Federal Register / Vol. 65, No. 105 / Wednesday, May 31, 2000 / Notices 34743

PENETRATION SEAL MATERIALS

AND OTHER MINOR CHANGES’’ (10

CFR PART 50) (WITS 199800128)’’

(PUBLIC MEETING) be held on May 25,

and on less than one week’s notice to

the public.

The NRC Commission Meeting

Schedule can be found on the Internet

at: http://www.nrc.gov/SECY/smj/

schedule.htm.

This notice is distributed by mail to

several hundred subscribers; if you no

longer wish to receive it, or would like

to be added to it, please contact the

Office of the Secretary, Attn: Operations

Branch, Washington, D.C. 20555 (301–

415–1661). In addition, distribution of

this meeting notice over the Internet

system is available. If you are interested

in receiving this Commission meeting

schedule electronically, please send an

electronic message to wmh@nrc.gov or

dkw@nrc.gov.

Dated: May 25, 2000.

William M. Hill, Jr.,

SECY Tracking Officer, Office of the

Secretary.

[FR Doc. 00–13674 Filed 5–26–00; 12:46 am]

BILLING CODE 7590–01–M

NUCLEAR REGULATORY

COMMISSION

Biweekly Notice; Applications and

Amendments to Facility Operating

Licenses Involving No Significant

Hazards Considerations

I. Background

Pursuant to Public Law 97–415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 97–415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from May 6, 2000,

through May 19, 2000. The last

biweekly notice was published on May

17, 2000 (65 FR 31354).

Notice of Consideration of Issuance of

Amendments to Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a

proposed determination that the

following amendment requests involve

no significant hazards consideration.

Under the Commission’s regulations in

10 CFR 50.92, this means that operation

of the facility in accordance with the

proposed amendment would not (1)

involve a significant increase in the

probability or consequences of an

accident previously evaluated; or (2)

create the possibility of a new or

different kind of accident from any

accident previously evaluated; or (3)

involve a significant reduction in a

margin of safety. The basis for this

proposed determination for each

amendment request is shown below.

The Commission is seeking public

comments on this proposed

determination. Any comments received

within 30 days after the date of

publication of this notice will be

considered in making any final

determination.

Normally, the Commission will not

issue the amendment until the

expiration of the 30-day notice period.

However, should circumstances change

during the notice period such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received before

action is taken. Should the Commission

take this action, it will publish in the

Federal Register a notice of issuance

and provide for opportunity for a

hearing after issuance. The Commission

expects that the need to take this action

will occur very infrequently.

Written comments may be submitted

by mail to the Chief, Rules Review and

Directives Branch, Division of Freedom

of Information and Publications

Services, Office of Administration, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555–0001, and

should cite the publication date and

page number of this Federal Register

notice. Written comments may also be

delivered to Room 6D22, Two White

Flint North, 11545 Rockville Pike,

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

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

written comments received may be

examined at the NRC Public Document

Room, the Gelman Building, 2120 L

Street, NW., Washington, DC. The filing

of requests for a hearing and petitions

for leave to intervene is discussed

below.

By June 30, 2000, 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 electronically

from the ADAMS Public Library

component on the NRC Web site, http:/

/www.nrc.gov (the Electronic Reading

Room). 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

34744 Federal Register / Vol. 65, No. 105 / Wednesday, May 31, 2000 / Notices

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. 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 electronically

from the ADAMS Public Library

component on the NRC Web site, http:/

/www.nrc.gov (the Electronic Reading

Room).

Duke Energy Corporation, 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 18,

2000.

Description of amendment request:

The amendments would revise

Technical Specifications (TS) 3.7.10 and

3.7.12 for Catawba Units 1 and 2. The

proposed changes address degraded

pressure boundaries on the Auxiliary

Building Filtered Ventilation Exhaust

System and the Control Room Area

Ventilation System. The proposed

changes in TS 3.7.10 and 3.7.12 would

add Notes which allow the affected

ventilation system boundaries to be

opened intermittently under

administrative controls. Also, it would

add a new condition in TS 3.7.10 and

3.7.12. This new condition will require

that the boundaries for these two

systems be returned to an operable

status within 24 hours, when both trains

of these systems are inoperable due to

an inoperable boundary.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

In accordance with the criteria set forth in

10 CFR 50.91 and 50.92, Duke Energy

Corporation has evaluated this license

amendment request and determined it does

not represent a significant hazards

consideration. The following is provided in

support of this conclusion.

1. Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

No. The Control Room Area Ventilation

System (CRAVS), Control Room pressure

boundary, the Auxiliary Building Filtered

Ventilation Exhaust System (ABFVES), or the

Emergency Core Cooling System (ECCS)

pump rooms area pressure boundary are not

assumed to be initiators of any analyzed

accident. Therefore, the proposed changes

contained in this LAR [license amendment

request] have no significant impact on the

probability of occurrence of any previously

analyzed accident.

The proposed new condition for the

CRAVS and ABFVES Technical

Specifications (TS) would permit a 24-hour

period to take action to restore an inoperable

pressure boundary to OPERABLE status. The

consequences of implementing the 24 hour

Completion Time are reasonable based upon:

(1) The low probability of a design basis

accident occurring during this time period,

(2) additional actions that are available to the

operator to minimize doses (e.g., self

contained breathing apparatus and alternate

control room air intakes), and (3) the

availability of an operable CRAVS/ABFVES

train to provide a filtered environment (albeit

with potential unfiltered leakage).

For cases where any of the affected control

room or ECCS pump room area/pump rooms

pressure boundaries are opened

intermittently under administrative controls,

appropriate compensatory measures would

be required by the proposed TS to ensure the

pressure boundary can be rapidly restored.

Based on the compensatory measures

available to the plant operators and the

administrative controls required to rapidly

restore an opened pressure boundary, the

accident consequences do not cause an

increase in dose above the applicable General

Design Criteria, Standard Review Plan, or 10

CFR 100 limits. The plant operators will

continue to maintain the ability to mitigate

a design basis event.

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

No. No changes are being made to actual

plant hardware which will result in any new

accident causal mechanisms. Also, no

changes are being made to the way in which

the plant is being operated. Therefore, no

new accident causal mechanisms will be

generated.

3. Does this change involve a significant

reduction in a margin of safety?

No. Margin of safety is related to the ability

of the fission product barriers to perform

their design functions during and following

accident conditions. These barriers include

the fuel cladding, the reactor coolant system,

and the containment system. The

performance of these barriers will not be

significantly degraded by the proposed

changes. The proposed changes would allow

affected pressure boundaries to be degraded

for a limited period of time (24 hours).

Federal Register / Vol. 65, No. 105 / Wednesday, May 31, 2000 / Notices 34745

However, the probability of a design basis

event occurring during this time is low and

additional actions (e.g., breathing apparatus)

would also be taken to minimize dose to the

plant operators. When the boundaries are

open on an intermittent basis, as permitted

by the changes proposed in this LAR,

administrative controls would be in place to

ensure that the integrity of the pressure

boundaries could be rapidly restored.

Therefore, it is expected that the plant, and

the operators, would maintain the ability to

mitigate design basis events and none of the

fission product barriers would be affected by

this change. Therefore, the proposed change

is not considered to result in 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.

Attorney for licensee: Ms. Lisa F.

Vaughn, Legal Department (PB05E),

Duke Energy Corporation, 422 South

Church Street, Charlotte, North Carolina

28201–1006.

NRC Section Chief: Richard L. Emch,

Jr.

FirstEnergy Nuclear Operating

Company, Docket No. 50–440, Perry

Nuclear Power Plant, Unit 1, Lake

County, Ohio

Date of amendment request: April 5,

2000.

Description of amendment request:

The proposed amendment implements

technical specification (TS) changes

associated with thermal-hydraulic

stability monitoring. New TS 3.3.1.3,

‘‘Oscillation Power Range Monitor

(OPRM) Instrumentation,’’ will provide

the minimum operability requirements

for the OPRM channels, the Required

Actions when they become inoperable,

and appropriate surveillance

requirements. The OPRMs will provide

automatic ‘‘detect and suppress’’ actions

to replace the administrative controls

currently in effect through operator

training and manual actions. The

amendment would remove monitoring

guidance from TS 3.4.1, ‘‘Recirculation

Loops Operating,’’ that will no longer be

necessary due to the activation of the

automatic OPRM instrumentation.

