20842 Federal Register / Vol. 61, No. 90 / Wednesday, May 8, 1996 / Notices

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555–0001, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

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

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

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

1–(800) 342–6700). The Western Union

operator should be given Datagram

Identification Number N1023 and the

following message addressed to John F.

Stolz: petitioner’s name and telephone

number, date petition was mailed, plant

name, and publication date and page

number of this Federal Register notice.

A copy of the petition should also be

sent to the Office of the General

Counsel, U.S. Nuclear Regulatory

Commission, Washington, DC 20555–

0001, and to J.W. Durham, Sr., Esquire,

Sr. V.P. and General Counsel,

Philadelphia Electric Company, 2301

Market Street, Philadelphia,

Pennsylvania 19101, attorney for the

licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

presiding 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).

For further details with respect to this

action, see the application for

amendment dated April 25, 1996, 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 located at the

Pottstown Public Library, 500 High

Street, Pottstown, Pennsylvania 19464.

Dated at Rockville, Maryland, this 3rd day

of May 1996.

For the Nuclear Regulatory Commission.

Frank Rinaldi,

Project Manager, Project Directorate I–2,

Division of Reactor Projects—I/II, Office of

Nuclear Reactor Regulation.

[FR Doc. 96–11431 Filed 5–7–96; 8:45 am]

BILLING CODE 7590–01–P

Biweekly Notice; Applications and

Amendments to Facility Operating

Licenses Involving no Significant

Hazards Considerations

I. Background

Pursuant to Public Law 97–415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 97–415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from April 13,

1996, through April 26, 1996. The last

biweekly notice was published on April

24, 1996 (61 FR 18162).

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 20555–0001, and

should cite the publication date and

page number of this Federal Register

notice. Written comments may also be

delivered to Room 6D22, Two White

Flint North, 11545 Rockville Pike,

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

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

written comments received may be

examined at the NRC Public Document

Room, the Gelman Building, 2120 L

Street, NW., Washington, DC. The filing

of requests for a hearing and petitions

for leave to intervene is discussed

below.

By June 7, 1996, the licensee may file

a request for a hearing with respect to

issuance of the amendment to the

subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714
Federal Register / Vol. 61, No. 90 / Wednesday, May 8, 1996 / Notices 20843

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555–0001, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington DC, by

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

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

1-(800) 342–6700). The Western Union

operator should be given Datagram

Identification Number N1023 and the

following message addressed to (Project

Director): petitioner’s name and

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555–0001, and to the attorney for

the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

factors specified in 10 CFR

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

For further details with respect to this

action, see the application for

amendment which is available for

public inspection at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, and at the local public

document room for the particular

facility involved.

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: March

28, 1996.

Description of amendments request:

Pursuant to 10 CFR 50.90, the Baltimore

Gas and Electric Company (BGE) hereby

requests an amendment to Operating

License Nos. DPR–53 and DPR–69 to

reduce the moderator temperature

coefficient (MTC) limit shown on

Technical Specification Figure 3.1.1–1.

This proposed change is necessary to

support changes in the safety analyses

made to accommodate a larger number

of plugged steam generator (SG) tubes

for future operating cycles. The

proposed limit will be more restrictive

than the existing limit to match the

analytical assumptions. In addition, the

licensee provided information to clarify

the relationship of the MTC to an

Anticipated Transient Without Scram

event in its licensing basis.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Would not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

The safety analyses for the current fuel

cycles assume 500 tubes per steam generator

(SG) are plugged and the maximum

beginning-of-cycle moderator temperature

coefficient (MTC) is assumed to follow the

curve in Technical Specification Figure

3.1.1.-1. For the fuel cycle to be installed in

Unit 1 in spring 1996, Baltimore Gas and

Electric Company (BGE) assumes in the

analyses that more SG tubes are plugged than

the current limit, and it is necessary to credit

a more restrictive (less positive) limit on the

maximum positive MTC to mitigate the
20844 Federal Register / Vol. 61, No. 90 / Wednesday, May 8, 1996 / Notices

Reactor Coolant System pressure and

temperature increase analyzed for these

events. Therefore, we are proposing a change

to the allowable positive MTC limits shown

on Technical Specification Figure 3.1.1–1.

The proposed limit will be more restrictive

than the existing limit to match the analytical

assumptions. Since the safety analyses

supporting an increase in the number of

plugged SG tubes are applicable to both Units

1 and 2, BGE is requesting this change for

both Units.

The proposed change makes the limit on

the maximum positive MTC more restrictive.

From an operational standpoint, a more

restrictive limit on MTC will help mitigate

the effect of plant transients on control of

plant parameters (e.g., reactor power,

pressurizer pressure, pressurizer level, etc.)

Therefore, the probability of a previously

analyzed accident will not be significantly

increased.

The reason for the proposed change is to

mitigate the effect (increased reactor coolant

temperatures) of increased SG U-tube

plugging on the results of the affected safety

analyses. Using the more restrictive limit on

the maximum positive MTC, the Loss of

Load, Loss of Feedwater Flow, Feed Line

Break, and Control Element Assembly

Withdrawal events were reanalyzed using

previously accepted methodologies. The

results of these analyses are within the

acceptance limits for these events. Therefore,

the consequences of a previously analyzed

accident will not be significantly increased.

The proposed change is similar to the

examples of amendments that are considered

not likely to involve significant hazards

considerations given in the Statements of

Consideration for 10 CFR 50.92 (51 FR 7744).

The example of interest is, ‘‘A change that

constitutes an additional limitation,

restriction, or control not presently included

in the technical specifications, e.g., a more

stringent surveillance requirement.’’ The

proposed change provides a more restrictive

limit on the positive MTC given in Technical

Specification Figure 3.1.1–1. Based on the

above arguments and the similarity to an

example in the Federal Register, BGE has

determined that the proposed 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.

The proposed change makes the limit on

the maximum positive MTC more restrictive.

The proposed change does not involve

installation of new or different equipment,

modify the interfaces with existing

equipment, change the equipment’s function,

or change the method of operating the

equipment. The proposed change does not

affect normal plant operations or

configurations. The more restrictive MTC

limit will help mitigate the effect of plant

transients on control of plant parameters.

Therefore, the proposed change does 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.

The proposed change provides for a more

restrictive limit for the allowable positive

MTC. The more restrictive limit on the

maximum positive MTC was evaluated using

previously approved methodologies and

compared to the existing acceptance criteria.

The analyses show that the proposed change

preserves the margin of safety by ensuring

that the results of the safety analyses for the

Loss of Load, Loss of Feedwater Flow, Feed

Line Break, and Control Element Assembly

Withdrawal events meet established NRC

acceptance limits for these events.

In addition, this proposed change is similar

to the example of amendments that are

considered not likely to involve significant

hazards considerations given in the

Statements of Consideration for 10 CFR 50.92

(51 FR 7744). The example of interest is, ‘‘A

change that constitutes an additional

limitation, restriction, or control not

presently included in the technical

specifications, e.g., a more stringent

surveillance requirement.’’ The proposed

change provides a more restrictive limit on

the positive MTC given in Technical

Specification Figure 3.1.1–1. Based on the

above arguments and the similarity to an

example in the Federal Register, BGE has

determined that the proposed change does

not involve a significant reduction in the

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 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: Susan F.

Shankman, Acting.

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

8, 1996.

Description of amendment request:

The proposed amendment would

remove Technical Specifications (TS)

3.3.4, Turbine Overspeed Protection; TS

3.7.12, Area Temperature Monitoring;

and TS 3.11.2.6, Gas Storage Tanks; and

their associated bases; and relocate them

to licensee-controlled documents, such

as the Final Safety Analysis Report. The

licensee revised the original amendment

request dated October 24, 1994, to

provide supplemental information to TS

6.8.4 for administrative control program

related to TS 3.11.2.6, by letters dated

August 31, 1995 and February 8, 1996.

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 was previously

presented in the Federal Register (59 FR

60397). The staff reviewed and

determined that the proposed license

amendment’s revisions do not alter the

original conclusion that no significant

hazards considerations exist pursuant to

10 CFR 50.92.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh,

North Carolina 27605.

Attorney for licensee: William D.

Johnson, Vice President and Senior

Counsel, Carolina Power & Light

Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: Eugene V.

Imbro.

Commonwealth Edison Company,

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

County Station, Units 1 and 2, LaSalle

County, Illinois

Date of amendment request:

December 21, 1995.

Description of amendment request:

The proposed amendments would

delete the requirement to place the

reactor mode switch in the Shutdown

position if a stuck open safety/relief

valve cannot be closed within two

minutes. The operator would still be

required to scram the reactor if

suppression pool average water

temperature reaches 110 degrees

Fahrenheit or greater. The licensee also

proposed changes to the TS index pages

to reflect Bases page changes that were

accepted by the NRC staff in a letter

dated May 23, 1995. Because the

changes to the index pages require a

license amendment, they have been

included as part of this submittal.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) Involve a significant increase in the

probability or consequences of an accident

previously evaluated because:

The proposed change does not involve a

significant increase in the probability or
Federal Register / Vol. 61, No. 90 / Wednesday, May 8, 1996 / Notices 20845

consequences of an accident previously

evaluated in the UFSAR. A stuck open SRV

event is a mild transient which neither

affects fuel limits nor radiological

consequences. The two minute requirement

to manually scram after a SRV becomes stuck

open is not assumed or used in any transient

or accident analysis in the FSAR. Removing

the two minute requirement to manually

scram after a SRV becomes stuck open does

not change the probability of any accident

evaluated in the FSAR. Removing the two

minute requirement to manually scram after

a SRV becomes stuck open also does not

change the capability of the suppression pool

during this event in case of any accident

involving reactor blowdown, because the

suppression pool average water temperature

limit in Technical Specification 3.6.2.1 is

still valid and enforced. The suppression

pool average water temperature limit is the

only requirement during operational

conditions 1 and 2 that assures sufficient

heat sink capacity in case of a LOCA in the

containment. Therefore, removing the two

minute requirement to manually scram after

a SRV becomes stuck open would not

increase the probability or consequences of

any postulated accident analyzed in the

FSAR.

(2) Create the possibility of a new or

different kind of accident from any accident

previously evaluated because:

The proposed change does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated in the UFSAR. This change does

not effect any hardware. This is a procedural

change to assure that the reactor will not be

unnecessarily scrammed by the operator after

a SRV is stuck open for two minutes. The

reactor will still be scrammed if suppression

pool average water temperature increases

above 110 degrees F. Since the design basis

of the suppression pool is protected by this

average water temperature limit, this

procedural change of removing the two

minute requirement to manually scram after

a SRV becomes stuck open introduces no

new accident or malfunction.

(3) Involve a significant reduction in the

margin of safety because:

The proposed change does not reduce the

margin as defined in the bases for any

Technical Specification. On the contrary, if

the two minute requirement to manually

scram after a SRV becomes stuck open is not

removed, the operator has to scram the

reactor thus challenging the RPS, the rector

vessel, and other associated components, and

reducing the related margin to safety. This

scram would be unnecessary if the

suppression pool average water temperature

is below the 110 degree F limit allowed by

the design basis of the suppression pool.

Reactor safety or suppression pool design

basis is not compromised because the

suppression pool average water temperature

limit alone guarantees that there would not

be any reduction in margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

requested amendments involve no

significant hazards consideration.

Local Public Document Room

location: Jacobs Memorial Library,

Illinois Valley Community College,

Oglesby, Illinois 61348.

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60603.

NRC Project Director: Robert A. Capra.

Duke Power Company, Docket Nos. 50–

369 and 50–370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg

County, North Carolina

Date of amendment request: March 4,

1996.

Description of amendment request:

The proposed amendments would

change the McGuire Units 1 and 2

Updated Final Safety Analysis Report to

delete the seismic qualification

requirement for the Containment

Atmosphere Particulate Radiation

Monitors.

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:

This proposed change has been evaluated

against the standards in 10 CFR 50.92 and

has been determined to involve no significant

hazards considerations, in that operation of

the facility in accordance with the proposed

amendment would not:

1. [I]nvolve a significant increase in the

probability or consequences of an accident

previously evaluated; or

EMF38(L) is not used directly for any

phase of power generation or conversion or

transmission, normal decay heat removal,

fuel handling, or the processing of

radioactive fluids. As such, it is not an

‘‘accident initiator’’. No ‘‘accident initiator’’

is affected by the change. Thus, the

probability of accidents evaluated in the

FSAR is not affected by the change. It is

determined that sufficient ability to

determine conditions inside containment

remain available for any earthquake up to

and including the SSE [safe-shutdown

earthquake]. Furthermore, should either

EMF38(L) or EMF39(L) be found to not be

functional following any earthquake,

including those smaller than the OBE

[Operating Basis Earthquake], the appropriate

steps will be taken; i.e., declare the

monitor(s) inoperable and apply the action

statement for TS [technical specification]

3.4.6.1 which may require that the associated

unit(s) be taken to Cold Shutdown (Mode 5)

if the minimum required Reactor Coolant

Leakage Detection Systems are not operable.

Cold Shutdown is a mode for which neither

the Emergency Core Cooling System nor the

containment safeguards are required. Finally,

no equipment provided to mitigate any

accident is adversely affected by the change.

For these reasons, the proposed change will

not involve a significant increase in the

probability or consequences of an accident

previously evaluated in the SAR [safety

analysis report].

2. [C]reate the possibility of a new or

different type of accident from any accident

previously evaluated; or

As stated above, no equipment used in

direct support of power generation or

conversion or transmission, normal decay

heat removal, fuel handling, or processing of

radioactive fluids is affected with the update.

No new failure modes are identified with the

change. The upper bound to an undetected

leak in the Reactor Coolant System is a Loss

of Coolant Accident [LOCA]. As noted above,

no equipment provided to mitigate a LOCA

is affected by the change. For these reasons,

the change will not create a new or different

type of accident from any accident

previously evaluated.

3. [I]nvolve a significant reduction in a

margin of safety.

It has been determined that sufficient

means remain at the disposal to the operators

to assess conditions within the containment

following any earthquake up to and

including the SSE. In particular, the ability

to determine leakage with the sensitivity

comparable to that of EMF38(L) can be

established. This meets the intent of the

Regulatory Position of RG [Regulatory Guide]

1.45. In addition, should it be determined

that either EMF38(L) or EMF39(L) is not

functional following any earthquake, the

appropriate steps will be taken; i.e, declare

the monitor(s) inoperable and apply the

action statement for TS 3.4.6.1 which may

require that the associated unit(s) be taken to

Cold Shutdown (Mode 5) if the minimum

required Reactor Coolant Leakage Detection

Systems are not operable. This brings the

unit(s) to a mode in which TS 3.4.6.1 does

not apply. It ensures that at least the

minimum required Reactor Coolant System

leakage detection systems will be functional

before power operations are continued

following a postulated earthquake smaller

than the OBE. It ensures protection of the

reactor coolant pressure boundary, one of the

fission product barriers. No other fission

product barrier is affected by the change.

Therefore, the margin of safety is not

reduced.

Therefore, based on the information

contained in this submittal, it is determined

that no significant hazard is associated with

the proposed change.

