60, No. 50 / Wednesday, March 15, 1995 / Notices

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).

Pennsylvania Power and Light

Company, Docket No. 50-388,

SusquehannaSteam Electric Station,

Unit 2, Luzerne County, Pennsylvania

Date of application for amendment:

February 7, 1995

Brief description of amendment: The

amendment changed the Technical

Specifications to allow continued

operation with one neutron flux monitor

system channel (≥B’’ channel)

inoperable and should the remaining

channel become inoperable to allow

continued plant operation for 7 days to

restore one of the two inoperable

channels.

Date of issuance: March 1, 1995

Effective date: March 1, 1995

Amendment No.: 115

Facility Operating License No. NPF22: Amendment revised the Technical

Specifications. Public comments

requested as to proposed no significant

hazards consideration: No. On February

8, 1995, the staff issued a Notice of

Enforcement Discretion, which was

immediately effective and remained in

effect until this amendment was issued.

The Commission’s related evaluation

of the amendments, finding of

emergency circumstances, consultation

with the Commonwealth of

Pennsylvania and final no significant

hazards considerations determination

are contained in a Safety Evaluation

dated March 1, 1995.

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts & Trowbridge 2300

N Street NW., Washington, D.C. 20037

Local Public Document Room

location: Osterhout Free Library,

Reference Department, 71 South

Franklin Street, Wilkes-Barre,

Pennsylvania 18071.

NRC Project Director: John F. Stolz

Dated at Rockville, Maryland, this 8th day

of March 1995.

For the Nuclear Regulatory Commission

Elinor G. Adensam,

Acting Director, Division of Reactor Projects

- III/IV, Office of Nuclear Reactor Regulation

[Doc. 95-6207 Filed 3-14-95; 8:45 am]

BILLING CODE 7590-01-F

[Docket No. 50–346]

Toledo Edison Company, et al.; Notice

of Withdrawal of Applications for

Amendments to Facility Operating

License

The U.S. Nuclear Regulatory

Commission (the Commission) has

granted the request of the Toledo Edison

Company, Centerior Service Company,

and the Cleveland Electric Illuminating

Company (the licensees) to withdraw its

March 13, 1992, September 11, 1992,

and February 17, 1993, applications for

proposed amendments to Facility

Operating License No. NPF–3 for the

Davis-Besse Nuclear Power Station, Unit

No. 1, located in Ottawa County, Ohio.

The proposed amendments would

have revised the facility technical

specifications by changing the venting

requirements for the Reactor Coolant

System, deleting figures in Section 5.1,

‘‘Design Features—Site,’’ and revising

the Safety Features Actuation System

and Steam and Feedwater Rupture

Control System Instrumentation

Setpoints.

The Commission had previously

issued Notice of Consideration of

Issuance of Amendment published in

the Federal Register on March 23, 1992

(57 FR 10050), for the March 13, 1992,

application; January 6, 1993 (58 FR 600)

for the September 11, 1992, application,

and June 23, 1993 (58 FR 34096), for the

February 23, 1993, application.

However, by letter dated February 10,

1995, the licensee withdrew the

proposed changes.

For further details with respect to this

action, see the applications for

amendment dated March 13, 1992,

September 11, 1992, and February 17,

1993, and the licensee’s letter dated

February 10, 1995, which withdrew the

applications for license amendments.

The above documents 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 located at the

University of Toledo Library,

Documents Department, 2801 Bancroft

Avenue, Toledo, Ohio 43606.

Dated at Rockville, Maryland, this 9th day

of March, 1995.

For the Nuclear Regulatory Commission.

Jon B. Hopkins,

Senior Project Manager, Project Directorate

III–3, Division of Reactor Projects III/IV, Office

of Nuclear Reactor Regulation.

[FR Doc. 95–6340 Filed 3–14–95; 8:45 am]

BILLING CODE 7590–01–M

OFFICE OF THE UNITED STATES

TRADE REPRESENTATIVE

Notice of Meeting of the Industry

Policy Advisory Committee

AGENCY: Office of the United States

Trade Representative.

AGENCY: Notice that the March 22, 1995

meeting of the Industry Policy Advisory

Committee will be held from 9:30 a.m.

to 2:30 p.m. The meeting will be closed

to the public from 9:30 to 1:00 p.m. The

meeting will be open to the public from

1:00 p.m. to 2:30 p.m.

SUMMARY: The Industry Policy Advisory

Committee will hold a meeting on

March 22, 1995 from 9:30 a.m. to 2:30

p.m. The meeting will be closed to the

public from 9:30 to 1:00 p.m. The

meeting will include a review and

discussion of current issues which

influence U.S. trade policy. Pursuant to

Section 2155(f)(2) of Title 19 of the

United States Code, I have determined

that this portion of the meeting will be

concerned with matters the disclosure

of which would seriously compromise

the development by the United States

Government of trade policy, priorities,

negotiating objectives or bargaining

positions with respect to the operation

of any trade agreement and other

matters arising in connection with the

development, implementation and

administration of the trade policy of the

United States. The meeting will be open

to the public and press from 1:00 p.m.

to 2:30 p.m. when trade policy issues

will be discussed. Attendance during

this part of the meeting is for

observation only. Individuals who are

not members of the committee will not

be invited to comment.

DATES: The meeting is scheduled for

March 22, 1995, unless otherwise

notified.

ADDRESSES The meeting will be held at

the Madison Hotel, located at 15th and

M streets, NW., Washington, DC., unless

otherwise notified.

FOR FURTHER INFORMATION CONTACT:

Michaelle Burstin, Director of Public

Liaison, Office of the United States

Trade Representative, (202) 395–6120.

Michael Kantor,

United States Trade Representatives.

[FR Doc. 95–6316 Filed 3–14–95; 8:45 am]

BILLING CODE 3190–01–M

POSTAL RATE COMMISSION

Notice of Facility Visit

March 10, 1995.

Members of the Commission and its

advisory staff will visit the offices of