Finally, the amendment would update

TS 5.6.5, ‘‘Core Operating Limits Report

(COLR),’’ to require the applicable

setpoints for the OPRMs to be included

in the COLR.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration which is presented below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed change specifies limiting

conditions for operation, required actions

and surveillance requirements for the

Oscillation Power Range Monitor (OPRM)

system, and allows operation in regions of

the power to flow map currently restricted by

the requirements of Interim Corrective

Actions (ICAs) and certain limiting

conditions of operation of Technical

Specification (TS) 3.4.1. The restrictions of

the ICAs and TS 3.4.1 were imposed to

ensure adequate capability to detect and

suppress conditions consistent with the onset

of thermal-hydraulic (T–H) oscillations that

may develop into a T–H instability event. A

T–H instability event has the potential to

challenge the Minimum Critical Power

(MCPR) safety limit. The OPRM system can

automatically detect and suppress conditions

necessary for T–H instability. With the

activation of the OPRM System, the

restrictions of the ICAs and TS 3.4.1 will no

longer be required.

The probability of a T–H instability event

is impacted by power to flow conditions

during operation inside specific regions of

the power to flow map, in combination with

power shape and inlet enthalpy conditions,

such that only under such conditions can the

occurrence of an instability event be

postulated to occur. Operation in these

regions may increase the probability that

operation with conditions necessary for a T–

H instability can occur. However, when the

OPRM is OPERABLE with operating limits as

specified in the Core Operating Limits Report

(COLR), the OPRM can automatically detect

the onset of significant local power

oscillations and generate a trip signal.

Actuation of a Reactor Protection System

(RPS) trip will suppress conditions necessary

for T–H instability and decrease the

probability of a T–H instability event. In the

event the trip capability of one or more of the

OPRM channels is not maintained, the

proposed change includes Required Actions

which limit the period of time before the

affected OPRM channel (or RPS system) must

be placed in the tripped condition. If these

actions would result in a trip function such

as a scram, or if the OPRM trip capability is

not maintained, an alternate method to detect

and suppress thermal hydraulic oscillations

is required, i.e., the same ICAs as are in place

today. In either case the duration of the

period of time allowed by the Required

Actions is limited, and the probability of a

T–H instability event during this limited time

is not significantly increased.

Several changes to TS 3.4.1 are made

which are more consistent with, or

conservative with, respect to the reviewed

and approved Standard Technical

Specifications for Boiling Water Reactors.

These generic changes are considered

applicable to the Perry Nuclear Power Plant.

They simply provide guidance on the

operator actions to be taken and the

associated time limits when the Specification

is entered, and do not impact the probability

of occurrence of an accident. For the above

reasons, the proposed change does not result

in a significant increase in the probability of

an accident previously evaluated.

An unmitigated T–H instability event is

postulated to cause a violation of the MCPR

safety limit. The proposed change ensures

mitigation of T–H instability events prior to

challenging the MCPR safety limit if initiated

from anticipated conditions, by detection of

the onset of oscillations and actuation of an

RPS trip signal. The OPRM also provides the

capability of an RPS trip being generated for

T–H instability events initiated from

unanticipated but postulated conditions.

These mitigating capabilities of the OPRM

system will become available as a result of

the proposed change and have the potential

to reduce the consequences of anticipated

and postulated T–H instability events. The

OPRM installation has been evaluated to not

adversely impact other installed equipment

such as the Average Power Range Monitors

(APRMs) or the RPS in a manner that could

prevent response to various postulated

events, so those events will not have

increased consequences due to the OPRMs.

Therefore, the proposed change does not

significantly increase the consequences of an

accident previously evaluated.

Therefore, the proposed change, which

specifies limiting conditions for operation,

required actions and surveillance

requirements for the OPRM system, and

allows operation in certain regions of the

power to flow map, does not significantly

increase either the probability or

consequences of an accident previously

evaluated.

2. The proposed change would not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed change specifies limiting

conditions for operation, required actions

and surveillance requirements of the OPRM

system, and allows operation in regions of

the power to flow map currently restricted by

the requirements of ICAs and TS 3.4.1. The

OPRM system uses input signals shared with

APRM and rod block functions to monitor

core conditions and generate an RPS trip

when required. Quality requirements for

software design, testing, implementation and

module self-testing of the OPRM system

provide assurance that new equipment

malfunctions due to software errors are not

created. The design of the OPRM system also

ensures that neither operation nor

malfunction of the OPRM system will

adversely impact the operation of other

systems and no accident or equipment

malfunction of these other systems could

cause the OPRM system to malfunction or

cause a different kind of accident. Therefore,

operation with the OPRM system does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

Operation in regions currently restricted by

the requirements of ICAs and TS 3.4.1 is

within the nominal operating domain and

ranges of plant systems and components, and

within the range for which postulated

accidents have been evaluated. Therefore

operation within these regions does not

34746 Federal Register / Vol. 65, No. 105 / Wednesday, May 31, 2000 / Notices

create the possibility of a new or different

kind of accident from any accident

previously evaluated. The changes to TS

3.4.1 to be more consistent, or conservative,

with respect to the reviewed and approved

Standard Technical Specifications, simply

provide guidance on the operator actions to

be taken and the associated time limits when

the Specification is entered, and also do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

Therefore, the proposed change, which

specifies limiting conditions for operation,

required actions and surveillance

requirements of the OPRM system, and

allows operation in certain regions of the

power to flow map, does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The proposed change will not involve a

significant reduction in the margin of safety.

The proposed change specifies limiting

conditions for operation, required actions

and surveillance requirements of the OPRM

system and allows operation in regions of the

power to flow map currently restricted by the

requirements of ICAs and TS 3.4.1.

The OPRM system monitors small groups

of LPRM [local power range monitor] signals

for indication of local variations of core

power consistent with T-H oscillations, and

generates an RPS trip when conditions

consistent with the onset of oscillations are

detected. An unmitigated T-H instability

event has the potential to result in a

challenge to the MCPR safety limit. The

OPRM system provides the capability to

automatically detect and suppress conditions

which might result in a T–H instability event,

and thereby maintains the margin of safety by

providing automatic protection for the MCPR

safety limit while reducing the burden on the

control room operators. Therefore, operation

with the OPRM system does not involve a

significant reduction in a margin of safety. In

the event an OPRM channel becomes

inoperable, the proposed change includes

actions which limit the period of time before

the affected OPRM channel (or RPS system)

must be placed in the tripp[ed] condition. If

these actions would result in a trip function

such as a scram (or if the OPRM trip

capability is not maintained), the alternate

method to detect and suppress thermal

hydraulic oscillations (the current ICAs) is

required to be put in place. The duration of

the period of time allowed by the Required

Actions is limited, and the probability of a

significant T–H instability event during this

limited time is not significantly increased.

Operation in regions currently restricted by

the requirements of ICAs and Technical

Specification [TS] 3.4.1 is within the nominal

operating domain and ranges of plant

systems and components, and within the

range assumed for initial conditions

considered in the analysis of anticipated

operational occurrences and postulated

accidents. Therefore, operation in these

regions does not involve a significant

reduction in the margin of safety. The

changes to TS 3.4.1 to be more consistent, or

conservative, with respect to the reviewed

and approved Standard Technical

Specifications, simply provide guidance on

the operator actions to be taken and the

associated time limits when the Specification

is entered, and also do not significantly

reduce the margin of safety.

Therefore, the proposed change, which

specifies limiting conditions for operation,

required actions and surveillance

requirements of the OPRM system, and

allows operation in certain regions of the

power to flow map, 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.

Attorney for licensee: Mary E.

O’Reilly, Attorney, FirstEnergy

Corporation, 76 South Main Street,

Akron, OH 44308.

NRC Section Chief: Anthony J.

Mendiola.

Florida Power and Light Company, et

al., Docket No. 50–335, St. Lucie Plant,

Unit No. 1, St. Lucie County, Florida

Date of amendment request: April 23,

2000.