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
20846 Federal Register / Vol. 61, No. 90 / Wednesday, May 8, 1996 / Notices

Church Street, Charlotte, North Carolina

28242.

NRC Project Director: Herbert N.

Berkow.

Entergy Operations, Inc., Docket Nos.

50–313 and 50–368, Arkansas Nuclear

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

Arkansas

Date of amendment request: April 11,

1996.

Description of amendment request:

The proposed technical specification

amendment modifies the reactor

building leak testing requirements per

Option B to 10 CFR 50, Appendix J.

Option B permits performance based

determination of the reactor building

leak testing frequency.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does Not Involve a Significant Increase

in the Probability or Consequences of an

Accident Previously Evaluated.

The proposed changes to the Technical

Specifications implement Option B of 10 CFR

50 Appendix J at ANO. The proposed

changes will result in increased intervals

between containment leakage tests

determined through a performance based

approach. The intervals between such tests

are not related to conditions which cause

accidents. The proposed changes do not

involve a change to the plant design or

operation. Therefore, this change does not

involve a significant increase in the

probability of any accident previously

evaluated.

NUREG–1493, ‘‘Performance-Based

Containment Leak-Test Program,’’

contributed to the technical bases for Option

B of 10 CFR 50 Appendix J. NUREG–1493

contains a detailed evaluation of the

expected leakage from containment and the

associated consequences. The increased risk

due to lengthening of the intervals between

containment leakage tests was also evaluated

and found acceptable. Using a statistical

approach, NUREG–1493 determined the

increase in the expected dose to the public

from extending the testing frequency is

extremely small. It also concluded that a

small increase is justifiable due to the

benefits which accrue from the interval

extension. The primary benefit is in the

reduction in occupational exposure. The

reduction in the occupational exposure is a

real reduction, while the small increase to

the public is statistically derived using

conservative assumptions. Therefore, this

change does not involve a significant

increase in the consequences of any accident

previously evaluated.

Therefore, this change does not involve a

significant increase in the probability or

consequences of any accident previously

evaluated.

2. Does Not Create the Possibility of a New

or Different Kind of Accident from any

Previously Evaluated.

The proposed change to the Technical

Specifications incorporates the performance

based approach authorized by Option B of 10

CFR 50 Appendix J. The interval extensions

allowed by this change do not involve a

change to the plant design or operation. No

safety related equipment or safety functions

are altered as a result of this change. The

reduced testing frequency does not affect the

testing methodology. As a result, the

proposed change does not affect any of the

parameters or conditions that could

contribute to initiation of any accidents. No

new accident modes are created by extending

the test intervals. Therefore, this change does

not create the possibility of a new or different

kind of accident from any previously

evaluated.

3. Does Not Involve a Significant

Reduction in the Margin of Safety.

The proposed change does not change the

performance methodology of the containment

leakage rate testing program. However, the

proposed change does affect the frequency of

containment leakage rate testing. With an

increased frequency between tests, the

proposed change does increase the

probability that a increase in leakage could

go undetected for a longer period of time.

Operational experience has demonstrated the

leak tightness of the containment buildings

has been significantly below the allowable

leakage limit.

The margin to safety that has the potential

of being impacted by the proposed change

involves the offsite dose consequences of

postulated accidents which are directly

related to containment leakage rates. The

limitation on containment leakage rate is

designed to ensure the BWN total leakage

volume will not exceed the value assumed in

our accident analysis. The margin to [sic]

safety for the offsite dose consequences of

postulated accidents directly related to

containment leakage is maintained by

meeting the 1.0 L. acceptance criteria. The

proposed change maintains the 1.0 L.

acceptance criteria.

Therefore, this change does not involve a

significant reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801.

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

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

20005–3502.

NRC Project Director: William D.

Beckner.

Entergy Operations, Inc., Docket No. 50–

368, Arkansas Nuclear One, Unit No. 2,

Pope County, Arkansas

Date of amendment request: April 11,

1996.

Description of amendment request:

The proposed technical specification

(TS) amendment adds low-temperature

overpressure protection (LTOP)

requirements to the TSs to resolve

Generic Issue 94 in accordance with

Generic Letter 90–06.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does Not Involve a Significant Increase

in the Probability or Consequences of an

Accident Previously Evaluated.

This proposed change provides additional

controls in the ANO–2 Technical

Specification [(TS)] for ensuring that LTOP

[low-temperature overpressure protection]

protection is available when required. The

limiting condition involving the

simultaneous injection of two HPSI [high

pressure safety injection] and three charging

pumps to an RCS [reactor coolant system]

water solid condition, was used in the

calculation of the ANO–2 proposed LTOP

setpoints. The methodology utilized in the

LTOP setpoint analysis is based on ASME

[American Society of Mechanical Engineers]

Code Case N–514. The code case establishes

a factor of 110 percent of the operating

pressure temperature curves instead of 100

percent. The safety factor utilized by the code

case provides a more reasonable vessel

overpressure allowance for conditions

expected under pressure loading from low

temperature transients. The SITs [safety

injection tanks] are required to be isolated, if

not depressurized, prior to entering the LTOP

enable temperature and are periodically

verified to be isolated when LTOP conditions

exist. The LTOP setpoint of the relief valves

proposed by this technical specification [TS]

change is not considered to be an initiator of

any transients, but is used to mitigate an

overpressure condition if such a transient

were to occur.

Therefore, this change does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. Does Not Create the Possibility of a New

or Different Kind of Accident from any

Previously Evaluated.

The design basis event for establishing

LTOP limits is the simultaneous injection of

two HPSI and three charging pumps to an

RCS water solid condition. The LTOP vent

size of 6.38 square inches and the valve

pressure setpoint of less than or equal to 430

psig are currently used for mitigation of low

temperature overpressure conditions. The

change in the enable setpoint was analyzed

by the application of Code Case N–514 and

determined to adequately ensure that this

temperature [sic] setpoint will mitigate a
Federal Register / Vol. 61, No. 90 / Wednesday, May 8, 1996 / Notices 20847

LTOP transient. The operator action to enable

the LTOP relief valves at 220 degrees ensures

that the RCS including the reactor vessel will

not undergo system pressures at low

temperature conditions beyond their design

limits. Therefore, there will not be any

impact to systems, structures or components

beyond their design requirements.

Therefore, this change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

3. Does Not Involve a Significant

Reduction in the Margin of Safety.

The addition of a new specification to the

ANO–2 Technical Specification [TS] will not

significantly reduce the margin of safety. The

LTOP safety factors are based on reanalyzed

conditions for 21 effective full power years

(EFPY) of operation utilizing methodology

contained in ASME Code Case N–514. The

LTOP evaluation under Code Case N–514 for

low temperature transients is considered

more appropriate than the ASME Section XI.

The code case establishes a factor of 110

percent of the operating pressure temperature

curves instead of 100 percent. The safety

factor utilized by the code case provides a

more reasonable vessel overpressure

allowance for conditions expected under

pressure loading from low temperature

transients. Although the proposed setpoint

may involve a slight reduction in a margin

of safety, the enable temperature setpoint

will provide an equivalent level of safety to

the reactor vessel during LTOP transients and

will satisfy the purpose of 10 CFR 50.60 for

fracture toughness. Therefore, based on the

refined methodology used to calculate ANO–

2 LTOP setpoints for 21 EFPY the margin of

safety will not be significantly reduced.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801.

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

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

20005–3502.

NRC Project Director: William D.

Beckner.

Entergy Operations, Inc., et al., Docket

No. 50–416, Grand Gulf Nuclear Station,

Unit 1, Claiborne County, Mississippi

Date of amendment request: April 18,

1996.

Description of amendment request:

The licensee has proposed to delete a

restriction on the 24-hour emergency

diesel generator operation test in

Surveillance Requirement 3.8.1.14 (Page

3.8–12) of the Technical Specifications

(TSs) for the Grand Gulf Nuclear

Station, Unit 1. The deletion would

allow the test to also be conducted

during power operation (i.e., during

Modes 1 and 2), instead of the current

requirement to only conduct the test

when the plant is shut down.

The frequency of conducting this test,

the conditions of the test, and the

criteria to pass the test are not being

changed by this amendment request.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration for the amendment

request, which is presented below:

Entergy Operations, Inc. [(EOI)] propose[d]

to change the current Grand Gulf Nuclear

Station [GGNS] Technical Specifications

[(TSs)]. The specific change is to modify note

2 to Surveillance 3.8.1.14. Presently, this

note prohibits the performance of the 24 hour

diesel maintenance run while the unit is in

either Mode 1 or 2. The proposed change

would remove this restriction thus allowing

the 24 hour run to be performed during any

mode of operation (i.e., modes 1, 2, 3, 4 or

5).

The Commission has provided standards

for determining whether a no significant

hazards considerations exists as stated in 10

CFR 50.92 (c). A proposed amendment to an

operating license involves no significant

hazards consideration if 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; (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.

Entergy Operations, Inc. [EOI] has

evaluated the no significant hazards

consideration in its request for this license

amendment and determined that no

significant hazards considerations results

from this change. In accordance with 10 CFR

50.91(a), Entergy Operations, Inc. [EOI] is

providing the analysis of the proposed

amendment against the three standards in 10

CFR 50.92(c). A description of the no

significant hazards consideration

determination follows:

I. The proposed change does not

significantly increase the probability or

consequences of an accident previously

evaluated.

The GGNS UFSAR [Updated Final Safety

Analysis Report] assumes that the AC

electrical power sources are designed to

provide sufficient capacity, capability,

redundancy and reliability to ensure that the

fuel, reactor coolant system and containment

design limits are not exceeded during an

assumed design basis event. Specifically, the

UFSAR assumes that the onsite EDG’s

[emergency diesel generator’s] provide

emergency power in the event offsite power

is lost to either one or all three ESF

[engineered safety feature] buses. In the event

of a loss of preferred power, the ESF

electrical loads are automatically connected

to the EDG’s in sufficient time to provide for

safe reactor shutdown and to mitigate the

consequences of a design basis accident such

as a LOCA.

The proposed change to permit the 24 hour

testing of the EDG’s during power operation

does not increase the chances or

consequences of any previously evaluated

accident. The capability of the EDG’s to

supply power in a timely manner will not be

compromised by permitting performance of

EDG testing during periods of power

operation. Design features of the EDG’s and

electrical systems ensures that if a LOCA

[loss of coolant accident] or LOP [loss of

offsite power] signal, either individually or

concurrently, should occur during testing

that the EDG would be returned to its readyto-load operation (i.e., EDG running at rated

speed and voltage separated from the offsite

sources) or separately connected to the ESF

bus providing ESF loads. As such, an EDG

being tested is considered to be Operable and

fully capable of meeting its intended design

function. Additionally, the testing of an EDG

is not a precursor to any previously evaluated

accidents.

Therefore, the proposed change allowing

testing of EDG’s during power operation will

not significantly increase the probability or

consequences of an accident previously

evaluated.

II. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

As previously discussed [above], the

proposed change to permit the performance

of EDG testing during power operation will

not affect the operation of any system or alter

any system’s response to previously

evaluated design basis events. The EDG’s will

automatically transfer from the test

configuration to the ready-to-load

configuration following receipt of a valid

signal (i.e., LOCA or LOP). In the ready-toload configuration, the EDG will be running

at rated speed and voltage separated from the

offsite source capable of automatically

supplying power to the ESF buses in the

event that preferred power is actually loss.

Surveillance Requirement 3.8.1.17

demonstrates that the EDG will automatically

override the test mode following generation

of a LOCA signal. In addition the ability of

the EDG’s to survive a full load reject is

verified by the performance of surveillance

requirement 3.8.1.10. These existing

surveillance requirements along with system

design features ensures that the performance

of EDG testing during power operation will

not create the possibility of a new or different

kind of accident from any previously

evaluated.

III. The proposed change does not involve

a significant reduction in a margin of safety.

The AC electrical power sources are

designed to provide sufficient capacity,

capability, redundancy, and reliability to

ensure the availability of necessary power to

ESF systems so that the fuel, reactor coolant

system and containment design limits are not

exceeded. Specifically, the EDG’s must be

capable of automatically providing power to

ESF loads in sufficient time to provide for

safe reactor shutdown and to mitigate the

consequences of a design basis accident in

the event of a loss of preferred power.
20848 Federal Register / Vol. 61, No. 90 / Wednesday, May 8, 1996 / Notices

Testing of EDG’s during power operation

will not affect the availability or operation of

any offsite source of power. In addition, the

EDG being tested remains capable of meeting

its intended design functions. Therefore the

proposed change to the Technical

Specification Surveillance Requirement

3.8.1.14 will not result in a reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

Location: Judge George W. Armstrong

Library, 220 S. Commerce Street,

Natchez, MS 39120.

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, N.W., 12th Floor,

Washington, DC 20005–3502.

NRC Project Director: William D.

Beckner.

GPU Nuclear Corporation, et al., Docket

No. 50–219, Oyster Creek Nuclear

Generating Station, Ocean County, New

Jersey

Date of amendment request: April 15,

1996 (TSCR No. 244).

Description of amendment request:

The proposed amendment would revise

Specification 5.3.1.B of the Oyster Creek

Technical Specifications. The current

specification prohibits handling a load

greater in weight than one fuel assembly

over irradiated fuel in the spent fuel

storage facility. The proposed change

will facilitate the off load of spent fuel

to the Oyster Creek Independent Spent

Fuel Storage Installation (ISFSI).

Specifically, the shield plug for the dry

shield canister (DSC) and the associated

lifting hardware will be moved over

irradiated fuel which is contained in the

DSC within the transfer cask located in

the Cask Drop Protection System

(CDPS).

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. State the basis for the determination that

the proposed activity will or will not increase

the probability of occurrence or

consequences of an accident.

The design features and capacity of the

reactor building crane provide a significant

safety factor. In addition, personnel training

and other administrative controls further

reduce risk. Thus, the dropping of the DSC

shield plug onto a loaded DSC and causing

damage to the spent fuel assemblies is not a

credible event. Therefore, it does not increase

the probability of or consequences of an

accident.

2. State the basis for the determination that

the activity does or does not create the

possibility of an accident or malfunction of

a different type than any previously

identified in the SAR [safety analysis report].

This activity will not create the possibility

of a new or different type of accident than

previously evaluated in the SAR because the

proposed heavy load handling exception

does not create a new credible accident

scenario. Dropping the shield plug on a

loaded DSC and damaging spent fuel

assemblies is not considered a credible event.

3. State the basis for the determination that

the margin of safety is not reduced.

This activity will not involve a significant

reduction in the margin of safety because the

proposed heavy load handling evolution does

not create a credible accident scenario.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Ocean County Library,

Reference Department, 101 Washington

Street, Toms River, NJ 08753.

Attorney for licensee: Ernest L. Blake,

Jr., Esquire. Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: John F. Stolz.

Maine Yankee Atomic Power Company,

Docket No. 50–309, Maine Yankee

Atomic Power Station, Lincoln County,

Maine

Date of amendment request: April 19,

1996.

Description of amendment request:

The proposed amendment would revise

Technical Specification 5.14 to add the

appropriate references identifying the

detailed methodology and conditions

for analyzing the Small Break Loss-ofCoolant Accident (SBLOCA) to the list

of the approved Core Operating Limits

Report 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. Does the Proposed Amendment involve

a significant increase in the probability or

consequences of an accident previously

evaluated?

These Proposed Changes are administrative

in nature and are consistent with the

guidance set forth in the NRC Generic Letter

88–16 identifying the requirements for the

inclusion of analytical methodology

references in Technical Specifications as

used in determining compliance with the

regulatory limits.

The references, as proposed to be included

in section 5.14 of the Technical

Specifications, have previously been

reviewed and approved by the NRC for

generic applicability to PWRs [Pressurized

Water Reactors]. The reports identified in the

Proposed Change have been accepted by the

NRC for referencing in plant licensing

applications.

Since the references listed in the Proposed

Change have previously been found to meet

the conditions of 10 CFR 50.46 and 10 CFR

Appendix K, and that the plant specific

safety analysis acceptance limits have not

changed or been modified, the use of these

references in the analysis of SBLOCA

accident for the Maine Yankee plant is

consistent with prior plant specific and

industry requirements and practices.

Therefore, we have concluded that the

Proposed Change will not result in a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. Does the Proposed Amendment create

the possibility for a new or different kind of

accident?

The Proposed Changes introduce no new

mode of plant operation; do not involve the

physical modification of any structure,

system, or component; do not affect the

function, operation or surveillance for any

equipment necessary for safe operation or

shutdown of the plant; and, do not involve

any changes to setpoints or limits or

operating parameters. The Proposed Changes

are administrative in nature only.

Therefore, we have concluded that the

Proposed Change cannot result in the

possibility of a new or different kind of

accident from that previously evaluated.

3. Does the Proposed Amendment involve

a significant reduction in a margin of safety?

The Proposed Changes are administrative

in nature, consistent with the guidance of

Generic Letter 88–12, and have been

reviewed previously by the NRC and found

acceptable with regard to the requirements of

10 CFR 50.46 and 10 CFR Appendix K.

Additionally, the plant specific safety

analysis acceptance criteria has not changed

from that used in the latest core reload

analysis.

Therefore, we have concluded that the

Proposed Change does not involve a

significant reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Wiscasset Public Library, High

Street, P.O. Box 367, Wiscasset, ME

04578.

Attorney for licensee: Mary Ann

Lynch, Esquire, Maine Yankee Atomic

Power Company, 329 Bath Road,

Brunswick, ME 04011.
Federal Register / Vol. 61, No. 90 / Wednesday, May 8, 1996 / Notices 20849

NRC Deputy Director: John A.

Zwolinski.

Niagara Mohawk Power Corporation,

Docket No. 50–220, Nine Mile Point

Nuclear Station Unit No. 1, Oswego

County, New York

Date of amendment request: February

7, 1996.

Description of amendment request:

The amendment would change the

operating license, the Technical

Specifications, and associated Bases to

permit the use of 10 CFR Part 50,

Appendix J, Option B, PerformanceBased Containment Leakage Rate

Testing in accordance with the

implementation guidance in NRC’s

Regulatory Guide 1.163 dated

September 1995. The change to the

operating license would delete, in

paragraph 2.D.ii, reference to certain

exemptions to Appendix J previously

granted by the NRC, which would no

longer be applicable once Option B is

implemented.

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 operation of Nine Mile Point Unit

2, in accordance with the proposed

amendment, will not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

NMP2 [Nine Mile Point, Unit 2] is

currently implementing Option A of

Appendix J of 10 CFR 50 for Type A, B and

C testing. The proposed change to the

Operating License, the Technical

Specifications and the Bases would

implement Option B to Appendix J of 10 CFR

50 at NMP2 for Type A, B and C testing.

Option B would allow increased testing

intervals after satisfying certain performance

based criteria. The proposed change also

corrects an inconsistency between the

restoration statements and the applicability

requirements of LCO [Limiting Condition of

Operation] 3.6.1.2. In addition, the proposed

change affects the testing intervals for the

verification of the interlocks on the primary

containment air lock and for the measuring

of the Hydrogen Recombiner System leakage

rate.

Appendix J describes the requirements for

leakage testing of the primary containment

and its components penetrating the primary

containment. The leakage testing interval of

the primary containment and its components

is not a precursor or initiator to an accident.

The primary containment and its

penetrations minimizes the leakage of

radioactivity into the environment during an

accident which pressurizes the primary

containment.

The testing intervals of the air lock

interlocks and of the Hydrogen Recombiner

System leakage rate are also not precursors or

initiators to an accident. The interlocks

function to provide assurance that at least

one air lock door will be closed and thereby

perform its accident mitigating function of

minimizing the leakage of radioactivity into

the environment during accident conditions.

The Hydrogen Recombiner System is

manually initiated following a loss-of-coolant

accident (LOCA) to maintain the hydrogen

concentration within the primary

containment below its flammable limit

during post-LOCA conditions.

An inconsistency exists between the

applicability statement of LCO 3.6.1.2 and

the requirement of the restoration statements

to restore prior to increasing reactor coolant

system temperature over 200 °F. Eliminating

this inconsistency does not diminish the

requirements contained in the Technical

Specifications.

Therefore, the proposed change does not

involve a significant increase in the

probability of an accident previously

evaluated.

The proposed change to the Operating

License, the Technical Specifications and the

Bases would replace the detailed and

prescriptive technical requirements

contained in Option A of Appendix J with

performance based requirements and

supporting regulatory/industry documents

contained in Option B of Appendix J. This

proposed change includes a description of

the 10 CFR 50 Appendix J Testing Program

Plan in Section 6.8.4.f of the Technical

Specifications.

This program plan, with one exception, is

consistent with RG [Regulatory Guide] 1.163.

This exception to the RG is acceptable as it

is technically equivalent to and replaces an

exemption that was applicable to Option A

of Appendix J. Therefore, this program plan

establishes leakage-rate test methods,

procedures, acceptance criteria and analyses

which comply with Option B of Appendix J

to 10 CFR 50.

The implementation of this program

continues to provide adequate assurance that

during a DBA [Design Basis Accident]-LOCA

the primary containment and its components

will continue to limit leakage rates to less

than the allowable leakage rates described in

the Technical Specifications and thereby

limit leakage consistent with the assumptions

of the accident analyses. Therefore, the

increased test intervals permitted by Option

B for the primary containment and its

penetrations will continue to implement the

safety objectives underlying the requirements

of Appendix J.

As discussed under the margin of safety,

the impact of the proposed change on the

consequences of a release is negligible. The

slight increase in the risk to the population

is compensated by the corresponding risk

reduction benefits associated with the

reduction in component cycling, stress, and

wear associated with increased test intervals.

At least one air lock door in each air lock

will continue to be closed during the onset

of an accident that would release

radioactivity into primary containment.

Therefore, the air lock interlocks continue to

provide assurance that at least one leak tested

barrier will limit leakage during accident

conditions.

The Hydrogen Recombiner System will

continue to operate to maintain the hydrogen

concentration within the primary

containment below its flammable limit

during post-LOCA conditions. This provides

assurance that primary containment integrity

will not be challenged by hydrogen burns.

Eliminating the inconsistency between the

restoration statements and the applicability

requirements of LCO 3.6.1.2 does not

diminish the requirements contained in the

Technical Specifications. The Technical

Specifications continue to require that the

leakage limits of LCO 3.6.1.2 be met prior to

entering OPERATIONAL CONDITIONS 1, 2,

or 3 (i.e., temperature greater than 200 °F).

Accordingly, operation with the proposed

change to the Operating License, the

Technical Specifications and the Bases will

not significantly increase the consequences

of an accident previously evaluated.

2. The operation of Nine Mile Point Unit

2, in accordance with the proposed

amendment, will not create the possibility of

a new or different kind of accident from any

accident previously evaluated.

The proposed change would implement

Option B of Appendix J of 10 CFR 50 for

Type A, B and C testing. Option B would

allow increased testing intervals after

satisfying certain performance based criteria.

The proposed change also corrects any

inconsistency between the restoration

statements and the applicability

requirements of LCO 3.6.1.2. In addition, the

proposed change affects the testing intervals

for the interlocks on the primary containment

air lock and for the measuring of the

Hydrogen Recombiner System leakage rate.

No new plant operating modes, system

operating configurations nor failure modes

are introduced by the proposed change. The

primary containment and its penetrations

will continue to perform their accident

mitigating function. The Hydrogen

Recombiner System will continue to function

to prevent hydrogen burns within primary

containment during post-LOCA conditions.

Accordingly, operation with the proposed

change will not create the possibility of a

new or different kind of accident from any

accident previously evaluated.

3. The operation of Nine Mile Point Unit

2, in accordance with the proposed

amendment, will not involve a significant

reduction in a margin of safety.

A regulatory impact analysis of

implementing performance-based

requirements indicates that relaxing the

frequency of Type A, B and C testing leads

to an increase in overall reactor risk of

approximately two percent. As indicated in

the Staff’s Regulatory Impact Analysis, this

increase is considered to be marginal to

safety.

As indicated above, increasing test

intervals can slightly increase the risk to the

population associated with the consequences

of a release; however, this is compensated by

the corresponding risk reduction benefits

associated with the reduction in component

cycling, stress, and wear associated with

increased test intervals. Therefore, when

considering the total integrated risk, the risk

associated with increased test intervals is

negligible.
20850 Federal Register / Vol. 61, No. 90 / Wednesday, May 8, 1996 / Notices

The proposed change is consistent with

current plant safety analyses. In addition, the

proposed change does not require revisions

to the design of NMP2. As such, the proposed

individual changes will maintain the same

level of reliability of the equipment

associated with containment integrity,

assumed to operate in the plant safety

analysis, or provide continued assurance that

specified parameters affecting leak rate

integrity, will remain within their acceptance

limits.

The as-left leakage after performing a

required leakage test continues to be less

than 0.60 La for combined Type B and C

leakage and less than or equal to 0.75 La for

Type A leakage. These as-left acceptance

criteria and the testing frequency as

established by the 10 CFR 50 Appendix J

Testing Program Plan provide assurance that

the measured leakage rate will not exceed the

maximum allowable leakage of La during

plant operation.

Visual examination of accessible interior

and exterior surfaces of the primary

containment continues to be performed prior

to initiating a Type A test. The total number

of visual examinations performed will

continue to be three times during a 10-year

period. Therefore, visual examinations of the

primary containment will continue to allow

for the timely uncovering of evidence of

structural deterioration and satisfy the

requirements of RG 1.163.

The primary containment air lock

interlocks will be tested prior to conducting

an air lock seal leakage test. This testing

requirement continues to provide adequate

assurance that at least one leak tested air lock

door in each air lock will be closed during

accident conditions.

The measuring of the Hydrogen

Recombiner System Leakage rate will

continue to be included as part of the overall

integrated leakage rate test. The test schedule

for measuring system leakage will also

continue to coincide with the schedule for

performing a Type A test.

The leakage limits of LCO 3.6.1.2 will

continue to be met prior to entering into

OPERATIONAL CONDITIONS 1, 2, or 3 (i.e.,

temperature greater than 200 °F). Satisfying

these leakage limits provides assurance that

the measured leakage rate will not exceed the

maximum allowable leakage rate of La during

plant operation. Therefore, operation with

the proposed change will not involve a

significant reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 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: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston & Strawn,

1400 L Street, NW., Washington, DC

20005–3502.

NRC Project Director: Susan Frant

Shankman, Acting

Niagara Mohawk Power Corporation,

Docket No. 50–410, Nine Mile Point

Nuclear Station, Unit 2, Oswego County,

New York

Date of amendment request: January

17, 1996.

Description of amendment request:

The proposed amendment would revise

the Technical Specifications (TSs)

including revisions to Specifications 3/

4.3.1, ‘‘Reactor Protection System

Instrumentation,’’ 3/4.3.2, ‘‘Isolation

Actuation Instrumentation,’’ 3/4.3.3,

‘‘Emergency Core Cooling System

Actuation Instrumentation,’’ 3/4.3.4.2,

‘‘End-of-Cycle Recirculation Pump Trip

System Instrumentation,’’ and the

associated Bases to relocate response

time limit tables from the TSs to the

Updated Safety Analysis Report

(USAR). The proposed revisions to the

TSs also include several administrative

changes.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The operation of Nine Mile Point Unit

2, in accordance with the proposed

amendment, will not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

The proposed amendment relocates Tables

3.3.1–2, ‘‘Reactor Protection System

Response Times,’’ 3.3.2–3, ‘‘Isolation System

Instrumentation Response Times’’ 3.3.3–3,

‘‘Emergency Core Cooling System Response

Times’’ and 3.3.4.2–3 ‘‘End-of-Cycle

Recirculation Pump Trip System Response

Time’’ from the Technical Specifications to

the USAR. The Technical Specification

Surveillance Requirements and associated

actions are not affected and remain in the

Technical Specifications. This change to the

reactor protection system instrumentation,

isolation actuation instrumentation, and

emergency core cooling system

instrumentation is being done in accordance

with the guidance provided in Generic Letter

93–08, ‘‘Relocation of Technical

Specification Tables of Instrument Response

Time Limits,’’ and the change to the end-ofcycle recirculation pump trip system

instrumentation is consistent with NUREG

1433, ‘‘Standard Technical Specifications,

BWR/4.’’ This change allows NMP2 [Nine

Mile Point Unit 2] to administratively control

subsequent changes to the response time

limits in accordance with 10CFR50.59.

Additionally, procedures which contain the

various response time limits are also subject

to the change control provisions of 10 CFR

50.59. Relocating this information does not

affect the initial conditions of a design basis

accident or transient analysis. The proposed

Technical Specification changes do not affect

the capability of the associated systems to

perform their intended functions within their

required response times. Since any

subsequent changes to the USAR or

procedures which contain the response time

limits are evaluated in accordance with

10CFR50.59, the proposed amendment does

not involve an increase in the probability or

consequences of an accident previously

evaluated.

2. The operation of Nine Mile Point Unit

2, in accordance with the proposed

amendment, will not create the possibility of

a new or different kind of accident from any

previously evaluated.

The proposed change would relocate the

response time limit tables from the Technical

Specifications to the USAR. Subsequent

changes to the USAR, or in procedures which

contain the various response time limits,

would be evaluated in accordance with the

requirements of 10CFR50.59, which would

evaluate the possibility of the creation of a

new or different kind of accident. The

proposed change does not involve any

physical alteration of the plant, change in a

Limiting Condition for Operation or change

in Surveillance Requirements. No new failure

modes are introduced. Therefore, this

proposed change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

3. The operation of Nine Mile Point Unit

2, in accordance with the proposed

amendment, will not involve a significant

reduction in a margin of safety.

The proposed change would relocate the

response time limit tables from the Technical

Specifications to the USAR. Future changes

to the response time limits in the USAR, or

in procedures which contain the various

response time limits, would be in accordance

with 10CFR50.59, which would evaluate the

proposed change to determine whether it

involved any reduction in the margin of

safety. The response time limits to be

transposed from the Technical Specifications

to the USAR are the same as the existing

Technical Specifications. Therefore, this

proposed change does not involve a

significant reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 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: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston & Strawn,

1400 L Street, NW., Washington, DC

20005–3502.