Description of amendment request:

The proposed license amendment (PLA)

is associated with the required timing

for containment hydrogen recombiner

post operation insulation resistance

testing. This PLA revises Unit 1

Technical Specification 3/4.6.4.2,

Electric Hydrogen Recombiners—W, to

clarify the requirement for the postoperation insulation resistance test of

Surveillance Requirement 4.6.4.2.b.4.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Operation of the facility in accordance

with the proposed amendment would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed amendment does not involve

an increase in the probability or

consequences of any accident previously

evaluated. This PLA provides a clarification

of the Technical Specification surveillance

requirements for verifying hydrogen

recombiner operability and reliability. This

PLA has no affect on the testing

requirements, test frequency, or acceptance

criteria for recombiner operability. This

change allows vendor recommended

guidance and in-house methodology to be

established when conducting recombiner

heater resistance testing. This will enable

consistency in testing and will allow

trending for determination of the material

condition of the recombiner heaters. The PLA

change provides clarification and preserves

the intent of the basis to monitor the material

condition of the recombiner heaters.

Additionally, this change provides

consistency and is identical with the Unit 2

Technical Specification surveillance. As

such, this change is considered

administrative in nature.

2. Operation of the facility in accordance

with the proposed amendment would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed amendment will not create

the possibility of a new or different kind of

accident from any accident previously

evaluated. This PLA is considered

administrative in nature and will not alter

the way in which the hydrogen recombiner

is operated or tested. This PLA allows vendor

recommended guidance to be established in

order to perform consistent testing and to

allow meaningful trending of the results to

verify recombiner operability. This PLA has

no affect on the testing requirements, test

frequency, or acceptance criteria for

recombiner operability. This PLA does not

result in any plant configuration changes or

new failure modes.

3. Operation of the facility in accordance

with the proposed amendment would not

involve a significant reduction in a margin of

safety.

The proposed amendment does not involve

a reduction in the margin of safety. This

administrative PLA clarifies the surveillance

requirement of the subject Technical

Specification by allowing the establishment

of vendor recommendations and in-house

testing methodology to provide consistent

testing conditions and allow meaningful

trending of results. This PLA has no affect on

the testing requirements, test frequency, or

acceptance criteria for recombiner

operability. As such, the assumptions and

conclusions of the accident analyses in the

UFSAR [Updated Final Safety Analysis

Report] remain valid and the associated

safety limits will continue to be met.

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.

Attorney for licensee: M.S. Ross,

Attorney, Florida Power & Light, P.O.

Box 14000, Juno Beach, Florida 33408–

0420.

NRC Section Chief: Richard P.

Correia.

Florida Power and Light Company,

Docket No. 50–251 and 50–252, Turkey

Point Units 3 and 4 in Miami-Dade

County

Date of amendment request: April 27,

2000.

Description of amendment request:

Florida Power and Light Company (FPL)

Federal Register / Vol. 65, No. 105 / Wednesday, May 31, 2000 / Notices 34747

requests to amend the Turkey Point Unit

3 Facility Operating License DPR–31

Fire Protection license condition 3.G,

and to amend the Turkey Point Unit 4

Facility Operating License DPR–41 Fire

Protection license condition 3.F. The

proposed revisions to the Facility

Operating Licenses are required to

incorporate references to NRC Safety

Evaluations issued in support of 10 CFR

50 Appendix R granted exemptions. In

addition, the proposed amendments

requests to modify Appendix A of the

Facility Operating Licenses DPR–31 and

DPR–41 of the Turkey Point Units 3 and

4 Technical Specifications (TS), Section

4.7.6.g. Due to an oversight, the

submittal for the request of License

Amendments Nos. 201 and 195 for

Section 6.0 ‘‘Administrative Controls,’’

L–99–056, dated March 8, 1999,

discussed revision to TS Section 4.7.6.g

on TS Page 3/4 7–21, but inadvertently

did not attach the revised marked up

Page 3/4 7–21.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) Operation of the facility in accordance

with the proposed amendments would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed amendments do not involve

an increase in the probability or

consequences of an accident previously

evaluated because the proposed changes are

administrative in nature adding references to

exemptions granted by the NRC and to reflect

relocation of record retention requirements

from the TS to the UFSAR. These

amendments will not involve a significant

increase in the probability or consequences

of an accident previously evaluated because

they do not affect assumptions contained in

plant safety analyses, the physical design

and/or operation of the plant, nor do they

affect Technical Specifications that preserve

safety analysis assumptions. Therefore, the

proposed changes do not affect the

probability or consequences of accidents

previously analyzed.

(2) Operation of the facility in accordance

with the proposed amendments would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed changes to the Facility

Operating Licenses and the Technical

Specifications are administrative in nature

and can not create the possibility of a new

or different kind of accident from any

previously evaluated since the proposed

amendments will not change the physical

plant or the modes of plant operation defined

in the facility operating license. No new

failure mode is introduced due to the

administrative changes since the proposed

changes do not involve the addition or

modification of equipment nor do they alter

the design or operation of affected plant

systems, structures, or components.

(3) Operation of the facility in accordance

with the proposed amendments would not

involve a significant reduction in a margin of

safety.

The operating limits and functional

capabilities of the affected systems,

structures, and components are unchanged

by the proposed amendments. The proposed

changes to the Facility Operating License

Conditions and the TS are administrative in

nature and do not reduce any of the margins

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.

Attorney for licensee: M.S. Ross,

Attorney, Florida Power & Light, P.O.

Box 14000, Juno Beach, Florida 33408–

0420.

NRC Section Chief: Richard P.

Correia.

North Atlantic Energy Service

Corporation, Docket No. 50–443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request: April 28,

2000.

Description of amendment request:

The Seabrook Station Technical

Specifications (TSs) are proposed to be

revised to implement the Relaxed Axial

Offset Control (RAOC) operating

strategy in support of the use of

upgraded Westinghouse fuel with

Intermediate Flow Mixers.

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. The NRC staff has

reviewed the licensee’s analysis against

the standards of 10 CFR 50.92(c). The

NRC staff’s review is presented below:

The proposed changes do not involve

a significant increase in the probability

or consequences of an accident

previously evaluated.

The proposed changes to TS 2.1.1,

3.2.1, 4.2.1.1, 4.2.2.2, 4.2.2.3, 4.2.2.4,

6.8.1.6.b, and changes to the

aforementioned TS Bases, are in support

of North Atlantic’s long-term operating

strategy to refuel and operate,

commencing with Cycle 8, with

Biweekly Notice Coordinator upgraded

Westinghouse fuel with Intermediate

Flow Mixers (VANTAGE+(w/IFMs)).

Evaluations/analyses of accidents which

are potentially affected by the

parameters and assumptions associated

with the fuel upgrade and RAOC

strategy have shown that all design

standards and applicable safety criteria

will continue to be met. The

consideration of these changes does not

result in a situation where the design,

material, and construction standards

that were applicable prior to the change

are altered. Therefore, the proposed

changes occurring with the fuel upgrade

will not result in any additional

challenges to plant equipment that

could increase the probability of any

previously evaluated accident.

The proposed changes associated with

the fuel upgrade and RAOC strategy do

not affect plant systems such that their

function in the control of radiological

consequences is adversely affected. The

actual plant configuration, performance

of systems, and initiating event

mechanisms are not being changed as a

result of the proposed changes. The

design standards and applicable safety

criteria limits will continue to be met

and therefore fission barrier integrity is

not challenged. The proposed changes

associated with fuel upgrade and RAOC

strategy have been shown not to

adversely affect the response of the

plant to postulated accident scenarios.

The proposed changes will therefore not

affect the mitigation of the radiological

consequences of any accident described

in the Updated Final Safety Analysis

Report (UFSAR).

The proposed changes to TS Table

2.2–1, TS 3.2.2, TS 3.2.3, and the title

on page 3/4 2–6 are editorial changes to

correct either typographical errors,

simplification of statements,

clarification of specific parameters

associated with temperature pressure

measurements, making some notations

consistent with improved Standard

Technical Specifications —

Westinghouse Plants, NUREG–1431,

Rev. 1, and relocating additional cyclespecific values for temperature, pressure

and time constants to the [Core

Operating Limits Report] COLR, or

correcting an erroneous title. These

changes do not result in a change to the

design basis of any plant structure,

system or component or parameters

currently specified in the COLR,

therefore, operation of the facility

within the prescribed limits of TS

remains unchanged.