NRC Project Director: Susan Frant

Shankman, Acting.
Federal Register / Vol. 61, No. 90 / Wednesday, May 8, 1996 / Notices 20851

Niagara Mohawk Power Corporation,

Docket No. 50–410, Nine Mile Point

Nuclear Station, Unit 2, Oswego County,

New York

Date of amendment request: January

25, 1996.

Description of amendment request:

The proposed amendment would

change a footnote in Table 3.3.3–1 and

the corresponding footnote in

surveillance Table 4.3.3.1–1 (both

referenced by Technical Specification 3/

4.3.3 ‘‘Emergency Core Cooling System

Actuation Instrumentation’’) to more

clearly define when, during cold

shutdown and refueling (i.e.,

Operational Conditions 4 and 5), the

Loss of Voltage and Degraded Voltage

relays associated with the 4.16 kV

Emergency Bus Undervoltage are

required to be operable. The footnotes

currently state: ‘‘Required when ESF

[Engineered Safety Features] equipment

is required to be OPERABLE.’’ The

proposed amendment would change the

footnotes to state: ‘‘Required when the

associated diesel generator is required to

be OPERABLE.’’

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 operation of Nine Mile Point Unit 2,

in accordance with the proposed

amendment, will not involve a significant

increase in the probability or consequence of

an accident previously evaluated.

The proposed change would require the

Loss of Power instruments to be OPERABLE

in Operational Conditions 4 and 5 only when

the associated diesel generator is required to

be OPERABLE. The Loss of Power relays

provide a support function to initiate the

associated diesel generator start and bus

unloading sequences. If that diesel generator

is not in service, the loss of power relays

perform no safety function. Therefore,

relating diesel generator OPERABILITY and

Loss of Power instrument OPERABILITY will

not involve an increase in the probability of

an accident previously evaluated.

The proposed change does not affect the

requirements of ESF OPERABILITY. The

change does not affect diesel generator

response to a loss of voltage or degraded

voltage on the Divisional 4.16 kV electrical

busses when the diesel generator is required

to be OPERABLE. Automatic response of the

ESF functions is unaffected by removing the

Loss of Power relays from service under these

conditions, therefore, the proposed change

will not involve a significant increase in the

consequences of an accident previously

evaluated.

The operation of Nine Mile Point Unit 2,

in accordance with the proposed

amendment, will not create the possibility of

a new or different kind of accident from any

accident previously evaluated.

The proposed change does not involve a

modification of plant equipment nor does it

change the way the equipment will be

maintained or operated. The revision to

Technical Specifications will continue to

require the Loss of Power instrumentation to

be OPERABLE when the associated diesel

generator is required to be OPERABLE. The

Loss of Power instruments will continue to

perform their safety function of initiating the

diesel generator start and bus unloading

sequences.

Therefore, this proposed change will not

create the possibility of a new or different

kind of accident from any previously

evaluated.

The operation of Nine Mile Point Unit 2,

in accordance with the proposed

amendment, will not involve a significant

reduction in a margin of safety.

The proposed change will not affect the

OPERABILITY, operation or reliability of any

ESF function including the diesel generators.

All ESF functions will remain available

during postulated accidents with a loss of

offsite electrical power. The change simply

clarifies when the Loss of Power instruments

are required to be OPERABLE during

Operational Conditions 4 and 5. Therefore,

the proposed change will not involve a

significant reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 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: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston & Strawn,

1400 L Street, NW., Washington, DC

20005–3502.

NRC Project Director: Susan Frant

Shankman, Acting.

Niagara Mohawk Power Corporation,

Docket No. 50–410, Nine Mile Point

Nuclear Station, Unit 2, Oswego County,

New York

Date of amendment request: March

15, 1996.

Description of amendment request:

The proposed amendment would revise

the surveillance requirements of

Technical Specification (TS) 4.6.2.1

‘‘Containment Systems—

Depressurization Systems—Suppression

Pool’’ to extend the time interval for

performing the containment drywell-tosuppression chamber bypass leakage

test from 18 months to an interval

corresponding to that required for the

Containment Integrated Leak Rate Test.

The provisions of TS 4.0.2 (which

would provide an extension of up to

25% of the specified surveillance

interval) will not apply. Specifically,

existing TS 4.6.2.1.d would become

subparagraphs d and e to require that

the suppression pool be demonstrated

operable:

d. At least once per 18 months by

conducting a visual inspection of the

exposed accessible interior and exterior

surfaces of the suppression chamber.*

e. At least every outage by requiring the

performance of a Containment Integrated

Leak Rate Test, as scheduled in conformance

with the criteria specified in the 10 CFR 50

Appendix J Testing Program Plan described

in Section 6.8.4.f, by conducting a drywellto-suppression chamber bypass leak test at an

initial differential pressure of 3 psi and

verifying that the [drywell-to-suppression

chamber bypass flow area] A/the square root

of K calculated from the measured leakage is

within the specified limit of 0.0054 square

feet.

1. If any drywell-to-suppression chamber

bypass leak test fails to meet the specified

limit, the test schedule for subsequent tests

shall be reviewed and approved by the

Commission.

2. If two consecutive tests fail to meet the

specified limit, a test shall be performed at

least each refueling outage until two

consecutive tests meet the specified limit, at

which time the original test schedule may be

resumed.

3. The provisions of Specification 4.0.2 do

not apply.

*Includes each vacuum relief valve and

associated piping.

The proposed changes would also add

a new surveillance requirement for the

testing of the bypass leakage path

containing the suppression chamber

vacuum breakers, with associated

acceptance criteria, which would be

performed each refueling outage that the

bypass leak test is not performed.

Specifically, a new TS 4.6.2.1f would

require that the suppression pool be

demonstrated operable:

f. During each refueling outage for which

the drywell-to-suppression chamber bypass

leak test in Specification 4.6.2.1.e is not

conducted, by conducting a test of the four

drywell-to-suppression chamber bypass leak

paths containing the suppression chamber

vacuum breakers at a differential pressure of

at least 3 psi and

1. Verifying that the total leakage area A/

the square root of K contributed by all four

bypass leak paths is less than or equal to 24%

of the specified limit, and

2. The leakage area for any one of the four

bypass leak paths is less than or equal to 12%

of the specified limit.

By separate action, the NRC has

provided notice of a proposed

amendment to change the frequency of

Containment Integrated Leak Rate Tests

in accordance with Option B of 10 CFR

Part 50 Appendix J . The proposed

changes described herein are intended
20852 Federal Register / Vol. 61, No. 90 / Wednesday, May 8, 1996 / Notices

to be consistent with the changes

proposed under Option B.

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 operation on Nine Mile Point Unit 2,

in accordance with the proposed

amendment, will not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

The proposed TS changes involve the

drywell-to-suppression chamber bypass leak

test frequency. There are no physical or

operational changes to the plant as a result

of these proposed TS revisions. Furthermore,

the primary containment acts as an accident

mitigator and not as an accident initiator.

Therefore, the proposed TS changes do not

affect the probability of any previously

evaluated accident.

The continued testing of bypass leakage

pathways containing the suppression

chamber vacuum breakers on a refueling

frequency, and the continued requirement for

visual inspection of containment structural

features assures that the bypass leakage path

will not degrade beyond the TS allowable

limit during the interval between

performance of the bypass leakage test.

Therefore, radioactivity release following an

accident will not be increased since the

pressure suppression capability of the

containment is not reduced from the existing

design, and there will be no significant

increase in the consequences of any accident

previously evaluated.

The operation of Nine Mile Point Unit 2,

in accordance with the proposed

amendment, will not create the possibility of

a new or different kind of accident from any

accident previously evaluated.

The proposed TS changes involve the

drywell to suppression chamber bypass leak

test frequency. There are no physical or

operational changes as a result of these

proposed TS changes. These proposed TS

changes also include a requirement to

continue performing a surveillance test on

the bypass leakage pathways containing the

vacuum breaker assemblies each refueling

outage for which the drywell-to-suppression

chamber test is not conducted. This test,

along with the visual inspection required

every refueling cycle, will ensure that

acceptable bypass leakage is maintained

during those intervals when the bypass leak

test is not required. Accordingly, the

possibility of a new or different type of

accident is not introduced. Therefore, the

proposed TS changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The operation of Nine Mile Point Unit 2,

in accordance with the proposed

amendment, will not involve a significant

reduction in a margin of safety.

The drywell-to-suppression chamber

bypass leak test data obtained during

previous testing at NMP2 [Nine Mile Point

Unit 2] demonstrates conformance, by a large

margin, to the TS and design leakage

requirements. The test data and engineering

evaluations indicate that there is negligible

risk that the bypass leakage will change

adversely in future years. Furthermore, the

proposed test frequency is judged to be

acceptable based on the small risk of bypass

leakage through paths other than those

containing the suppression chamber vacuum

breakers.

A test of the bypass leak pathways

containing the vacuum breakers will be used

to verify acceptable bypass leakage during

those outages when the bypass leak test is not

performed. The proposed test of the bypass

leak pathways containing the vacuum

breakers, with stringent acceptance criteria,

combined with the other negligible potential

leakage areas provide an acceptable level of

assurance that the bypass leakage can be

measured. This capability ensures that an

adverse condition can be detected and

corrected such that the existing level of

confidence that the primary containment will

function as required during a LOCA [loss-ofcoolant accident] is maintained. Therefore,

the proposed TS changes do not involve a

significant reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston & Strawn,

1400 L Street, NW., Washington, DC

20005–3502.

NRC Project Director: Susan Frant

Shankman, Acting.

Niagara Mohawk Power Corporation,

Docket No. 50–410, Nine Mile Point

Nuclear Station, Unit 2, Oswego County,

New York

Date of amendment request: March

20, 1996.

Description of amendment request:

The proposed amendment would revise

Tables 3.3.1–1 and 4.3.1–1 of Technical

Specification 3/4.3.1 ‘‘Reactor

Protection System Instrumentation’’ to

delete the operability requirement for

the Average Power Range Monitor

(APRM) Neutron Flux-Upscale,

Setdown and Inoperative functions in

Operational Conditions (OCs) 3 (Hot

Shutdown) and 4 (Cold Shutdown).

These same functions would also be

revised for OC 5 (Refueling) to indicate

that operability will only be required

during shutdown margin

demonstrations performed per TS

3.10.3.

Basis for proposed no significant

hazards consideration determination:

The revisions to the APRM functions are

proposed to support licensee’s plans to

replace Local Power Range Monitors

during the next refueling outage. The

revisions also provide for the eventual

replacement of the existing APRM

System with the Nuclear Measurement

Analysis and Control Power Range

Neutron Monitoring System, and the

eventual installation of the Oscillation

Power Range Monitor system for the

detection of reactor instability

conditions. These modifications are

based upon Report NEDO–31960, ‘‘BWR

Owners’ Group Long-Term Solutions

Licensing Methodology, approved by

the Commission July 12, 1993; the

licensee’s response of November 8,

1994, selecting Option III in NEDO–

31960 for Nine Mile Point, Unit 2; NRC

Generic Letter 94–02, ‘‘Long-Term

Solutions and Upgrade of Interim

Operating Recommendations for

Thermal-Hydraulic Instabilities in

Boiling Water Reactors’’ dated July 11,

1994; and General Electric Licensing

Topical Report, NEDC–32410P–A,

‘‘Nuclear Measurement Analysis and

Control Power Range Neutron Monitor

(NUMAC–PRNM) Retrofit Plus Option

III Stability Trip Function,’’ which was

approved by the Commission September

5, 1995.

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 operation of Nine Mile Point Unit 2 in

accordance with the proposed amendment

will not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The Reactor Protection System (RPS)

initiates a reactor scram when one or more

monitored parameters exceed their specified

limits to preserve the integrity of the fuel

cladding and the Reactor Coolant System and

to minimize the energy that must be absorbed

following a loss-of-coolant accident. The

proposed changes will revise the OCs in

which the APRM Neutron Flux-Upscale,

Setdown and Inoperative RPS

Instrumentation is required. These changes

do not affect the probability of precursors of

any accidents previously evaluated, and

therefore, do not increase their probability.

During normal operation in OCs 3 and 4,

all control rods are fully inserted and the

reactor mode switch position control rod

withdrawal blocks do not allow control rods

to be withdrawn. Therefore, the RPS APRM

functions are not required. Specification

3.9.10 does allow one control rod to be

removed from the core in OC 4 by placing the

mode switch in the refuel position. However,

with the reactor mode switch in the refuel

position, refueling interlocks are in place

(i.e., one-rod out, etc.), which together with
Federal Register / Vol. 61, No. 90 / Wednesday, May 8, 1996 / Notices 20853

adequate shutdown margin will preclude

unacceptable reactivity excursions. The

APRM Neutron Flux-Upscale, Setdown

function is not required during OC 5 except

during shutdown margin demonstrations.

The SRMs [source range monitors], IRMs

[intermediate range monitors], and refueling

interlocks provide adequate protection from

reactivity excursions during OC 5. The

exception is during the shutdown margin

demonstration when more than one control

rod will be withdrawn and the APRMs will

continue to be required to be operable as a

backup to the IRMs. Testing of the RPS

APRM functions will continue to be

performed in those OCs for which operability

is required. Consequently, the reliability and

performance of the RPS APRM functions in

these OCs will not be adversely affected.

Therefore, the proposed change will not

result in a significant increase in the

consequences of any accidents previously

evaluated.

The operation of Nine Mile Point Unit 2 in

accordance with the proposed amendment

will not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed changes will revise the

applicable OCs in which the APRM neutron

Flux-Upscale, Setdown and Inoperative RPS

instrumentation is required. Changes to OC

requirements will not introduce any new

accident precursors and will not involve any

physical alternations to plant configurations

which could initiate a new or different kind

of accident. NMP2 is analyzed for a single

control rod withdrawal error during

refueling. Since the core is designed to meet

shutdown requirements with the highest

worth rod withdrawn, the core remains

subcritical even with one rod withdrawn.

The one-rod-out interlock which allows only

one control rod to be withdrawn in OC 5 is

not affected by the proposed changes.

Consequently, the proposed changes do not

create an accident different than the

previously analyzed single control rod

withdrawal error event. Surveillance testing

will continue to be performed to assure

reliability and maintain current performance

levels. Therefore, the proposed change will

not create the possibility of a new or different

kind of accident from any previously

evaluated.

The operation of Nine Mile Point Unit 2 in

accordance with the proposed amendment

will not involve a significant reduction in a

margin of safety.

The proposed changes to the RPS APRM

function instrumentation Technical

Specification requirements will not adversely

affect the design or the performance

characteristics of the RPS instrumentation

nor will it affect the ability of the RPS APRM

instrumentation to perform its intended

function. As discussed above, the subject

RPS instrumentation is not required in OC 3,

4, and 5 except for shutdown margin

demonstrations. Accordingly, deletion of the

requirement to have these functions operable

in these OCs will not significantly reduce a

margin of safety. Surveillance testing will

continue to be performed for those OCs in

which the instrumentation is required to

assure reliability. 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 request

involves no significant hazards

consideration.

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston & Strawn,

1400 L Street, NW., Washington, DC

20005–3502.

NRC Project Director: Susan Frant

Shankman, Acting.

Northeast Nuclear Energy Company, et

al., Docket No. 50–336, Millstone

Nuclear Power Station, Unit No. 2, New

London, Connecticut

Date of amendment request: March

28, 1996.

Description of amendment request:

The proposed amendment would

change Technical Specification Section

3.7.7, ‘‘Sealed Source Contamination,’’

by making the criteria for testing sealed

sources for contamination and leakage

at Millstone Unit No. 2 the same as

those at Millstone Unit No. 3, the

Haddam Neck Plant, and Seabrook

Station. Specifically, the sealed sources

that are required to be free of greater

than or equal to 0.005 microcuries of

removable contamination would be

those that would exceed ‘‘100

microcuries of beta and/or gamma

emitting material or 5 microcuries of

alpha emitting material.’’ The Bases

Section 3/4.7.7, ‘‘Sealed Source

Contamination,’’ would also be changed

to reference the appropriate section of

10 CFR 70.39.

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 (SHC), which is presented

below:

Pursuant to 10 CFR 50.92, NNECO

[Northeast Nuclear Energy Company] has

reviewed the proposed changes and

concludes that the changes do not involve a

significant hazards consideration (SHC) since

the proposed changes satisfy the criteria in

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

changes do not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The changes make the criteria for testing

sealed sources for contamination and leakage

at Millstone Unit No. 2 the same as those at

Millstone Unit No. 3, the Haddam Neck Plant

and Seabrook Station. Although the leakage

criteria for sealed sources that are to be tested

is being changed, the allowable leakage

remains small. Any leakage that is identified

would not cause a significant radiation

exposure. The source storage area is routinely

surveyed by Health Physics in accordance

with Health Physics Department procedures

and any significant leakage would be

detected. Therefore, this change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed change in the criteria for

testing sealed sources for contamination and

leakage will not change the way the sources

are used. Therefore, this change will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Involve a significant reduction in a

margin of safety.

The possible radiation exposure to both the

workers and the public from this change is

very small. All protective systems which

would detect any release of material from the

site remain in place so there is no reduction

in safety for the public. Likewise, all

protective systems for the workers remain in

place. Workers using the sources routinely

pass through the whole body contamination

monitors. In addition, the source storage

areas are surveyed routinely by Health

Physics in accordance with Health Physics

Department procedures, and any significant

leakage would be detected. The bases section

is being revised to reference the appropriate

section of 10 CFR 70.39. Therefore, there is

no 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: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360, and Waterford

Library, Attn: Vince Juliano, 49 Rope

Ferry Road, Waterford, CT 06385.

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

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

NRC Project Director: Phillip F.

McKee.

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 amendment request: March

12, 1996.
20854 Federal Register / Vol. 61, No. 90 / Wednesday, May 8, 1996 / Notices

Description of amendment request:

The proposed changes would remove a

requirement to interconnect two or more

accumulators for the purpose of cross

checking instrumentation in the event

that one of the two pressure or level

instrument channels on an accumulator

is declared inoperable.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) Does the proposed license amendment

involve a significant increase in the

probability or consequences of an accident

previously analyzed?

Response: The design basis accident for

which the accumulators were designed is the

double ended guillotine break of a cold leg.

Interconnecting or not interconnecting

accumulators does not have any effect on the

probability of occurrence of this event. By

eliminating the requirement to interconnect

accumulators, the proposed amendment

assures that a minimum of three

accumulators are available, as assumed in the

safety analyses, to mitigate the consequences

of a large-break loss-of-coolant [LBLOCA]

accident. Therefore, the proposed

amendment does not involve a significant

increase in the probability or consequences

of an accident previously analyzed.

(2) Does the proposed license amendment

create the possibility of a new or different

kind of accident from any accident

previously evaluated?

Response: The proposed amendment does

not involve any physical changes to plant

equipment or setpoints and does not create

the possibility of a new or different kind of

accident. Eliminating the requirement to

interconnect accumulators ensures that the

plant configuration is maintained consistent

with that assumed in the safety analysis and

no new failure modes are created.

(3) Does the proposed amendment involve

a significant reduction in a margin of safety?

Response: There is no margin of safety

specified in the Technical Specifications for

these instrument channels. There are no

setpoints or allowable values associated with

these instrument channels which affect

Safety Limits or Limiting Safety System

Settings. The proposed amendment ensures

that the safety analysis assumption regarding

the accumulators remains valid and the

resulting peak fuel clad temperature meets

specified acceptance criteria. The proposed

amendment does not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 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: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10601.

Attorney for licensee: Mr. Charles M.

Pratt, 10 Columbus Circle, New York,

New York 10019.

NRC Project Director: Susan Frant

Shankman, Acting.

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 amendment request: March

14, 1996.

Description of amendment request:

The proposed changes would allow a

one-time extension of the inspection

interval for the steam generator tubes

that is due in July 1996.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) Does the proposed license amendment

involve a significant increase in the

probability or consequences of an accident

previously evaluated?

Response: The proposed license

amendment does not involve a significant

increase in the probability or consequences

of an accident previously evaluated. As

stated in the Basis of the IP3 [Indian Point

Unit 3] Technical Specifications, the program

for inservice inspection of steam generator

tubes regarding equipment, procedures, and

sample selection is based upon the guidance

and recommendations in Regulatory Guide

1.83 and NRC Generic Letter 85–02. The

addition of the footnote to extend the

surveillance due date will not increase the

deviation from the guidance and

recommendation stated above, and, therefore

will not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

(2) Does the proposed license amendment

create the possibility of a new or different

kind of accident from any accident

previously evaluated?

Response: The proposed license

amendment does not create the possibility of

a new or different kind of accident from any

accident previously evaluated. The proposed

change does not involve the addition of any

new or different type of equipment, nor does

it involve the operation of equipment

required for safe operation of the facility in

a manner different from those addressed in

the Final Safety Analysis Report. Therefore,

the proposed change will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

(3) Does the proposed amendment involve

a significant reduction in a margin of safety?

Response: The proposed license

amendment does not involve a significant

reduction in a margin of safety. The proposed

change does not adversely affect any safety

related system or component operation or

operability, instrument operation, or safety

system setpoints and does not result in

increased severity of any of the accidents

considered in the safety analysis. This

change has no adverse effect on any margin

of safety and, therefore, does not create a

significant reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10601.

Attorney for licensee: Mr. Charles M.

Pratt, 10 Columbus Circle, New York,

New York 10019.

NRC Project Director: Susan Frant

Shankman, Acting

Power Authority of the State of New

York, Docket No. 50–333, James A.

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of amendment request: March

22, 1996.

Description of amendment request:

The amendment proposes changes to

the Technical Specifications (TS) to

establish operability requirements for

avoidance and protection from thermal

hydraulic instabilities to be consistent

with Boiling Water Reactor Owners

Group long-term solution Option I–D.

Editorial changes are also made to

support the revised specifications,

improve readability of Bases sections,

and enhance the presentation of

requirements for single loop operation.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Operation of the FitzPatrick plant in

accordance with the proposed Amendment

would not involve a significant hazards

consideration as defined in 10 CFR 50.92,

since it would not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated because:

The implementation of BWR Owners’

Group long-term stability solution Option

I–D at FitzPatrick does not modify the

assumptions contained in the existing

accident analysis. The use of an exclusion

region and the operator actions required to

avoid and minimize operation inside the

region do not increase the possibility of an

accident. Conditions of operation outside of

the exclusion region are within the analytical

envelope of the existing safety analysis. The

operator action requirement to exit the
Federal Register / Vol. 61, No. 90 / Wednesday, May 8, 1996 / Notices 20855

exclusion region upon entry minimizes the

possibility of an oscillation occurring. The

actions to drive control rods and/or to

increase recirculation flow to exit the region

are maneuvers within the envelope of normal

plant evolutions. The flow referenced scram

has been analyzed and will provide

automatic fuel protection in the event of an

instability. Thus, each proposed operating

requirement provides defense in depth for

protection from an instability event while

maintaining the existing assumptions of the

accident analysis.

2. Create the possibility of a new or

different kind of accident from those

previously evaluated because:

The proposed operating requirements

either mandate operation within the

envelope of existing plant operating

conditions or force specific operating

maneuvers within those carried out in

normal operation. Since operation of the

plant with all of the proposed requirements

are within the existing operating basis, an

unanalyzed accident will not be created

through implementation of the proposed

change.

3. Involve a significant reduction in the

margin of safety because:

Each of the proposed requirements for

plant thermal hydraulic stability provides a

means for fuel protection. The combination

of avoiding possible unstable conditions and

the automatic flow referenced reactor scram

provides an in depth means for fuel

protection. Therefore, the individual or

combination of means to avoid and suppress

an instability supplements 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: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Attorney for licensee: Mr. Charles M.

Pratt, 1633 Broadway, New York, New

York 10019.

NRC Project Director: Susan Frank

Shankman, Acting.

Power Authority of the State of New

York, Docket No. 50–333, James A.

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of amendment request: March

22, 1996.

Description of amendment request:

The amendment proposes to revise

Technical Specification (TS) Table 3.2–

2, ‘‘Core and Containment Cooling

System Initiation and Control

Instrumentation Operability

Requirements.’’ The proposed changes

will revise allowed outage times (AOTs)

for 4kV Emergency Bus Undervoltage

Trip Functions. The AOTs for these trip

functions were extended by

Amendment 227; however, the AOT

extensions for these trip functions were

not consistent with the requirements of

Standard Technical Specifications

(STS), NUREG–1433, and differed from

the recommendations in the associated

Licensing Topical Report. Additional

changes are proposed to TS Table 3.2–

2 and to TS Table 4.2–2, ‘‘Core and

Containment Cooling System

Instrumentation Test and Calibration

Requirements.’’ These changes will: (1)

replace the generic actions for

inoperable instrument channels with

function-specific actions, (2) replace the

generic test AOT with function-specific

test AOTs, and (3) relocate selected trip

functions from the TS to an Authority

controlled document.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Operation of the FitzPatrick plant in

accordance with the proposed Amendment

would not involve a significant hazards

consideration as defined in 10 CFR 50.92,

since it would not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed changes are limited to

replacement of the generic actions and test

AOT with function-specific actions and test

AOTs, and relocation of selected trip

functions from the TS to an Authority

controlled document. The changes do not

introduce any new modes of plant operation,

make any physical changes, or alter any

operational setpoints. Therefore, the changes

do not degrade the performance of any safety

system assumed to function in the accident

analysis. Consequently, there is no effect on

the probability or consequences of an

accident.

2. Create the possibility of a new or

different kind of accident from those

previously evaluated.

The proposed changes do not introduce

any new accident initiators or failure

mechanisms since the changes do not

introduce any new modes of plant operation,

make any physical changes, or alter any

operational setpoints. Therefore the changes

do not create the possibility of a new or

different kind of accident.

3. Involve a significant reduction in the

margin of safety.

The proposed changes do not alter the

manner in which safety limits, limiting safety

system settings, or limiting conditions for

operation are determined. The relocated

requirements do not satisfy the 10 CFR 50.36

criteria for inclusion in the Technical

Specifications. Therefore, the changes do not

involve a significant reduction in the margin

of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Attorney for licensee: Mr. Charles M.

Pratt, 1633 Broadway, New York, New

York 10019.

NRC Project Director: Susan Frant

Shankman, Acting.

Power Authority of the State of New

York, Docket No. 50–333, James A.

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of amendment request: March

27, 1996.

Description of amendment request:

The amendment proposes to revise the

Technical Specifications to support

adoption of the primary containment

leakage rate testing requirements of

Option B to 10 CFR 50, Appendix J at

the FitzPatrick plant, and clarify the

numerical value of the allowable

containment leakage rate (La) as 1.5

percent per day.

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 Authority has evaluated the proposed

TS Amendment and determined that it does

not represent a significant hazards

consideration. Based on the criteria for

defining a significant hazards consideration

established in 10 CFR 50.92, operation of the

James A. FitzPatrick Nuclear Power Plant in

accordance with the proposed amendment

will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated because:

The proposed changes do not involve a

change to the design or operation of the

plant. The systems affected by this proposed

TS change are not assumed in any safety

analyses to initiate any accident sequence.

Therefore, the probability of any accident

previously evaluated is not increased by this

proposed TS change. The clarification of the

allowable containment leakage rate (La) is

consistent with the accident analyses. There

is no change to the consequences of an

accident previously evaluated because

maintaining leakage within limits assumed in

the accident analyses ensures that the dose

consequences resulting from an accident are

not increased. The proposed TS changes

maintain an equivalent level of reliability
20856 Federal Register / Vol. 61, No. 90 / Wednesday, May 8, 1996 / Notices

and availability for all affected systems. The

ability of the affected systems associated with

maintaining leak rate integrity to perform

their intended function is unaffected by the

proposed TS changes. Implementation of

these changes will provide continued

assurance that specified parameters

associated with containment integrity will

remain within acceptance limits, and as

such, will not significantly increase the

consequences of a previously evaluated

accident.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated because:

The proposed changes allow adoption of

those requirements specified in Option B to

10 CFR 50, Appendix J, and do not involve

a change to the plant design and operation.

As a result, the proposed changes do not

affect the parameters or conditions that could

contribute to the initiation of any accidents.

The methods of performing primary

containment leakage rate testing are not

changed. No new accident modes are created

by allowing extended intervals for Type A, B

and C testing, or by clarifying the numerical

value of the allowable containment leakage

rate (La). No safety-related equipment or

safety functions are altered, or adversely

affected, as a result of these changes. The

proposed changes will not introduce failure

mechanisms beyond those already

considered in the current plant safety

analyses. Extension of the test intervals, and

clarification of the allowable leakage rate,

does not contribute to the possibility of a

new or different kind of accident or

malfunction from those previously analyzed.

3. Involve a significant reduction in the

margin of safety because: The proposed

changes affect the frequency of primary

containment leakage rate testing, and the

numerical definition of the allowable

containment leakage rate (La). The design of

the FitzPatrick plant is not changed. The

methodology for test performance is

unchanged and Type A, B and C tests will

continue to be performed at ≥Pa. The

proposed changes provide sufficient controls

to ensure that proper maintenance and

repairs are performed on the primary

containment, and systems and components

penetrating the primary containment. The

reliability of containment systems assumed

to operate in the plant safety analyses is not

reduced. The numerical value of La specified

in Specification 6.20 is consistent with the

accident analyses, therefore, the dose

consequences of any analyzed accidents are

not increased. Therefore, the proposed

changes provide continued assurance of the

leak tightness of the containment without

adversely affecting the public health and

safety and, as such, will not involve a

significant reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Attorney for licensee: Mr. Charles M.

Pratt, 1633 Broadway, New York, New

York 10019.

NRC Project Director: Susan Frant

Shankman, Acting.

Public Service Electric & Gas Company,

Docket Nos. 50–272 and 50–311, Salem

Nuclear Generating Station, Unit Nos. 1

and 2, Salem County, New Jersey

Date of amendment request: April 22,

1996.

Description of amendment request:

The amendments would change the

Technical Specifications to implement

10 CFR Part 50, Appendix J, Option B,

for the Type A test by referring to

Regulatory Guide 1.163, ‘‘PerformanceBased Containment Leakage-Test

Program.’’