The proposed change to TS 3.2.1,

ACTION a.2, to delete the need to

reduce the power range neutron flux

high trip setpoints subsequent to

reducing rated thermal power (RTP) to

less than 50% whenever axial flux

difference (AFD) is outside of the

applicable limits specified in the COLR,

does not significantly increase the

34748 Federal Register / Vol. 65, No. 105 / Wednesday, May 31, 2000 / Notices

probability or consequences of an

accident previously evaluated.

Therefore, for the reasons stated

above, the probability or consequences

of an accident previously evaluated are

not significantly increased for all the

proposed TS changes presented herein.

2. Create the possibility of a new or

different kind of accident from any

accident previously evaluated.

The possibility for a new or different

type of accident from any accident

previously evaluated is not created

since the proposed changes associated

with the fuel upgrade and RAOC

strategy do not result in a change to the

design basis of any plant structure,

system or component. These proposed

changes do not cause the initiation of

any accident nor create any new failure

mechanisms. Equipment important to

safety will continue to operate as

designed. Component integrity is not

challenged. The proposed changes do

not result in any event previously

deemed incredible being made credible.

The proposed changes are not

expected to result in conditions that are

more adverse and are not expected to

result in any increase in the challenges

to safety systems.

Therefore, the proposed changes do

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 proposed changes will assure

continued compliance within the

acceptance limits previously reviewed

and approved by the NRC for use of

upgraded fuel features with RAOC. All

of the appropriate acceptance criteria for

the various analyses and evaluations

will continue to be met.

The proposed editorial changes do not

change the current limits specified in

Technical Specifications.

Removing the requirement for

manually reducing the power range

neutron flux high trip setpoint does not

result in a significant reduction in a

margin of safety. There are other levels

of trip protection to terminate a rapid

rise in power excursion, such as the

overtemperature delta-temperature (OT–

T) trip function and previous power

range neutron flux high trip setpoint. In

addition, a rapid rise in power to greater

than 50 percent RTP with AFD outside

limits does not immediately create an

unacceptable situation. The increased

potential for a reactor trip caused by the

manual manipulation of the setpoint

needlessly exposes the plant to an

unnecessary trip with the potential for

an undesirable plant transient which

may unnecessarily challenge safety

systems.

Therefore, the proposed

aforementioned TS changes do not

involve a signification reduction in a

margin of safety.

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.

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

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

NRC Section Chief: James W. Clifford.

Northeast Nuclear Energy Company, et

al., Docket No. 50–423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of amendment request: February

1, 2000, as supplemented by letter dated

April 13, 2000.

Description of amendment request:

The proposed amendment proposes

changes to the cable spreading room

technical specifications to permit

pressurizing the cable spreading room to

a pressure that exceeds the pressure of

the adjacent control room envelope area

during testing.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

In accordance with 10 CFR 50.92, NNECO

has reviewed the proposed changes and has

concluded that they do not involve a

significant hazards consideration (SHC). The

basis for this conclusion is that the three

criteria of 10 CFR 50.92(c) are not

compromised. The proposed changes do not

involve an SHC because the changes would

not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed Technical Specification and

Bases changes to exclude the requirements of

Surveillance Requirements (SRs) 4.7.7.e.2,

4.7.8.c.2, and 4.7.8.c.3 during pressurization

testing of the Cable Spreading Room (CSR)

will not increase the probability of an

accident previously evaluated. Operation of

the Control Room Emergency Air Filtration

System and the Control Room Envelope

Pressurization System cannot cause an

accident to occur.

The proposed Technical Specification and

Bases changes to exclude the requirements of

SRs 4.7.7.e.2, 4.7.8.c.2, and 4.7.8.c.3 during

pressurization testing of the CSR may

adversely impact the consequences of

previously evaluated accidents. During CSR

pressurization testing, the Control Room

Emergency Air Filtration and the Control

Room Envelope Pressurization Systems may

not be able to pressurize and maintain the

Control Room envelope at a positive pressure

with respect to adjacent areas and the outside

atmosphere. As a result, radioactivity

released from a design basis accident may

enter the Control Room envelope. However,

since the CSR area will actually be at a higher

pressure than the outside atmosphere (during

CSR pressurization testing), radioactive

leakage into the CSR area, and subsequently

into the Control Room envelope, should not

occur after the temporary fan has been

stopped. Administrative controls will be

established to immediately stop the

temporary fan and rapidly depressurize the

CSR area in the event Control Building

isolation is necessary. Once the CSR area is

depressurized, the Control Room Emergency

Air Filtration System and the Control Room

Envelope Pressurization System will be able

to function as designed to mitigate the

consequences of the accident. In addition,

the probability of a design basis accident

(DBA) occurring while the CSR is pressurized

is low. Therefore, exempting the

requirements of SRs 4.7.7.e.2, 4.7.8.c.2, and

4.7.8.c.3 during CSR pressurization testing

will not result in a significant increase in the

consequences of an accident previously

evaluated.

The proposed Technical Specification and

Bases change to clarify the mode of operation

of the Control Room Emergency Air Filtration

System when the pressurization requirement

of SR 4.7.7.e.2 applies, will have no adverse

effect on plant operation, or the availability

or operation of any accident mitigation

equipment. The plant response to the design

basis accidents will not change. In addition,

the proposed change can not cause an

accident. Therefore, there will be 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 accident

previously evaluated.

The proposed Technical Specification and

Bases changes to exclude the requirements of

SRs 4.7.7.e.2, 4.7.8.c.2, and 4.7.8.c.3 during

pressurization testing of the CSR, and to

clarify the mode of operation of the Control

Room Emergency Air Filtration System when

the pressurization requirement of SR 4.7.7.e.2

applies, will not alter the plant configuration

(no new or different type of permanent

equipment will be installed) or require any

new or unusual operator actions. Temporary

equipment will be utilized to pressurize the

CSR, and administrative controls, using

additional personnel beyond the normal shift

complement, will be implemented to restore

the CSR to a configuration that will allow the

Control Room Emergency Air Filtration

System and the Control Room Envelope

Pressurization System to function as

designed to mitigate the consequences of an

accident. The temporary equipment and

administrative controls that will be

implemented to perform the CSR

pressurization testing will not introduce any

new failure modes that could result in a new

accident. Also, the response of the plant and

the operators following these accidents is

unaffected by the changes. Therefore, the

proposed changes will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

Federal Register / Vol. 65, No. 105 / Wednesday, May 31, 2000 / Notices 34749

3. Involve a significant reduction in a

margin of safety.

The proposed Technical Specification and

Bases changes to exclude the requirements of

SRs 4.7.7.e.2, 4.7.8.c.2, and 4.7.8.c.3 during

pressurization testing of the CSR may

adversely impact the ability of the Control

Room Emergency Air Filtration System and

the Control Room Envelope Pressurization

System to function as designed to protect the

Control Room Operators following a DBA,

and to use other accident mitigation

equipment contained within the Control

Room envelope. However, the administrative

controls that will be established to

immediately stop the temporary fan and

rapidly depressurize the CSR area if Control

Building isolation is necessary will provide

reasonable assurance that the habitability of

the Control Room envelope will be

maintained. Therefore, exempting the

requirements of SRs 4.7.7.e.2, 4.7.8.c.2, and

4.7.8.c.3 during CSR pressurization testing

will not result in a significant reduction in

a margin of safety.

The proposed Technical Specification and

Bases change to clarify the mode of operation

of the Control Room Emergency Air Filtration

System when the pressurization requirement

of SR 4.7.7.e.2 applies will have no adverse

effect on plant operation, or the availability

or operation of any accident mitigation

equipment. The plant response to the design

basis accidents will not change. Therefore,

there will be no significant reduction in a

margin of safety.

The proposed changes do not alter the

design, function, or operation of the

equipment involved. The impact of the

proposed changes has been analyzed, and it

has been determined they do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated, do not create the possibility of a

new or different kind of accident from any

accident previously evaluated, and do not

involve a significant reduction in a margin of

safety. Therefore, NNECO has concluded the

proposed changes do not involve an SHC.