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

Containment leak rate testing is not an

initiator of any accident. The proposed

changes do not make any physical changes to

the containment. The proposed changes do

not affect performance of the containment,

reactor operations or accident analysis.

Therefore, the proposed changes will not

involve an increase in the probability of any

previously evaluated accident.

Since the allowable leakage rate is not

being changed and since the analysis

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

concludes that the impact on public health

and safety due to extended intervals is

negligible, the proposed changes will not

involve an increase in the consequences of

any previously evaluated accident. Therefore,

adoption of a performance-based verification

of leakage rates for the overall containment

boundary will provide an equivalent level of

safety and does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed change makes no physical

changes to the plant. Since no physical

changes are involved and since the analysis

documented in NUREG–1493 confirms that

the performance based schedule continues to

maintain a minimal impact on public risk, it

can be concluded that the effect of the

containment on any accident will not change.

The proposed change does not affect normal

plant operations or configuration, nor does it

affect leak rate test pressure.

Therefore, the proposed amendment will

not create the possibility of a new or different

kind of accident from any previously

evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

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

levels of reliability of containment integrity.

The performance-based approach to leakage

rate testing recognizes that historically good

results of containment testing provide

appropriate assurance of future containment

integrity. This supports the conclusion that

the impact on the health and safety of the

public as a result of extended test intervals

is negligible. Since the analysis documented

in NUREG–1493 confirms that the

performance based schedule continues to

maintain a minimal impact on public risk, it

can be concluded that the margin of safety is

not significantly affected by the proposed

changes.

The test history at Salem Units 1 and 2 (no

ILRT failures) provides continued assurance

of the leak tightness of the containment

structure.

Therefore, the proposed amendment will

not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Salem Free Public library, 112

West Broadway, Salem, New Jersey

08079.

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW, Washington,

DC 20005–3502.

NRC Project Director: John F. Stolz.

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of amendment request: April 4,

1996 (TS 96–01).

Description of amendment request:

The proposed change would revise the

appropriate technical specifications,

surveillances, and bases as needed for

the conversion from Westinghouse

nuclear fuel to Framatome Cogema

Mark-BW17 nuclear fuel.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:
Federal Register / Vol. 61, No. 90 / Wednesday, May 8, 1996 / Notices 20857

TVA has evaluated the proposed technical

specification (TS) change and has determined

that it does not represent a significant

hazards consideration based on criteria

established in 10 CFR 50.92(c). Operation of

Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The analyses provided in Topical Report

BAW–10220P show that the changes do not

significantly change the results of previously

evaluated events. These analyses provide the

template for accident analyses assumptions

that must be met by the cycle-specific reload

analysis.

The SQN Units 1 and 2 Cycle 9 reload

cores with Mark-BW fuel will be designed to

operate within the approved limits for

accident analysis. The limits provided in the

TS and described in the Updated Final Safety

Analysis Report (UFSAR) provide the

framework for accident analyses. By

maintaining these limits, the probability or

consequences of accidents related to the core

changes do not significantly change. Thus, it

is concluded that there is no significant

increase in the probability or consequences

of an accident previously evaluated.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

The change to Mark-BW fuel cores and

mixed (transition) cores has been evaluated

in the Topical Report BAW–10220P. It was

concluded that the change did not create new

or different kinds of accidents. The change in

fuel suppliers has been evaluated for

consideration of the effects of power

distribution and peaking factors such that

there are no restrictions on the use of MarkBW fuel assemblies beyond those already

established in the UFSAR and TS. Adherence

to the safety analysis limits restricts the

possibility of new or different accidents.

Historically, new accidents have not been

associated with changes in fuel suppliers as

long as safety analysis limits continue to be

met. It is concluded that transition to MarkBW fuel does 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 margin of safety is established by the

acceptance criteria used by NRC. Meeting the

acceptance criteria assures that the

consequences of accidents are within known

and acceptable limits. The loss-of-coolant

accident (LOCA) acceptance criteria are

unchanged: peak cladding temperature of

≤2200 degrees Fahrenheit, peak cladding

oxidation of ≤17 percent, average clad

oxidation of ≤1 percent, and long-term

coolability. These requirements continue to

be met. The methods used to demonstrate

conformance with these limits have changed,

and were reviewed to assure that the

methods, as well as the results, are

acceptable. The acceptance criteria for

Departure from Nucleate Boiling (DNB)

events has not changed and is still the 95

percent probability and 95 percent

confidence interval that DNB is not occurring

during the transient. The DNB correlation,

and methods used to demonstrate that DNB

limits are met, have changed, and these

changes were reviewed to assure

conformance with acceptable practices. Other

changes, as well as the changes discussed

above, have been evaluated in the referenced

safety analyses and are shown to meet

applicable acceptance criteria. Other

margins, such as avoiding fuel centerline

melting are not significantly changed. Based

on these results, it is concluded that the

margin of safety is not significantly reduced.

The NRC has reviewed the licensee’s

analysis and, based on this review, it

appears that the three standards of 10

CPR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga,

Tennessee 37402.

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H,

Knoxville, Tennessee 37902.

NRC Project Director: Frederick J.

Hebdon.

Union Electric Company, Docket No.

50–483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application request: April 12,

1996.

Description of amendment request:

The proposed amendment would

change Technical Specification (TS) 3/

4.4 and its associated Bases to address

the installation of laser welded tube

sleeves in the Callaway Plant steam

generators.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The elevated tubesheet LWS [laser welded

sleeve] configuration has been designed and

analyzed in accordance with the

requirements of the ASME [American Society

of Mechanical Engineers] Code. The applied

stresses and fatigue usage for the sleeve and

weld are bounded by the limits established

in the ASME Code. ASME Code minimum

material property values are used for the

structural and plugging limit analysis.

Ultrasonic inspection is used to verify that

minimum weld fusion zone thickness are

produced. Mechanical testing has shown that

the individual joint structural strength of

Alloy 690 LWS under normal, upset and

faulted conditions provides margin to the

acceptance limits. These acceptance limits

bound the most limiting (3 times normal

operating pressure differential) burst margin

recommended by RG [Regulatory Guide]

1.121. Therefore, each individual joint

provides for structural integrity exceeding RG

recommendations.

Leakage testing for 7⁄8′′ and 3⁄4′′ tube sleeves

has demonstrated that no unacceptable levels

of primary to secondary leakage are expected

during any plant condition, including the

case where the seal weld is not produced in

the lower joint of the tubesheet sleeve.

Similar tests of 11/16’’ tube sleeves will be

completed prior to Refuel 8.

The sleeve minimum acceptable wall

thickness (used for developing the depthbased plugging limit for the sleeve) is

determined using the guidance of Regulatory

Guide 1.121 and the pressure stress equation

of Section III of the ASME Code. The limiting

requirement of Regulatory Guide 1.121,

which applies to part throughwall

degradation, is that the minimum acceptable

wall must maintain a factor of safety of three

against tube failure under normal operating

(design) conditions. A bounding set of design

and transient loading input conditions was

used for the minimum wall thickness

evaluation in the generic evaluation.

Evaluation of the minimum acceptable wall

thickness for normal, upset and postulated

accident condition loading per the ASME

Code indicates these conditions are bounded

by the design condition requirement

minimum wall thickness.

A bounding tube wall degradation growth

rate per cycle and an eddy current

uncertainty has been assumed for

determining the sleeve TS plugging limit.

The sleeve wall degradation extent

determined by eddy current examination,

which would require plugging sleeved tubes,

is developed using the guidance of RG 1.121

and is defined in WCAP–14596 to be 39

percent throughwall of the sleeve nominal

wall thickness.

The consequences of failure of the sleeve

joint are bounded by the current steam

generator tube rupture analysis included in

the Callaway FSAR. Due to the slight

reduction in diameter caused by the sleeve

wall thickness, primary coolant release rates

would be slightly less than assumed for the

steam generator tube rupture analysis

(depending on the break location), and

therefore, would result in lower total primary

fluid mass release to the secondary system.

The proposed change does not adversely

impact any other previously evaluated design

basis accident of the results of LOCA and

non-LOCA accident analyses for the current

TS minimum reactor coolant system flow

rate. The results of the analyses and testing

demonstrate that the sleeve assembly is an

acceptable means of maintaining tube

integrity. Furthermore, per Regulatory Guide

1.83, ‘‘Inservice Inspection of Pressurized

Water Reactor Steam Generator Tubes’’

recommendations, the sleeved tube can be

monitored through periodic inspections with

present eddy current techniques. These

measures demonstrate that installation of

sleeves spanning degraded areas of the tube

will restore the tube to a condition consistent

with its original design basis.

Corrosion testing of laser welded sleeve

joints indicates that the corrosion resistance
20858 Federal Register / Vol. 61, No. 90 / Wednesday, May 8, 1996 / Notices

(relative to roll transition control samples)

can be increased by greater than a factor of

ten with the application of a post weld heat

treatment [PWHT]. All free span laser welds

will receive a post weld heat treatment.

Therefore, rapid corrosion degradation of the

free span laser weld joint region is not

expected. Recently performed corrosion

testing of LWS joints in locked (at the first

TSP [tube support plate] structure) tube

conditions indicates that the PWHT, the

stress corrosion cracking initiation potential

in the weld region of the parent tube is

reduced and the cracking resistance is

enhanced. Similar test results and

conclusions would be expected for Callaway

based on the similarity of designs and

expected tube far field residual stresses.

Conformance of the sleeve design with the

applicable sections of the ASME Code and

results of the leakage and mechanical tests,

support the conclusion that installation of

LWS will not increase the probability or

consequences of an accident previously

evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

Sleeving will not adversely affect any plant

component. Stress and fatigue analysis of the

repair has shown that the ASME Code and

Regulatory Guide 1.121 criteria are not

exceeded. Implementation of LWS maintains

overall tube bundle structural and leakage

integrity at a level consistent to that of the

originally supplied tubing during all plant

conditions. Leak and mechanical testing of

sleeves support the conclusions of the

calculations that each sleeve joint retains

both structural and leakage integrity during

all conditions. Sleeving of tubes does not

provide a mechanism resulting in an accident

outside of the area affected by the sleeves.

Any accident as a result of potential tube or

sleeve degradation in the repaired portion of

the tube is bounded by the existing tube

rupture accident analysis.

Implementation of LWS will reduce the

potential for primary to secondary leakage

during a postulated steam line break while

not significantly impacting available primary

coolant flow area in the event of a LOCA. By

effectively isolating degraded areas of the

tube through repair, the potential for steam

line break leakage is reduced. These

degraded intersections now are returned to a

condition consistent with the Design Basis.

While the installation of a sleeve reduces

primary coolant flow, the reduction is far

below that caused by plugging. Therefore, far

greater primary coolant flow area is

maintained through sleeving versus plugging.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The LWS repair of degraded steam

generator tubes has been shown by analysis

to restore the integrity of the tube bundle

consistent with its original design basis

condition, i.e., tube/sleeve operational and

faulted condition stresses are bounded by the

ASME Code requirements and the repaired

tubes are leaktight. The safety factors used in

the design of sleeves for the repair of

degraded tubes are consistent with the safety

factors in the ASME Code used in steam

generator design. The design of the tubesheet

sleeve lower joints for the 3⁄4’’ and 7⁄8’’

sleeves have been verified by testing to

preclude leakage during normal and

postulated accident conditions. Similar tests

of 11⁄16’’ sleeves will be completed prior to

Refuel 8. The portions of the installed sleeve

assembly which represent the reactor coolant

pressure boundary can be monitored for the

initiation and progression of sleeve/tube wall

degradation, thus satisfying the requirements

of Regulatory Guide 1.83. The portion of the

tube bridged by the sleeve joints is effectively

removed from the pressure boundary, and the

sleeve then forms the new pressure

boundary. The areas of the sleeved tube

assembly which require inspection are

defined in WCAP–14596.

In addition, since the installed sleeve

represents a portion of the pressure

boundary, a baseline inspection of these

areas is required prior to operation with

sleeves installed. The effect of sleeving on

the design transients and accident analyses

has been reviewed based on the installation

of sleeves up to the level of steam generator

tube plugging coincident with the minimum

reactor flow rate and the Callaway Safety

Analysis.

Provisional requirements cited in other

NRC Safety Evaluation Reports addressing

the implementation of sleeving have required

the reduction of the individual steam

generator normal operation primary to

secondary leakage limit from 500 to 150 gpd.

Consistent with these evaluations, Union

Electric will reduce the per steam generator

leak rate limit of 500 gpd in TS 3.4.6.2.c to

150 gpd. The establishment of this leakage

limit at 150 gpd provides additional safety

margin.

Finally, Union Electric will reduce the tube

plugging limit from 48 percent through wall

to 40 percent through wall to be consistent

with NUREG–1431. The establishment of the

plugging limit at 40 percent through wall

provides additional safety margin.

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: Callaway County Public

Library, 710 Court Street, Fulton,

Missouri 65251.

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, N.W.,

Washington, D.C. 20037.

NRC Project Director: William H.

Bateman.

Union Electric Company, Docket No.

50–483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application request: April 12,

1996.

Description of amendment request:

The proposed amendment would

change Technical Specification (TS) 3/

4.4 and its associated Bases to address

the installation of electrosleeves in the

Callaway Plant steam generators.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The electrosleeve configuration has been

designed and analyzed in accordance with

the requirements of the ASME [American

Society of Mechanical Engineers] Code. The

applied stresses and fatigue usage for the

sleeve are bounded by the limits established

in the ASME Code. ASME Code minimum

material property values are used for the

structural and plugging limit analysis.

Mechanical testing has shown that the

structural strength of nickel electrosleeves

under normal, upset and faulted conditions

provides margin to the acceptance limits.

These acceptance limits bound the most

limiting (3 times normal operating pressure

differential) burst margin recommended by

RG [Regulatory Guide] 1.121. Leakage testing

for 5⁄8′′, 7⁄8′′ and 3⁄4′′ tube sleeves has

demonstrated that no unacceptable levels of

primary to secondary leakage are expected

during any plant condition. Similar tests of

11⁄16′′ tube electrosleeves will be completed

prior to Refuel 8.

The sleeve nominal wall thickness (used

for developing the depth-based plugging

limit for the sleeve) is determined using the

guidance of Regulatory Guide 1.121 and the

pressure stress equation of Section III of the

ASME Code. The limiting requirement of

Regulatory Guide 1.121, which applies to

part throughwall degradation, is that the

minimum acceptable wall must maintain a

factor of safety of three against tube failure

under normal operating (design) conditions.

A bounding set of design and transient

loading input conditions was used for the

minimum wall thickness evaluation in the

generic evaluation. Evaluation of the

minimum acceptable wall thickness for

normal, upset and postulated accident

condition loading per the ASME Code

indicates these conditions are bounded by

the design condition requirement minimum

wall thickness.

A bounding tube wall degradation growth

rate per cycle and an NDE [nondestructive

examination] uncertainty has been assumed

for determining the sleeve TS plugging limit.

The sleeve wall degradation extent

determined by NDE, which would require

plugging sleeved tubes, is developed using

the guidance of RG 1.121 and is defined in

BAW–10219P to be 20 percent throughwall.

The consequences of failure of the sleeve

are bounded by the current steam generator

tube rupture analysis included in the

Callaway FSAR [final safety analysis report].

Due to the slight reduction in diameter

caused by the sleeve wall thickness, primary

coolant release rates would be slightly less
Federal Register / Vol. 61, No. 90 / Wednesday, May 8, 1996 / Notices 20859

than assumed for the steam generator tube

rupture analysis (depending on the break

location), and therefore, would result in

lower total primary fluid mass release to the

secondary system.

The proposed change does not adversely

impact any other previously evaluated design

basis accident or the results of LOCA [lossof-coolant accident] and non-LOCA accident

analyses for the current TS minimum reactor

coolant system flow rate. The results of the

analyses and testing demonstrate that the

electrosleeve is an acceptable means of

maintaining tube integrity. Furthermore, per

Regulatory Guide 1.83 recommendations, the

sleeved tube can be monitored through

periodic inspections with present NDE

techniques. These measures demonstrate that

installation of sleeves spanning degraded

areas of the tube will restore the tube to a

condition consistent with its original design

basis.

Conformance of the electrosleeve design

with the applicable sections of the ASME

Code and results of the leakage and

mechanical tests, support the conclusion that

installation of electrosleeves will not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

Electrosleeving does not represent a

potential to adversely affect any plant

component. Stress and fatigue analysis of the

repair has shown that the ASME Code and

Regulatory Guide 1.121 criteria are not

exceeded. Implementation of electrosleeving

maintains overall tube bundle structural and

leakage integrity at a level consistent to that

of the originally supplied tubing during all

plant conditions. Leak and mechanical

testing of electrosleeves support the

conclusions of the calculations that each

sleeve retains both structural and leakage

integrity during all conditions. Sleeving of

tubes does not provide a mechanism

resulting in an accident outside of the area

affected by the sleeves. Any accident as a

result of potential tube or sleeve degradation

in the repaired portion of the tube is bounded

by the existing tube rupture accident

analysis.

Implementation of sleeving will reduce the

potential for primary to secondary leakage

during a postulated steam line break while

not significantly impacting available primary

coolant flow area in the event of a LOCA. By

effectively isolating degraded areas of the

tube through repair, the potential for steam

line break leakage is reduced. These

degraded intersections now are returned to a

condition consistent with the Design Basis.

While the installation of a sleeve reduces

primary coolant flow, the reduction is far

below that caused by plugging. Therefore, far

greater primary coolant flow area is

maintained through sleeving versus plugging.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The electrosleeve repair of degraded steam

generator tubes has been shown by analysis

to restore the integrity of the tube bundle

consistent with its original design basis

condition, i.e., tube/sleeve operational and

faulted condition stresses are bounded by the

ASME Code requirements and the repaired

tubes are leaktight. The safety factors used in

the design of sleeves for the repair of

degraded tubes are consistent with the safety

factors in the ASME Code used in steam

generator design. The portions of the

installed sleeve assembly which represent

the reactor coolant pressure boundary can be

monitored for the initiation and progression

of sleeve/tube wall degradation, thus

satisfying the requirements of Regulatory

Guide 1.83. The portion of the tube bridged

by the sleeve is effectively removed from the

pressure boundary, and the sleeve then forms

the new pressure boundary. The areas of the

sleeved tube assembly which require

inspection are defined in BAW–10219P.

In addition, since the installed sleeve

represents a portion of the pressure

boundary, a baseline inspection of these

areas is required prior to operation with

sleeves installed. The effect of sleeving on

the design transients and accident analyses

has been reviewed based on the installation

of sleeves up to the level of steam generator

tube plugging coincident with the minimum

reactor flow rate and the Callaway Safety

Analysis.

Provisional requirements cited in other

NRC Safety Evaluation Reports addressing

the implementation of sleeving have required

the reduction of the individual steam

generator normal operation primary to

secondary leakage limit from 500 to 150 gpd.

Consistent with these evaluations, Union

Electric will reduce the per steam generator

leak rate limit of 500 gpd in TS 3.4.6.2.c to

150 gpd. The establishment of this leakage

limit at 150 gpd provides additional safety

margin.

Finally, Union Electric will reduce the tube

plugging limit from 48 percent through wall

to 40 percent through wall to be consistent

with NUREG–1431. The establishment of the

plugging limit at 40 percent through wall

provides additional safety margin.

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: Callaway County Public

Library, 710 Court Street, Fulton,

Missouri 65251.

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, N.W.,

Washington, D.C. 20037.

NRC Project Director: William H.

Bateman.

Vermont Yankee Nuclear Power

Corporation, Docket No. 50–271,

Vermont Yankee Nuclear Power Station,

Vernon, Vermont

Date of amendment request: April 4,

1996.

Description of amendment request:

The proposed amendment would revise

the Technical Specifications regarding

secondary containment integrity

including addition of required actions

in the event secondary containment

integrity is not maintained when

required. It would also require

surveillance of the secondary

containment isolation valves under the

licensee’s in-service testing program.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) The proposed changes do not result in

any hardware changes. The requirements for

Secondary Containment integrity are not

assumed in the initiation of any analyzed

event. The proposed changes establish and

maintain adequate assurance that Secondary

Containment Integrity will be maintained as

assumed in analyses for the mitigation of

accident consequences. Not requiring

Secondary Containment Integrity when the

reactor coolant system is not vented in the

Cold Shutdown condition or the Refuel Mode

does not involve an increase in previously

evaluated accident consequences since no

mechanism exists to impart additional

fission-products into the reactor coolant.

Under these conditions, activities for which

the reactor coolant system would not be

vented would be strictly controlled and

monitored. As a result, leaks or pipe breaks

would typically be detected before significant

inventory loss occurred. These activities

would typically be performed after refueling

when few noncondensible gases remain in

the reactor coolant. The temperature

limitation of 212°F will ensure that water,

not steam, would be emitted from the

postulated leak or pipe break. In addition,

under these conditions, stored energy is

sufficiently low that even with loss of

inventory following a recirculation line

break, core coverage would be maintained by

the low pressure emergency core cooling

systems required per Specification 3.5.H and

the fuel would not exceed its peak clad

temperature limit. As a result, the potential

for failed fuel and a subsequent increase in

reactor coolant activity is minimized and

significant releases of radioactive material to

the environment would not be expected to

occur. Therefore, these changes will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

(2) The proposed changes do not involve

a physical alteration of the plant (no new or

different type of equipment will be installed)

or changes in parameters governing normal

operation and will not alter the method used

by any system to perform its design function.

The proposed changes to not allow plant

operation in any mode that is not already

evaluated and will still ensure Secondary

Containment Integrity is maintained when

required. Thus, these changes do not create

the possibility of a new or different kind of
20860 Federal Register / Vol. 61, No. 90 / Wednesday, May 8, 1996 / Notices

accident from any accident previously

evaluated.

(3) The proposed changes to Secondary

Containment Integrity requirements have no

impact on any safety analysis assumptions.

Secondary Containment Integrity will be

maintained as assumed in the safety analyses

and as stated in current Bases 3.7.B and

3.7.C. Not requiring Secondary Containment

Integrity when the reactor coolant system is

not vented in the Cold Shutdown condition

or the Refuel Mode does not involve

significant reduction in a margin of safety

since no mechanism exists to impart

additional fission products into the reactor

coolant. Under these conditions, activities for

which the reactor coolant system would not

be vented would be strictly controlled and

monitored. As a result, leaks or pipe breaks

would typically be detected before significant

inventory loss occurred. These activities

would typically be performed after refueling,

at low decay levels, and with reactor coolant

temperature less than or equal to 212°F. In

addition, under these conditions, stored

energy in the reactor core is very low. The

reactor pressure vessel would rapidly

depressurize in the event of a large primary

system leak and the low pressure emergency

core cooling systems required per

Specification 3.5.H under these conditions

would be adequate to keep the core flooded.

This would ensure that the fuel would not be

uncovered and would not exceed the peak

clad temperature limit.

As a result, the potential for failed fuel and

a subsequent increase in reactor coolant

activity is minimized and significant releases

of radioactive material to the environment

would not be expected to occur. Therefore,

these changes do not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

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

Ropes and Gray, One International

Place, Boston, MA 02110–2624.

NRC Project Director: Susan Frant

Shankman.

Vermont Yankee Nuclear Power

Corporation, Docket No. 50–271,

Vermont Yankee Nuclear Power Station,

Vernon, Vermont

Date of amendment request: April 4,

1996.

Description of amendment request:

The proposed amendment would revise

the surveillance requirements for

control rod over-travel to remove the

specific testing methodology from the

Technical Specifications to

administratively controlled documents.

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 control rod drive mechanism overtravel is not considered to be the initiator of

any previously analyzed accident.

Verification of coupling of the control rods

and drive mechanisms is performed by other

means and continues to be required in the

same manner, so there is no significant

increase in the probability of a rod drop

accident. The over-travel indication is also

not considered in the mitigation of

consequences of any previously analyzed

accident, and the removal of a specific

surveillance of the indication will not affect

the response of the control rods or the reactor

protection system to these accidents.

Therefore, this change will not significantly

increase the probability or consequences of

any previously analyzed accident.

(2) The proposed change does not

necessitate a physical alteration of the plant

(no new or different type of equipment will

be installed) nor changes in parameters

governing normal plant operation. The

proposed change will continue to provide

effective methods to assure the control rods

and their drive mechanisms are coupled and

preserve the safety functions associated with

the prevention or automatic mitigation of

design basis accidents. Thus, this change

does not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

(3) The proposed changes continue to

provide an appropriate method for

verification of the capability of the overtravel indication to perform its function.

Therefore, this change will not significantly

reduce a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

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

Ropes and Gray, One International

Place, Boston, MA 02110–2624.

NRC Project Director: Susan Frant

Shankman.

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: April 15,

1996.

Description of amendment request:

The proposed changes will clarify the

applicability of the quadrant power tilt

ratio (QPTR) requirements.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Operation of Surry Power Station in

accordance with the proposed Technical

Specifications change will not:

1. Involve a significant increase in the

probability of occurrence or the

consequences of an accident previously

evaluated.

The application of the QPTR limits, as

proposed, will assure that the gross core

radial power distribution remains consistent

with design limits above 50% power. At or

below 50% rated thermal power, there is

insufficient stored energy in the fuel or

insufficient energy being transferred to the

reactor coolant to require implementation of

a QPTR limit on the distribution of core

power. Therefore, the proposed change to

clarify the applicability of the QPTR

requirements has no impact on the

probability of an accident occurrence and

does not increase the consequences of any

design basis accident.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

There are no plant modifications or

changes in methods of plant operation

introduced by the proposed change. The

change would limit the application of QPTR

limits to operation at power levels >50% to

preclude core power distributions from

occurring which would violate fuel design

criteria previously analyzed. At or below

50% rated thermal power, there is no impact

to core power distributions which could

affect the fuel design criteria. Therefore, the

proposed change does not create the

possibility for an accident or malfunction of

a different type than that previously

evaluated in the safety analysis report.

3. Involve a significant reduction in a

margin of safety.

The proposed change only affects the

applicability of the QPTR limits. The QPTR

limits remain unchanged to preclude any

violation of previously analyzed fuel design

criteria. Adherence to the QPTR limits, hot

channel factors, and applicable Limiting

Conditions for Operation will continue.

Therefore, the margin of safety as described

in the Bases Section of any part of the

Technical Specifications is not reduced.

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,
Federal Register / Vol. 61, No. 90 / Wednesday, May 8, 1996 / Notices 20861

Riverfront Plaza, East Tower, 951 E.

Byrd Street, Richmond, Virginia 23219.

NRC Project Director: Eugene V.

Imbro.

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.

Duke Power Company, et al., Docket

Nos. 50–413 and 50–414, Catawba

Nuclear Station, Units 1 and 2, York

County, South Carolina

Date of amendment request: April 3,

1996.

Description of amendment request:

The proposed amendments would

revise the hydrogen mitigation system

Technical Specifications (TS). The

change would provide that, if neither

the Train A or Train B igniter is

operable in any one containment region,

then there is an allowance of 7 days to

restore one hydrogen igniter to

OPERABLE status, or be in Hot

Shutdown within the next 6 hours. This

would be consistent with the guidance

of the Standard TS for Westinghouse

plants, NUREG–0431.

Date of publication of individual

notice in Federal Register: April 16,

1996 (61 FR 16649).

Expiration date of individual notice:

May 16, 1996.

Local Public Document Room

location: York County Library, 138 East

Black Street, Rock Hill, South Carolina.

Houston Lighting & Power Company,

City Public Service Board of San

Antonio, Central Power and Light

Company, City of Austin, Texas, Docket

No. 50–498, South Texas Project, Unit 1,

Matagorda County, Texas

Date of amendment request: January

22, 1996, as supplement by letter dated

April 4, 1996.

Brief description of amendments: The

proposed amendment would modify the

steam generator tube plugging criteria in

Technical Specification 3/4.4.5, Steam

Generators, and the allowable leakage in

Technical Specification 3/4.4.6.2,

Operational Leakage, and the associated

Bases. The proposed amendment would

allow the implementation of steam

generator voltage-based repair criteria

for the tube support plate (TSP)/tube

intersections for Unit 1.

Date of individual notice in the

Federal Register: April 16, 1996 (61 FR

16651)

Expiration date of individual notice:

May 16, 1996.

Local Public Document Room

location: Wharton County Junior

College, J. M. Hodges Learning Center,

911 Boling Highway, Wharton, TX

77488.

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 amendment request: March

14, 1996.

Description of amendment request:

The proposed amendment would revise

the Technical Specifications for Indian

Point Nuclear Generating Unit No. 3 to

allow a one-time extension of the test

intervals for the pressurizer safety valve

setpoint and snubber functional testing

that is due in May 1996.

Date of publication of individual

notice in Federal Register: April 3, 1996

(61 FR 14835)

Expiration date of individual notice:

May 3, 1996.

Local Public Document Room

location: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10601.

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.

Connecticut Yankee Atomic Power

Company, Docket No. 50–213, Haddam

Neck Plant, Middlesex County and

Northeast Nuclear Energy Company, et

al., Docket Nos. 50–245, 50–336, and

50–423, Millstone Nuclear Power

Station, Units 1, 2, and 3, New London

County, Connecticut

Date of application for amendments:

November 22, 1995.

Brief description of amendments: The

amendments delete from the Technical

Specifications certain review

responsibilities of the Plant Operations

Review Committee and the Site

Operations Review Committee relating

to the Emergency Plan and the Security

Plan and their respective implementing

procedures. The proposed changes are

consistent with the guidance of Generic

Letter 93–07.

Date of issuance: April 24, 1996

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment Nos.: 189, 94, 196, and

128

Facility Operating License Nos. DPR–

61, DPR–21, DPR–65, AND NPF–49:

Amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 14, 1996 (61 FR

5812)

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated April 24, 1996.

No significant hazards consideration

comments received: No.
20862 Federal Register / Vol. 61, No. 90 / Wednesday, May 8, 1996 / Notices

Local Public Document Room

location: Russell Library, 123 Broad

Street Middletown, Connecticut 06457,

for the Haddam Neck Plant, and the

Learning Resources Center, Three Rivers

Community-Technical College, 574 New

London Turnpike, Norwich, CT 06360,

and the Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

CT 06385, for Millstone 1, 2, and 3.