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.

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

P.O. Box 270, Hartford, Connecticut.

NRC Section Chief: James W. Clifford.

Northern States Power Company,

Docket No. 50–263, Monticello Nuclear

Generating Plant, Wright County,

Minnesota

Date of amendment request: May 4,

2000.

Description of amendment request:

The proposed amendment would add

new sections to the Technical

Specifications (TSs) addressing missed

surveillance test requirements and

establishing a TS Bases control program,

revise TS Chapter 6 to allow use of

generic titles in lieu of plant-specific

titles, allow an alternative when the

radiation protection manager does not

meet the qualifications of Regulatory

Guide 1.8, relocate sections of TS

Chapter 6 pertaining to onsite and

offsite review and special inspections to

the Operational Quality Assurance Plan,

and correct typographical errors.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed amendment will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Operation of the Monticello plant in

accordance with the proposed changes does

not involve a significant increase in the

probability or consequences of an accident

previously evaluated. None of the proposed

changes involves a physical modification to

the plant, a new mode of operation or a

change to the Updated Safety Analysis

Report transient analysis. These proposed

amendments conform to the guidance of

NUREG–1433, Revision 1, which was

previously issued by the NRC.

The proposed changes do not reduce the

level of qualification or training and the

aggregate knowledge of the plant staff

remains intact. In total, these changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed amendment will not

create the possibility of a new or different

kind of accident from any accident

previously analyzed.

The proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated because the proposed changes do

not introduce a new mode of plant operation,

surveillance test requirement or involve a

physical modification to the plant. These

proposed amendments generally conform to

the guidance of NUREG–1433, Revision 1,

which was previously issued by the NRC.

The proposed changes are administrative

in nature. The changes propose to relocate

specifications from the Technical

Specifications to the Operational Quality

Assurance Plan through which adequate

control is maintained.

The proposed changes do not alter the

design, function or operation of any plant

component and therefore no new accident

scenarios are created. Therefore, the

possibility of a new or different kind of

accident from any accident previously

evaluated would not be created by these

amendments.

3. The proposed amendment will not

involve a significant reduction in the margin

of safety.

The proposed changes do not involve a

significant reduction in a margin of safety

because the current Technical Specification

requirements for safe operation of the

Monticello plant are maintained. The

proposed changes are administrative in

nature and do not involve a physical

modification to the plant, a new mode of

operation or a change to the Updated Safety

Analysis Report transient analyses. The

proposed changes do not alter the scope of

equipment currently required to be operable

or subject to surveillance testing nor does the

proposed change affect any instrument

setpoints or equipment safety functions.

Therefore, a significant reduction in the

margin of safety would not be involved with

these proposed changes.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Jay E. Silberg,

Esq., Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW,

Washington, DC 20037.

NRC Section Chief: Claudia M. Craig.

Southern California Edison Company, et

al., Docket Nos. 50–361 and 50–362,

San Onofre Nuclear Generating Station,

Units 2 and 3, San Diego County,

California

Date of amendment requests: May 3,

2000 (PCN–516).

Description of amendment requests:

The amendment application proposes to

revise the San Onofre Nuclear

Generating Station, Units 2 and 3,

Technical Specification (TS) 3.4.3, ‘‘RCS

Pressure and Temperature (P/T)

Limits,’’ and the associated Bases. The

proposed change would reduce the

minimum boltup temperature from 86

°F to 65 °F for the reactor coolant system

during the period of time when the

reactor vessel head bolts are in tension.

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:

Will operation of the facility in accordance

with this proposed change involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

Response: No.

This proposed change is a request to revise

Technical Specification 3.4.3, ‘‘Pressure

Temperature Limits.’’ The proposed change

reduces the Minimum Boltup Temperature

(MBT) from 86°F to 65°F. During operations

below 86°F, the plant is in a shutdown mode,

open to the atmosphere, and depressurized.

34750 Federal Register / Vol. 65, No. 105 / Wednesday, May 31, 2000 / Notices

This proposed change does not affect the

shape of the Pressure Temperature Limits

when Reactor Coolant System (RCS)

temperature is above 86°F. Therefore, the

probability or consequences of an accident

previously evaluated will not be increased by

operating the facility in accordance with this

proposed change.

Will operation of the facility in accordance

with this proposed change create the

possibility of a new or different kind of

accident from any accident previously

evaluated?

Response: No.

This proposed change does not change the

design or configuration of the plant.

Therefore, this proposed change will not

create the possibility of a new or different

kind of accident from any accident that has

been previously evaluated.

(3) Will operation of the facility in

accordance with this proposed change

involve a significant reduction in a margin of

safety?

Response: No.

This proposed change involves reducing

the MBT from 86°F to 65°F. This proposed

change meets the American Society of

Mechanical Engineers (ASME) Code

requirements for establishing the minimum

temperature in the reactor pressure vessel

flange region when the pressure does not

exceed 20% of the pre-operational

hydrostatic test pressure. All margins of

safety established by the ASME Code

requirements are maintained. The operation

of the facility in accordance with this

proposed change will not involve a

significant reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment requests involve no

significant hazards consideration.

Attorney for licensee: Douglas K.

Porter, Esquire, Southern California

Edison Company, 2244 Walnut Grove

Avenue, Rosemead, California 91770.

NRC Section Chief: Stephen Dembek.

Tennessee Valley Authority, Docket No.

50–390 Watts Bar Nuclear Plant, Unit

1(WBN), Rhea County, Tennessee

Date of amendment request: April 10,

2000 (TS 99–013).

Description of amendment request:

The proposed amendment requests

NRC’s approval to use the F* alternate

repair criterion in the tubesheet region

of the steam generator (SG). The F*

criterion addresses the action required

when degradation has been detected in

the top of the mechanically expanded

portion of SG tubes within the SG

tubesheet.

The proposed change designates a

portion of the tube for which tube

degradation of a defined type does not

necessitate remedial action, except as

dictated for compliance with tube

leakage limits as set forth in the WBN

Technical Specifications (TS). The

proposed amendment would modify the

TS which provide tube inspection

requirements and acceptance criteria to

determine the level of degradation for

which the tube may remain in service.

The proposed amendment would add

definitions required for the F* alternate

plugging criterion and prescribe the

portion of the tube subject to the

criterion.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The proposed amendment does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The presence of the tubesheet enhances the

tube integrity in the region of the hardroll by

precluding tube deformation beyond its

initial expanded outside diameter. The

resistance to both tube rupture and tube

collapse is strengthened by the presence of

the tubesheet in that region. Hardrolling of

the tube into the tubesheet results in an

interference fit between the tube and the

tubesheet. Tube rupture cannot occur

because the contact between the tube and

tubesheet does not permit sufficient

movement of tube material. In a similar

manner, the tubesheet does not permit

sufficient movement of tube material to

permit buckling collapse of the tube during

postulated loss-of-coolant-accident (LOCA)

loadings.

The type of degradation for which the

alternate plugging criterion, F*, has been

developed (cracking with a circumferential

orientation) can theoretically lead to a

postulated tube rupture event, provided that

the postulated through-wall circumferential

crack exists near the top of the tubesheet. An

evaluation including analysis and testing has

been performed to determine the resistive

strength of roll expanded tubes within the

tubesheet. That evaluation provides the basis

for the acceptance criteria for tube

degradation subject to the F* criterion.

The F* length of roll expansion is

sufficient to preclude tube pullout from tube

degradation located below the F* distance,

regardless of the extent of the tube

degradation. The existing technical

specification leakage rate requirements and

accident analysis assumptions remain

unchanged in the unlikely event that

significant leakage from this region does

occur. As noted above, tube rupture and

pullout are not expected for tubes using the

F* alternate plugging criterion. Any leakage

out of the tube from within the tubesheet at

any elevation in the tubesheet is fully

bounded by the existing steam generator tube

rupture analysis included in the WBN Unit

1 Final Safety Analysis Report (FSAR). The

proposed alternate plugging criterion (F*)

does not adversely impact any other

previously evaluated design basis accident.

B. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

Implementation of the proposed F*

tubesheet alternate plugging criterion does

not introduce any significant changes to the

plant design basis. Use of the criterion does

not provide a mechanism to result in an

accident initiated outside of the region of the

tubesheet expansion. A hypothetical accident

as a result of any tube degradation in the

expanded portion of the tube would be

bounded by the existing tube rupture

accident analysis. Tube bundle structural

integrity and leaktightness are expected to be

maintained. Therefore, the proposed change

does not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

C. The proposed amendment does not

involve a significant reduction in a margin of

safety.

The use of the tubesheet alternate plugging

criterion has been demonstrated to maintain

the integrity of the tube bundle

commensurate with the requirements of

Regulatory Guide 1.121, ‘‘Bases for Plugging

Degraded PWR Steam Generator Tubes,’’ for

indications in the free span of tubes and the

primary to secondary pressure boundary

under normal and postulated accident

conditions. Acceptable tube degradation for

the F* criterion is any degradation indication

in the tubesheet region, more than the F*

distance below either the bottom of the

transition between the roll expansion and the

unexpanded tube, or the top of the tubesheet,

whichever is lower. The safety factors used

in the verification of the strength of the

degraded tube are consistent with the safety

factors in the American Society of

Mechanical Engineering (ASME) Boiler and

Pressure Vessel Code used in steam generator

design. The F* distance has been verified by

testing to be greater than the length of roll

expansion required to preclude both tube

pullout and significant leakage during

normal and postulated accident conditions.

Resistance to tube pullout is based upon the

primary to secondary pressure differential as

it acts on the surface area of the tube, which

includes the tube wall cross-section, in

addition to the inside diameter-based area of

the tube. The leak testing acceptance criteria

are based on the primary to secondary

leakage limit in the technical specifications

and the leakage assumptions used in the

FSAR accident analyses.

Implementation of the alternate tubesheet

plugging criterion will decrease the number

of tubes which must be taken out of service

with tube plugs or repaired with sleeves.

Both plugs and sleeves reduce the RCS flow

margin; thus, implementation of the F*

alternate plugging criterion will maintain the

margin of flow that would otherwise be

reduced in the event of increased plugging or

sleeving. Based on the above, it is concluded

that the proposed change does not result in

a significant reduction in a loss of margin

with respect to plant safety as defined in the

FSAR or the bases of the WBN technical

specifications.

Federal Register / Vol. 65, No. 105 / Wednesday, May 31, 2000 / Notices 34751

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.

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 10H,

Knoxville, Tennessee 37902.

NRC Section Chief: Richard P.

Correia.

Tennessee Valley Authority (TVA),

Docket No. 50–390 Watts Bar Nuclear

Plant (WBN), Unit 1, Rhea County,

Tennessee

Date of amendment request: April 10,

2000 (TS 99–014).

Description of amendment request:

The proposed amendment would revise

the WBN Unit 1 Technical Specification

(TS) to incorporate new requirements

associated with steam generator (SG)

tube inspection and repair. The new

requirements establish an alternate

voltage based SG tube repair criteria at

tube support plate and Flow

Distribution Baffle plate intersections.

This change is consistent with NRC

Generic Letter 95–05 ‘‘Voltage-Based

Repair Criteria for Westinghouse Steam

Generator Tubes Affected By Outside

Diameter Stress Corrosion Cracking.’’

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

A. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Tube burst criteria are inherently satisfied

during normal operating conditions due to

the proximity of the tube support plate. Test

data indicates that tube burst cannot occur

within the tube support plate (TSP), even for

tubes which have 100 percent through-wall

electric discharge machining (EDM) notches,

0.75 inches long, provided that the TSP is

adjacent to the notched area. Since tube to

tube support plate proximity precludes tube

burst during normal operating conditions,

use of the criteria must retain tube integrity

characteristics which maintain a margin of

safety of 1.43 times the bounding faulted

condition [main steam line break (MSLB)]

differential pressure of 2405 psig. As

previously stated, the Regulatory Guide (RG)

1.121 criterion requiring maintenance of a

safety factor of 1.43 times the MSLB pressure

differential on tube burst is satisfied by 3⁄4-

inch diameter tubing with bobbin coil

indications with signal amplitudes less than

VSL = 6.03 volts, regardless of the indicated

depth measurement. At the flow distribution

baffle (FDB), a safety factor of 3 against the

normal operating condition DP is applied;

here a voltage of VSL = 3.81 volts satisfies the

burst capability recommendation.

The upper voltage repair limit (VURL) will

be determined prior to each outage using the

most recently approved NRC database to

determine the tube structural limit (VSL). The

structural limit is reduced by allowances for

nondestructive examination (NDE)

uncertainty (VNDE) and growth (VG) to

establish VURL. As an example, the NDE

uncertainty component of 20 percent and a

voltage growth allowance of 30 percent per

full power year can be utilized to establish

a VURL of 3.71 volts for TSP indications, and

2.34 volts for the FDB indications. The 20

percent NDE uncertainty represents a squareroot-sum-of-the-squares (SRSS) combination

of probe wear uncertainty and analyst

variability. The flaw growth allowance

should be an average growth rate or 30

percent per effective full power year,

whichever is larger. The 30 percent growth

allowance used to determine VURL is

conservative for the current conditions at

WBN Unit 1. The most current NRC

approved database, contained in EPRI

[Electric Power Research Institute] NP–7480–

L, Addendum 2, was used to establish the

VURL values for the FDB and TSP

intersections. Once approved by NRC, the

industry protocol for updating the database

will be followed by TVA, ensuring that the

most current database is utilized for future

applications of the criteria.

Relative to the expected leakage during

accident condition loadings, it has been

previously established that a postulated

MSLB outside of containment but upstream

of the main steam isolation valves (MSIV)

represents the most limiting radiological

condition relative to the alternate voltage

based repair criteria. In support of

implementation of the revised repair limit, it

will be determined whether the distribution

of cracking indications at the tube support

plate intersections during future cycles are

projected to be such that primary to

secondary leakage would result in site

boundary doses within a fraction of the 10

CFR 100 guidelines or control room doses

within the 10 CFR 50, Appendix A, General

Design Criteria (GDC)–19 limit. A separate

calculation has determined this allowable

MSLB leakage limit to be 10 gallons per

minute (gpm) in the faulted loop assuming a

reactor coolant system (RCS) dose equivalent

iodine concentration of 1.0 mCi/gm. The

establishment of the 10 gpm leak rate value

is controlled by the 0 to 2 hour offsite doses

at the site boundary for the accident initiated

iodine spike case, not control room dose.

The methods for calculating the

radiological dose consequences for this

MSLB are consistent with FSAR Chapter 15

and therefore, the WBN licensing basis. TVA

bases the calculated thyroid dose

consequences on conversion factors from the

International Commission on Radiation

Protection (ICRP) Publication 2.

In summary, the calculated radiological

consequences of the exclusion area boundary

and the low population zone are larger than

previously reported for the postulated

steamline break event due to the increased

leakage and more conservative iodine spiking

factors. However, the calculated radiological

consequences remain in compliance with the

guidelines in the Standard Review Plan,

Chapter 15, 10 CFR 50, Appendix A, GDC–

19 and 10 CFR 100 reported for the

postulated steamline break event. Therefore,

it is concluded that the proposed changes do

not result in a significant increase in the

radiological consequences of an accident

previously analyzed.

Consistent with the guidance of Section 2.c

of Generic Letter (GL) 95–05, the WBN Unit

1 MSLB leak rate analysis performed prior to

returning the SGs to service may be

performed based on the projected next endof-cycle (EOC) voltage distribution or the

actual measured distribution at a given

outage. The method to be used for the first

outage when outside diameter stress

corrosion cracking (ODSCC) indication

growth rates are available will be based on

the indications found during that outage. As

noted in GL 95–05, it may not always be

practical to complete EOC calculations prior

to returning the SGs to service. Under these

circumstances, it is acceptable to use the

actual measured bobbin voltage distribution

instead of the projected EOC voltage

distribution to determine whether the

reporting criteria is being satisfied.

Therefore, as implementation of the 1.0

volt voltage-based repair criteria at WBN Unit

1 does not adversely affect steam generator

tube integrity and implementation is shown

to result in acceptable radiological dose

consequences, the proposed TS change does

not result in a significant increase in the

probability or consequences of an accident

previously evaluated within the WBN Final

Safety Analysis Report (FSAR).

B. The proposed amendment does not

create the possibility of a new or different

kind of accident from previously analyzed.

Implementation of the proposed steam

generator tube alternate voltage based repair

criteria (1.0 volts) does not introduce any

significant changes to the plant design basis.

Neither a single or multiple tube rupture

event would be expected in a steam generator

in which the repair limit has been applied

(during all plant conditions).

The bobbin probe voltage-based tube repair

criteria of 1.0 volt is supplemented by:

enhanced eddy current inspection guidelines

to provide consistency in voltage

normalization, a 100 percent eddy current

inspection sample size at the tube support

plate elevations, and rotating pancake coil

(RPC) inspection requirements for the larger

indications left in service to characterize the

principal degradation as ODSCC.

TVA will implement a maximum normal

operating condition primary to secondary

leakage rate limit of 600 gallons per day (gpd)

total primary to secondary leakage and 150

gpd primary to secondary leakage per steam

generator to help preclude the potential for

excessive leakage during all plant conditions.

The 150 gpd leakage limit is more restrictive

than the current TS operating leakage limit

(of 500 gpd) and is intended to provide

additional margin to accommodate a stress

corrosion crack which might grow at a greater

than expected rate or unexpectedly extend

outside the thickness of the tube support

34752 Federal Register / Vol. 65, No. 105 / Wednesday, May 31, 2000 / Notices

plate. Leakage trending capability consistent

with EPRI Report TR–104788, ‘‘Primary-toSecondary Leak Guidelines’’ has been

implemented at WBN Unit 1.

As steam generator tube integrity upon

implementation of the 1.0 volt repair limit

continues to be maintained through inservice inspection and primary to secondary

leakage monitoring, the possibility of a new

or different kind of accident from any

accident previously evaluated is not created.

C. The proposed amendment does not

involve a significant reduction in a margin of

safety.

The use of the voltage-based bobbin probe

tube support plate elevation repair criteria at

WBN Unit 1 maintains steam generator tube

integrity commensurate with the criteria of

Regulatory Guide (RG) 1.121. RG 1.121

describes a method acceptable to the NRC

staff for meeting GDCs 14, 15, 31, and 32 by

reducing the probability or the consequences

of steam generator tube rupture. This is

accomplished by determining the limiting

conditions of degradation of steam generator

tubing, as established by in-service

inspection, for which tubes with

unacceptable cracking should be removed

from service. Upon implementation of the

proposed criteria, even under the worst case

conditions, the occurrence of ODSCC at the

tube support plate elevations is not expected

to lead to a steam generator tube rupture

event during normal or faulted plant

conditions. The EOC distribution of crack

indications at the tube support plate

elevations is confirmed to result in

acceptable primary to secondary leakage

during all plant conditions and that

radiological consequences are not adversely

impacted.

As a preventative measure, a total of 214

tubes are excluded from the application of

the ODSCC criteria because of the combined

effects of loss-of-coolant-accident (LOCA) +

safe shutdown earthquake (SSE) on the steam

generator component (as required by GDC 2).

It was determined that tube deformation or

through-wall cracks could occur in these

tubes.

As noted previously, implementation of

the tube support plate intersection voltagebased repair criteria will decrease the

number of tubes which must be repaired. The

installation of steam generator tube plugs

reduces the RCS flow margin. Thus,

implementation of the 1.0 volt repair limit

will maintain the margin of flow that would

otherwise be reduced in the event of

increased tube plugging.

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.

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 10H,

Knoxville, Tennessee 37902.

NRC Section Chief: Richard P.

Correia.

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

electronically from the ADAMS Public

Library component on the NRC Web

site, http://www.nrc.gov (the Electronic

Reading Room).

Carolina Power & Light Company, et al.,

Docket Nos. 50–325 and 50–324,

Brunswick Steam Electric Plant, Units 1

and 2, Brunswick County, North

Carolina

Date of application for amendments:

November 23, 1999.

Brief description of amendments: The

amendments changed Technical

Specification 5.5.7.c.1, ‘‘Ventilation

Filter Testing.’’ The testing criteria have

been changed to be consistent with the

NRC request in Generic Letter 99–02,

‘‘Laboratory Testing of Nuclear-Grade

Activated Charcoal.’’

Date of issuance: May 16, 2000.

Effective date: May 16, 2000.

Amendment Nos.: 209 and 237.

Facility Operating License Nos. DPR–

71 and DPR–62: Amendments change

the Technical Specifications.

Date of initial notice in Federal

Register: December 29, 1999 (64 FR

73086) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

May 16, 2000.

No significant hazards consideration

comments received: No.

Commonwealth Edison Company,

Docket Nos. STN 50–454 and STN 50–

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

Ogle County, Illinois

Docket Nos. STN 50–456 and STN 50–

457, Braidwood Station, Unit Nos. 1 and

2, Will County, Illinois

Date of application for amendments:

December 22, 1999, as supplemented on

March 1, 2000.

Brief description of amendments: The

amendments relocate Reactor Coolant

System (RCS) related cycle-specific

parameter limits from the technical

specifications to, and thus expanding,

the Core Operating Limits Reports

(COLRs) for Byron, Units 1 and 2, and

Braidwood, Units 1 and 2.

Date of issuance: May 15, 2000.

Effective date: Immediately, to be

implemented within 30 days.

Amendment Nos.: 113 and 106.

Facility Operating License Nos. NPF–

37, NPF–66, NPF–72 and NPF–77: The

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 23, 2000 (65 FR 9003).

The March 1, 2000, submittal provided

additional clarifying information that

did not change the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated May 15, 2000.

No significant hazards consideration

comments received: No.

Commonwealth Edison Company,

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

County Station, Units 1 and 2, LaSalle

County, Illinois

Date of application for amendments:

July 14, 1999, as supplemented on

January 21, February 15, February 23,

March 10, March 24, March 31 (two

letters), April 7 and April 14, 2000.

Brief description of amendments: The

amendments increase the maximum

reactor core power level to 3489

megawatts thermal; an increase of 5

percent of rated core thermal power, for

Federal Register / Vol. 65, No. 105 / Wednesday, May 31, 2000 / Notices 34753

LaSalle County Station, Units 1 and 2.

In addition, the proposed amendments

correct a non-conservative value in the

upper limit for drywell and suppression

chamber internal pressure that was

discovered during the course of the

power uprate review.

Date of issuance: May 9, 2000.

Effective date: For Unit 1,

immediately, to be implemented within

60 days; for Unit 2, immediately, to be

implemented prior to start up of cycle

9.

Amendment Nos.: 140 and 125.

Facility Operating License Nos. NPF–

11 and NPF–18: The amendments

revised the license and Technical

Specifications.

Date of initial notice in Federal

Register: August 25, 1999 (64 FR 46427).

The letters dated January 21, February

15, February 23, March 10, March 24,

two letters on March 31, April 7, and

April 14, 2000, contain supplemental,

clarifying information that did not

change the staff’s initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated May 9, 2000.

No significant hazards consideration

comments received: No.

Commonwealth Edison Company,

Docket No. 50–374, LaSalle County

Station, Unit 2, LaSalle County, Illinois

Date of application for amendment:

February 28, 2000, as supplemented on

April 28, 2000

Brief description of amendment: The

amendment increases the Technical

Specification safety limit for the

Minimum Critical Power Ratio from

1.08 for two loop operation and 1.09 for

single loop operation to 1.11 and 1.12,

respectively.

Date of issuance: May 17, 2000.

Effective date: Immediately, to be

implemented within 60 days.

Amendment No.: 126.

Facility Operating License No. NPF–

18: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: March 22, 2000 (65 FR 15377).

The April 28, 2000, submittal provided

clarifying information that did not

change the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 17, 2000.

No significant hazards consideration

comments received: No.

Duke Energy Corporation, 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:

August 4, 1999, as supplemented by

letter dated April 19, 2000.