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 11, 1996, as supplemented by

letter dated April 2, 1996.

Brief description of amendments: The

amendments revise Technical

Specification Table 3.6–1, Table 3.6–2a

and Table 3.6–2b to delete references to

process penetration M308 and service

water system (RN) valves RN–429A and

RN–432B from the lists of secondary

containment bypass valves and

containment isolation valves. The RN

valves are no longer in service and are

planned to be removed in forthcoming

outages. The penetration will then be

capped with blank flanges.

Date of issuance: April 23, 1996

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment Nos.: 143 and 137

Facility Operating License Nos. NPF–

35 and NPF–52: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: February 14, 1996 (61 FR

5813) The April 2, 1996, letter provided

additional information that did not

change the scope of the January 11,

1996, 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 April 23, 1996.

No significant hazards consideration

comments received: No

Local Public Document Room

location: York County Library, 138 East

Black Street, Rock Hill, South Carolina

29730.

Duke Power Company, Docket Nos. 50–

269, 50–270, and 50–287, Oconee

Nuclear Station, Units 1, 2, and 3,

Oconee County, South Carolina

Date of application of amendments:

December 7, 1995.

Brief description of amendments: The

amendments revise Secondary Decay

Heat Removal Technical Specification

(TS) 3.4.2 and TS Table 4.1–1 to delete

the requirement of having the main

feedwater pump discharge header

pressure switch provide an input to

actuate the Anticipatory Reactor Trip

System and Emergency Feedwater

System.

Date of Issuance: April 15, 1996.

Effective date: As of the date of

issuance to be implemented within 30

days

Amendment Nos.: 216, 216, 213.

Facility Operating License Nos. DPR–

38, DPR–47, and DPR–55: The

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 22, 1996 (61 FR 1628).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated April 15, 1996.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Oconee County Library, 501

West South Broad Street, Walhalla,

South Carolina 29691.

Florida Power and Light Company, et

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

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

County, Florida

Date of application for amendments:

October 25, 1993, as supplemented

August 31, 1994, and October 5, 1995.

Brief description of amendments: The

amendments modify the surveillance

requirements related to dune survey and

mangrove swamp monitoring and

relocate them to the Final Safety

Analysis Report

Date of Issuance: April 11, 1996.

Effective Date: April 11, 1996.

Amendment Nos.: 142 and 82.

Facility Operating License Nos. DPR–

67 and NPF–16: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: December 22, 1993 (58 FR

67844) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

April 11, 1996

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort

Pierce, Florida 34954–9003

Maine Yankee Atomic Power Company,

Docket No. 50–309, Maine Yankee

Atomic Power Station, Lincoln County,

Maine

Date of application for amendment:

August 31, 1995, as supplemented

February 29, 1996.

Brief description of amendment: The

amendment revises License Condition

2.B(6)(c), Fire Protection, and relocates

fire protection requirements from the

Maine Yankee Atomic Power Station

Technical Specifications to the Maine

Yankee Fire Protection Plan. The

amendment is consistent with the

guidance of NRC Generic Letters 86–10,

Implementation of Fire Protection

Requirements, and 88–12, Removal of

Fire Protection Requirements, from the

Technical Specifications.

Date of issuance: April 5, 1996.

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment No.: 156.

Facility Operating License No. DPR–

36: Amendment revised the Technical

Specifications and License.

Date of initial notice in Federal

Register: October 11, 1995 (60 FR

52932) The February 29, 1996, letter

provided document dates that did not

change the initial proposed no

significant hazards consideration

determination. The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

April 5, 1996.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Wiscasset Public Library, High

Street, P.O. Box 367, Wiscasset, ME

04578.

Northeast Nuclear Energy Company,

Docket No. 50–245, Millstone Nuclear

Power Station, Unit 1, New London

County, Connecticut

Date of application for amendment:

October 25, 1995.

Brief description of amendment: The

amendment changes the Technical

Specification regarding the average

power range monitor (APRM) setpoints.

These changes establish limiting

conditions for operations and

surveillance requirements for the APRM

flow-biased scram and rod block

setpoints. The amendment also

incorporates several editorial changes

and renumbered pages, removal of blank

pages, revised Table of Contents, and

modified Bases section for APRM

setpoint requirements.

Date of issuance: April 15, 1996.

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment No.: 93.

Facility Operating License No. DPR–

21. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: December 20, 1995 (60 FR

65682).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 15, 1996.

No significant hazards consideration

comments received: No.
Federal Register / Vol. 61, No. 90 / Wednesday, May 8, 1996 / Notices 20863

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360 and at the

temporary local public document room

located at the Waterford Library, ATTN:

Vince Juliano, 49 Rope Ferry Road,

Waterford, CT 06385.

Omaha Public Power District, Docket

No. 50–285, Fort Calhoun Station, Unit

No. 1, Washington County, Nebraska

Date of amendment request: August 4,

1995, as supplemented by letter dated

January 22, 1996.

Brief description of amendment: This

amendment revised the Technical

Specifications (TS) for the requirements

for the containment radiation high

signal (CRHS) and the safety injection

and refueling water (SIRW) tank low

signal (STLS) contained in TS 2.15,

Tables 2–3 and 2–4. Table 3–2 of TS 3.1

will also be revised to include

administrative changes to the CRHS

surveillance methods to be consistent

with the applicable surveillance

functions. The Basis of TS 2.15 is being

revised to clarify that the number of

installed channels for CRHS is two. The

term ‘‘SOURCE CHECK’’ is being

deleted from the Definitions section.

Date of issuance: April 24, 1996.

Effective date: April 24, 1996.

Amendment No.: 173.

Facility Operating License No. DPR–

40. The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: August 30, 1995 (60 FR

45182).

The January 22, 1996, supplemental

letter provided additional clarifying

information and did not change the

initial no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 24, 1996.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska

68102.

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:

March 1, 1995, as supplemented by

letter dated April 16, 1996.

Brief description of amendments: The

amendments change the concentration

of calibration gas required to calibrate

the Hydrogen and Oxygen Analyzers,

and support the requirements of

Limerick Generating Station Transient

Response Implementation Plan (TRIP)

T–102, ‘‘Primary Containment Control

Bases.’’

Date of issuance: April 23, 1996.

Effective date: Both units, as of date

of issuance, to be implemented within

45 days.

Amendment Nos.: 116 and 78.

Facility Operating License Nos. NPF–

39 and NPF–85. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: April 26, 1995 (60 FR 20525)

The April 16, 1996 letter requested a

new effective date and did not change

the initial proposed no significant

hazards consideration determination

nor the Federal Register notice.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated April 23, 1996.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania

19464.

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of application for amendments:

August 21, 1992; supplemented

September 3, 1993, and March 28, 1996

(TS 92–07).

Brief description of amendments: The

amendments revise the allowable value

for the reactor coolant system loss of

flow reactor trip setpoint from greater

than or equal to 89.4 percent to greater

than or equal to 89.6 percent.

Date of issuance: April 26, 1996.

Effective date: April 26, 1996.

Amendment Nos.: 221 and 212.

Facility Operating License Nos. DPR–

77 and DPR–79: Amendments revise the

technical specifications.

Date of initial notice in Federal

Register: September 30, 1992 (57 FR

45090). The September 3, 1993 and

March 28, 1996 supplemental letters

provided clarifying information which

did not change the proposed no

significant hazards consideration.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 26, 1996.

No significant hazards consideration

comments received: None.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402.

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50–346, Davis-Besse Nuclear Power

Station, Unit No. 1, Ottawa County,

Ohio

Date of application for amendment:

February 5, 1996.

Brief description of amendment: This

amendment clarifies TS 3/4.3.2.1, Table

3.3–3, Safety Features Actuation System

Instrumentation, and revises Bases 3/

4.3.1 and 3/4.3.2, Reactor Protection

System and Safety System

Instrumentation, to accurately reflect

the design and actuation logic of the

diesel generator load sequencer and the

essential bus undervoltage relays.

Date of issuance: April 23, 1996.

Effective date: April 23, 1996.

Amendment No.: 211.

Facility Operating License No. NPF–3:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 13, 1996 (61 FR 10397).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 23, 1996.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of Toledo, William

Carlson Library, Government

Documents Collection, 2801 West

Bancroft Avenue, Toledo, Ohio 43606

Virginia Electric and Power Company,

Docket Nos. 50–280 and 50–281, Surry

Power Station, Unit Nos. 1 and 2, Surry

County, Virginia

Date of application for amendments:

November 20, 1995, as supplemented

March 14, 1996.

Brief description of amendments:

These amendments would permit the

use of 10 CFR Part 50 Appendix J,

Option B, performance-based

containment leakage rate testing.

Date of issuance: April 18, 1996.

Effective date: April 18, 1996.

Amendment Nos. 208 and 208.

Facility Operating License Nos. DPR–

32 and DPR–37: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: December 20, 1995 (60 FR

65686) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

April 18, 1996.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Swem Library, College of

William and Mary, Williamsburg,

Virginia 23185.
20864 Federal Register / Vol. 61, No. 90 / Wednesday, May 8, 1996 / Notices

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

example, in derating or shutdown of a

nuclear power plant or in prevention of

either resumption of operation or of

increase in power output up to the

plant’s licensed power level, the

Commission may not have had an

opportunity to provide for public

comment on its no significant hazards

consideration determination. In such

case, the license amendment has been

issued without opportunity for

comment. If there has been some time

for public comment but less than 30

days, the Commission may provide an

opportunity for public comment. If

comments have been requested, it is so

stated. In either event, the State has

been consulted by telephone whenever

possible.

Under its regulations, the Commission

may issue and make an amendment

immediately effective, notwithstanding

the pendency before it of a request for

a hearing from any person, in advance

of the holding and completion of any

required hearing, where it has

determined that no significant hazards

consideration is involved.

The Commission has applied the

standards of 10 CFR 50.92 and has made

a final determination that the

amendment involves no significant

hazards consideration. The basis for this

determination is contained in the

documents related to this action.

Accordingly, the amendments have

been issued and made effective as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the application for

amendment, (2) the amendment to

Facility Operating License, and (3) the

Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment, as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, the Gelman Building, 2120 L

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

local public document room for the

particular facility involved.

The Commission is also offering an

opportunity for a hearing with respect to

the issuance of the amendment. By June

7, 1996, the licensee may file a request

for a hearing with respect to issuance of

the amendment to the subject facility

operating license and any person whose

interest may be affected by this

proceeding and who wishes to

participate as a party in the proceeding

must file a written request for a hearing

and a petition for leave to intervene.

Requests for a hearing and a petition for

leave to intervene shall be filed in

accordance with the Commission’s

‘‘Rules of Practice for Domestic

Licensing Proceedings’’ in 10 CFR Part

2. Interested persons should consult a

current copy of 10 CFR 2.714 which is

available at the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington, DC and

at the local public document room for

the particular facility involved. If a

request for a hearing or petition for

leave to intervene is filed by the above

date, the Commission or an Atomic

Safety and Licensing Board, designated

by the Commission or by the Chairman

of the Atomic Safety and Licensing

Board Panel, will rule on the request

and/or petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

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
Federal Register / Vol. 61, No. 90 / Wednesday, May 8, 1996 / Notices 20865

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 20555–001, 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 20555–001, 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).

Tennesse Valley Authority, Docket Nos.

50–259, 50–260, and 50–296, Browns

Ferry Nuclear Plant, Limestone County,

Alabama

Date of application for amendment:

April 14, 1996.

Brief description of amendment: The

proposed amendment clarifies

operability requirements for reactor

vessel water level instrumentation to

permit testing of components required

by technical specifications.

Date of issuance: April 16, 1996.

Effective date: April 16, 1996.

Amendment Nos.: 229, 244, and 204.

Facility Operating License Nos. DPR–

33, DPR–52 and DPR–68: Amendment

revises the technical specifications.

The Commission’s related evaluation

of the amendment, finding of emergency

circumstances, and final determination

of no significant hazards consideration,

are contained in a Safety Evaluation

dated April 16, 1996. Public comments

requested as to proposed no significant

hazards consideration: No.

Local Public Document Room

location: Athens Public library, South

Street, Athens, Alabama 35611.

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H,

Knoxville, Tennessee 37902

NRC Project Director: Frederick J.

Hebdon.

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50–346, Davis-Besse Nuclear Power

Station, Unit No. 1, Ottawa County,

Ohio

Date of application for amendment:

April 18, 1996.

Brief description of amendment: The

amendment approves the use of the

station black out diesel generator in lieu

of the emergency diesel generator

associated with decay heat removal loop

2 during the tenth refueling outage. This

condition will continue as long as no

work is performed in the switchyard or

on the SBODG or the remaining

emergency diesel generator and a

shutdown risk contingency plan is

developed to ensure challenges to spent

fuel pool cooling are minimized. This

condition is expected to last for no more

than seven days.

Date of issuance: April 19, 1996.

Effective date: April 19, 1996.

Amendment No.: 210.

Facility Operating License No. NPF–3:

This amendment approved a one-time

change to the design basis as described

in the Updated Safety Analysis Report.

Public comments requested as to

proposed no significant hazards

consideration: No.

The Commission’s related evaluation

of the amendment, finding of emergency

circumstances, and final determination

of no significant hazards consideration

are contained in a Safety Evaluation

dated April 19, 1996.

Local Public Document Room

location: University of Toledo, William

Carlson Library, Government

Documents Collection, 2801 West

Bancroft Avenue, Toledo, Ohio 43606.

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Gail H. Marcus.

Dated at Rockville, Maryland, this 1st of

May 1996.

For the Nuclear Regulatory Commission.

Steven A. Varga,

Director, Division of Reactor Projects—I/II,

Office of Nuclear Reactor Regulation.

[FR Doc. 96–11295 Filed 5–7–96; 8:45 am]

BILLING CODE 7590–01–P

Privacy Act of 1974, as Amended;

Establishment of a New System of

Records

AGENCY: Nuclear Regulatory

Commission.

ACTION: Establishment of a new system

of records.

SUMMARY: The Nuclear Regulatory

Commission (NRC) is proposing to

establish a new Privacy Act System of

Records, NRC–41, ‘‘Tort Claims and

Personal Property Claims,’’ to maintain

records needed to evaluate, settle, refer,

pay, and/or adjudicate claims filed by

individuals against the NRC.

EFFECTIVE DATES: The new system of

records will become effective without

further notice on June 17, 1996, unless

comments received on or before that

date cause a contrary decision. If

changes are made based on NRC’s

review of comments received, NRC will

publish a new final notice.

ADDRESSES: Send comments to the

Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555–0001, Attention:

Docketing and Services Branch. Hand

deliver comments to 11555 Rockville

Pike, Rockville, Maryland, between 7:45

am and 4:15 pm Federal workdays.

Copies of comments received may be

examined, or copied for a fee, at the

NRC Public Document Room at 2120 L

Street, NW., Lower Level, Washington,

DC.

FOR FURTHER INFORMATION CONTACT: Jona

L. Souder, Freedom of Information/

Local Public Document Room Branch,

Division of Freedom of Information and