Brief description of amendments: The

amendments revise the Technical

Specifications (TS) Limiting Conditions

for Operation for Reactor Coolant

System (RCS) Subcooling Margin

Monitor in TS Table 3.3.3–1 and revise

the functions associated with

surveillance requirements for RCS

Loops-Test Exceptions in TS 3.4.17. By

letter dated April 19, 2000, the licensee

withdrew the proposal to relocate the

Auxiliary Feedwater Loss of Offsite

Power function from TS 3.3.2–1 to TS

3.3.2–1. The other changes requested by

August 4,1999, application were

addressed under separate

correspondence.

Date of issuance: May 19, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment Nos.: 186 and 179.

Facility Operating License Nos. NPF–

35 and NPF–52: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: September 8, 1999 (64 FR

48861)

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated May 19, 2000.

No significant hazards consideration

comments received: No

Entergy Operations, Inc., Docket No. 50–

313, Arkansas Nuclear One, Unit No. 1,

Pope County, Arkansas

Date of amendment request: August

18, 1999, as supplemented by letter

dated April 20, 2000.

Brief description of amendment: The

requested change would revise

Technical Specification 3.5.3, ‘‘Safety

Feature Actuation System Setpoints,’’

and its associated Bases to allow for an

increase to the low reactor coolant

system pressure setpoint. This setpoint

change was requested to account for

additional instrument uncertainties

associated with cable insulation

resistance effects and to allow for the

plugging of up to 1200 tubes in each

steam generator.

Date of issuance: May 10, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment No.: 207.

Facility Operating License No. DPR–

51: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 26, 2000 (65 FR 4270).

The April 20, 2000, letter provided

clarifying information that did not

change the scope of the August 18,

1999, application and the initial

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 10, 2000.

No significant hazards consideration

comments received: No.

Entergy Operations, Inc., Docket No. 50–

382, Waterford Steam Electric Station,

Unit 3, St. Charles Parish, Louisiana

Date of amendment request: July 15,

1999, as supplemented by letters dated

March 29, 2000, April 13, 2000, April

25, 2000, and May 9, 2000.

Brief description of amendment: The

amendment changed the Technical

Specifications to institute a Technical

Specification Bases Control Program

and to provide for record retention as

specified in the Quality Assurance

Program Manual.

Date of issuance: May 9, 2000.

Effective date: As of the date of

issuance and shall be implemented 60

days from the date of issuance.

Amendment No.: 161

Facility Operating License No. NPF–

38: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: January 26, 2000 (65 FR 4274).

The supplements dated March 29, 2000,

April 13, 2000, April 25, 2000, and May

9, 2000, did not change the scope of the

initial proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 9, 2000. No

significant hazards consideration

comments received: No.

Entergy Operations, Inc., Docket No. 50–

382, Waterford Steam Electric Station,

Unit 3, St. Charles Parish, Louisiana

Date of amendment request: July 19,

1999.

Brief description of amendment: The

proposed change modifies Technical

Specification 4.5.2.f.2 by increasing the

performance requirement for the low

pressure safety injection pumps.

Date of issuance: May 10, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 60 days from the date of

issuance.

Amendment No. 162.

34754 Federal Register / Vol. 65, No. 105 / Wednesday, May 31, 2000 / Notices

Facility Operating License No. NPF–

38: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: January 26, 2000 (65 FR 4277).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 10, 2000.

No significant hazards consideration

comments received: No.

Entergy Operations, Inc., Docket No. 50–

382, Waterford Steam Electric Station,

Unit 3, St. Charles Parish, Louisiana

Date of amendment request: July 29,

1999.

Brief description of amendment: The

amendment changed the Technical

Specifications (TS) to extend the

allowable outage time to seven days for

one containment spray system (CSS)

train inoperable. A new ACTION has

been added to provide a shutdown

requirement for the inoperability of two

CSSs. The associated changes to TS

Bases are included. However, the

licensee requested MODE 4 end state for

TS 3.6.2.1 is being deferred.

Date of issuance: May 15, 2000.

Effective date: As of the date of

issuance and shall be implemented 90

days from the date of issuance.

Amendment No.: 163.

Facility Operating License No. NPF–38:

The amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 9, 2000 (65 FR 6406).

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated May 15, 2000.

No significant hazards consideration

comments received: No.

PECO Energy Company, Docket Nos.

50–352 and 50–353, Limerick

Generating Station (LGS), Units 1 and 2,

Montgomery County, Pennsylvania

Date of application for amendments:

June 22, 1999, as supplemented January

3, 2000.

Brief description of amendments: The

amendments remove the recirculation

system motor generator set stop

surveillance requirement from the LGS

Units 1 and 2 Technical Specifications.

Date of issuance: May 8, 2000.

Effective date: Both units—As of date

of issuance, to be implemented within

30 days.

Amendment Nos.: 142 and 104.

Facility Operating License Nos. NPF–

39 and NPF–85. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: September 8, 1999 (64 FR

48864). The January 3, 2000, letter

provided clarifying information that did

not change the initial proposed no

significant hazards consideration

determination or expand the scope of

the original Federal Register Notice.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated May 8, 2000.

No significant hazards consideration

comments received: No

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

electronically from the ADAMS Public

Library component on the NRC Web

site, http://www.nrc.gov (the Electronic

Reading Room).

The Commission is also offering an

opportunity for a hearing with respect to

the issuance of the amendment. By June

14, 2000, 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

Federal Register / Vol. 65, No. 105 / Wednesday, May 31, 2000 / Notices 34755

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

electronically from the ADAMS Public

Library component on the NRC Web

site, http://www.nrc.gov (the Electronic

Reading Room). If a request for a hearing

or petition for leave to intervene is filed

by the above date, the Commission or an

Atomic Safety and Licensing Board,

designated by the Commission or by the

Chairman of the Atomic Safety and

Licensing Board Panel, will rule on the

request and/or petition; and the

Secretary or the designated Atomic

Safety and Licensing Board will issue a

notice of a hearing or an appropriate

order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) The nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses. Since the Commission has

made a final determination that the

amendment involves no significant

hazards consideration, if a hearing is

requested, it will not stay the

effectiveness of the amendment. Any

hearing held would take place while the

amendment is in effect.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555–0001, Attention:

Rulemakings and Adjudications Staff or

may be delivered to the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, by the above date. 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 the

factors specified in 10 CFR

2.714(a)(1)(i)–(v) and 2.714(d).

Detroit Edison Company, Docket No.

50–341, Fermi 2, Monroe County,

Michigan

Date of amendment request: May 8,

2000.

Description of amendment request:

The amendment revises Technical

Specification Surveillance Requirement

3.8.1.9 to increase the limit for the peak

transient voltage measured following a

full-load rejection by the emergency

diesel generator that is being tested.

Date of issuance: May 9, 2000.

Effective date: As of its date of

issuance and shall be implemented

within 2 days.

Amendment No.: 140.

Facility Operating License No. NPF–43:

Amendment revises the Technical

Specifications. 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 May 9, 2000.

Attorney for licensee: John Flynn,

Esq., Detroit Edison Company, 2000

Second Avenue, Detroit, Michigan,

48226.

NRC Section Chief: Claudia M. Craig.

Dated at Rockville, Maryland, this 24th day

of May 2000.

For the Nuclear Regulatory Commission.

John A. Zwolinski,

Director, Division of Licensing Project

Management, Office of Nuclear Reactor

Regulation.

[FR Doc. 00–13518 Filed 5–30–00; 8:45 am]

BILLING CODE 7590–01–P

OCCUPATIONAL SAFETY AND

HEALTH REVIEW COMMISSION

Agency Information Collection

Activities: Proposed Collection;

Comment Request; Evaluation of the

‘‘E–Z Trial’’

ACTION: Notice.

SUMMARY: As part of its effort to reduce

paperwork and the burden placed on

survey recipients, the Occupational

Safety and Health Review Commission

(OSHRC) is conducting a preclearance

consultation to provide the general

public and Federal agencies with an

opportunity to comment on a proposed

collection of information in accordance

with the Paperwork Reduction Act of

1965, Public Law 104–13. OSHRC is

soliciting comment concerning an

information collection required to

evaluate the Review Commission’s ‘‘E–

Z–Trial’’ program.